NRC Generic Letter 1979-15

NRC Generic Letter 1979-015: Operating Problems Have Occurred in Pressurized Water Reactor Steam Generators

ML031320280
Person / Time
Issue date:	03/21/1979
From:	Eisenhut D G Office of Nuclear Reactor Regulation
To:
References
NUREG-0523, GL-79-015
Download: ML031320280 (89)

so 5tk~ ~ ~UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D. C. 20656REGULATORY DOCKET FILE COPYMarch 21, 1979GL-7'7-ALL PRESSURIZED WATER REACTOR LICENSEESGentlemen:This letter is being sent to all licensees authorized to operate or constructa pressurized water power reactor and to all applicants for a license tooperate or construct a pressurized water power reactor.Operating problems have occurred in Pressurized Water Reactor (PWR) steamgenerators. The enclosed report, "Summary of Operating Experience withRecirculating Steam Generators," NUREG 0523, focuses on the problemsassociated with steam generators of the recirculation type, i.e., thosemanufactured by Combustion Engineering and Westinghouse. The report dis-cusses the NRC staff's evaluation of these problems and the programs forresolving these problems.The NRC has recently identified steam generator degradation as an UnresolvedSafety Issue deserving the highest priority for resolution. However, forthe reasons identified in the report, the NRC staff has concluded that con-tinued operation of existing plants and licensing of new plants with recir-culation type steam generators, pending completion of our review, does notconstitute an undue risk to the health and safety of the public and there-*fore may continue.It should be noted that a number of research efforts are currently underway which will improve our knowledge of steam generator degradation mech-anisms. The information presented in the report represents our currentunderstanding of each issue. Comments on this report and information* C9f p.-2- March 21, 1979related to steam generator degradation mechanisms are encouraged andshould be forwarded to Dr. Boen-Dar Liaw, Engineering Branch, Divisionof Operating Reactors, U.S. Nuclear Regulatory Commission, Washington,D.C. 20555.hy DirectorDivision of Operating ReactorsEnclosure:Summary of Operating Experiencewith Recirculating SteamGenerators, January 1979,NUREG 0523cc w/o enclosure:Service List Mr. William J. Cahill, Jr.Consolidated Edison Company of New York, Inc.cc: White Plains Public Library100 Martine AvenueWhite Plains, New York 10601Joseph D. Block, EsquireExecutive Vice PresidentAdministrativeConsolidated Edison Companyof New York, Inc.4 Irving PlaceNew York, New York 10003Edward J. Sack, EsquireLaw DepartmentConsolidated Edison Companyof New York, Inc.4 Irving PlaceNew York, New York 10003Anthony Z. RoismanNatural Resources Defense Council917 15th Street, N.W.Washington, D. C. 20005Dr. Lawrence R. QuarlesApartment 51Kendal at LongwoodKennett Square, Pennsylvania 19348Theodore A. RebelowskiU. S. Nuclear Regulatory CommissionP. 0. Box 38Buchanan, New York 1051

&4fe L :/j;Y2-13 -?/- /--' NUREG-0523SUMMARY OF OPERATING EXPERIENCEWITH RECIRCULATING STEAM GENERATORSD. G. EisenhutB. D. LiawJ. StrosniderOffice of Nuclear Reactor RegulationU. S. Nuclear Regulatory Commission7911 090 2 4 8 X,,/,, Ac 3 -o--/ -/'0.9NUREG-0523SUMMARY OF OPERATING EXPERIENCEWITH RECIRCULATING STEAM GENERATORSD. G. EisenhutB. D. LiawJ. StrosniderManuscript Completed: January 1979Date Published: January 1979Division of Operating ReactorsOffice of Nuclear Reactor RegulationU.S. Nuclear Regulatory CommissionWashington, D.C. 20555

% UCONTENTSPageACKNOWLEDGMENT ...................................................... vii1. INTRODUCTION .................................................. 12. DESCRIPTION OF TYPES OF OPERATIONAL PROBLEMS .................. 32.1 Caustic Stress Corrosion and Wastage ..................... 32.2 Denting and U-Bend Cracking .............................. 42.3 Tube Support Plate Cracking .............................. 92.4 Anti-Vibration Bar Wear or Fretting ...................... 93. OPERATING EXPERIENCE .......................................... 133.1 Plants Designed by Westinghouse .......................... 133.1.1 Caustic Stress Corrosion and Wastage .............. 133.1.2 Denting and U-Bend Cracking ....................... 203.1.3 Tube Support Plate Cracking ......................... 253.1.4 Anti-Vibration Bar Wear or Fretting ................. 263.2 Plants Designed by Combustion Engineering ................ 263.2.1 Caustic Stress Corrosion and Wastage .............. 283.2.2 Denting ........................................... 283.2.3 Tube Support Plate Cracking ....................... 324. CORRECTIVE ACTIONS AND REPAIRS ...334.1 Short-Term Program and Licensing Requirements ..334.1.1 Turkey Point Units 3 and 4 and Surry Units 1and 2 ........................................... 344.1.2 Indian Point Unit 2 .. 374.1.3 San Onofre Unit 1 .. 374.1.4 Millstone Unit 2 .. 384.1.5 Other CE Facilities .. 394.2 Long-Term Repairs ........................................ 394.2.1 Tube Sleeving ..................................... 394.2.2 Steam Generator Repair ............................ 404.2.3 Condenser Integrity ............................... 424.2.4 Condensate Polishers .............................. 424.2.5 Steam Generator Tube Repair ........................ 43i CONTENTS (continued)Page5. RELATED RESEARCH PROGRAMS ..................................... 445.1 Westinghouse Electric Corporation ........................ 445.2 Combustion Engineering ................................... 445.3 NRC-Funded Research Programs ............................. 456. CONCLUSIONS ................................................... 466.1 Basis for Continued Operation ............................ 466.2 Basis for Continued Operation of Plants with SevereDegradation ............................................ 476.3 Licensing of New PWR Facilities .......................... 47APPENDIX A -PWR DESIGN CONFIGURATION .............................. 49APPENDIX B -TASK ACTION PLANS ..................................... 53APPENDIX C -CONDENSER TUBE MATERIALS FOR OPERATING PLANTS (PWR) ... 79GLOSSARY ........................................................... 81ii ALIST OF FIGURESFigure Title Page1 Problem Areas in PWR Steam Generator. 22 Typical Denting Mechanism. 53 Flow Slot Deformation. 64 Flow Slot "Hourglassing". 75 Support Plate Cracking at Edge of Flow Slot. 86 Schematics of U-Tube Ovalization .107 Cross Section of Dented Tube Showing Locationof Leakage .118 Steam Generator Support Plate "Islanding" .129 A Typical Westinghouse Steam Generator ..... ............ 16JO Drilled Tube Support Design .......... ................... 1711 CE Steam Generator Egg Crate Tube Support PlateDesign .1812 Summary of Secondary Water Chemistry Treatmentin Operating Westinghouse Plants ...... ............... 1913 Typical Tube Support Plate Hard Spots ..... ............. 2214 A Typical Combustion Engineering Steam Generator ....... 2715 Summary of Secondary Water Chemistry Treatmentin Operating Combustion Engineering Plants .2916 Cross Section of Steam Generator Tube Array .3517 Steam Generator Tube Sleeve .41A-1 Pressurized Water Reactor (PWR) Cooling Cycles .50A-2 Schematic of Reactor Coolant System for PWR .51iii LIST OF TABLESTable Title Page1 Summary of Steam Generator Adverse Experience 142 Steam Generator Tube Plugging Summary ..... ........ 153 Denting in Westinghouse Steam Generators ..... ..... 234 Denting in CE Steam Generators ...... .............. 30iv

1.ACKNOWLEDGMENTThe NRC staff gratefully acknowledges the permission granted bylicensees and vendors to use the following figures:Figure 4Figure 5Figure 9Figure 14San Onofre Unit 1, Southern California Edison andSan Diego Gas and Electric CompanyIndian Point Unit 2, Consolidated Edison CompanyWestinghouse Electric CorporationCombustion Engineeringv SUMMARY OF OPERATING EXPERIENCEWITH RECIRCULATING STEAM GENERATORS1. INTIODUCTIONOperating problems have occurred in the steam generators of each of thethree manufacturers of pressurized water reactors (PWR) nuclear steamsupply systems (NSSS): Babcock & Wilcox, Combustion Engineering, andWestinghouse Electric Corporation. This report focuses on the problemsassociated with steam generators of the recirculation type that are designedby Westinghouse and Combustion Engineering. It identifies the operationalproblems observed to date, including the NRC staff's evaluation of suchproblems, and provides a status report summarizing the NSSS, licensee, andstaff programs for the resolution of each problem. (Figure 1 shows themajor types of degradation for recirculation type steam generators.)Information related to the cause of these problems is discussed to theextent that such information is known and available. It should be notedthat a number of research efforts related to these problems are currentlyunder way. Therefore, some of the causal information included in thissummary represents our current understanding of each issue. For those whoare not completely familiar with PWRs, Appendix A briefly describes thefunctions of the various coolant systems of pressurized water reactors.-1- FIGURE 1. PROBLEM AREAS IN PWR STEAM GENERATORI

2. DESCRIPTION OF TYPES OF OPERATIONAL PROBLEMS2.1 Caustic Stress Corrosion and WastageInconel-600 tubing is typical of that found in most operating recircu-lation types (U-tube) of steam generators. Intergranular stress corrosioncracking and localized tube wall thinning (wastage) are the major types ofdegradation that affect the exterior surface of the tubing. "Pitting"(that is, relatively deep, small volume wastage of the exterior surface ofsteam generator tubing) has also been experienced.Wastage has occurred when a coordinated phosphate treatment of the secondarycoolant has been utilized and is attributed to the local concentration ofresidual acidic phosphates. In some cases, these acidic phosphates havenot been completely removed after a changeover from a phosphate treatmentto all-volatile treatment (AVT)* of the secondary coolant water. Approxi-mately a dozen plants have experienced some degree of wastage while operatingwith phosphate water treatment. Since the establishment of AVT chemistrycontrol, both the evidence and the extent of wastage have diminished andno further substantial tube degradation due to this mechanism is expectedto occur. Caustic stress corrosion cracking is caused mainly by eitherthe formation of caustic compounds in the secondary coolant (i.e., fromhydrolysis of trisodium phosphate) or by caustic-forming impurities carriedinto the steam generator by the feedwater.The principal cause of serious corrosion damage from either wastage orcaustic stress corrosion cracking is the local concentration of aggressivechemicals within the secondary side of steam generators. The major sourceof these impurities is in-leakage of condenser cooling water. Because ofthis, the boundary between the secondary coolant system and the condensercooling system is of significance. The concentration of these impuritiesis affected by thermal and mechanical design parameters of steam generators,by accumulations of chemicals and corrosion products within the steamgenerators as plants age, and by the nominal and transient variations inwater and air environments to which steam generator internals are exposed.Both types of corrosion generally occur where regions of restricted waterflow and high heat flux tube surfaces cause impurities to concentrate orphosphates to precipitate (hideout). These high concentrations may occurat crevices between the tubing and the tube support plates or the tubesheet, and in areas where sludge deposits have built up on the tube sheetor tube support plates."This chemistry control is called AVT because the chemicals injected intothe secondary water eventually volatilize and escape with steam.-3-

