AECL-5013

ATOMIC ENERGYOF CANADA LIMITED

L'ENERGIE ATOMIQUEDU CANADA LIMITEE

STEAM GENERATOR TUBE FAILURES:

WORLD EXPERIENCE IN WATER-COOLED

NUCLEAR POWER REACTORS IN 1973

by

P.D. STEVENS-GUILLE and M.G. HARE

Chalk River Nuclear Laboratories

Chalk River, Ontario

January 1975

STEAM GENERATOR TUBE FAILURES:

WORLD EXPERIENCE IN WATER-COOLED NUCLEAR POWER REACTORS IN 1973

by

P.D. Stevens-Guille and M.G. Hare

Chalk River Nuclear LaboratoriesAtomic Energy oi' Canada Limited

Chalk River, Ontario

January 1975

\ECL-50H

Defaillance des tubes de ehaudiere:

experience mondiale acquise en 1973 dans les reacteurs

de puissance refroidis par eau

par

P.D. Stevens-Guille et M.G. Hare

Resume

Onze reactears refroidis par eau sur 4l) ont en des defaillances de tubes de chaudiere en 1973. Le presentrapport passe en revue les causes de ces defaillances et les mesures a prendre pour qu'elles ne se produisentpas Les principals causes sont la corrosion et les vibrations. Quant aux mesures preventives elles les-sortissent surtout a la chimie de Peau secondaire et au concept des tubes.

L'Encrgie Atomique du Canada, LimiteeLaboratoires Nucleaires de Chalk River

Chalk River, Ontario

Janvier 1075

AECL-5013

STEAM GENERATOR TUBE FAILURES:

WORLD EXPERIENCE IN WATER-COOLED NUCLEAR POWER REACTORS IN 1973

by

P.D. Stevens-Guille and M.G. Hare

ABSTRACT

lileven of 49 water-cooled reactors experienced failures of steam generator tubes in l'>7.i. This icpi.nsummarizes these t'ai'ures. Corrosion and vibration were ihe main causes. Secondary-water chemistry anildesign are the two most important aspects in preventing tube failures.

Chalk River Nuclear LaboratoriesAtomic Energy of Canada Limited

Chalk River, Ontario

January 1'I7S

AWL-SO I

THE CONCERN FOR STEAM GENERATOR IN-TEGRITY

Steam generators arc key component separatingthe primary and secondary coolant systems inpros, uri/ed water reactors (I'VVR). They provideoperational and safety advantages, but like manyinli.'iface components, their failure affects bothsystems and incapacitates the whole reactor.

The year |V73 was important lor manulaciurcrsand operators of nuclear steam generators. Tubefailures occurred in epidemic proportions in tworeactiMs and to a lesser degree in nine others. Theprevious year, steam generator and condenser tubeleaks were 'he largest shglc cause of reaeior down-time iii the United S ta t e s ' ' ' with outages becomingmore a v l l y with the "energy crisis". The impli-cations ol steam generator reliability were soonreali/.etl. Concern to,' theii integrity was expressed byleading nuclear journals, reactor vendors andutilities*--U

I he single most important lad to emerge in ! ' i /3w.is the role played by water chemistry in thesecumljr\ system and its inllueike on tube coirosion.Major manulacturers adopted opposite views whichweie not resolved until late in 1^74 when the use olall volatile (zero solids) water treatment was speLified!v, most vendors.

By the end of ll)7.1. 4') water-cooled reactors withsteam generators were in operation'-", eight morethan in |l.i72. l-.leven experienced lube failures.

This report is a summary of the 1 •)73 steamgenerator experience and follows two previousreports on tlie l n 7 l and 1<)72 evpchence'V (>).1 ibleI summarizes the response to a questionnaire sent toreactoi operators and a search of I lie publishedliterature on steam generators. No data weie ;!ahlefrom the USS't or the German Democratic Republic.

'I he distinction iictween tubes with leaks anddelects is not male in this report. Sir.ce it is in theinterests ol operators to maintain steam raisingcapacity and minimize repairs, it is reasonable toassume that there aie always good reasons to plugtubes that have defects.

TABLE I - Reactors with Steam Generator Tube Failures in 1973

ReactorTypePower(MWe) netCountry

NumberManufacturer

Sieam Generators

TypeAreaConditions

Number oftubes

Tube SizeTube Material

Number of Nature of Tube FailuresTubeFailures

Bez.nau-1PWR350Switzerland

Westinghouseverticald 145 m2

is.5 MI'a315"C

:t>04::.: mm ou1 2 mm thickInconel MO

p i t t i n g c o r r o s i o n m

wel l -de f ined pat lorn

in h«'l leg. pnn i c ross

flow s u s p e c t e d , slihlge

Imiiiil nil liihcslieel

Dresden-1PWR

:uoUSA

GariglianoBWRI 50

lti.lv

4FosterWheeler

K.M.F.Stork

vertical(.05 in2

7.0 MI'a:K5°C

vertical148(i m2

7.0 MI'a277"(

] K ( ) 1

15.1.) mm OIJStainless Stecl-304

I7S5!l» mm Ol)1.')? mm iliuk

Mnnel-40U

unknown

lew

TABLE 1 Continued

ReactorType

Power

(MWe) net

Country

NumberManufacturer

Tvpe

AreaConditions

Number of

tubes

Tube Si.'.e

Tube Material

Number of

Tube

Failures

Nature of Tube Failure

Haddam Necx 4

I'WK Westinghousc

575

I'SA

vertical2573 m7

13.8 MPa

379419 mm ODinconel-(>00

10 stress corrosion and

fretting, 8 in center

of hot leg above tube-

sheet, 2 in U-bend,

sludge found on

tubeshcet

ll.B. Robinson-:

