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1996

Stress corrosion crack detection in alloy 600 in high temperature caustic.

Brisson, Bruce W.

Monterey California. Naval Postgraduate School http://hdl.handle.net/10945/9017

This publication is a work of the U.S. Government as defined in Title 17, United States Code, Section 101. Copyright protection is not available for this work in the United States.

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DUDL. sHARY NAVAL POSTGRADUATE SCHOOL MONTEREY CA 93943-5101

Stress Corrosion Crack Detection in Alloy 600 in High Temperature Caustic by Bruce W. Brisson B.S. Mechanical Engineering

Syracuse University

SUBMITTED TO THE DEPARTMENT OF NUCLEAR ENGINEERING IN PARTIAL AUEFICEMENTOF THE REQUIREMENTS FOR THE DEGREE OF

NUCLEAR ENGINEER and MASTER OF SCIENCE IN NUCLEAR ENGINEERING AT THE MASSACHUSETTS INSTITUTE OF TECHNOLOGY

JUNE 1996 © 1996 Bruce W. Brisson. All Rights Reserved.

The author hereby grants to MIT permission to reproduce and to distribute publicly paper and electronic copies of this thesis document in whole or in part.

nil \t cy WMV i А ғ Ty - PA ғ LIBRAR

ИН АКЕП SCHOOL MUNICMEY CA 93943-5101

Prrees Corrosion Crack Detection in Alloy 600 In High Temperature Caustic

by Bruce William Brisson

submitted to the Department of Nuclear Engineering on May 1996 in partial fulfillment of the

Requirements for the Degree of Nuclear Engineer and Master of Science in Nuclear Engineering

Abstract

Alloy 600, the material used for pressurized water reactor steam generator tubing, is susceptible to environmentally assisted stress corrosion cracking. Intergranular stress corrosion cracking (IGSCC) attacks the tubes in areas of high residual stress, and in crevice regions. No method has been successfully developed to monitor steam generator tubing in- situ for crack initiation and growth. Essentially all available published IGSCC crack growth data for alloy 600 is based on non-tubing material. Although it is very likely that the current data base is applicable to tubing processing, differences between tube and other geometries make a comparison between tubing and other data important for verification purposes. However, obtaining crack initiation and growth data from tubing is difficult due to the geometry and the thin wall thickness.

In this research the goal was to develop a system to monitor crack initiation and growth under in-situ steam generator conditions in tubing and, having developed such a system, to then obtain data on actual alloy 600 tubing for comparison to published results for non-tubing material.

A nickel autoclave system capable of testing pressurized tube samples was constructed. The system is capable of operating in either recirculating or static mode as assembled. The operating environment was 10% NaOH (with 0.1% Na,CO, added)

with a hydrogen over-pressure. The test sample was internally pressurized and electrically isolated from the rest of the system. Crack initiation and growth was monitored using an alternating current potential drop (ACPD) method especially developed for this purpose. Sample polarization to +150mV with respect to nickel was imposed to accelerate crack initiation.

Mill annealed alloy 600 tubing was chosen as the test material. A special heat of material, fabricated for industry-wide research by the Electric Power Research Institute was used.

The material, heat 96834, was mill annealed at 927+14°C.

The program was divided into two phases: (1) system/methods development, and (2) crack growth measurement.

A multi-frequency ACPD system was developed for the detection of initiation and propagation of short cracks in alloy 600 tubing in high temperature caustic environments typical of an operating steam generator tube to tube sheet crevice. The system was then used to monitor crack initiation and growth in alloy 600 tubing.

Limited IGSCC crack growth data for the tube samples was obtained. Average values of K; ranged from 4.2 to 17,7 МРаүт ПИШЕ growth rates from 1.6 to 12.3 mm/yr.

The data obtained compare well with other, non-tubing, data in the sense that there is overlap between the data from this program and the general body of data in the literature. However, the data developed here indicates (limited as the data is) that a power law dependence of crack growth rate on K may be appropriate. The general body of literature data is too scattered to show any such dependence.

Thesis Supervisor: Ronald Ballinger Title: Associate Professor of Nuclear Engineering and Materials Science and Engineering

Acknowledgments

MEcSunotesutticiently express my gratitude for the material support and advice given by numerous companies and individuals in constructing and operating the equipment for this thesis. I would especially like to thank the Electric Power Research Institute and Dr. McIlree for their generous financial support and advice, Peter Stahle, our lab engineer, for his expertise and guidance in the many disciplines it required to make things go, and Dr. Hui Au for his technical expertise in software development.