2.2 Denting and U-Bend CrackingIn December 1975, the NRC was informed by Westinghouse that several plantsdesigned by them had experienced steam generator tube deformation in theform of a reduction in tube diameter. This reduction in tube diameter waslater termed "denting."Later laboratory reports of dented tubes indicated that the annulus betweentubes and support plates was filled with hardened corrosion products (asshown in Figure 2) that continue to form by the corrosion of the supportplates and, therefore, exert sufficient forces to "dent" the tube diametri-cally. Severe buildup of corrosion products has caused cracking of thetube support plate ligaments between the tube holes and the water circulationflow holes. The phenomenon of denting in Westinghouse plants has beenattributed to acid chloride salts that concentrate in the annulus betweenthe tubes and the tube support plates. The first incidence of dentingoccurred shortly after steam generator secondary water chemistry controlwas switched from phosphate treatment to an all-volatile treatment (AVT).Contamination of the secondary coolant by inleakage of condenser coolingwater was believed to have caused a catalytic reaction with residualphosphates.The simultaneous presence of residual phosphate in the tube/tube supportplate annulus and chloride in the condenser cooling water caused acceleratedcorrosion of carbon steel support plates present in most plants. Thecorrosion product from the carbon steel support plate occupies approximatelytwice the volume of the material corroded. The continuing corrosionproduct exerts sufficient forces to dent the tube and/or crack the tubesupport plate ligaments between the tube holes and the water circulationflow holes. These dented tubes thus become subject to higher strains;however, they have otherwise generally retained their integrity. (Thatis, there have been relatively few leaks at the dent locations and norapid failures at dent locations.) Denting has occurred more recently atplants that have used AVT exclusively.Along the chord of the innermost rows of tubes in Westinghouse-designedsteam generators, there is a row of rectangular flow slots in the tubesupport plate. These slots are approximately 16 inches long by 2-3/4inches wide and are spaced about 20 inches center to center (see Figure 3).Because of the pressure built up in the tube support plate due to thedenting phenomenon, the flow slots in the tube support plates have beenobserved to deform (the "hourglassing" effect); that is, the centralportion of the parallel flow slot walls has moved closer, so that someflow slots are now narrower in the center than at the ends. Figures 4and 5 are photographs of hourglassed flow slots from San Onofre Unit 1and Indian Point Unit 2, respectively. Because the initial parallel slotwalls have moved closer, the tube support plate material supporting the-4 -

0.014"1-70.75"4- 0.014" (DEPOSIT)P CORROSION PRODUCT0.75"FIGURE 2. TYPICAL DENTING MECHANISM

FIGURE 3. FLOW SLOT DEFORMATION I

IIHOTLEGSi DECOLDLEGSi DEIIIIFIGURE 4. FLOW SLOT "HOURGLASSING"

IFIGURE 5. SUPPORT PLATE CRACKING AT EDGE OF FLOW SLOT

tubes nearest this central portion of these flow slots has also movedinward, which consequently forces an inward displacement of the legs ofthe tubes at these locations. When this inward movement of the legs ofthe tubes has occurred at the upper support plate, it has been shown tocause an increase in the hoop strain at the tube U-bend apex. This effectis shown in Figure 6. It is this additional increase in strain at theapex of the U-bend that is believed to be the additional factor requiredto initiate and increase the susceptibility of Inconel-600 alloy tubingexposed to PWR reactor coolant to stress corrosion cracking at the top ofthe U-bend.Because of tube denting or ovalization (non-uniform denting, see Figure 7),tubes at tube/tube support plates have developed small stress corrosioncracks in the longitudinal direction of the tube. These small cracks aremasked by the support plates. During normal operation, small leaks throughthese cracks have occurred in a few plants where severe tube denting hasoccurred.2.3 Tube Support Plate CrackingAs a consequence of continuing magnetite growth in the tube-to-tube supportplate annulus and the subsequent cracking of support plates, portions(small pieces) of the support plate material in Westinghouse-designed steamgenerators have moved with tubes into the flow slots to cause the so-called"islanding" phenomenon; i.e., broken support pieces moving into flow slots(see Figure 8).This phenomenon could lead to the possible loss of lateral support of someinner-row tubes. Concern about this problem has been alleviated in manyplants by the fact that many tubes in inner rows have been plugged as partof the preventive plugging programs.2.4 Anti-Vibration Bar Wear or FrettingSteam generator design in several plants use anti-vibration bars in theupper "curved" portion of U-bend steam generator tube bundles to providelateral support. Inspections in two plants have shown that these barshave caused fretting between the tubes and the bars. Recent inspectionsof some removed bars have revealed serious degradation. The cause appearsto be primarily mechanical and is affected by the material (i.'e., carbonsteel bars versus Inconel bars in new designs), the shape of the bars, theclearances, and the bar support design. Only two operating nuclear powerplants in the United States have ant4-vibration bar design that haveexperienced degradation and, thus, this is not considered to be a widespreadproblem.-9 -

IlSECTION A-AEXTRADOSEINITIALTHICKNESS0.050"IT OINTRADOSEFIGURE 6. SCHEMATICS OF U-TUBE OVALIZATION

/7 1NSCALE: OUTSIDE DIAMETER = 0.875 INCHESWALL THICKNESS = 0.050 INCHESFIGURE 7. CROSS SECTION OF DENTED TUBE SHOWING LOCATION OF LEAKAGE(TURKEY POINT UNIT 4,4TH TUBE SUPPORT PLATE ELEVATION).

FLOW SLOTBROKEN PORTION OF SUPPORTPLATE DISPLACED INTO FLOWSLOT TO FORM "ISLAND"TUBE HOLEAND TUBE FLOW HOLE0 000 0 ° °00FIGURE 8. STEAM GENERATOR SUPPORT PLATE "ISLANDING"

3. OPERATING EXPERIENCEAt this time (December 1978), there are thirty-three operating PWR unitswith recirculation type steam generators in the United States.* Of these,seventeen have been found to have one or more forms of tube degradation.These total numbers do not include eight PWRs using the once-through steamgenerators designed by Babcock & Wilcox. Table 1 identifies the Westinghouseand Combustion Engineering units that have been found to have significantforms of degradation. Table 2 summarizes the extent of steam generatortube plugging that has been performed to date as a result of all modes ofdegradation.3.1 Plants Designed by WestinghouseAll commercially operating Westinghouse-designed steam generators are thevertical shell recirculation type units (Figure 9). All use Inconel-600tubing except for the Yankee Rowe unit, which uses stainless steel tubing.A major consideration in all Westinghouse designs of operating steamgenerators is that they use several fully extended drilled support plates(Figure 10). The drilled support plates are significant because theannular space between the steam generator tube and the support plate isrelated to several forms of degradation. By comparison, the typicalCombustion Engineering design in operating reactors uses both drilledsupport plates and "egg crate" support plates (Figure 11).The controlling parameter for the various corrosion mechanisms that leadto tube degradation appears to be related to steam generator secondarywater chemistry control. The operating history and method of secondarywater chemistry control for these plants is shown in Figure 12. Thepredominant method of chemistry control prior to 1975 was coordinatedpH-phosphate control. In late 1974 through early 1975, thirteen Westing-house plants converted from phosphate control to all-volatile treatment(AVT). Nine newer plants started up with AVT control. Two plants electednot to convert to AVT because of concern for condenser tube integrityproblems and their particular satisfactory operating history with respectto steam generator tube corrosion. Both plants did make minor changes toensure a more restrictive range of phosphate concentration. Tube corrosion(thinning) at both plants is continuing but at a much slower rate.3.1.1 Caustic Stress Corrosion and WastageThe purpose of the chemistry changeover for the operating plants fromphosphate (P04) chemistry control to the AVT method was principally toarrest tube thinning (wastage) that primarily occurred near the tubesheet.*Indian Point Unit 1 is not included.-13 -

TABLE 1SUMMARY OF STEAM GENERATOR ADVERSE EXPERIENCE12/29/78SECONDARY SIDE DENTING CONSIDERATIONU-BEND CRACKING __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _FORNSSS PLANT NAME WASTAGE FRETTING HIGH CYCLE TUBE SP HOUR- SP CRACKING LEAKING U-BEND REPLACEMENTNSSS PLAN NAM WASAGEOR RETUBINGFATIGUE SCC* DENTING GLASSING OR ISLANDING DENTS CRACKCE MAINE YANKEE X.MINORMILLSTONE 2 X-MODERATE XPALISADES X X X-MINOR X (Tube Sleeving)ST. LUCIE 1 X-MINORW HADDAM NECK X X X X.MINORR.E. GINNA 1 X X X-MINORINDIAN POINT 2 X-MODERATE X X XINDIAN POINT 3 X-MINOR XPOINT BEACH 1 X X X-MODERATEPOINT BEACH 2 X X-MODERATE X(?IH.B. ROBINSON 2 X X X-MINORSAN ONOFRE I X X X-EXTENSIVE X X XSURRY1 X X-EXTENSIVE X X X X XSURRY 2 X X-EXTENSIVE X X X x xTURKEY POINT 3 X X.EXTENSIVE X X X XTURKEY POINT 4 X X-EXTENSIVE X X X X XYANKEE ROWE X*SCC -CAUSTIC STRESS CORROSION CRACKINGNOTES: 1. TO DATE THERE ARE 33 OPERATING PWR UNITS (NOT INCLUDING INDIAN POINT 1) WHICH UTILIZE RECIRCULATION TYPE OF STEAMGENERATORS.2. 17 HAVE BEEN FOUND TO HAVE ONE OR MORE FORM(SI OF DEGRADATION. AS SUMMARIZED ABOVE.3. TROJAN AND D. C. COOK HAVE HAD INDICATIONS OF LIMITED DEGRADATION IN RECENT INSPECTIONS.

TABLE 2STEAM GENERATOR TUBE PLUGGING SUMMARY12/29/78NSSS PLANT NAME PERCENTAGE OF TUBES PLUGGED LEAKING/LOST l REMARKS_ _ _ _ Pl10 2 0 l 30 PLUGSCE Arkansas 2 _Calvert Cliffs 1 0%,1/78Calvert Cliffs 2 0%/, 1/78Fort Calhoun 1 0%, 11/77Maine Yankee 0%. 4/77, some dentingMillstone 2 2-78Palisades 1 _ 2-78 35 tubes are sleevedSt. Lucie 1 _I 0%. 2/78 some dentingW Beaver Valley 1Cook 1 0%, 2/77Cook 2Farley 1 _ 0%, 5/78Ginna 1 _477Haddam Neck 10-77 <1%Indian Point 2 5-78 <1%Indian Point 3Kewaunee 0%, 4/78North Anna 1Point Beach 1 2-78 LostPoint Beach 2 3-78 <1%Prairie Is. 1 0%, 3/77Prairie Is. 2 11-77 <1%Robinson 2 2-78Salem 1San Onofre 1 477 _ _ weSurry 1 12-78 LeakingSurry 2 -X 7-78 -LzakingTrojan 0%, 5/77, 1 tube leakingTurkey Point 3 12-77 -_ _ LostTurkey Point 4 , ..78 LostYankee Rowe 7-77Zion 1 =- 0%, 3/78Z io n 2 _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

FIGURE 9. A TYPICAL WESTINGHOUSE STEAM GENERATOR

SUPPORTPLATESECTIONFIGURE 10. DRILLED TUBE SUPPORT DESIGN

FIGURE 11. CE STEAM GENERATOR EGG CRATE TUBE SUPPORT PLATE DESIGN

lofln IQR7 i196 19R6 1Q70 1071 1972 1973 1974 1975 1976 1977 1978 1979.-,5. .....1I.. ' " I-' 1 ......... ......... ......../z,0,r,/z,/.,0Vzz...!. !-:-lwi; ;;; e ;e.....; ;-; ;-- .i i -_ *- --1 ---------j ---F,- -, .--.. -.-X ..----------: ;;;. ;1 1 : : ';!.!.!.!.!.! !-----: :-;: : : ':- 1-: -: -: -f:: :-: : : :, : .: .: : ! 0... i. ... ..---------- ------.... ---------- ----- -- ---- -.....-.......F-.LT :: :: : ..:.... ;.;.. ;.. i.;... ... .......;..;..;. .-b::.:::, :, j ;; J _lI::::: .: :.::::%::::T ::: ...lX ...... ., ..:;::: ........... ,, ,,,,,, ,.,,,,.,,, r.,,,...,.. .... Y/X. Y/// r/x///g//:l4;:;;;..; _ : ' i ' .1.S .S -_ _ tr _ _ _ s _ _ _ f _ _ _;;.....: ::. ..::: m:., '. ./ // E / I/IlI I I I I g _ _ _ _. _ _ _ Is _ _ v V w... Y, %/77ZzV0z Z_ : .i _ ^ b _ _ _ i _ _ _ _ ._ S _ fl _ _---i .........: K77/y 7 7 74, 4 --- r ri n r-....: ..._- ... .l ...:w:yX77$X777AX$XAXXX,*- ----- ----rYankee RoweSan Onofre 1Haddam NeckGinna 1H.B. Robinson 2Point Beach 1Point Beach 2Surry 1Turkey Point 3Surry 2Indian Point 2Turkey Point 4Zion 1Prairie Is. 1Zion 2KewauneePrairie Is. 2Cook 1TrojanIndian Point 3Beaver Valley 1Salem 1Farley 1Cook 2North Anna 1-.. ....I e ...- --- I -. -. ---- ------ --Io .jO- /i:-- &t -7 --Y 7 7 11 n i niL :::: I: '/ / -/ e / / ' --I- ---I... : .~,t/77/Y77/7774 V/fr, , .277,1/2'7A'AAA</7,,, -, -.I.He ------VZ/VZZIV//wr a,, Us,, -IVA///<g7A'77LLegend: I:2.lPO47/lAVTFIGURE 12. SUMMARY OF SECONDARY WATER CHEMISTRY TREATMENTIN OPERATING WESTINGHOUSE PLANTS