PWR700USA

3w'esimghouse

vertical

4128 n v

15.4 MPa

317°C

32dO

22.2 mm OD

1.2 mm thick

lnconcl-600

two by intergranular

corrosioii 1 m above

tubeshcet in hot (eg,

one by erosion at

U-bend

Indian Point-1

PWR

2(>5USA

KWOI'WR

j';8(iermany

Miluma-II'WK3 20Japan

N-ReactorI W(IR

sooI'SA

Palisades-1I'WR700USA

4Babcock &WiUox

:

('lUteliolfnu.g-•ihutte

Slerkrade ACJtUalck AC,

:

( omhustionEngineering

CombustionEngineering

CnmbustiiinEngineering

horizontal10.3 MPa270°C

vertical2750 n r14.7 MPa3H)"C

vertical3381 n r1 5.4 MPa322"C

liori/ontalI4SO m2

10.3 MPa:?7°c

vertical

8 i l

25.4 mm ODStainless Steel-304

260722 mm OD1.2 mm thickInonel-iiOO

442h14 mm OD1.42 mm thickInconel-GOO

l l)l(.IS 9 mm 01)1.5 mm thickStainless Sleel-30-1

851922 2 mm OP1.2 mm ihickInciind-hOU

iar. Onolre-II'WR

430

rs\

CombustionEngineering

vertical2575 , „ -

14.5 MI'a304" C

19

1.4

Inc

37V4mm Oi)

mm thick

onel-i-iOO

153 stress corrosioncracking suspectedat supports at U4)cnd

I'' intergranular stresscorrosion crackingabove tubesheet incentre of bundle

2013 wall thinning by localizedthermal hydraulicconditions at bendregion under anti-vibration straps

I unknown, no failures inInconel-dOO steamgenerators

-1700 wall thinning bylocalized thermalhydraulic conditions in11 rows from dividerplate at U-bend at anti-vibration straps andlateral supports

restrictions at firsttube support in hotleg. fretting at U-hend

SURVEY OF 1973 TUBE FAILURES

Beznau-1, -2, Switzerland

Following one of the largest tube plugging opera-tions ever performed, operation of Beznau-1 has beensatisfactory. During a scheduled refuelling shutdownin March 1973, eddy-current testing revealed onlyeight interpretable defects which were then plugged.This follows some 950 defects located the previousyear. However some defects not detected by eddy-current testing were discovered when two tubes wereremoved for laboratory examination from Units 1and 2. Wall thinning and pitting were discoveredimmediately above the tubesheet. The deepest pitswere only 0.2 mm, less than 20% of the tubethickness, and probably below the range of detectionhy eddy-current testing. No evidence of intergranularattack was found, indicating that caustic attack wasunlikely. Although some doubt remained as to thecause, it was thought that the delects could haveoriginated from localized acid conditions, by eva-porative concei trations of corrodants or by a metal-phosphate reac'ion. The similarity between theBeznau defects md those in Palisades and Mihama-1cannot be discounted. It is likely that the Beznaudefects are caused by phosphate wastage, a pheno-menon that has affected at least eight other reactorsand was the largest single cause of corrosion failuresin 1973.

Wastage is believed to be a corrosion phenomenonconnected with phosphate Vv'ater treatment and notdue to erosion or cavitalion despite the similurappearance of the damage area. Recent explanationsof tube wastage have placed emphasis on the type andoperating limits of secondary-water treatment.

Low frequency eddy-current tests were made tomeasure the height of sludge above the tubesheet onthe hot leg (primary inlet side). The sludge profileranged from zero at the peripheral tubes to 120 mmat the centre. The profile correlated with the Utility'sidea on local How patterns above the tubesheet. In1973 tin; sludge vis mechanically cleaned tium butliUnits 1 and 2. Sludge was also found in the cold legof all steam generators.

Both Beznau-1 and -2 operated with sodiumphosphate secondary-water chemistry. An adjustmentof the molar ratio was the only change made in 1973.k full report on the history of the Beznau steamgenerators appears in reference* 7). [j,,<,h units haveoperated satisfactorily during the first half ol

Haddam Neck (Connecticut Yankee), USA

Two eddy-current examinations were made atHaddam Neck in 1973. Both used a remotely con-trolled probe positioning machine and motorizedprobe feeder to reduce radiation dose to operators.All sleain generators were partially inspected in areaswhere failure was thought likely to occur based onprevious inspections at Haddam Neck and SanOnofre-I. Indications were found in the region 10 to250 mm above the centre section of the tjbesheet onthe inlet side in three of the four steam generators.Corrosion from the outer surface was thought to be(he cause of the indications. No indications werefound on the outlet side in the region above thetubesheet. Further testing in the Li-bend regionrevealed defeel indications adjacent to iiiiti-vihruliunbars llu.'.ight to be caused by fretling between thetubes and bars. Ten tubes were explosively plugged inone steam generator.