I extend my sincere thanks and gratitude to Professor Ronald Ballinger, my thesis advisor, for his guidance and persistent encouragement that “things really would work”. Without his expertise and constant remedies to apparent dead ends, this thesis would not have been possible.

Finally, I would like to dedicate this work to my wife Mary and my father, Marcel:

To Mary, I owe an unpayable debt for her constant encouragement, endless hours proof reading my papers and thesis, and somehow managing our four small children when I needed to study, write or stay at the lab late. I will never forget the love and support.

To my father, Marcel, who encouraged me from first grade on and was always there when I needed him, but passed away before I started at MIT, Thanks Dad.

Table of Contents

A ПИ EE. sacs aee cM s s ee ee 2 O ее. 4 TEA O AS NA ЕТИ aste veteres e iere eie e 5 Е ее. 7 В... 10 ее... 11 о. 14 2.1 Nuclear Steam Generator Materials and Environment ........... 14 EE mcam ene raros (COnstruetioln КОО ООС КО ое. 14 PEE cSrsameCenerator Chemistty..... v... o isa i nbn URL RU. ROS ee 16 2.1.3 Material Properties of Alloy 600 Steam Generator Tubes...18

MEM Iss Corrosion Cracking Theories .....-:4.9..- e ae e eins 22 2 Rupture theory Lor Crack Initiation... sten 22 Ир 2erackeEOpasation bycAnodiesDissolution..-..:. 7.2: 4.99245 26 Naci Emopagation by Мчетотота Formation... .... <<... 28 2.2.4 Impurities in Secondary Steam Generator Watef ............ 30

Е о ЕЕ ea a a ne 32 DcraelMeasurement bwePetential Drop Methode .......2- 9*5 34 Is Ey o ЮЕ Росава Drop а. ое ре у. 34 D ору ACUPorterntaedl Drop. x EE. rixa T RET TS 36 DNCEHRKocentehResedg5Gi ve шшш инт ез. ш RH ео ео оа еее ES 43 DENMUDOGEISESRMISISUEATCAPDPARATUSS C. eee ee spei us ees i s 45 эм есап са бус сеп ресс роп ш 4 m mem e sue than nn 45 В Еос Гамевава Heating System ааа 47 о теи Ка оп Боор ао ена a 48 ЕО ак пе Е Е еа ое аре а 49 аа зашрге Pressumiodtion System... в 50 СОПОЛ еШ e e ee A vie ЭЯ оа ее оу згеше A e 52 м Е... ее 54

a 54

Е ВЕЕР ВОО. 57

Seo Alternating Current Potential Drop (ACPD) System ............ 58 E FERIMENTAL PROCEDURE een een a 61 DN cDDESSLOOI otSbréinedbple Testing o. uer] Ee e m m 61 астас па сТатасп апа Скомсп Меавпкелепеа .2.................. 68 Le esa cree crete resto Mu EE I. esie sae ees ste s s d s 70 EE С БИО ког Рштелрје тезе то fot. weds een 70 Em Crack initiation and бкомсп Меавикетпепез .................... 78 Som BBG (sO) Nimes ice USE Uu uS reete eum e Vor M ere re s 88 ООО Е Erunciple а 88 саске Бабе ee ee 0% 91 DEXCONCLUSIONS ...:-9 92s Peer ate tet S S T S i re 95 ее... 96 КЕЕ ЕМС ЕЗИ О ТАТО ао о о оаа а xS 97 БТ SYSTEM OPERATING CHARACTERISTICS re... en seen 101 A-1-1 Omega CN9000A Heater Controller Parameter Settings ....... 101 A TS RET QUE Ius eir rur eee ws IOS Ae Aoc lavemHceadiSea MRIN Farlutes ... none... osos 104 ОН НЕЕ ЕКО a A e шз Rr e ens 106 а Eten start-zup Procedure = Static Operation ...... 0.000. 107 Pad 20 Stem etsbt upMDPOcedure Recirculation. een e e eea 110 A=2-3 System Shutdown Procedure - Static Орегатісп ............. 113 522 Sowsbem Shutdown Procedure Recibculatqon .3.59 4.9 «ds 114 2122-59 Emeneeneyschürdewnzbrocedurese An eres «Sale = eR rss s IIS А-2-6 Башр е Реевешк зар ош System Operation ана 116 2-7 Па ено оп эуеш реба оа ао он То 2-9 lec rro les ste kei latina Technique a није ........ 119 AUS DE RARUE De Ss UMM DRAW LNG GS 1. 9999999533. 2 a a ei 121