Turkey Point Units 3 and 4 and R. E. Ginna, whose steam generators hadexhibited extensive thinning while on P04, experienced significantlyreduced rates of wall thinnings following conversion. Improvements intechnology in performing steam generator tube inspections using eddycurrent techniques (ECT) over the last few years have resulted inidentification of an apparent increased rate of wall thinning.The objective of using P04 control was to buffer inleakage of impuritiesfrom the condenser and to prevent formation of boiler scale on the steamgenerator tubes. Control of caustic level was also of concern. In fact,improper use of phosphate often leads to caustic stress corrosion. Withthe changeover to AVT control, caustic stress corrosion has remained aconcern. Stress corrosion cracking in plants that converted from P04 toAVT control is related to previous P04 concentration and possibly tomakeup water contamination. Plants with only short periods of P04 controlbefore conversion and plants that initially started with AVT have notexperienced operational problems due to tube wall thinning or causticstress corrosion cracking.A second significant effect of the conversion to AVT upon wastage hasoccurred due to a change in the character of steam generator sludgedeposits. In steam generators using phosphate, the sludge is coarse,granular material that forms a cohesive mass on the tube sheet. Plantsthat converted after d short period of phosphate treatment have exhibiteda finely divided sludge of dense particles that are more easily removed bywater-lancing procedures. The sludge is similar in metal composition tothe phosphated sludges because the iron impurities in the feedwater areunchanged. The improved ability to remove the "AVT sludge" should help tominimize wastage of steam generator tubes because the wastage is mostsevere within areas of sludge deposits.3.1.2 Denting and U-Bend CrackingDenting is caused by the buildup of corrosion products in the crevicesbetween the tubes and tube support plates or between tubes and the tubesheet. The corrosion products, which are primarily derived from thecarbon steel support plate and consist mainly of iron magnetite (Fe304),expand volumetrically (about 2:1) to fill the crevice and, therefore,exert forces on the tubes and on the tube support plates. Phenomenadirectly associated with denting include the following:Tube diameter reductionTube leakageTube support plate hole distortion.Tube support plate flow hole distortion (flow slot hourglassing)Tube support plate expansion and crackingWrapper distortion-20 -

Denting has resulted in greater than about 0.25 inches reduction in tubediameter in the most severely affected units (3/4- and 7/8-inch tubes).The reduction in tube diameter is generally not concentric. This isdramatically illustrated in Figure 7, which shows the cross section of adented tube removed from an operating facility. Areas of high tensilestress on either outside or inside surfaces in dented tubes are susceptibleto stress corrosion cracking, and small leaks have occurred in steamgenerators with severe denting.Areas of the tube support plates located near the edges of flow slots andthe support plate periphery, which do not have flow holes, are stifferthan the rest of the plate. These areas, termed "hard spots," haveexperienced more severe denting than other regions of the steam generator.These areas are shown in Figure 13. In less-stiff areas of the supportplate, it is somewhat easier for the plates to deform than it is for thetubes to be dented. Distortion of the flow holes, flow slots, and plateperiphery have occurred as the volume of the corrosion products and tubesupport plate increases. Figure 4 shows the hourglassing (see Section 2.2)type of deformation of the support plate in the flow slot area. In someinstances, extreme support plate deformation has resulted in cracking ofthe support plates. Figure 5 is a photograph of cracking that occurs atthe flow slot edge. Cracking of the support plate behind the first row oftubes has also been observed. This form of cracking (also known as islanding,see Secton 2.3) causes a portion of the support plate and the tubes containedin it to move into the slot (see Figure 8).The denting phenomena are believed to be directly related to the secondarywater treatment history of a plant. Plants that converted to AVT afterextended use of phosphate water treatment have had severe denting. However,recent experience indicates that ingress of chlorides through condenserleaks may be a significant contributor to denting. Thus far, denting hasnot been significant at plants with low chloride in the condenser coolingwater although denting has been found to occur in plants with very lowamounts of chlorides. The most severe cases are plants with brackish orseawater condenser cooling. Table 3 lists plants discussed herein andtheir types of condenser cooling water and condenser tube material.Except for Indian Point Unit 3, denting has not been reported in Westing-house plants with all AVT or limited phosphate history. The extent ofdenting of Indian Point Unit 3 is minor with average denting of about 3 to4 mils. The success of AVT can be attributed to the close control ofchloride ingress. Westinghouse AVT chemistry specifications establishstrict chemistry guidelines regarding cation conductivity and chloridelevels. When condenser problems occur, the AVT chemistry guidelines areexceeded and plants that implement timely corrective actions have avoidedsevere denting problems.-21 -

LEAKER0 HARD SPOTSUqFIGURE 13. TYPICAL TUBE SUPPORT PLATE HARD SPOTS

TABLE 3DENTING IN WESTINGHOUSE STEAM GENERATORSEXTENT OFDENTING* CONDENSER TUBE MATERIALPLANTCOOLING WATERSurry Units 1 & 2 Extensive 90 Cu 10 Ni BrackishTurkey Point 3 & 4 Extensive Al-brass SeawaterSan Onofre Unit 1 Extensive 2 boxes-titanium Seawater(stabilized) 2 boxes-90-10 CuNiPoint Beach Units Moderate Admiralty with stainless steel Fresh water (lake)1 & 2 -impingement areaIndian Point Unit 2 Moderate Admiralty BrackishR.E. Ginna Unit 1 Minor Admiralty with stainless steel Fresh water (lake)impingement areaConnecticut Yankee Minor Admiralty and stainless steel Fresh water (river)in impingement areaH.B. Robinson 2 Minor Admiralty and stainless steel Fresh water (lake)in impingement area*See Glossary at end of report for definitions of terms.

Turkey Point Units 3 and 4 and Surry Units 1 and 2 began commercial opera-tion from mid-1972 to mid-1973. Like almost all units with U-tube designsteam generators, these units began operation using a sodium phosphatesecondary water chemistry treatment. This treatment was designed toremove precipitated or suspended solids by blowdown and was successful asa scale inhibitor. However, during early use, many PWR U-tubed steamgenerators with Inconel-600 tubing experienced stress corrosion cracking.The cracking was attributed to free caustic that can be formed when theNa/PO4 ratio exceeds the recommended limit of 2.6. In addition, some ofthe insoluble metallic phosphates, formed by the reaction of sodium phosphateswith the dissolved solids in the feedwater, were not adequately removed byblowdown. These precipitated phosphates tended to accumulate as sludge onthe tube sheet and tube supports at the central portion of the tube bundlewhere restricted water flow and high heat flux occur. Phosphate concentration(hideout) at crevices in areas of the steam generator, noted above, causedlocalized wastage resulting in thinning of the tube wall. The problem ofstress corrosion cracking was corrected by maintaining the Na/PO4 ratiobetween 2.6 and 2.3. Although the recommended Na/PO4 ratio was maintainedin some units, it did not correct the phosphate hideout problem thatcaused wastage of the Inconel-600. Largely to correct the wastage andcaustic stress corrosion cracking encountered with the phosphate treatment,most PWRs with a U-tube designed steam generator using a phosphate treatmentfor the secondary coolant have converted to an all-volatile chemistry.Surry Units 1 and 2 and Turkey Point Units 3 and 4 converted to AVT inmiddle to late 1974.In 1975, deformation, or the so-called "denting," of steam generator tubesoccurred in several PWR facilities, including Surry Units 1 and 2 andTurkey Point Units 3 and 4, after 4 to 14 months operation. This occurredafter the conversion from a sodium phosphate treatment to an AVT chemistryfor the steam generator secondary coolant. Tube denting occurs predominantlyin rigid regions, or so-called "hard spots," in the tube support plates.These hard spots are located in the tube lanes between the six rectangularflow slots in the support plates near the center of the tube bundle andaround the peripheral locations of the support plate where the plate iswedged to the wrapper and shell. The hard-spot areas do not contain thearray of water circulation holes found elsewhere in the support plates.The Surry Units 1 and 2 have experienced severe denting and support platedeformation throughout their steam generators. On September 15, 1976,during normal operation, one U-tube in steam generator "A" at Surry Unit 2suddenly developed a primary-to-secondary leak of about 80 gpm. Subsequent.investigations revealed that the leak resulted from an axial crack, approxi-mately 4-1/4 inches in length, in the U-bend apex of an inner-row tube.It was also established that the crack initiated from the primary side ofthe tubing. Hourglassing of the flow slots in the upper tube supportplate "pulled the legs" of the U-bend closer together thereby causing-24 -

higher stresses in the tube material in the "U" area, which resulted instress corrosion cracking. This ovalization phenomenon is shown inFigure 6. As a result of the event, the innermost row of tubes wasremoved from service by plugging. This action was taken in all SurryUnit 2 steam generators that exhibited a large degree of hourglassing inthe upper support plate flow slots. The potential for dent-relatedcracking (at U-bends and support plates) has necessitated the preventiveplugging of over 20 percent of the tubes in the Surry Units 1 and 2 steamgenerators (as of this report). The approach of removing the innermostrow of tubes from service by plugging because of hourglassing has alsobeen used at other facilities.The Turkey Point Units 3 and 4 have also experienced severe denting-relatedphenomena throughout their steam generators and have plugged approximately12 and 17 percent respectively of their tubes as part of their preventiveplugging programs (as of this report). At San Onofre Unit 1, only two ofthe three steam generators have been affected to date and only in the hotleg of the lowest two (of a total of four) support plates. No significantdenting has been detected in the third steam generator. In addition, itappears that the denting phenomena have slowed down in the San Onofrefacility. Other Westinghouse units that have observed various levels ofdenting include R. E. Ginna, Indian Point Units 2 and 3, Point BeachUnits 1 and 2, Haddam Neck, and H. B. Robinson Unit 2.3.1.3 Tube Support Plate CrackingExcessive deformation as a result of continued magnetite growth hasresulted in cracking of the tube support plates at Surry Units 1 and 2,Turkey Point Unit 4, San Onofre Unit 1, and Indian Point Unit 2.At Surry Units 1 and 2 and Turkey Point Unit 4, portions of the supportplate have moved, along with tubes, into the flow slots. This phenomenontermed islanding (see Sections 2.3 and 3.1.2) could potentially lead tothe possible loss of lateral support of some inner-row tubes. Concern forthis lack of support was eliminated by preventive plugging of many of thetubes in the inner rows.Support plate cracking has also been observed at San Onofre Unit 1.Deformation, hourglassing, and cracking was found in the bottom two tubesupport plates in steam generators "A" and "C" by inspections. During theApril 1978 inspection, no evidence of deformation, hourglassing, or crackingexisted in the upper two support plates in steam generators "A" and "C" orany of the four support plates in steam generator "B". Physical measurementso'f the three flow slots nearest the upper-hand hole entry were made insteam generator "C". These measurements indicated that the flow slots inthe top support plate did not deviate from their manufactured condition.-25 -