Eddy-current tests were also made to measure theheight of deposits above the tubesheel. They rangedfrom zero at the peripheral tubes to 225 mm at thecentral region of the inlet side. This is similar to thesludge patterns at Beznau-1; in high flow regionsthere is little o. none, while in the central, low flowregion the deposit height is a maximum.

During !Q73 Haddi'm Neck used phosphatesecondary-water treatment, with PO4 between 10 and80 ppm, pi I between X.5 and 10.6 with a Na:I'O.,ratio of 2.0 to 2.d.

H.B. Robinson-2. USA

After 32 tube failures in the II.B. RnbiiiMin-2reactor in l l '72. improved steam generator chemistivcontrols were instituted that included continuousphosphate addition and continuous blovvdown. Ai>eddy-current inspection in April and May of nearly1009? of tubes in the inlet side showed only twodefects in the region above the tubesheel. InNovember a leak was detected in the "C" steamgenerator. An eddy-current inspection survey ofabout 200 tubes was made with a remote probe-positioning machine. The leak was found io be from50 io 150 mm above ihc upper lube suppoii in theU-bcnd region. The remaining tubes it-sled revealedno deteiioration since the April-May inspection. Tlit-leaking luhe was explosively plugged and thesecondary side hydriista'.iailly pressure tested at 5.5MPa(KOOpsig).

Further tube leaks developed in ll»''4. and afurther 35 tubes were explosive,\ plumed. I-ldJy•current tests and laboratory examination of ;i ic-moved tube showed gcneia! wall iinmniijj ovei a sh'irt

section of the tubes as opposed to inteigranular stresscorrosion cracking. This was a new form of corrosiveannck^ .

Caicfui examination was made ol the steamcenerator chemistry history. A Na:PO., ratio o( 2.2 to2.4 was maintained during the first part of 1973 theratio then dropped to less than 2.2. On the advice ofthe Westmghou.se Electric Corporation, the ratio wasthen increased to 2.3 to 2.6 after the discovery of achemical invariant point at a ratio of 2.18. Below thispoint a corrosive phosphate solution could tonn inlow How regions of the steam generators. Failures inll>74 in the tubesheet and U-bcnd regions wereattributed to this form of corrosion.

Indian Point-1, USA

Steam generator problems plagued the operationoi' Indian Point-I in l*-)7.>. 153 lubes were pluggedbringing the total to 249. In one steam generator, 104tubes were plugged in the lower seven rows of thehorizontal tube bundle. Failure wa.; attributed to themovement of a tube support plate which probablyresulted in increased stress and accelerated failurefrom stress corrosion cracking in the U-bend region.Other failures were distributed apparently at randomon the tubesheet. All tubes were eddy-current tested.Most repairs were made with explosive plugs, but theresults were not satisfactory: some were not leak-tightand were repaired by manual welding. The reactoroperated with volatile chemistry control in thesecondary system.

A second problem with the Indian Point-1 steamgenerators arose in April 1973. Defects were foundby radiography in 6 of 21, 6-inch downcomer pipes.Field repairs were performed in radiation fields of 6to 15 R/'h in difficult locations. Closed circuit tele-vision, remote welding and cutting machines andfull-scale mockups were all used to minimize radia-tion dose to the repair crews. Repairs took four tofive months, 1500 men were involved and the totalradiation exposure was 3500 man-rem of which 509;was received by boilermakers and welders" 0).

Inadequate access, lack of reliable remotelyoperated macliines and insufficient decontaminationtechniques severely hampered repairs. It is note-worthy that at one stage virtually all the availablewelders in the New York City area were used (topersonnel exposure limits) on this repair. Utilitiesoperating in more sparsely populated areas than NewYork City could suffer much greater hardships anddelays with a repair of this magnitude.

KWO, Federal Republic of Germany

K.W0 had one shutdown to repair tube leaks In11>73< I' I A total of 19 tubes were plugged; a notabledecrease from the preceding year. In September1973. eddy-current tests were made and comparedwith results obtained in 1972. Only a small increase-in defect propagation in 43 lubes was found. Tubeswith, defects of N0% or more of the tube wallthickness were plugged. KWO continues to usevolatile sccondary-watei treatment. Leaks in the maincondenser which could lead to the ingress of im-purities are carefully monitored. Condenser tubeleaks can be plugged without reducing reactor power.

Mihama-1, Japan

The first evidence of tube leakage at the Mihama-IPlant was in June 1972. During a five-month shut-down for repairs in thai year. 110 tubes were pluggedand ten were removed alter a lOO'/r eddy-currentexamination of both steam generators. In March !973a re-inspection of tubes in the bend region revealedtube wall thinning. During a four-month outage afurther 2013 tubes were plugged.