A-4. AXIAL FATIGUE PRE-CRACK SAMPLE PREPARATION PROCEDURE . 142

A Dr ВЕРОНА e Оаа аана 146

List of Figures

ДЕД

2-2

2-4 25

Beh 325 326

379 4-1 JU

TYPICAL AREAS OF STEAM GENERATOR TUBE FAILURE BY STRESS CORIO Si ON ARA stele ten es uL Ure doe a bes о ES

EVANS DIAGRAM FOR ALLOY 600 SHOWING THE EFFECT OF CHROMIUM CONCENTRATION ON CORROSION CURRENT DENSITY AND

PASS TVA E nee RN s п о. 20 SCHEMATIC OF A STEAM GENERATOR TUBE SHOWING TYPICAL

ЛЕРД О ОЕШ ЕТ СНЕ ЕСЕ ДШ ОЕК СС o a P e v ew eoo go s 25 ШЕР Wem CNN Cal Ole ee een 38

CURRENT DENSITY AS A FUNCTION OF DISTANCE FROM THE EXTERIOR WALL OF A TUBULAR CONDUCTOR FOR VARIOUS VALUES

Оо sien cue ont gai eee ek este clase VETERE ee ee ree 39 Ос СОТ октор BY PICK-UP WIRES ON THE ТОВЕ a a 41 AUTOCLAVE SYSTEM SHOWING RECIRCULATION, STATIC AND

DISCOUNT ION VG DEMS es ee ee ee оО AUTOCLAVE WITH HEATING MANTLE SHOWING RECIRCULATION AND

STANI CRCONNECTIONS E T T mms 47 Уа РОВ СО СТВ ООН А ее S 53 EDX OF ALLOY 600 HEAT 96834 (Low TEMPERATURE ANNEAL) ........ 56 THREE SDIMENSTONAL SIM MICROGRAPH OF HEAT 96334 „un scene 56 HEAT 96834 SEM MICROGRAPH SHOWING INTRAGRANULAR

БЕК ITA TES MM re ere a a ao aa V e DU SIMPLIFIED SCHEMATIC OF TEST SAMPLES ILLUSTRATING

MACHINING DIBPERENCES [Ni THE ОБЕ SECTIONS ... . 3 2.44 .60.06. O ТИЕ ЕРТ ОТАМЕЗ 225,05. о, e cae Caster a. In 59 А РЮ® © МС ТЕМ ОСНЕМАЛТ СИУ (a ен ео о. 60 ПАРАНА БӨЛ ЗАЛИВЕ О Ое кта ва ае 61 SCHEMATIC OF SAMPLE TUBE INSTALLED INTO AUTOCLAVE HEAD ......... 63

SAMPLE ASSEMBLED INTO AUTOCLAVE HEAD SHOWING ACPD INPUT AND OUTPUT WIRES, THERMOCOUPLE, LEVEL SENSING AND

ЕЕЕ REFERENCE E LECTRODE: „ы... a. 64 POLARIZATION CURVE FOR ALLOY 600 AT 300°C In 10% NAOH........ 67 POTENTIAL DROP FOR THE TEST AND REFERENCE ÁREAS ............... ү

SAMPLE TEMPERATURE AND WALL STRESS (PERCENT ОЕ 0.2% ТЕО ТЕО. Lux CUM I iS d gau cun nes 21

TEST AREA POTENTIAL DROP NORMALIZED WITH RESPECT TO THE REFERENCE PARES FOR. EIECURE 5-2,2 ГТ ое. TZ

Srl 5212 Sika]. 3

223 5-24 5-23

SAMPLE TEMPERATURE AND WALL STRESS (PERCENT OF 0.2%

MIES TRES II A a e УЗ POTENTIAL DROP FOR TEST AND REFERENCE ÁREAS .................. 74 NORMALIZED POTENTIAL DROP FOR БІСПЕН 5-5.................... Z4

STEREOSCOPIC COMPOSITE OF SAMPLE SURFACE SHOWING CRACK EMANATING FROM BENEATH THE REFERENCE АВЕА РЕОВЕ ............... 793