Measurements of the flow slots in the lower support plates in steamgenerators "A" and "C" were not made. Results of the latest steam generatorinspection, completed in October 1978, indicated that tube denting andsupport plate cracking may have stabilized in the San Onofre Unit 1 steamgenerator.During a recent inspection at Indian Point Unit 2, cracks were found inthe third flow slot from the manway side in the second tube support platein one steam generator. The cracks are located in the ligaments betweenthe flow slot and first-row tube holes near the center of the flow slot.Figure 5 is a photograph of such support plate cracking. In addition, itwas discovered that a tube support plate is in contact with the wrapper inthat steam generator. During the inspection, the licensee also removed asection of the lowest tube support plate from another steam generator.The sample was removed as part of a chemical cleaning feasibility study atIndian Point Unit 2. The support plate section contained the first tworows of tubes in columns 3 through 13 occupying an area approximately 14by 5 inches. The section was cut out using electrical discharge machining(EDM) and was removed through a 6-inch-diameter hand hole below the lowestsupport plates. While the specimen was being removed from the steamgenerator, parts of the support plate and 6 of the 22 tubes broke loosefrom the specimen. Measurements of the flow hole elongation between thefirst and second rows and the second and third rows account for virtuallyall the hot leg flow slot hourglassing.3.1.4 Anti-Vibration Bar Wear or FrettingA few plants originally equipped with round, carbon steel anti-vibrationbars (AVBs) have experienced varying degrees of tube wear at the AVBlocations. The cause appears to be primarily mechanical and is dependenton the material (that is, carbon steel bars), the shape of the bars, theclearances, and the bar support design.The most severely affected plant is San Onofre Unit 1 where, between April1975 and October 1976, a large number of tubes exhibited a substantialincrease in wear rate. An additional AVB array using square, chromium-plated Inconel bars on unworn surfaces near the original set of AVBs wasinstalled in all three steam generators. Severely degraded tubes wereremoved from service by plugging.Connecticut Yankee, and a foreign plant with similar AVB design, have alsoexperienced tube wear. However, the extent of the wear has been small andonly a few tubes have been removed from service at each of these facilities.3.2 Plants Designed by Combustion EngineeringAll commercial operating Combustion Engineering (CE) steam generators areof the vertical shell type with recirculating Inconel-600 U-bend tubingand integral steam separation equipment (see Figure 14). The CE design-26 -

1IN>\NFIGURE 14. A TYPICAL COMBUSTION ENGINEERING STEAMGENERATOR

has a combination of drilled carbon steel partial support plates similarto the Westinghouse design but without flow holes (Figure 10) and carbonsteel "egg crates" (Figure 11) for tube supports. With the exception ofthe Palisades Plant, the drilled plates are two partial plates locatednear the top of the tube bundle.The methods used for secondary steam generator water treatment for CE-designedoperating plants are shown in Figure 15. Palisades began operation with aphosphate treatment for the secondary water and converted to AVT to arrestthe tube wastage problem encountered with phosphate. All other facilitiesutilized AVT chemistry for the secondary coolant from the beginning ofoperation. (Fort Calhoun operated a short time on phosphate control priorto commercial operation.)3.2.1 Caustic Stress Corrosion and WastagePalisades conversion from phosphate to AVT secondary water chemistrycontrol to correct the tube thinning or wastage problem seems to haveeradicated the major occurrences of wastage-type tube degradation.Table 1 summarizes the present status of wastage experienced with CE steamgenerators.3.2.2 DentingFour Combustion Engineering plants are now known to have experienced somedegree of "denting" (Table 4). Palisades, which uses water from Lake Michiganin a closed-cycle cooling system with cooling towers, used phosphate treatmentprior to AVT. Operation at the other three plants has been totally onAVT. These plants include Maine Yankee, Millstone Unit 2, and St. Lucie.Most of the tube supports in these units are of the egg-crate design, withonly two partial support plates of the drilled-hole design in each steamgenerator. The Maine Yankee plant has 90-10 CuNi condenser tubes, whereasMillstone Unit 2 has Al-brass. The main condenser at Millstone Unit 2has experienced operational difficulties and has been retubed. The dentingat Maine Yankee is limited to the drilled-hole partial support plates. Inaddition, inspection at Millstone in late 1977 showed some dent-likesignals within the egg-crate region. It is unclear what phenomenon isoccurring at these locations.The Maine Yankee plant became operational in October 1972. In July 1973,steam generator 3 was inspected and only a few tubes in the partiallydrilled support plate were examined at that time with no indication oftube degradation. In July 1974, steam generators 1 and 2 were inspected,with the same negative result. In June 1975, steam generator 3 was inspectedfor the second time. No degradation of any kind was seen, although only afew tubes in the partial support plate were examined. At that time,sludge was present on the tube sheet, up to 3-1/2 inches deep. In May-28 -

1972 1973 1974 1975 1976 1977 19781979----- --------- -t --- -t -t -Vls < os s~r/ --- ----------------- i-, ,-,-,V- ----- ----1 7Z/4, ... ...... ................. ......... .i-.-?-1-T--T77 1 /AV7/ZPalisadesMaine YankeeFort CalhounCalvert Cliffs 1Millstone 2St. Lucie 1Calvert Cliffs 2Arkansas 2----, l--I/L~L 1~//,//.--- -, .-. -, -ILegend: .P04l AVTFIGURE 15. SUMMARY OF SECONDARY WATER CHEMISTRY TREATMENTIN OPERATING COMBUSTION ENGINEERING PLANTS ---S

TABLE 4DENTING IN CE STEAM GENERATORSEXTENT OFDENTING*PLANTCONDENSER TUBE MATERIALrnni Mr. WATPQPalisades Minor Admiralty changed to Closed-cycle cooling90/10Cu-Ni (lake water)Maine Yankee Minor Al-brass and some Brackish water70/30 CuNiSt. Lucie 1 Minor Al-brass Seawater(70/30 Cu-Ni in Air RemovalSection)Millstone 2 Moderate Al-brass changed to Seawater90/10 Cu-Ni*See Glossary at end of report for definitions of terms.

1977, steam generators 1 and 2 were reexamined, with relatively shallowdents found in the drilled partial support plates, averaging about 1 mil.Sludge on the tube sheet had reached a maximum depth of 4 to 4-1/2 inchesover a few tubes with a little more occurring in the hot leg than in thecold leg, but no sludge was found on the drilled support plates. Thepartial drilled-hole support plates are at the seventh and ninth supportlevels with a total of about 30 percent of the tubes passing through oneor the other or both of them. Radial clearance in the drilled holes is8 mils by design.In addition, chloride (Cl-) and cation conductivity have been monitored atthe condensate at Maine Yankee and Cl- is checked daily in the steamgenerators. Minor condenser leaks have occurred about once a month. WhenCl- has exceeded 0.05 ppm in the condensate, power has been reduced, andthe leak has been located and fixed in 1 to 1-1/2 days. Therefore, secondarywater quality has generally been controlled and maintained, although 1 ppmCl- in the blowdown was exceeded about 60 times during the operatinglifetime, and, on a few occasions early in operation, rose to some tens ofppm's for short periods. On the other hand, the experience at Maine Yankeeis believed significant because denting occurred in spite of significantefforts to keep the secondary water relatively free of impurities. However,denting at Maine Yankee after 53 months of operation is less than atMillstone Unit 2 after 23 months of operation where condenser problems aremore severe.At Millstone Unit 2, the initial denting in the drilled plates was moderateand widespread, measuring 7 to 13 mils, but no leaks were found in thesteam generator. The plant has had a relatively high level of Cl- in thesecondary water, because of almost continuous low-level inleakage throughthe condensers. The Millstone Unit 2 plant became operational in August1975. The aluminum-brass (Al-brass) main condenser has since been retubed,because of condenser problems. A minimum of 0.15 ppm Cl- has been reported,with a typical blowdown analysis in 1976 of Na = 0.2 ppm, Cl- = 0.6 ppm,and conductivity = 6 pmho/cm. Contaminants were only a little less in thefirst part of 1977. Since startup, Cl- has exceeded 1 ppm on 76 days.Approximately 2,200 tubes in Millstone Unit 2, which pass through the twotop support plates, were eddy current tested (ECT) in each steam generatorwith a 0.540-inch-diameter probe at 400 Hz. Essentially all the tubesinspected in both steam generators had dent indications at each of thedrilled support plates. The majority of tubes not passing the 0.540-inchECT probe are in regions of the support plate periphery adjacent to thelugs supporting the plates. Dent indications (1.0 mil) at tube/egg-crateintersections were observed in approximately 70 percent of the tubesinspected in both steam generators.-31 -

St. Lucie Unit 1 has conducted steam generator inspections during theirfirst refueling outage. The unit, which began commercial operation inDecember 1976, has CE steam generators similar in design to MillstoneUnit 2 and has operated continually with AVT secondary water chemistry.The inspection program to date included 380 hot leg tubes and 110 cold legtubes in one steam generator. Preliminary results indicated 55 tubes withdent signals after one fuel cycle of operation. The average dent magnitudeis in the range of 1 to 2 mils with the maximum dent magnitude observed being4 mils.3.2.3 Tube Support Plate CrackingAt Millstone Unit 2, tube denting has caused the two top partial supportplates in both steam generators to expand against the "hard spots" atsupporting lugs on the tube bundle shroud. The stresses induced by theexpanding support plates has caused cracking of the ligaments between thetube holes and circulation flow holes in corner areas of the uppermostsupport plate along the outer band of tubes adjacent to the rim of solidmetal at the outer periphery of the plates. Shear stresses have causedcracking along the inner boundary of the solid rim section and shifting atthe corner areas of the plate. This produced a shearing action on thetubes and deformed the tube wall of about 20 outer peripheral tubes locatedin the corner areas of the plate.The tube support plate ligament cracks observed at Millstone Unit 2 resultedfrom high support plate strains. These strains are the result of corrosionproduct growth in the annular clearance between the tubes and tube holeswithin the plate. The aggregate tube ligament strains are relieved byplate expansion within certain limits. At Millstone Unit 2, plate expan-sion was constrained by solid unyielding attachments or wedges between theplate and steam generator tube bundle shroud. As a result, the ligamentsand tubes became deformed and the plate cracked to relieve the stress.-32 -

4. CORRECTIVE ACTIONS AND REPAIRS4.1 Short-Term Program and Licensing RequirementsIncreased steam generator tube inservice inspection (ISI) frequency,preventive tube plugging, and more stringent technical specificationslimiting operation with steam generator tube leakage have formed the basesfor continued safe operation of degraded steam generators. Other require-ments, such as reduced primary coolant radioactivity limits, have alsobeen required for severely degraded steam generator plants.The Standard Technical Specifications for Westinghouse and CE facilitiesrequire a steam generator ISI every 12 to 24 calendar months unless twoconsecutive ISIs indicate good condition of the steam generators, in whichcase the interval may be increased to 40 months. In the event that steamgenerator degradation is observed, the inspection frequency must return tothe original 12- to 24-month schedule. In the event of severe steamgenerator degradation, the NRC has required plants to perform ISI morefrequently.Tube-plugging criteria addressing wastage types of degradation areroutinely included in plant technical specifications. These pluggingcriteria are based on the guidelines in Regulatory Guide 1.121 and aredesigned to ensure the integrity of degraded tubes during normal oraccident conditions. Plugging criteria for tube denting is not specifi-cally addressed in the current revision of Regulatory Guide 1.121 and ismore difficult to establish. Dented tubes are susceptible to stresscorrosion cracking (SCC), which is dependent on stress level, time, andenvironment. Tests have shown that dented tubes with small through-wallcracks near the support plate have adequate margins against tube burst orcollapse under normal operation, transients, and postulated accidents.Severe SCC could, however, reduce the margins to an unacceptable level.Therefore, tube-plugging criteria for SCC of severely dented tubes must bebased on the magnitude of denting (stress level), operating time, and therate of degradation. The objective of the tube-plugging criteria is toremove from service any tubes that might develop through-wall cracks orbecome severely degraded before the next ISI. (The present extent of thetube plugging for Westinghouse and CE plants is summarized in Table 2.)The continued integrity of the steam generator tubing is further monitoredby a reactor primary coolant-to-secondary system leak rate limit in planttechnical specifications. The technical specification leak rate limitcorresponds to a leak rate from a sufficiently small defect so that itwill not result in a sudden tube rupture under design basis accidents.For units with severely degraded steam generators, this limit has beenreduced to as low as 0.3 gpm. Following an occurrence of leakage of0.3 gpm or greater at such facilities, the facility is required to shut-33 -