At the time, the Mihama-1 failures were differentfrom those in other reactors. Tubes removed from thesteam generators and other tubes inspected by in-situradiographyf'-) showed local wall thinning at theintersection with anti-vibration straps. This wastagewas thought to be due to local thermal-hydraulicconditions. Alternate wetting and drying of tubesurfaces exposed to concentrated free caustic (fromcondenser in-leakage) was thought to result in rapidlocal corrosion.

Chemistry control of Mihnma-1 initially involvedtrisodium phosphate additions with a Na:PO4 ratio of3.0. This was followed by co-ordinated phosphatetreatment (Na:PO4 ratio 2.4) for a short time prior tothe tube failures in 1972. In 1973 phosphoric acidwas added to achieve a Na:PO3 ratio of 2.1. Nofurther changes were made in 1973.

Palisades, USA

Paiisades experienced two outages for tube repairin 1973. The first occurred in January and lasted1-1/2 months. Extensive eddy-current testing andin-situ radiography revealed tube wastage of the samenature and at similar locations to that found inMihama-1. About 1400 tubes were plugged duringthis outage. Inconel plugs were manually inserted andwelded with a single-pass no-filler process. All tubesin the first 1 1 rows of each steam generates wereplugged in this way.

Free caustic in Ihe secondary water from con-denser tube failure was (hough! to be the cause of thetube wastage at thai time.

In August, primary lo secondary leak ratesexceeded the 68 K/h licensing limit and a secondoutage started ivhich continued for over a year.Extensive eddy-current testing revealed tube wastageat lateral supports on the inlet side and up IO 90 mmof sludge was measured on the tubesheet by eddy-current methods.

Combustion Engineering concluded that the tubewastage phenomenon would be slopped by changingfrom a co-ordinated phosphate to a volatilescenndary-water treatment. It also reported! 13) (]lailaboratory tests conducted in similar phosphatechemistry conditions can duplicate the tube wastageobserved in Palisades.

Co-ordinated phosphate chemistry was used inPalisades from start-up until the beginning of large-scale tube failure. It was then changed by addingdisodium phosphate to achieve aNa:POj ratio of 2.0(similar to Mihama-1) and the use of maximumblowdown. In 1073, Combustion Engineeringrecommended a volatile or "zero solids" watertreatment. Prior to this change, a warm water rinse ofthe secondary side of the steam generators was madeto remove phosphates and corrosion products.Approximately 500 kg was removed! '3).

Subsequently, a new corrosion problem, notrelated to wastage has affected at least 50 tubes! 14).It is thought to be due to a form of intergranularattack which has previously been observed in nuclearsubmarine reactor steam generators with stainlesssteel tubes.

R.E. Ginna-l.USA

Although no tube failures occurred in 1973, aJanuary 1974 inspection revealed localized deteriora-tion of a significant number of tubes due to corrosionfrom the secondary side just above the tubesheet. Theexact nature of the corrosion attack resulting in tubethinning was not known; however, it was believedthat prolonged operation beiii-sth a sludge surfacewith a Na:PC»4 molar ratio of 2.2 or less contributedto ihe failures. The majority of defective tubes werefound in the hot leg, but some were found in the coldleg. Between 12 and 18 tubes were pluggedv 15).

SanOnofre-i, USA

San Onofrc-1 experienced seven outages to repairsteam generator tubes to the end of 1973: three ;>fwhich occurred in 1973. A major eddy-current

examination was made during the refuelling outage inJune 1973 when over 50'^ of all lubes were tested.Defect indications were found in 170 tubes. Acomparison was made with defect indications dis-covered the previous year. Some increases werefound. The majority of defects were in the U-bendregion under the anti-vibration bars of peripheraltubes. A smaller number of defects were found 380mm above the tubesheet. Tube sections from bothther.e locations were removed for laboratory investi-gations. Defects in the U-bend region were found u>be caused by fretting wear between tubes and ann-vibration bars. Wall thinning defects above the lube-sheet were attributed to chemistry problems. Lowfrequency eddy-current tests were made to measurethe height of sludge above the tubesheet on thesecondary side. A total of 15 tubes were explosivelyplugged before returning to service.

During a second shutdown in Octobei lc)73. afurther 15 tubes were plugged due to major re-strictions found at the first tube support. The eddy-current probe could not pass by these restrictionswhich may have been caused when the steamgeneratoi was dropped during installation. Thereactor operated with congruent phosphate secondary-water treatment with a Na:P04 ratio between2.3 and 2.6. Earlier in 1973 this ratio had beenbetween 2.0 and 2.6. This change followed recom-mendations from Wesiinghotisc fclcvlric Corporation,the steam generator supplier.

LOCATION AND CAUSE OF FAILURE

Table 2 lists the iailures and their cause. It can beseen that corrosion is far more significant thanvibration as a tube failure mechanism. Figure 1 showsa typical nuclear steam generator with lnconel-600tubes, Regions where failure occurred in 1973 arcshown on the left, and those for 1972 on I lie right.Failures in stainless steel arid Monel-400, and sevcialother lube failures are not shown because the steamgenerator construction is different oi because thelocation of Ihe failure is not known.