SEM COMPOSITE MICROGRAPH OF THE REFERENCE AREA FRACTURE

SURFACE O O о си ево ван 75 SEM COMPOSITE MICROGRAPH OF THE OPPOSITE SIDE OF THE

KEBERENCE AREA, MPRACTURE. GUREACE . 0... ME: оао ое ооо ан 76 SEM MICROGRAPH OF REGION 1 CLEARLY SHOWING IGSCC IN

IN cis aia arta A Re ERR, EE e 76 SEM MICROGRAPH OF AREA 2 SHOWING DUCTILE FAILURE REGION ....... du DUCTTER TE SCCETRANSTTITONSBRECLON DSG 9 0 О cn Z7

SEM CoMPOSITE MICROGRAPH OF IGSCC FRACTURE SURFACE SHOWDINEIBERTETICALSCRACH ERONT SE a a 000% 79

NORMALIZED ACPD FOR SAMPLE SHOWING CRACK INITIATION

PENIS BASED ONFENEREASING Ока POTENTIAL. DROP... 3.60 + ses ose ans 79 SCHEMATIC OF TUBE SAMPLES WITH FATIGUE PRE-CRACKS IN

ESTATE. ne ren are ee 80 SAMPLE TEMPERATURE AND APPLIED WALL STRESS (PERCENT OF

ООО ЕНЕС ee ee TS 81 POTENTIAL DROP VERSUS TIME FOR TEST AND REFERENCE AÁREAS......... 82

STEREO MICROSCOPE IMAGE OF THE THROUGH-WALL CRACK SEEN ARABE DAS ESOR DAESDIACHAINED NOTCA соо о ооо ооо 83

Low POWER STEREO MICROSCOPE IMAGE SHOWING THE RELATIVE POSITION OF THE THROUGH WALL. CRACK TO THE PROBES ....... +... u. 83

NORMALIZED POTENTIAL DROP (TEST AREA WITH RESPECT TO THE REFERENCE AREA) SHOWING AN INCREASING SIGNAL FROM THE

О Е A ME ee ee en 84 SEM COMPOSITE MICROGRAPH OF FRACTURE SURFACE CLEARLY

АРА ет 85 MAGNIFIED SEM MICROGRAPH OF IGSCC AREA OUTLINING Two

POSSIBLE СВАЛОК (GROWTH ARRAS еее 85 MICROGRAPH OF THE PRE-IATIGUBEMNEEA l.l. 2433 nen nen 86 MICROGRAPHUOF THE DUCTILE БАГОВ ВОО ое рае. 86

MicROGRAPH SHOWING THE PRE-FATIGUE TO IGSCC TRANSITION РР ее 86

А- 1-3 A-1-4 2-1 29-2

À-3-4 AS => A3-6 Ago / A-3-8 A239

A=3-10 yl Dog 12 A913 A-4-1 A-4-2

CALCULATED TE OTENTIAL ПЕОР МЕБСОПЕ АСТПА, .................... CALCULATED P EOTENDIAL DROP VERSUS ACTUAL e oa een unse

ALLOY 600 TUBE CRACK GROWTH VERSUS STRESS INTENSITY ЕО НАСОС О а еее

COMPARISON OF CRACK GROWTH DATA VERSUS STRESS INTENSITY

(CR) RR ARI IU EA. v he Du» USDODECRACKIOROWIH RATES VERSUS PH oe aa nee se ee

HEATUP 950726 SHOWING OSCILLATIONS OF THE HEATER A A TT DEM

HEATUP PLOT SHOWING CONTRAST IN TIME BETWEEN BERIRCHEATTONFAÄND STATIC MODE E 2. 1 Ао.

ВОО ПАСЕ TO "O-RING FROM TESD 950903... Was о с се. а ОСОО СУ ЕИ р бо

SAMPLE PRESSURIZATION CONTROL BOARD TUBING AND TEMPLATE DV WE Xt MEE naue NN aA e aes is

RECIRCULATION AND STATIC SYSTEM CONTROL BOARD TUBING AND РЕА РЕ ЕТА она

ENIECIMON@ Oe oTEM (COMPONENT = lc ое. порно ооо

EMERGENCY DEPRESSURIZATION AND SAFETY COLLECTION SYSTEM COM WEN E ааа е

ДЇ @®УЛ\УЕШН ЕКА» ЕТА aos erras OR Meme eee arte AUTOCLAVE HEAD AND SAMPLE ASSEMBLY DETAIL... ооо стон BESES AMPE E ASSRMBDEY CDETATES. «ctr eer eterne ntur re ve e IDETATILEDOGONTROE WIRING SCHEMATICT. or efc s ea I ERE ETUR ERIS e FORK woes NOTCH DEPTH VERSUS SURFACE CHORD LENGTH еее вена o лесе к

SCHEMATIC OF FIXTURE USED WITH THE MTS TESTING SYSTEM TO INDUCE EATICUE CRACKS IN THE ТЕСТ SAMPLES ЕИ Ос.