down, to plug the leaking tube, often to conduct a steam generator tubeinspection, and occasionally to request NRC approval prior to restart.This approach helps to ensure that severely degraded tubes (that is, evento the point of leakage) are removed from service and that other tubes areinspected to detect any other continued degradation.4.1.1 Turkey Point Units 3 and 4 and Surry Units 1 and 2Operation of the Turkey Point Units 3 and 4 and Surry Units 1 and 2 isbeing closely regulated by the NRC. These units have been conductinginspections about every six months in order to carefully monitor the rateof steam generator tube degradation. Each inservice inspection includeseddy current inspections, tube gauging, and support plate examinations.The numbers and locations of tubes to be gauged are established using afinite element computer model of the tube support plates. As a result ofeach inspecton, tubes that may be susceptible to SCC and that may begin toleak prior to the next ISI are plugged. The typical plugging criteriaoutlined below are based on operational experience and are essentially thesame for these four units.1. All tubes that do not pass the 0.540-inch ECT probe will beplugged (nominal inside diameter of virgin tube is 0.770 inch).2. Additionally, if attempting to justify operation for six monthsor longer on a severely degraded facility, two tubes beyond(that is, higher row numbers) any tube in columns 15-79 thatdoes not pass the 0.540-inch probe will be plugged in thetube-lane region; for such tubes in column 1-14 and 80-94, fivetubes beyond will be plugged on the hot leg side and four tubesbeyond will be plugged on the cold leg side in the tube-laneregion (see Figure 16 for column and row numbering).3. All tubes that do not pass the 0.610-inch probe will be plugged.No surrounding tubes are plugged by this step.4. Those tubes in any column, for which plugging under criteria 1,2, or 3 above is-implemented in the tube-lane region, will alsobe plugged in the lower-numbered row of tubes back to the tubelane if not already plugged.5. As a conservative measure, all tubes immediately surrounding anyknown leaky tubes, including the diagonally adjacent tubes, willbe plugged if they are not already covered by the foregoingcriteria.6. In any given column that is surrounded by columns containingtubes with significant tube restriction or prior plugging-34 -

l93 91 19 87 35 83 31 79 77 75 73 71 6 67 65 63 61 59 57 56 53 51 49 47 45 43 41 39 37 35 33 31 29 27 25 23 21 19 17 15 1311 994 9,21"18818lgcl841.,211017gl76l74I721 7,lk8ll~ll4l,2llSl~sll66l64l5,2 l5@4814614,14l4214013813C13l413l2130 2l111Cl12.412l12.o111l11 114 12111 lISI3 1.1 1COLUMNS-46-4.44-43-42-41-40 39-37-36-34 332-26 27-25-24__ _ _ 23-22__ __ 21-20-19-17-16-is-14 13-122 11-10-a-7-6-4-2 3ROWS;, L I1t1 1 R, 1 I............. i1 i i~u l tm w 7tt l T ! I _ LL I I f1j L L_ _,_,1, I I__ _ __NWAY NOZZhllU f ULTLWJ jLLE U~~IVINWAYNOZZLE-~-FIGURE 16. CROSS SECTION OF STEAM GENERATOR TUBE ARRAY

(thereby creating a "plugging valley" in the pattern), engineeringjudgment will be used to fill the bottom of the valley. In theperipheral tube-lane areas near the three and nine o'clockwedges, tubes surrounded by previously plugged tubes or tubesexhibiting high deformation activity will be plugged based onengineering judgment.7. Additional preventive plugging will be implemented at the hotleg wedge locations. This plugging will include all tubes that:a.b.C.Restrict the 0.610-inch probe, orRestrict the 0.650-inch probe at the periphery, orSurround leakers and tubes that restrict the 0.540-inchprobe including the diagonally adjacent tubes.8. Application of the criteria specified in 7, above, willon the basis of engineering judgment for cold leg wedgebe madelocations.9. Additional preventive plugging will be implemented in the patch-plate region. This plugging will include all tubes that:a. Restrict the 0.610-inch probe, orb. Surround leakers and tubes that restrict the 0.540-inchprobe including the diagonally next tube, orc. Lie on either side of the patch plate boundary (plateperimeter on one side and plug welds on the other three)and restrict the 0.650-inch probe.The above criteria indicate -hard spot" areas of the tube support plateswhere the tubes are more susceptible to denting. The exact column or rownumbers to bound regions for tube plugging depend on the lateral supportarrangement of the support plates.In addition to the aforementioned conservative criteria for tube pluggingin plants with severely degraded steam generator tubing, several otherrequirements are generally in place at these facilities. Typical otherrequirements include the following:1. A technical specification requiring plant shutdown if a leakingdent exceeds 0.3 gpm in a steam generator. This requirement isintended to require plugging of the leaking tube and usuallyrequires additional tube ISI.2. A technical specification requiring plant shutdown foradditional tube inspection if any two separate dented tubes are-36 -

found to leak in any 20-day period regardless of the leakagelevel of each tube. This requirement is intended to require aninspection to explore the rate of degradation in steam generatorsbecause, with the conservative plugging pattern, it is notexpected to have two leakages in such a short period of time.3. More restrictive limits have been incorporated in technicalspecifications to limit normal operation radioactivity levels inthe primary coolant. This requirement is part of our defense-in-depth approach and is meant to limit the consequence of a hypo-thetical major loss-of-coolant accident if one postulated thatmajor steam generator tube leakages occurred simultaneously withthe accident.4.1.2 Indian Point Unit 2Indian Point Unit 2 is one of the six PWR facilities that were initiallyidentified to have suffered steam generator tube denting and that havebeen under close monitoring since the latter part of 1976. Steam generatorinspection conducted in February 1978 revealed minor but somewhat progres-sive denting. The 1978 inspection also revealed two cracks at a flow slotin the second support plate of one steam generator. The cracks had notbeen previously observed. In addition, it was observed that a tube supportplate was in contact with the wrapper in the same generator. During theinspection, a section of the lowest tube support plate was removed as apart of a chemical cleaning feasibility study. While the specimen wasbeing removed from the steam generator, parts of the support plate brokeloose from the specimen. The inspection results indicated that activecorrosion of the carbon steel support plate was continuing, but at aslower rate in comparison with other units.The plugging criterion that was implemented consisted of plugging of anytube that would not pass a 610 mil or smaller ECT probe. In addition, thereactor coolant-to-secondary leakage limit was reduced to 0.3 gpm and theaforementioned "two leakages in 20 days" requirement was imposed. Withthese corrective actions and licensing conditions, 16 equivalent full-powermonths of operation were justified for Indian Point Unit 2.4.1.3 San Onofre Unit 1Photographs and videotapes taken in September 1977 and from inspectionsprior to that time showed cracking at the edge of the flow slots in thebottom two support plates of two of the three steam generators at SanOnofre Unit 1. A steam generator ISI was therefore conducted during April1978 to determine if the degradation was progressing. The results of thisinspection established that no perceptible progression of denting orchange in support plate condition had occurred between October 1976 andApril 1978. Nine tubes that would not pass a 0.460-inch probe in April-37 -

1978 were plugged. Because the San Onofre steam generator tubes have athicker wall and smaller diameter, this degree of denting corresponds tothe same tube wall hoop strain as a Surry or Turkey Point steam generatortube dented to about a 0.500-inch inside diameter. The 0.3 gpm technicalspecification leak rate limit was also imposed. San Onofre Unit 1 steamgenerators were again inspected (ISI) during the scheduled September 1978refueling. The results of this inspection did not alter the conclusionsreached following the April 1978 inspection.4.1.4 Millstone Unit 2Millstone Unit 2, which has the most severe denting problem of all the CEplants, has performed extensive repairs to minimize the progression ofsupport plate cracking and shifting and further tube damage. Approxi-mately 80 percent of the plate constraint is attributable to the lugssupporting the plates. Analyses have shown that compressive and shearstresses associated with plate constraint would cause further cracking andshifting of the partial support plate. This condition would cause addi-tional deformation of the peripheral tubes. The results of finite elementanalysis indicated that stresses in the plate adjacent to the rim would bereduced by removing the lugs and a portion of the peripheral plate rimadjacent to the tube bundle shroud. Therefore, the following modificationswere made at the Millstone Unit 2 facility:1. Removal of all lugs at each support plate and a portion of theperipheral solid rim in the uppermost plate to reduce "hardspots" and minimize the possibility of further cracking andshifting of the plates in each steam generator.2. Preventive plugging of all peripheral tubes adjacent to thesolid rim that have the greatest potential for failure, includ-ing additional tubes near the periphery in the corner regions ofboth support plates.3. Plugging of all tubes not passing the 0.540-inch ECT probe andthose surrounding the restricted tube.4. Avoiding and minimizing unfavorable chemistry conditions.5. Exclusion of seawater ingress by means of assuring condensertube integrity (i.e., retubing the condenser with 90-10 CuNi)and a full-flow condensate polishing system, which will beavailable during cycle 2 operation.In addition to the above tube-plugging pattern in items 2 and 3, thefollowing preventive plugging was performed based on the critical tubehoop strain predicted by the finite element analysis of the tube supportplate:-38 -

1. Any tubes that were damaged during the rim and support lugremoval operation were plugged.2. All tubes that lie along an apparent continuous series of liga-ment cracks in the plates were plugged.3. All tubes not passing the 0.540-inch ECT probe and those surround-ing the restricted tube were plugged.Implementation of the plugging criteria resulted in plugging 290 tubes insteam generator 1 and 352 tubes in steam generator 2.4.1.5 Other CE FacilitiesMaine Yankee and Arkansas Unit 2 are other CE facilities that have removedall lugs at each drilled support plate and a portion of the peripheralsolid rim in the uppermost plate to reduce "hard spots" and minimize thepotential for cracking and shifting of the plates in each steam generator.At Arkansas Unit 2, these modifications were performed prior to initialstartup. Other CE units, Calvert Cliffs Units 1 and 2 and Fort CalhounUnit 1, have not experienced any form of tube degradation with an AVTchemistry. Recent inspections of St. Lucie Unit 1 steam generatorsrevealed a buildup of corrosion products in the annulus between the tubeand tube support plate, creating a potential for future denting. A proposalto chemically clean the St. Lucie Unit 1 steam generators to remove thesecorrosion products is being reviewed by the NRC.4.2 Long-Term Repairs4.2.1 Tube SleevingCombustion Engineering presently has under way a program to demonstratethe feasibility of installing sleeves as an alternate measure to tubeplugging at the Palisades facility. The operation of the Palisades steamgenerators has, in the past, resulted in localized corrosive attack on theoutside (secondary side) of the steam generator tubing. The reduction insteam generator tube wall thickness due to this corrosive attack mayprogress to the point of causing tubes to leak during operation. Inaddition, reduction in tube wall thickness may lessen the ability of thetube to continue to perform its function as a primary coolant pressureboundary during design accident conditions such as a loss-of-coolantaccident (LOCA) or a main steam line break (MSLB).Historically, the corrective action taken where steam generator tube walldegradation has been identified was to install welded plugs at the inletand outlet of the steam generator tube when the reduction in wall thick-ness exceeded the plugging limit. This value of wall reduction requiring-39 -

plugging was calculated such that adequate tube strength remained toprevent failures of the steam generator tubes during normal operation andpostulated accident conditions.Installation of tube plugs in a steam generator tube removes the heattransfer surface of the tube from service. The technique for installationof steam generator sleeves eliminates this negative aspect of steamgenerator tube plug installation. The sleeves are installed at the localarea of tube wall reduction and impose only a minor restriction to primarycoolant flow. Thus, while providing a corrective response to the weakeningeffect of tube wall reduction, the effects on heat transfer and primarycoolant flow are minimized.The steam generator sleeving concept consists of installing, inside thesteam generator tube, a smaller diameter Inconel-600 tube to span thedegraded area of the parent steam generator tube. This system is shown inFigure 17. Both ends of the inserted sleeve are hydraulically expandedinto an interference fit with the parent tube. The rationale forinstalling the sleeves in this manner is to restore the mechanicalstrength of the degraded tube to a level adequate to prevent ruptureduring postulated accident conditions. By installing a sleeve to span thedegraded area, the structural integrity of the tube is reestablished.Although the sleeving process has been used only on a limited scale at thePalisades steam generators, the sleeving process may be applicable to allPWR generators. To qualify the sleeve for other applications, specificsizing and environmental conditions would have to be examined to ensureapplicability.4.2.2 Steam Generator RepairExtensive preventive plugging as a result of continuing tube denting cancause excessive steam generator inspections and reduction in unit avail-ability. For these reasons, Florida Power and Light Company (FPL) andVirginia Electric Power Company (VEPCO) are planning replacement of thelower portion of the steam generators at Turkey Point and Surry, respec-tively.FPL and VEPCO are currently pursuing engineering and licensing activitiesto effect these steam generator repairs. Repair of the first such rteamgenerator is tentatively scheduled to begin in early 1979. The existingsteam generators are expected to be cut apart at the transition piece tothe upper section of the shell. The upper section of the steam generatorwill then be stored inside the containment and joined to the new lowersteam generator assembly, which will include the new tube bundle. Thelower assemblies, including the old tube bundles, will exit the contain-ment via the equipment hatch.-40 -