Failures continue in predominate in the hot logabove the tubesheet and in the U-bend region. Allfailures propagated I rum the secondary side. Thelubeshcel region is iifien marked by JUMH cioss-llowleading to sludge buildup and possible phuiphaic In hewastage. Many operators have UM/d Inw fiequ'-MUveddy-current testing to measure (lie height of sludge

TABLE 2 - Cause of Steam Generator TubeFailures in 1973

Cause

Corrosion

Vibration

Unknown orOther

Reactor

Beznau-1H.B. RobinsonIndian Point-1KWOMihama-1PalisadesSan Onofre-1

Haddam NeckSan Onofre-1

Desden-1GariglianoN-ReactorSan Onofre-1

Number ofFailures

S3

153lc)

2013-1700

14

101

6"a few"

115

.Wli<Hl«TION 11IAP ItOIOH - - .__ \

PLI610M i.US!..*J0Jl.(.UI( i l l l t l ,,

{» fL»lE IEGIQH

FIGURE 1 FAIIURE REGIONS IN INCOME', 600 TUBED 5TFAM G E N C H A I O R S

on tubesheets. Some operators have resorted tomanual methods to remove it.

The bend region is marked by high tube stressesand complex two-phase cross-flow. Failures in thisarea were largely due to stress corrosion cracking orvibration of the tube against support or anti-vibrationbars. Dryout may also occur here.

Failure patterns indicate that good thermal i."irthydraulic design is of utmost importance in steamgenerator construction. The nature of the failuresindicates that good chemistry control is of utmostimportance in operating steam generators.

LEAK LOCATION, INSPECTION AiND REPAIR

Le-\ks are usually identified by radioactivity in thesecondary system: in steam, boiler blowdovvn orcondenser-air exhaust. They are then located visuallyafter hydrostatically pressurizing the secondarysystem.

Inspection of the leaking tubes is consideredessential to determine the nature of the leak soremedial action may be taken to prevent further

failures. Table 3 shows that eddy-current testing isthe preferred method. It was used by 9 of 11operators who experienced leaks in 1973. Theycarried out inspections ranging from 10 to 100% ofall the steam generator tubes. Operators without tubefailures also performed periodic eddy-current tests forprevenlative maintenance.

Tubes with significant wall thinning or penetrationare plugged. Steam generators usually have sufficientreserve heat transfer surface (~25%) area so thatplugging several percent of tubes does not reducesteam raising capacity. However, tube failures inpossibly three reactors may have been so extensive asto reduce steam raising capacity. The criterion forwhen to plug a defective tube varies between opera-tors and has been the subject of debate with the USAtomic Energy Commissionf 13,16), A recent reportevaluated the safety of 3/4 inch tubes under loss ofcoolant accident (LOCA) conditions and concludedthat tubes with a minimum wall thickness of 25%(75% penetration) v/ould withstand all the loadsimposed upon themO7). Based on this reportoperators at the R.E. Ginna plant plug tubes withdefect indications with 50% or more wall deteriora-tion, assuming that the 25% difference between 75and 50% could occur between successive inspec-tionsO5).

Most operators use explosive plugs for repair. Onlyone case of poor performance ,"ium these plugs hasbeen reported. Most of 104 plugs leaked at IndianPoint-1 and had to he repaired by manual welding.The expense of manual welding in terms of labour,repair time and nian-rcm will undoubtedly meanwider use of explosive plugging in the future. Mostmanufacturers or operators have now developedtechniques for explosive plugging.

Radiation dose to repair crews on sieumgenerators are usually not separated from the totaldose to the reactor si ,ff in general but where they areseparated they point to the high cost of such repairs.Table 3 gives details on some of these doses. Forcomparison, a PWR operator usually budgets for500-700 man-rcm per reactor year for operation andmaintenance.

TABLE 3 - Inspection and Repair of Steam Generators in 1973

Reactoir

Beznau-1

Dresden-1

GiTigliano

HaddamNeck

H.B.Robinson-2

Inspection

- 6 0 % ECT*Defects >75%penetrationplugged

notavailable

notavailable

- 3 0 % ECTDefects >50%penetrationplugged

-100% ECT

Repair

Explosiveplugs 200mm long

notavailable

notavailable

Explosiveplugs

Explosiveplugs

No. ofTubes

Repaired

8

6

few

10

3

RadiationFields

6-9 R/liinside SG.1 20 man-rein dosein 1973

notavailable

notavailable

notavailable

~15 R/hinside SG.

Remarks

1 5 tubes removedin total,up to 100 mm sludgeon lubesheet

1972 ECT resultscompared with 1973,sludge measuredby ECT

Several tubesremoved

54 man-remtotal dose

IndianPoint-1

100% ECT 49 rolled innipples104 explosiveplugs

Reweld pipedown comers

153 Hot leg:lOR/hcold leg:4.8R/h

To 15 R/haverage faR/h.1500 men3500 man-remdose

Some explosiveplugj leaked

Lack of automatic& remote equipmenthampered repairs

TABLE 3 - (Continued)

Reactor

KWO

Miluivu-1

N-Reactor

Palisades

San Onofre-1

Inspection

-30% ECTDetects >80%penetrationplugged

100% ECT

- 8 0 % ECTDefects >hO*penetrationplugged

- 5 0 % ECTDefects >50%penetrationplugged

Repair

Explosiveplugs

Rolled andhand weldedplugs

Mechanical

plugs

Hand weldedInconelplugs

Explosiveplugs

No. ofTubes

Repaired

1')