SIDE PHOTO OF SAMPLE AND FIXTURE MOUNTED IN THE MTS м о.

FRONT VIEW OF SAMPLE AND FIXTURE INSTALLED IN THE MTS BRUM REMISE MEL cuu wow S CERT ET SD E ER UE RUN QS TA

List of Tables

Sl 922 DEI 52 AIl

ЕА 22-2 Ae 3 А-3-1 41-2 А22-3 А-а Bao = 5 A= 3-6

Д-3-7 B

MECHANICAL PROPERTIES OF HEAT 96834

CHEMICAL COMPOSITION OF ALLOY 600

оо ө ө ө ө ө ө ө ө © © © © © © ә е ө ө ө ө ө ө ө

TABULATED RESULTS FOR STRESS INTENSITY AND GROWTH RATES........ TABULATED RESULTS FOR STRESS INTENSITY AND CROWTH RATES........

CN9000A PID CONTROLLER PARAMETERS FOR STATIC AUTOCLAVE а

STATIC OPERATION STARTUP VALVE LINEUP

RECIRCULATION OPERATION STARTUP VALVE LINEUP

INJECTION SYSTEM VALVE LINEUP

ALLOY 600 TEST PLATFORM COMPONENTS SUMMARY

SAMPLE PRESSURIZATION SYSTEM COMPONENT ID..

STATIC SYSTEM COMPONENT IDENTIFICATION

RECIRCULATION SYSTEM COMPONENT ID

INJECTION SYSTEM COMPONENT ID

DEPRESSURIZATION AND SAFETY COLLECTION SYSTEM COMPONENT ID

ооо о о о о о о о о о ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө ө o. ө е ө ө ө ө

CONTROL SYSTEM WIRING DETAIL

TEST DATA WITH ERROR CALCULATIONS

10

i1. Introduction

Since 1967, most pressurized water reactor steam generators have been constructed with tubes made of alloy EUN Alloy 600 is a strong, corrosion resistant nickel based alloy. However, environmentally assisted cracking (EAC) has occurred within tube sheets, just above tube sheets and in tight radius U-bends in this material.

Today, steam generator tube degradation is the life- limiting phenomenon in pressurized water nuclear reactors with U-tube steam generators. Comprising over half the primary to secondary pressure boundary surface area, the Steam generator tubes are of critical importance to reactor safety. Extensive investigations by numerous researchers have failed to produce a cohesive failure theory for the cracking phenomenon, known as Stress Corrosion Cracking’. Stress Corrosion Cracking has been observed on both the primary and secondary side of the steam generator tubes, елап [ат Stress Corrosion Cracking (IGSCC) on the secondary side periodically occurs at sites of Intergranular Attack (IGA), though either IGA or IGSCC can occur independently.

Because IGSCC is of such concern for reactor safety, frequent shutdowns for steam generator tube inspections must occur to ensure tube integrity. Each maintenance period not only costs millions of dollars in lost power generation, but many man-rem of maintenance personnel exposure. Development of an on-line steam generator tube monitoring system would

decrease the periodicity of shutdowns for tube inspections

! J.T. Adrian Roberts,Structural Materials in Nuclear Power Systems,

(New York: Plenum Press, 1981), р. 340. Herbert H. Uhlig and R. Winston Revie, Corrosion and Corrosion Control, 3rd Ed, (New York: John Wiley & Sons, 1985), p. 366.

i

and provide a more reliable warning of impending tube failure.

The scope of this thesis is to provide a proof of peanciple for detection of stress corrosion cracks in actual steam generator tube material under high temperature, pressure and corrosive environmental conditions utilizing previously developed alternating current potential drop (ACPD) crack detection techniques. If successful, the techniques used to detect the stress corrosion cracks can be expanded and refined into an on-line system in the future.

In order to demonstrate a proof of principle for crack detection by ACPD techniques in a high temperature, corrosive environment, an autoclave system with both static and recirculation capability was constructed. Under in-situ conditions, steam generator tubes require years for stress corrosion cracking to occur. To expedite the process, tube specimens were exposed to highly caustic conditions (10% sodium hydroxide and 0.1% Na,CO, added to subcooled deionized water) at 315°C. Boiling was prevented by maintaining pressure greater than saturation. Prevention of boiling is critical to preventing local concentration of Caustic .