--x 0.032-RADIALDEFORMATION0.010" TYP.12"DEGRADATIONSUPPORT/ STRUCTUREINTERFERENCEMECHANICALJOINTSTEAMGENERATOR -TUBE3/4" OD x 0.048WALLt1" TYP.1/4" TYP.-J.FIGURE 17. STEAM GENERATOR TUBE SLEEVE

Several secondary side design changes will likely be made in the repairedsteam generators that will enhance their resistance to the previouslydiscussed forms of degradation. A flow distribution baffle will be incor-porated to improve flow velocities and circulation across the tube sheet.By directing flow to the blowdown pipe location, the effectiveness ofsludge removal by blowdown would be increased. A new tube support platedesign using a four-lobed broach tube hole, called a "Quatrefoil," willmaximize flow along the tubes and reduce the susceptibility to corrosionproduct buildup and tube denting. The new plates are also fabricated fromcorrosion-resistant Type 405 ferritic stainless steel with Inconel-600tubing being heat-treated to increase its resistance to stress corrosioncracking.4.2.3 Condenser IntegrityRecent experience indicates that ingress of chlorides through condenserleaks is a principal contributor to the denting problems. Elimination ofcondenser leaks is, therefore, a primary concern in ensuring steam generatorintegrity. Improved integrity of rolled joints and selection of better*materials will increase condenser reliability and reduce contaminant inputto the steam generators. In particular, the use of titanium tubing inseawater cooled condensers and stainless steel tubing in fresh watercooled condensers offers improved corrosion resistance, enhances condenserintegrity, and reduces the source of copper to the condensate. The use ofcopper-based alloys in condenser tubing is being approached with extremecaution because the presence of soluble copper and/or nickel may promotethe chemical reaction that causes denting. Of the copper-based alloys,CuNi alloys offer improved corrosion resistance over brasses or bronzescommonly used in older units.The same concern for reducing corrosion and the source of copper ions incondensers applies to the feedwater heaters and moisture separator reheaters(MSR). Westinghouse suggests the use of 300 series stainless steel forthe feedwater heaters and either carbon steel or carefully chosen types offerritic stainless steel for the MSR. The use of 90-10 CuNi alloy isacceptable in the low-oxygen environment of the MSR because significantcopper pickup is not expected.4.2.4 Condensate PolishersAcceptable secondary coolant chemistries have been maintained both withand without the use of condensate polishing. The levels of sludge accumu-lation of the tube sheets are comparable in plants with or without polishers;however, there has been no evidence of caustic-induced corrosion in anyplant equipped with condensate polishing. Condensate polishing can be anasset in maintaining a contaminant-free secondary coolant chemistry.-42 -

T44.2.5 Steam Generator Tube RepairAnother option presently being examined is a method of removing only steamgenerator tubes as opposed to replacing the entire lower section (shell,tube sheet, and tube bundle) from degraded steam generators and replacingthem with new tubes. This option would reuse the existing steam generatorupper and lower vessel shell and the steam generator tube sheet.-43 -

5. RELATED RESEARCH PROGRAMS5.1 Westinghouse Electric CorporationResearch into the causes of denting include operational testing, destruc-tive analysis of tubing and support plate samples, and laboratory experi-mentation. The initial information on denting phenomenon was derived fromexamination of tube/support plate samples that revealed thick oxide buildup,tube diameter reduction, and chemical makeup of the crevice-filling materials.Only minor corrosive attack on the tube material was observed. The crevicecontained a thick layer of almost pure magnetite (Fe3O4).Westinghouse conducted a series of tests on the crevices with contaminantsand have been able to produce denting in the laboratoy. Denting hassubsequently been reproduced in model boilers equipped with plant-typegeometrical configurations.The presence of chloride has been found to be a common factor in reproduc-ing denting. Nickel chloride solutions and ferrous or cupric chlorideshave produced measureable denting. Thus far, test data indicate thatcertain substances, e.g., phosphates, calcium hydroxide, zinc oxide, andborates, seem to retard the denting process. Morpholine, an AVT additive,has shown a beneficial effect by reducing the corrosion rate of carbonsteel.Westinghouse has also found a correlation between net hydrogen (H2)generation in the steam generator and the existence of denting. Testingthe effects of lithium borate and boric acid additions to the steam generatorshas been combined with a program in which the H2 produced by corrosion hasbeen monitored. Some reduction in the H2 generation rate has been seenwith boric acid injection. Tests to verify and quantify the borate andboric acid effects are in progress.5.2 Combustion EngineeringModel boiler tests have also been used by Combustion Engineering to simulatethe denting mechanism as parts of the EPRI overall steam generator program.The findings are basically in agreement with those reported by Westinghouseand both experiments correlate with Potter and Mann's observations* ofmagnetite formation in the crevices of carbon steel in high-temperaturepure water containing iron or nickel or copper chloride salts. TheCombustion Engineering experiments indicate that denting could occur dueto the corrosion of copper-based condenser tubes and condenser inleakageE. C. Potter and G. M. W. Mann, "The Fast Linear Growth of Magnetite onMild Steel in High Temperature Aqueous Conditions," p.26, BritishCorrosion Journal, Vol. 1 (1965).-44 -

of cooling water containing chloride ions. The type of copper-basedtubing would be important in the accumulation of CuO2 in the secondaryside of the steam generator; i.e., copper-nickel alloys being morecorrosion resistant. The CuO2 "builds up" in the crevices between steamgenerator tubes and tube support plates and acts as a concentrator forchloride ions. As a result of acid chloride concentration, the carbonsteel support plate corrosion is accelerated with corrosion productbuildup causing denting.The Combustion Engineering tests also indicate that the phosphates to AVTtransition was not necessary to initiate the denting process, even beforeplants with pure, uncontaminated AVT environments were found to havesuffered denting. In additon to reproducing the denting phenomenon, theCombustion Engineering tests are also being used to demonstrate chemicalcleaning as a means to remove corrosion products to arrest the dentingprocess.5.3 NRC-Funded Research ProgramsThe NRC is funding two research programs related to steam generatorproblems. The first program is investigating stress corrosion cracking ofPWR steam generator tubing. The objectives of this program are (1) todevelop data that will enable prediction of stress corrosion cracking ofInconel-600 in terms of factors such as temperature, stress, materialhistory, and environment, and (2) to evaluate the aqueous environment andmetallurgical structures under which stress corrosion cracking occurs.The second program is directed toward an investigation of steam generatortube integrity. The objective of this program is to develop a validatedmodel, based upon experimental data, for prediction of margin-to-failureunder burst and collapse pressure of degraded steam generator tubing. Theexperimental work will be performed on designated steam generator tubespecimens from our operating PWR steam generator. It is expected thatthis pogram will provide the necessary corroboration for calculations thatindicate adequate safety margins exist under theoretical accidentconditions with degraded steam generator tubes.-45 -

6. CONCLUSIONSThe resolution of operational problems related to the PWR steam generatorsis a complex task that requires the joint efforts of the vendors andoperators of these plants and the NRC staff. In this regard, the staffformulated Task Action Plans (TAP) A-3 and A-4 (copies of which are enclosedin Appendix B) for plants designed by Westinghouse and Combustion Engineering,respectively, to organize and give priority to NRC staff efforts in thefinal resolution of problems related to the operation of PWR steam generators.6.1 Basis for Continued OperationFor PWRs with recirculation types of steam generators, the NRC staffconcluded that, pending completion of the Task Action Plans A-3 and A-4,continued operation does not constitute an undue risk to the health andsafety of the public for the following reasons:1. Primary-to-secondary leakage rate limits, and associated surveillancerequirements, have been established to assure that the occurrenceof tube cracking during operation will be detected and appropriatecorrective action will be taken before any individual crackbecomes unstable under normal operating, transient, or accidentconditions.2. Inservice inspection requirements and preventive tube pluggingcriteria have been established so that the great majority ofdegraded tubes will be identified and removed from servicebefore leakage develops.3. On a case-by-case basis, additional measures have been taken (a)to minimize contamination of the secondary coolant by inleakageof condenser cooling water (for example, condenser tubes withimproved corrosion resistance have been installed), and (b) tominimize buildup in the steam generators of corrosion productsgenerated in the secondary system (for example, feedwater heaterswith improved corrosion resistance have been installed). Controlsor monitoring of parameters that affect steam generator waterchemistry are being considered to provide additional assurancethat the potential for tube degradation during operation isminimized.4. Observed through-wall cracks at dented locations (that is, tube/support plate intersections) have been small and stable (norapid failures) during normal operation. In addition, becausesuch cracks are constrained by the support plates, they are notanticipated to become unstable (burst) during postulated accidents.-46 -

I5. Even if a LOCA or a MSLB were to occur during operation and sometubes were in a state of incipient failure, the radiologicalconsequences of such an event would not be severe.6. Continuous feedback from operating experience and the TAP effortswill be utilized to update interim criteria and requirements.6.2 Basis for Continued Operation of Plants With Severe DegradationFor plants experiencing severe degradation, the following additionalinterim bases were also considered:1. The probability of a design basis accident occurring duringnormal operation is small, and the probability that the accidentwould occur during the short period of time while the plant wasoperating with either a slightly leaking tube or a tube at thepoint nearing leakage is even smaller.2. Even if an accident occurs when there are cracked tubes, theconservatively calculated consequences are still acceptablysmall.3. A small amount of leakage (e.g., less than the technicalspecification limit) can be tolerated during normal operationwithout exceeding the offsite dosage limits of 10 CFR Part 20.6.3 Licensing of New PWR FacilitiesThe preceding rationale, which constitutes the basis for continued operationof licensed Westinghouse and Combustion Engineering PWR facilities, alsosupports continued licensing of new facilities. Furthermore, to theextent that is practicable, depending on the status of the design, fabrication,and installation of the steam generators for facilities not yet licensedfor operation, "state-of-the-art" design improvements and operating proceduresthat eliminate, or at least reduce, the potential for steam generator tubedegradation are required by the staff. The following design and operationalfactors are considered by the staff in the conduct of its reviews:1. Designs to provide improved circulation to eliminate low flowareas and to facilitate sludge removal.2. Designs to minimize flow-induced vibration and cavitation.3. Designs to provide increased flow around the tubes at the supportplates.-47 -