2013

1

-1700

29

RadiationFields

notavailable

not available

— 1 man-rem

9-10 R/hat man-way—50 man-rem dose

Remarks

1972 ECT resultscompared with 1973,crack propagationnoted

10 Kibes removed,extensive repaireffort necessary

no failures in10 lnconel-600SG's

Wi month shutdownin Jan. 73, 1 yearshutdown startedin Aug. 1973

1972 ECT resultscompared with 1973,some additionalfretting noted

*ECT = Eddy-Current Testing

TUBE FAILURE RATES

Tables 4 and 5 list all the water-cooled powerreactors in the world using volatile and phosphatesecondary-water chemistry. Table 6 lists boiling waterreactors with secondary-steam generators. The tableslist the operating experience from the stait ofoperation and are intended to contrast experiencewith the two types of secondary-water chemistry.

The tables show the eflective full power days fromstartup, the number of tube defects in 1973. and thecumulative total to January ll)74. Mean timebetween failures (MTBF) per 1000 tubes and persteam generator are calculated by dividing thepioduct of either tube operating years or steam

generator operating years by total tube failures.Operating years is defined as the time during whichthe reactor generated electricity. These times weregathered from various sources, including the survey;they are not tabulated. The MTBF is used as an indexof merit; its absolute value may not have practicalsignificance.

The relative merit of the two commonly usedmethods of secondary-water treatment, volatile andphosphate, are much debated. Excluding reactorswith less than 100 effective full power days (EFPD)of operation. Table 4 shows that of 15 reactors with"olatile secondary chemistry, five have had tubefailures. Three of the five have 100 or more. The first,Indian Point-1. has a long history of stress corrosionfailures with its stainless steel tubes. The second.KWO. experienced many failures on the primary side.

obviously unrelated to secondary-water chemistry.Tin' thin!, Shippingpoi t-2, experienced most lubelailures before changing to volatile treatment. Thesewere actually sodium phosphate category lailures.

Table 5 lists reactors wilh sodium-phosphate treat-ment. Ten of IX reactors have had tube failures. Fourof the ten have had 100 or more. F.xtrcmelynumerous failures early in reactor life arc regarded asepidemics. Two reactors. Miliama-I and Palisades, arem this category. Bc/.naii-l appears to be undercontrol, although it once appeared epidemic.

This comparison between the two methods ofsecondary-water treatment suggests that volatile treat-

ment is superior to sodium phosphate. This tact hasrecently been recognized and all major vendors in theUSA have changed their prescribed water treatmentto the all-volatile method.

Table d shows the experience of boiling waterreactors using secondary-steam generators. Thisdesign of system has been superseded and so data forthese systems is of secondary interest. The Tarapurreactors were retubed with Inconel in 1968 prior toservice when their stainless steel tubes experiencedgross lailure due to stress corrosion cracking. Noreactor in this category has had poor operatim;experience in lc)73. Secondary-water chemistry dauwas unavailable for these units.

TABLE 4 - Steam Generator Experience to January 1. 1974: Reactors Using Volatile Secondary Water Treatmtnt

Tube Materialand Reactor

Stainless Steel

Ardennes

Indian Point-1MZFRN-ReactorTrino VercelleseYankee Rowe

Inconel-600

Fort Calhoun

KWOMaine YankeeN-Reactor

Oconee-1Oconee-2

Shippingport-2

Monel-400

KanuppPickeiin^-1

Piekering-2Pickering-3Pickering-4

Rapp

EFPD*

fromStartup

110019001400

-200014003500

601400

170-2000

10010

2200

200700500400200

50

TubeDefects

in 1973

0153

0100

01900000

0000

0

0

173

TotalDefects

0249

NA-15

041

0100

000

0141

00

00U

0

546

MTBF**/

1000 tubes(Years)

0.1

1.5

1.7

0.2

0.3

1.0

MTBF/steamgenerator(Years)

0.1

o.s

1.0

<0.I

0.2

0.5

EFPD - Effective Full Power Davs M 1 B 1 M e a n T i m e B e t w e e n I -a l lure- .

TABLE 5 - Steam Generator Experience to January 1, 1974: Reactors With Sodium Phosphate Secondary-Water Treatment

Tube Materialand Reactor

lnconel-600

Be/naii-1

Be/.nau-2lladdam NeckH.B. Robinson-:lndian-Point-2Jose CabreraMihama-1Mihama-2NPD

PalisadesPoint Beach-1Point Beach-2Prairie Island-1R.E. GinnaSan Onofre-1Surrey-1Surrey-2Turkey Point-3Turkey Point-4Zion-1Zion-2

Inconel-800

BorsseleKKS

Monel-400

Douglas Point

EFPD*fromStartup

l>00

(-.001600700

101250600350

2800

250750250

10001500

200ISO200

8030

h0450

1050

Tubedefectsin 1973

8

010

00

20 i 300

-17000000

30000000

00

0

TubeDefects

%0

02635

0

217301

-1700lc)3

000

54000000

00

1

MTBF**/1000 tubes(Years)