The specimens were internally pressurized to obtain a hoop stress up to 145% of yield stress. Based on previous Studies of stress corrosion on alloy 600. this allowed crack

WR : : 3 : А initiation in weeks . Further reductions in crack

? M. Payne and P. McIntyre, Influence of Grain Boundary Microstructure

on Susceptibility of Alloy 600 to Intergranular Attack and Stress corrosion Cracking., Corrosion=NACE, XLIV.No. У. (Мау 1988), pp. 314- 319.

12

initiation time were realized by polarizing the sample to +150 mv with respect to a nickel electrode’.

Crack detection was achieved using ACPD techniques previously developed’. Pick-up probes were attached on either side of an expected crack initiation site. An additional set of pickup probes were attached away from the test area to serve as a reference signal. The site of crack initiation was controlled by plating the tube specimen with nickel in every area but the crack initiation site. Probe spacing, frequency and current were determined empirically and through prior analytical results to yield optimum AC potential drop sensitivity.

Static system tests were performed to (i) test system imeprity to the caustic solution, (ii) determine the ability of the selected temperatures, pressures and solution pH to initiate specimen cracking in a reasonable time frame, (iii) verify the ability of the system to initiate stress corrosion cracking in the specimen and (iv) verify the ability of the ACPD system to identify crack initiation and growth under these conditions. Subsequent testing obtained stress corrosion cracking data for the tube specimens for

comparison to data available in the literature.

* J. B. Lumsden,, S.L. Jeanjaquet, J.P.N. Paine and A. McIlree,

"Mechanism and Effectiveness of Inhibitors for SCC in a Caustic Environment", (Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol Il EsrSckenrjdoe, СО: NACE International. August 7-10,1995), pp. 317-325. ?^ R.G. Ballinger and I.S. Hwang, "Characterization of Microstructure and IGSCC of Alloy 600 Steam Generator Tubing", (Final Report, EPRI TR- 101983, Palo Alto: Electric Power Research Institute, (February 1993),p. 22865

3

2. Background

2.1 Nuclear Steam Generator Materials and Environment

The first nuclear steam generator tubes were constructed using austenetic stainless steel'. Corrosion related degradation occurred at a relatively rapid rate. The next generation of generators were built with alloy 600 tubing and an evolving water chemistry control program . Though the general corrosion problems were mitigated to a great extent, stress corrosion and other localized corrosions continued to

(охехел SS

2.1.1 Steam Generator Construction

The tubes of a typical steam generator used in nuclear applications are of either a recirculating, or U-tube, design (Westinghouse, Combustion Engineering) or a once-through design (Once Through Steam Generator or OTSG manufactured by Babcock and Wilcox)”. A typical steam generator has approximately ШОС straight-through tubes (Babcock and Wilcox design) or approximately 8000 to 9000 U-tubes in recirculating designs to provide adequate surface area for heat transfer from the primary to secondary fluid. The tubes can extend in length to A meters (68 feet) in a straight through design. Approximately 15 support plates in both designs provide vertical and lateral support and vibration reduction over the

tube length. The tubes are normally not welded onto the

1 J.R. Cels, Caustic Stress Corrosion Cracking Studies at 288C(550F) Using

the Straining Electrode Technique, Corrrosion-NACE, V34, No.6 (June 1978), 198.

James A. Adams and Eric S. Peterson, Steam Generator Secondary pH During A Steam Generator Tube Rupture, Nuclear Technology, Vol 102 (June 1993), p. 205.

У J.T.A. Roberts, Structural Materials in Nuclear Power Systems, (New York: Plenum Press, 1981), p. 323.

14

support plates and have clearances up to 0.25 mm (.010 in) between the support plate and tube to allow for thermal ОЕ Оо 7238 mm (1,5 in) thick tube sheet separates the primary from the secondary (one upper and one lower in a once- through design). The steam generator tubes are normally on the order of 12 mm to 18 mm in diameter and are heat fitted and tungsten inert gas welded into the tube sheet on the primary side. On the OTSGs manufactured by Babcock and Wilcox, the entire steam generator is stress-relieved in a furnace after final fabrication and welding of the hemispherical heads are completed”.