4. Selection of material for tube support plates that demonstratesimproved corrosion resistance.5. Material selection (composition), processing, and heat treatmentto minimize the susceptibility of tubes to stress corrosioncracking.6. Secondary water system chemistry control.7. Designs to allow for installation of an ion exchanger (condensatedemineralizer) in the secondary water system to minimize feedwatercontamination.In view of the above, the staff concluded that issuance of ConstructionPermits (CP) and Operating Licenses (OL), pending completion of genericstudies, can continue with reasonable assurance that operation will notpresent an undue risk to the health and safety of the public.-48 -

aAPPENDIX APWR DESIGN CONFIGURATIONNuclear power plants using the pressurized water reactor (PWR) designconcept contain three separate cooling-cycles. The three cooling cyclesof.a typical PWR are shown in Figure A-1. The first cooling cycle com-prises the primary coolant system that pumps pressurized coolant waterthrough the heat-generating core of the reactor where it picks up heat.The second cooling cycle consists of large heat exchangers called steamgenerators, a steam-driven turbine generator, a steam condenser, feedwaterpumps, and associated piping systems. Heat generated in the primarycoolant system is transferred to the secondary system through steam genera-tors. The water in the secondary coolant system boils in the steam generatorcreating steam that is used to drive the turbine generator. After itpasses through the turbine generator, the steam is condensed back intowater in the steam condenser. The secondary cooling water is returned tosteam generators by the feed pumps, thereby completing the cycle. Thethird cooling cycle is the condenser cooling water system that providescold water to condense the steam back to water in the steam condenser.The basis for this closed-cycle system is to ensure that the radioactiveprimary coolant water, the secondary cooling water, and the condensercooling water are separated from each other. The steam generator is theconnecting link between the radioactive primary and nonradioactive secondarycoolant system and is, therefore, a principal part of the reactor coolantpressure boundary. Figure A-2 shows the major components of the reactorcoolant system.Two major types of steam generators are currently in use in pressurizedwater reactors in the United States. These are the recirculating type,which is manufactured by Westinghouse and Combustion Engineering, and theonce-through type, which is manufactured by Babcock & Wilcox. Typicalrecirculating types of steam generators manufactured by Westinghouse andCombustion Engineering are thoroughly described in the main report. Inthe recirculating type of steam generator, hot coolant water from thereactor enters the steam generator through the primary coolant inletnozzle and flows into the inlet side of the steam generator lower plenum.The coolant water then flows through a large number of U-tubes to theoutlet side of the lower plenum where it exits the steam generator throughthe primary coolant outlet. A vertical divider plate separates the inletand outlet plenums. The secondary coolant water enters the steam generatorthrough the feedwater inlet and flows through the feedwater ring into anannulus area between the wrapper and the shell. It then flows to thebottom of the steam generator and up through the tube bundle region.-49 -

E I 1 l TURBINE.GENERATOR.w *v... .,CONDENSER.l COOLINGWATERFIGURE A-1. PRESSURIZED WATER REACTOR (PWR) COOLING CYCLESi, I I .STEAM OUTLET (TO TURBINE)STEAM OUTLET(TO TURBINE)FEEDWATER INLET(FROM CONDENSER)PRESSURIZE1FIGURE A-2. SCHEMATIC OF REACTOR COOLANT SYSTEM FOR PWR

Several thousand tubes (the number of tubes depends on the design) arewelded into and are vertically supported by the tube sheet that is 20inches thick (or more) and is perforated with thousands of holes for thesteam generator tubes. At various higher elevations, the tubes penetratethrough holes in tube support plates that provide lateral support and, inthe U-bend area, anti-vibration bars are sometimes laced through the tubesto minimize flow-induced vibrations. The steam generator tubes are approxi-mately 7/8 to 3/4 inch in outside diameter with a wall thickness of about0.050 inch. The tubes provide the heat transfer surfaces between theprimary coolant water and the secondary coolant water. The steam generatortubes constitute over 50 percent of the area of the total primary coolantsystem pressure-retaining boundary. Above the U-tubes, in the upperportion of the steam generator, there is a moisture-separating system forimproving the quality of the steam generated that is sent to the turbinegenerator.-52 -

IAPPENDIX BTAP A-3Westinghouse Steam Generator Tube IntegrityTAP A-4Combustion Engineering Steam Generator Tube Integrity(Data in this Appendix is taken from USNRC Report NUREG-0371,"Task Actin Plans for Generic Activities, Category A," November1978, pp. A-3/1 through A-3/11, pp. A-4/1 through A-4/10.)-53 -

I3'iITask A-3WESTINGHOUSE STEAMLead NRR Organization:Lead Supervisor:Task Manager:Applicability:Projected Completion Date:GENERATOR TUBE INTEGRITYDivision of Operating Reactors(DOR)Darrell G. Eisenhut, A/D forSystems and Projects, DORB.D. Liaw, EB/DORWestinghouse Pressurized Water ReactorsDecember 1979-55 -

Task A-3Rev. No. 1May 19781. DESCRIPTION OF PROBLEMPressurized water reactor steam generator tube integrity can bedegraded by corrosion induced wastage, cracking, reduction in thetube diameter (denting) and vibration induced fatigue cracks. Theprimary concern is the capability of degraded tubes to maintaintheir integrity during normal operation and under accident condi-tions (LOCA or a main steam line break) with adequate safety margins.Westinghouse steam generator tubes have suffered degradation due towastage ana stress corrosion cracking. Both types of degradationhave been nominally arrested; however, degradation due to dentingwhich leads to primary side stress corrosion cracks is the major .problem at present, and the principal focus of this technicalactivity.2. PLAN FOR PROBLEM RESOLUTIONThe major portion of the NRC staff efforts related to the resolutionof the denting problem will consist of evaluation of the results ofinvestigations by Westinghouse, EPRI, and EPRI supported contractors.In addition, NRC supported technical assistance and confirmatoryresearch programs will be used as the basis for evaluation ofapplicant supplied data.The specific activities directed at resolution of the denting prob-lem in Westinghouse steam generators consist of the following issuesand tasks:A. Generic Evaluation of ISI ResultsReview and evaluate the various eddy current inspection results;i.e., experience from operating reactors and evaluate thesedata as they relate to the generic determination of failureprobability of degraded tubes. In addition, evaluate the testprograms and analytical studies to provide staff understandingsufficient to continue to provide justification for continuedsafe operation of operating reactors.B. Evaluation of Transients and Postulated AccidentsEvaluation of failure consequences under postulated accidentconditions (LOCA and MSLB) to determine the acceptable levelsof primary to secondary leakage rates and the effect on ECCSperformance. The results will be used to define the acceptable-57 -

Task A-3Rev. No. 1May 1978number of tube failure that may be necessary as a licensingbasis considering predicted fuel behavior and radiologicaldose during transients and postulated accident conditions.C. Evaluation of Steam Generator Tube Structural IntegrityReview and evaluate te structural integrity of steam generatortubes under normal operating and postulated accident conditions(LOCA, SSE and MSLB) including licensee and Westinghouse anal-yses where appropriate to generic conclusions.D. Establish Tube Plugging CriteriaEstablish a generic tube plugging criteria that is consistentwith the determined allowable leak rate, tube structural integ-rity and degradation rates. These results will allow assess-ment of the adequacy of the requirements defined in RegulatoryGuide 1.121.E. Secondary Coolant Chemistry RequirementsEvaluate the mechanism of tube degradation. The results willbe used to define the requirements for secondary coolantchemistry control including considerations for condenserin-leakage.F. Evaluation of ISI MethodsReview the development of improved eddy-current probes, coilsand multi-frequency techniques to better quantify dents andgrowth of dents and increase sensitivity of detecting cracksin dented regions.G. Establish Criteria for Revision of Regulatory Guide 1.83Integrate experience from inservice inspection results, theresults from the evaluation of various ISI improvements andthe plugging and secondary water chemistry requirements intocriterion for possible revision of Regulatory Guide 1.83.H. Steam Generator Replacement (Prototype)Review and evaluate plans for initial steam generator replace-ment as generic basis for subsequent replacement actions.-58 -

Task A-3Rev. No. 1May 1978I. Review Design Criteria for Plants Not Yet LicensedReview and evaluate design modifications proposed by applicantsand Westinghouse to prevent denting in plants not yet licensedfor operation.3. BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLE_TION OF TASKThe safety issue addressed by this Task Action Plan is applicable toselected Pressurized Water Reactors with Westinghouse-designed (W)steam generators.For W PWRs currently licensed for operation, we have concluded that,pending completion of this TAP, continued operation does not constitutean undue risk to the health and safety of the public for the followingreasons:Primary to secondary leakage rate limits, and associated surveil-lance requirements, have been established to provide assurancethat the occurrence of tube cracking during operation will bedetected and appropriate corrective action will be taken suchthat an individual crack will not become unstable under normaloperating, transient or accident conditions.Augmented inservice inspection requirements and preventativetube plugging criteria have been established to provide assurancethat the great majority of degraded tubes will be identified andremoved from service before leakage develops.Steam generator water chemistry control requirements are beingconsidered to provide additional assurance that the potentialfor tube degradation during operation is minimized. On a case-by-case basis, additional measures have been taken to (1) minimizecontamination of the secondary coolant by in-leakage of condensercooling water (e.g., condenser tubes with improved corrosionresistance have been installed) and (2) minimize buildup in thesteam generators of corrosion products generated in the secondarysystem (e.g., full flow condensate demineralizers have beeninstalled).Observed through-wall cracks at dented locations, i.e., tube/support plate intersections, have been small and stable (norapid failures) during normal operation. In addition since suchcracks are constrained by the support plates, they are notanticipated to become unstable (burst) during postulated accidents.-59 -

Task A-3Rev. No. 1May 1978Even if a LOCA or a MSLB were to occur during operation andsome tubes were in a state of incipient failure, the radio-logical consequences of such an event would not be severe.Continuous feedback from operating experience and the TAPefforts will be utilized to update interim criteria andrequirements.For plants experiencing severe degradation, the following additionalinterim bases were also considered:The probability of the design basis accident during normaloperation is small and the probability that the accident wouldoccur during the short period of time between the leak detectionand the plant shutdown is even smaller.Even if an accident occurs when there are cracked tubes, theconservatively calculat'ed consequences are still acceptablysmall until plant shutdown.A small amount of leakage (e.g., less than the TechnicalSpecification limit) can be tolerated during normal operationwithout exceeding the offsite dosage limits of 10 CFR Part 20.The above-montioned rationale which constitutes the basis forcontinued operation of licensed W PWR facilities also supportscontinued licensing of new facilities. Further, to the extent thatis practicable, depending on the status of the design, fabricationand installation of the steam generators for facilities not yetlicensed for operation, "state-of-the-art" design improvements andoperating procedures which eliminate or at least minimize thepotential for steam generator tube degradation are required by thestaff. The following design and operational factors are consideredby the staff in the conduct of its reviews:Designs to provide improved circulation to eliminate low flowareas, and to facilitate sludge removal.Designs to minimize flow induced vibration and cavitation.Designs to provide increased flow around the tubes at thesupport plate.Selection of material for tube support plates with improvedcorrosion resistance.-60 -

Task A-3Rev. No. 1May 1978Material selection (chemistry), processing and heat treatmentto minimize the susceptibility of tubes to stress corrosioncracking.Secondary system water chemistry control.Secondary side material selection (condensers, feedwater,heaters turbine discs and blades, elbows, etc.), and watercleanup system to minimize erosion and the resulting sludgeand corrosion product buildup in the steam generators.Designs to allow for installation of an ion exchanger (conden-sate demineralizer) in-the secondary system to minimize feed-water contamination.* Condenser leakage detection systems.In view of the above, we conclude that issuance of ConstructionPermits and Operating Licenses, pending completion of this TAP, cancontinue with reasonable assurance that operation will not presentan undue risk to the health and safety of the public.4. NRR TECHNICAL ORGANIZATIONS INVOLVEDA. Engineering Branch, Division of Operating Reactors, has theprimary lead responsibility for the overall review and evalu-ation of steam generator tube integrity. This includes opera-tional experiences, tube failure mechanisms and potentialrepairs, plugging criteria, ISI requirements, tube failureprobability, leakage rate limits, and secondary coolant systemcontrol. This also includes the lead responsibility fordetermining the probability of LOCA and MSLB initiating eventsand the probability of tube failures during these events andresponsibility for determining the number of tubes assumed tofail in LOCA and MSLB analyses. The Engineering Branch alsohas lead responsibility for the review of prototype steamgenerator tube replacement.Manpower Estimates: 0.1 man-year FY 1977; 1.0 man-year FY 1978;1.0 man-year FY 1979.B. Environmental Evaluation Branch, Division of Operating Reactors,has the lead responsibility for the review and evaluation ofthe offsite dosage related to the consequence or probabilityof a Main Steam Line Break (MSLB) accident or LOCA given thephysical conditions determined in item A, above. EEB will-61 -