<0 1

3.00.6

3.4Epidemic

15.0

Epidemic~0.1

1.0

—

50

MTBF/SG(Years)

<0.1

0.80.6

1.3Epidemic

7.1

Epidemic~0.1

-

0.3

-

30

iiFPD Effective Full Power Days

MTBF- Mean Time Between Failures

10

TABLE 6 Steam Generator Experience to January 1, 1974: Boiling Water Reactors

Tube Materialand Reactor

Stainless Steel

Dresden-1KRBK.WL

inconel-600

Tarapur-1Tarapur-2

Monel-400

Gavigliano

EFPD*fromStartup

270018001200

800800

2300

TubeDefectsin 1973

C

0

00

"few"

TotalDefects

490

- 2 0

>115

MTBF**/1000 tubes(Years)

1.4

<1.8

<0.2

MTBF/s learngenerator(Years)

0.8

<0.3

<0.1

* EFPD - Effective Full Power Days

** MTBF Mean Time Between Failures

CONCLUSIONS

Improvements to the Design of Steam Generators

Table 2 shows that corrosion and vibrationaccount for the majority of tube failures. Figure 2shows that these failures occur in well-defined areasin the secondary side of steam generators. Theseobservations suggest that much more attention mustbe given to both thermal and hydraulic conditions.Evidence from the massive lube failure rates inMihaina-1 and Palisades suggests that local thermalconditions at the intersection of tubes and anti-vibration straps led to alternate wetting and drying ofthe lube surface. Fxtremely high concentrjtions ofcaustic may occur under these conditions leading torapid tube failure*'8). Hydraulic conditions resultingin poor cross-flow in the region above the tubesheeicoupled with inadequate blowdown led to theformation of thick sludge in Palisades, Haddain Neck.Be/nau-1 and San Onofre-1. Aggressive chemicals,often from condenser in-leakage. brought aboutprematuie corrosion failures. In some steamgenerators, failed tubes formed well-defined patternson the tubesheet, possibly delineating these areas olpoor cross-How.

Major steam gencratoi manufacturers regardinformation about thermal and hydraulic conditionsas proprietary. Very littie data is made puhlic otherthan gross performance characteristics, powerfulanalytical methods aie now becoming available tocalculate a variety of parameters in the three-dimensional, two-phase flow conditions found insteam generators"4 '. Nuclcai steam supply systemvendors in Canada and the USA are now using thesemethods, which arc expected to lead to betteisecondary-side flow conditions.

Corrosion, being a fickle mechanism, is likely tocause failures in the foreseeable future. Thus internalobstructions such as doors, hars and so on. a.* well asexternal access space, should always he designed withrepairs in mind.

Influence of Tube Material

E a r l i e r s t e a m g e n e r a t o r s u r v e y s ' ^ - ' 1 ' i n d i c a t e t h a tc e r t a i n l u b e a l l o y s a p p e a r t o h a v e a m u c h l o w c if a i l u r e r a t e t h a n o t h e r s . A n e x a m i n a t i o n ol i h ep i e s e n t s u r v e y d a t a . d o c . n o t s u p p o r t t h i s . | - o ii n s t a n c e , t h e e a r l y w i d e - s p r e a d failure*, o l - . i . n u l c ^ -s t c e l t u ' . e d - . t e a m g e n e r a t o r . - , is nov . l a i g e K u n d e ic o n t r o l a n d o p e r a t o r s m a y e x p e c t l o n g s c i v u r w i t hc a r e f u l u l i e m i c a l c o n t i o l <>| t h e s e c o n d e r s w a l c i .

1!

Secondary-Water Chemistry REFERENCES

h is clear from the operating duta ol some .Vrcuii-i^ with ovei 100 L 1PD ofopeKiiion that thosewith \>'!.itile secondary-water treatment have had•siaiiilu .intly lewer tube failures that ihose withsodunn-phosphate treatment. No tube failureepidemic have occurred in reactors with volatilewater treatment in contrast to at K'asl uvo in reactorswith phosphate treatmen!.

However, the authors arc aware of the complexityof corrosion phenomena in steam generators andchoose to draw no conclusions other than the simpleone above. The interested reader is directed to arecent bibliography water- and steam-sidecorrosion in boilers covering 247 relerences'-'-"during ll>5 I-1L>73. Pioblems specific to nuclear steamgenerators have been discussed a', several recentcoherences*->•-->.

Improvements to Maintenance

Equipment for inspecting and repairing sieamgenerators is indispensable. As the frequency oi'repairs and the level of radiation fields increase,operators find that remote and automatic machineryfor tube inspection and repair become more neces-sary. Eddy-current testing is regarded as the bestmethod for inspection, a fact reinforced by the recentUSAEC regulatory guide which specifies its use indetail'-3). The next development in this field is toapply systems engineering techniques to process,diagnose and store eddy-current signals from tubes.This will enable operators to anticipate tube failure insome degree and allow preventative tube pluggingduring scheduled reactor shutdown. The applicationof systems engineering cwi also yield remotemachine^ that are capable of carrying out inspection,plugging and proof testing with a minimum of manualintervention and hence low radiation dose to theoperators.