As can be seen in Figure 2-1, crevices between the secondary side of the tube sheet and tubes and on each

side of the support plate

tubes exist due to the en

-———— ALLUY 600 TUBE

Benserutetion deseribed. In ^

-CREVICE REGIDN

addition to the crevice, these b m areas have high residual or

working stresses due to the

press fit or constraining Figure 2-1: Typical Areas of

nature of the design. These Steam Generator Tube Failure By

: | Stress Corrosion Attack crevices are typically the sites for intergranular attack (IGA) and stress corrosion cracking (IGSCC) and are the inherent flaw in steam generator design from a corrosion perspective. Additionally, residual stresses during U-tube bending can increase the localized

stress levels many times higher than the hoop and radial

stresses induced from the primary to secondary side

P Babcock & Wilcox, Steam: its generation and use (New York: Babcock & Wilcox, 1975) р. 23-11.

15

differential pressure. These residual stress areas will be

deseribed in more detail later.

2.1.2 Steam Generator Chemistry

The secondary side of nuclear steam generators operates in an alkaline environment with a pH typically in the 8.5 to 9.2 range for recirculating steam generators and 8.8 to 9.6 for Babcock and Wilcox once-through steam generators. This pH range maintains a low general corrosion rate for all components in the steam generator since many components are either low alloy or carbon steel, including the steam generator shell.

There are four principal additives used to maintain secondary chemistry’. The principal pH control additive is ammonia (NH,OH), selected for its compatibility with all materials used in steam generator construction. Ammonia is very volatile and only minimal residuals remain on the heat transfer surfaces and in crevices during power operation. This allows contaminants (such as chlorine or sodium) to concentrate in tube crevices and create a corrosive environment. Other additives, such as morpholine, boric acid and hydrazine are added to provide buffering in the crevices. Morpholine, being less volatile than ammonia, has provided protection of carbon steel components (steam and condensate piping outside the steam generator especially) from erosion/corrosion in areas of two phase flow. Boric acid has been added to reduce the occurrence of “denting” caused by the growth of magnetite in the tube support plate crevices in generators with carbon steel support

plates and has also mitigated intergranular attack of alloy 600

Е James P. Adams and Eric S. Peterson, "Steam Generator Secondary pH During A Steam Generator Tube Rupture," Nuclear Technology, CII, (June 1993), pp. 304-306.

ĉ Ibid, p. 305.

16

by high caustic/. Finally, hydrazine has been added as an oxygen scavenger. Although alloy 600 is not susceptible to BN ide stress Corrosion cracking, it is still susceptible to oxygen-associated localized corrosion.

Chemistry in a steam generator crevice is not known. The case for a high caustic condition stems from much research and experimentation in this regime. R. Bandy, et.al., conducted tests on alloy 600 in high temperature caustic environments and concluded that the concentration of alkaline species was one of the more important factors governing intergranular corrosion . It should be noted, however, that although the general environment of the tests was alkaline, no measurements of the chemistry in the crevice area were performed and Bandy concluded that “a more thorough knowledge of the crevice chemistry is required”. S.M. Payne, et al., reported that impurities from the secondary water may concentrate in superheated hot leg crevices to 50% by weight of sodium hydroxide. Another possible source for free caustic Formation ie from ionic impurities introduced from corrosion products in the feed, condensate and chemical addition system and leakage in the condenser. These impurities include sulfur,

phosphates, and chlorides, among others’. These impurities

“Boric Acid Application Guidelines for Intergranular Corrosion Inhibition,” EPRI NP 5558, Electric Power Research Institute (1984).

James P. Adams and Eric S. Peterson, “Steam Generator Secondary pH During Pecteam Generator Tube Rupture,” Nuclear Technology, Gil, (June 1993), р. 305.

í R. Bandy, R. Roberge, and D. van Rooyen, “Intergranular Failures of alloy 600 in High Temperature Caustic Environments," Corrosion - NACE", XVI no. IEEE MArchol985).pp. 142, 149-150.

R. Bandy, R. Roberge, and D. van Rooyen, pp. 142-151.

S. M. Payne and P. McIntyre, "Influence of Grain Boundary Microstructure on the Susceptibility of alloy 600 to Intergranular Attack and Stress Corrosion Cracking,” Corrosion-NACE, XLIV no. V, (Мау 1988), рр. 314-319.

J.T.A. Roberts, Structural Materials in Nuclear Power Systems, (New York: Pienum Press, 1981).pp. 336-337:

11

17

can concentrate in the low flow regions, such as the tube sheet crevice, or in sludge at the top of the tube sheet, and cause Keeallly high alkaline conditions in high heat flux, low flow

regions (typically where intergranular corrosion occurs).