Chemical and mechanical methods of decon-taminating primary circuits and removing depositsfrom secondary circuits may also be necessary ifradiation fields continue to increase with operatingtime. Such methods have implied benefits and riskswhich should be weighed carefully before use.

ACKNOWLEDGEMENT

This survey was only possible with the help ofmany reactor operators. The authors gratefullyacknowledge theii co-operation.

1. D.J. Skowholl, "Performance and Reliability ^Ojieialing I'niieii States Nuclear Power Plants".IA1A-SM-I7N 101. |W74.

2. "Watch the Wa'er Chemistry". Nuclear L:n\i-necring International. V. IN. n. 211. p. 'M 5-9,6,Dec. ll>73.

.'. 'industry Increasingly Concerned Over SteamGenerator Problems". Nucleonics Week. V. 14.n. Ml. p. .vd, Dec. 13. l°73.

4. "Power and Reseaich Reactors in MemberStates". ll*74 Edition. International AtomicEnergy Agency. Vienna, 11>74.

>. P.D. Stevens-Guille, "Steam Generator TubeFailures: A World Survey of Water-CooledNuclear Power Reactors During 1972". AtomicEnergy of Canada Limited report AECL-4753.March 1974.

6. P.D. Stevens-Guille, "Steam Generator TubeFailures: A World Survey of Water-CooledNuclear Power Reactors to the end of 1971".Atomic Energy of Canada Limited reportAECL-4449. April 1973.

7. H. Baschek, H. Sandana, "The Steam GeneratorFailure History of the Nuclear Power PlantsBeznau-1 and -2," presented to 1974 Educa-tional Seminar, "Colloquium on Steam Genera-tor Tubing Failures", Southwest Researchinstitute. San Antonio, Texas, USA, 1974.

8. Nuclear Newsletter from Switzerland. No. 28,Summer 1974.

9. H.B. Robinson Plant, Unit 2 Semi-AnnualOperations Report No. 7, July 1, 1973 to Dec.31, 1973 (Carolina Power and Light Co. RaleighN.C. (USA)), United States Atomic EnergyCommission Docket 50261-325, 26 Feb. 1974.

10. A.S. Cheifetz, "Health Physics Aspects of aNuclear Boiler Downcomer Repair Operation",American Nuclear Society Annual Meeting.Philadelphia. 1974.

11. A. Mayr, E. Pickel. "Investigation of" SteamGenerators in Nuclear Power Plant Obrigheim(KWO)", Reactor Conference 1974 of theGerman Atom Forum/KTG, Berlin (in German).

12

12. k. Stone, "Inspection of Tube Thinning inSteam Generators", Topical Meeting, Radio-graphy and Complementary Nondestructivelesling in the Nuclear Industry, Portland.Oregon, Aug. 1<)72.

13. "Response to March 7, ll)74 Requests forAdditional Information Concerning PalisadesSteam Generator Tube Examination and RepairReport" (Non Proprietary Version) ConsumersPower, Michigan, United States Atomic EnergyCommission Docket 50255-234, Mar-h, 11>74.

14. Nucleonics Week. V. 15, n. 32, p. X-1), Aug. X.

22. Electric Puwer Research Institute and AmericanSociety ot Mechanical Engineers. InternationalWorkshop on Steam Generator1.. Palo <\h<>Calif. I SA, July, 1<J74.

23. "Regulatory Guide 1.83", United States AtomicEnergy Commission, ll)74.

15. "Description of Present Status of SteamGenerators". Ruches ter Gas and ElectricCorporation, R.E Ginna Nuclear Power PlantUnit No. 1, United States Atomic EnergyCommission Docket 50244-25d, March l'>74.

16. "Palisades Nuclear Plant. Report CEN-2(P)-01.Steam Generator Tube Examination andRepair". Consumers Power Co., Jackson. Mich.(USA). United States Atomic Energy Com-mission Docket 50255-224, 26 Feb. 1974.

17. P. De Rosa, W. Rinne, ll.W. Massic. Jr.. P.Mitchell, "Evalualion of Steam Generator Tube.Tube Sheet, and Divider Plate Under CombinedLOCA and SEE Conditions". WestingliouscElectric Corporation. WCAP-7K32.. W 3 .

IS. "Steam Generator Problems at PWR's", UnitedStates Atomic Energy Commission Report. ROE73-8, Aug. l°-73.

I1). S.V. Patankar. D.B. Spaulding. "A CalculationalFiocedure lor Heat, Mass and MomentumTransfer in Three-Dimensional Parabolic Flows".Int. J. Heat Mass Transfer, V. 15, p. 1787-1806.

20. "Water- and .Steam-Side Corrosion of Boilers".Central Electricity Generating Board. Report C'EBibliography 233, 1C73.

21. National Association of Corrosion Engineers. 4-SMarch 1474. Chicago. 111.. USA.

The International Standard Serial Number

CN ISSN 0067-0367

has been assigned to this series of reports.

To identify individual documents in the serieswe have assigned an AECL- number.

Please refer to the AECL- number whenrequesting additional copies of this document

from

Scientific Document Distribution OfficeAtomic Energy of Canada Limited

Chalk River, Ontario, Canada

KOJ 1J0

Pr ice - 5 0 $ per c o p y

238-75