2.1.3 Material Properties of Alloy 600 Steam Generator Tubes

¡Aloy 600 is a nickel-chromium-iron alloy containing at Memote72, nickel (plus cobalt), 14 to 17% chromium, 6-10% iron, OS carbon (maximum, nominal carbon content is in the 0.08% range ) and small percentages of manganese, sulfur, silicon and copper. These concentrations are nominal and actual concentrations vary within this range. The importance of this variation and the small percentages of other materials added in the course of alloy 600 manufacture will play a role in the ability of a particular heat to resist intergranular attack and stress corrosion cracking. Typically in older vintage steam generators, following milling, the tubes are heated to a temperature to allow rearrangement of the individual atoms into a more stable structure. The tubes are then allowed to cool in the furnace. This process is known as mill annealing and softens the tubes to impart added ductility!*. Steam generator tubes conform to American Society for the Testing of Materials (ASTM) standard B-163 and American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III (Nuclear Vessels) with a minimum tensile strength of 552 MPa (80000 psi) and minimum 0.2* yield strength of 241 MPa (35000 psi) in a mill-annealed condition. Alloy 600 is ductile at

room temperature with Charpy V-Notch impact strengths of

i Wayne Z. Friend, Corrosion of Nickel and Nickel Based Alloys (New York: John Wiley & Sons, 1980), p. 143.

L. H. Van Vlack, Elements of Material Science and Engineering, 3rd Ed., (Reading, MA: Addison-Wesley Publishing Co., 1975), pp. 374-375.

18

approximately 250 Joules at typical steam generator secondary EndcEoberating temperatures of 224'C to 273'G (inlet and

outlet, respectively)" for mill-annealed material. It does not embrittle after long exposure at high temperatures.

Fatigue strength is affected by grain size and condition and by temperature, with typical values of fatigue strength greater

than 325 MPa for 10° cycles”

Nickel provides alloy 600 with a resistance to corrosion bvariety of organic and inorganic compounds. It is especially resistant to caustics and caustic solutions, the corrosion resistance being proportional to the nickel content

in sodium hydroxide.

Chromium is added to provide resistance to corrosion in high temperature oxidizing environments, and resistance to sulfur compounds (sulfur compounds act as catalyst poisons that ease access of hydrogen into a metal lattice resulting in cracking due to hydrogen embrittlement”). The effect of chromium additions to nickel is illustrated in Figure 2-2. The active region of Figure 2-2 is defined as the potential range at whieh significant corrosion occurs. The point at which the current suddenly drops to a value orders of magnitude lower is called the critical current density and the lower current density is the passive current density. The vertical region of Figure 2-2 where this lower current density exists is known as

the passive region and the potential in this range is the

> Neil E. Todreas and Mujid S. Kazimi, Nuclear Systems I (Bristol, PA: taylor Francis, 1990), p. 5. : Inconel, Inco Alloys International (Huntington,WV: Inco Alloys International, Inc).

Herbert H. Uhlig and R. Winston Revie, Corrosion and Corrosion Control (New York: John Wiley & Sons, 1985), p. 142.

је

passive potential’. Passivity for a metal is simply the state where the rate of dissolution in a given environment under steady state conditions becomes less as the electrode potential is increased than the rate at some lower potential”. At chromium concentrations greater than 9%, chromium in nickel significantly reduces the current density in the active erosion region. For chromium concentrations greater than 10%, the critical current density is reduced, the passive

potential widens and the passive current density is reduced”.

ALLOY 600 300*C, 20mV/min

Reduced Cr decreases the passive potential

10% Na OH

= 0.8

а

»

d!

a 96 Reduced Cr increases

= the passive current density z

ul

5 0.4

a.

o ғә

о Reduced Cr increases the critical current density

“1072 io^! 10° 10! 10? CURRENT DENSITY (mA/cm?)

Figure 2-2: Evans Diagram for Alloy 600 Showing the Effect of Chromium Concentration on Corrosion Current Density and Passivity.

1? Herbert H. Uhlig and R. Winston Revie, Corrosion and Corrosion Control,

ОТО Ed (New York: John Wiley & Sons, 1985), pp. 61-65. C. Wagner, Discussions at the First International Symposium on Passivity,

Heiligenberg, West Germany, 1957, Corrosion Science, Vol 5, (1965). 9.751. e Wayne Z. Friend, Corrosion of Nickel and Nickel Based Alloys (New York:

Schu Wiley 8 Sons, 1980), p. 136.