Task A-3Rev. No. 1May 1978also consult with EB and provide support for the probabilisticevaluation of MSLB and LOCA initiating events, the probabilityof tube failures during these postulated events and evaluationof environmental aspects of steam generator tube replacement.Manpower Estimates: 0.1 man-year FY 1977; 0.2 man-year FY 1978;0.2 man-year FY 1979.C. Reactor Safety Branch, Division of Operating Reactors, has thelead responsibility for the review and evaluation of: (1) theECCS performance related to secondary-to-primary leakage as aconsequence of a LOCA, and (2) the effect of primary-to-secondary leakage during a MSLB accident on fuel failures.Manpower Estimates: 0.1 man-year FY 1977; 0.13 man-year FY 1978;0.13 man-year FY 1979.D. Mechanical Engineering Branch/Materials Engineering Branch,Division of Systems Safety, has lead responsibility for thereview of new design/material concepts and new system compo-nent requirements. This will apply to PWR facilities not yetlicensed for operation.The activities involved will include the review and evaluationof applicant's and Westinghouse's proposed improvements on thedesign and/or operation of the steam generators for items suchas secondary coolant chemistry, design modifications to avoiddenting, condenser design to avoid inleakage, ISI requirements,recommendation for revision of Regulatory Guides, and provisionsfor access opening and space in the containment to facilitatesteam generator inspections.Manpower Estimates: 0.1 man-year FY 1977; 0.5 man-yearFY 1978; 0.5 man-year 1979; 0.5 man-yearFY 1980.E. Analysis Branch, Division of Systems Safety, has the leadresponsibility in developing analytical capabilities (computercode, etc.) to evaluate the effects of steam generator tuberupture(s) concurrent with various reactor transients thatinclude MSLB and LOCA accidents. The purpose is to determinethe equivalent number of tube failures that can be toleratedduring transient events. This information will then be fac-tored into the overall program of determining an adequatesample plan for tube inspections.-62 -

Task A-3Rev. No. 1May 1978Manpower Estimates: 0.1 man-year FY 1977; 0.2 man-year FY 1978;0.2 man-year FY 1979.F. Reactor Systems Branch, Division of Systems Safety. Has theresponsibility of implementing new procedures on CP/OL safetyanalyses for plants yet to be licensed should any be requiredas the result of this technical activity.Manpower Estimates: 0.1 man-year FY 1979; 0.3 man-year FY 1980.G. Environment Project Branch No. 1, Division of Site Safety andEnvironmental Analysis. Responsible for the review of thenonradiological environmental aspect of steam generator replace-ment for the lead unit.Manpower Estimate: 0.2 man-year FY 1978.5. TECHNICAL ASSISTANCEA. Contractor: Brookhaven National Laboratory (BNL) -DOR, DSSFunds Required: $98K FY 1977; $125K FY 1978; $225K FY 1979.This effort is funded as part of an overall program at BNLapplicable to the three Category A Technical Activities (A-3,A-4, and A-5) related to PWR steam generators. Funding valuesunder DORSAT are not included.This program is needed to obtain technical consultation andassistance to review information in areas of water chemistryand corrosion analysis, monitored jointly by EB/DOR and MTEB/DSS.Stress and/or burst strength calculations are funded in partunder DORSAT contract on an as-needed basis. This program willprovide assistance in accomplishing Tasks 2C, 2E, and 2G.B. Contractor: Idaho National Engineering Laboratory (INEL) -DSSFunds Required: $75K FY 1977; $100K FY 1978.This effort is generic in nature and will be applicable to thethree Category A Technical Activities (A-3, A-4, and A-5)related to PWR steam generators.The purpose of this program is to determine the effect of steamgenerator tube plugging on the predicted peak clad temperaturesfollowing a postulated LOCA. The primary activity is to producea reliable computer code to aid the evaluation of the effects-63 -

Task A-3Rev. No. 1May 1978of tube plugging on the ECCS performance. An addition to theprogram will be needed to consider steam generator tube failuresconcurrent with MSLB or a LOCA. This program will provideassistance in accomplishing Tasks 2B and 2D.C. Contractor: Sandia Laboratories, DORFunds Required: $50K FY 1977; $100K FY 1978; $150K FY 1979.This work is of generic nature, and will be applicable to allPWR steam generators.The purpose of.this program is to perform a statistical analysisof steam generator tube failures in operating reactors in orderto establish the bases for the sampling plan for inserviceinspection. This is a new program to augment staff effort insteam generator safety reviews and will assist in addressingTasks 2A, 2F, and 2G. t6. ASSISTANCE REQUIREMENTS FROM OTHER NRC OFFICESA. Office of Nuclear Regulatory Research, Division of ReactorSafety Research, Metallurgy and Materials Branch and Probabi-listic Analysis Branch.RES has funded, at the request of NRR, a major confirmatoryexperimental program at Pacific Northwest Laboratory. Theactivity of this program consists of a series of tests to verifythe burst and cyclic strengths of degraded steam generator tubesand the leakage rate data. This program is managed by Metallurgyand Materials Branch, (Task 2C).RES has funded, at the request of NRR, a program, to addressthe factors which determine Inconel 600 susceptibility to stresscorrosion cracking in primary water. Metallurgical condition,chemistry, temperature, stress and environment will be considered,(Task 2E).B. Office of Standards Development, Division of Engineering Stand-ards, Structures and Components Standards Branch.OSD has funded a confirmatory research program at BattelleColumbus Laboratory to evaluate eddy current methods for inspect-ing steam generator tubes as a subcontract to Brookhaven NationalLaboratory, (Part of Task 2F).-64 -

Task A-3Rev. No. 1May 1978C. Office of the Executive Director for Operations, AppliedStatistic Group.Provide assistance to EB/DOR for statistical assessment ofsteam generator tube integrity, (Part of Tasks 2A, 2F, and 2G).D. ACRSThis task is closely related to one of the generic items iden-tified by the ACRS and, accordingly, will be coordinated withthe committee as the task progresses.7. INTERACTIONS WITH OUTSIDE ORGANIZATIONSA. Licensee(s) of Westinghouse (W) Nuclear FacilitiesAt present all W plants experiencing tube denting will bemonitored for the progress of denting. Each licensee willsubmit an analysis of the consequences of tube denting ontube integrity and demonstrate that adequate safety marginsexist for continued safe operation. The Turkey Point andSurry licensees will be closely monitored relative to steamgenerator replacement.B. WestinghouseThe primary interaction with Westinghouse has been and continuesto be on the investigation program for the resolution of theproblems at Westinghouse designed plants and their genericimplication such as the licensing bases or justifications forcontinued operation of Westinghouse plants with known tubedegradations. For interim periods of operation before thecause of denting is identified and corrective measures imple-mented, the interaction will be needed to ensure that Westing-house develops and improves capabilities for the evaluation ofECCS performance under postulated accidents concurrent withtube failures should such a licensing basis become necessary.Review and evaluate new designs proposed to prevent denting infacilities not yet licensed for operation.C. EPRI, PWR Owner Group, etc.Interactions with other organizations such as the ElectricPower Research Institute (EPRI) and the "ad hoc" organizationof PWR owners may also be required because of mutual interestsin the safe operation of steam generators in general and, inparticular, the various problems associated with the operationof steam generators.-65 -

Task A-3Rev. No. 1May 1978The purpose for interactions with these organizations is toexchange information on the research works sponsored by NRC andthese outside organizations in identifying potential problemsor solutions to existing problems associated with the operationof steam generators. Current programs in this area include anEPRI sponsored steam generator program in conjunction withCombustion Engineering. One aspect of this program is designedto define the mechanism of tube denting, and its goal is toprovide corrosion-related information for improved steam gener-ator coolant system technology and operation. The technologywill be applied to the operation of plant systems and componentsthat affect the reliability of steam generators. Additionally,EPRI had underway an ISI round robin test program for steamgenerator tubes to determine the effectiveness of various ISItechniques and methods for tube inspection.8. POTENTIAL PROBLEMSExcept for steam generator replacement, there is no apparent shortterm resolution of tube denting in affected Westinghouse plants.The many programs underway to resolve tube denting in presentlyoperating plants may bring about a partial solution, by arrestingdenting through a cleaning program, sometime early in 1979.However, by establishing quantitative plugging criteria for dentedtubes, and requiring scheduled inspections varying with the degreeof denting observed, safety concerns can be minimized to the pointwhere continued operation can be justified.Finally, completion of many of the indicated tasks will depend onthe scheduled completion of programs sponsored by organizationsoutside NRR. As with most experimental investigations, periodicaldelays can be expected, which may delay completion of some of thetasks indicated in the Task Action Plan.-66 -

Task A-4COMBUSTION ENGINEERING STEAM GENERATOR TUBE INTEGRITYLead NRR Organization: Division of Operating Reactors(DOR)Lead Supervisor: Darrell G. Eisenhut, A/D forSystems and Projects, DORTask Manager: Frank M. Almeter, EB/DORApplicability: Combustion Engineering PressurizedWater ReactorsProjected Completion Date: December 31, 1979-67 -

Task A-4Rev. No. IMay 19781. DESCRIPTION OF PROBLEMPressurized water reactor operating experience during the past fiveyears has shown that steam generator tube integrity can be degradedby corrosion induced wastage, cracking, reduction in tube diameter(denting) and vibration induced fatigue cracks. Since the steamgenerator tubes are an integrated part of the reactor coolant pres-sure boundary in the PWR system, the primary concern is the capabil-ity of degraded tubes to maintain their integrity during normaloperation and under accident conditions (LOCA or a main steam linebreak) with adequate safety margins.Palisades has been the only Combustion Engineering designed plant toexperience tube degradation due to wastage and secondary side stresscorrosion cracking with the use of a phosphate treatment for thesecondary coolant. Both types of degradation have been nominallyarrested by conversion to AVT chemistry control. However, tubedegradation due to denting (but to a lesser degree than the Westing-house steam generators) occurred after the conversion to an AVTchemistry Recent inservice inspections at two sea coast facilitieswith CE designed steam generators, which used an AVT chemistry forthe secondary coolant since initial startup, have shown that theprior use of phosphates is not a necessary precursor to causedenting in steam generator tubing. Denting which leads to primaryside stress corrosion cracking is the major problem at present andthe principal focus of this technical activity. However, as steamgenerator operating experience is accumulated and interpreted, ithas become evident that condenser cooling water in-leakage resultingfrom the corrosion of condenser tubes can contaminate the secondarywater of PWR steam generators and may be the principle source leakingto all types of steam generator tube degradation. It has also becomeevident that the maintenance of secondary coolant water quality can-not be achieved if condenser in-leakage is allowed. Because thecondenser is an important component of the PWR secondary system, anapproach must be developed to minimize condenser in-leakage to ensureadequate steam generator tube integrity.2. PLAN FOR PROBLEM RESOLUTIONThe problem will be resolved by reviewing the type and mechanism oftube degradation in operating reactors to evaluate the effects of tubestructural integrity and failure probability under normal operationand accident conditions (LOCA, SSE and MSLB). Assessment of theeffects of degraded tubes on postulated accident conditions will befactored into the development of new criteria for tube plugging,acceptable levels of primary to secondary leakage, and ISI require-ments to ensure the safe operation of operating pressurized water-69 -

Task A-4Rev. No. 1May 1978reactors. To minimize tube degradation, priority areas whereimprovements in steam generator design and criteria for the secondarycoolant system are needed will be identified to develop licensingpositions for the CP/OL review of new plants.The specific activities directed at resolution of the denting problemin Combustion steam generators consist of the following issues andtasks:A. Generic Evaluation of ISI ResultsReview and evaluate the various eddy current inspection results;i.e., experience from operating reactors and evaluate these dataas they relate to the generic determination of failure probabil-ity of degraded tubes. In addition, evaluate the test programsand analytical studies to provide staff understanding suffi-cient to continue to provide justification of continued safeoperation of operating reactors.B. Evaluation of Transients and Postulated AccidentsEvaluation of failure consequences under postulated accidentconditions (LOCA and MSLB) to determine the acceptable levelsof primary to secondary leakage rates and the effect on ECCSperformance. The results will be used to define the acceptablenumber of tube failures that may be necessary as a licensingbasis considering predicted fuel behavior and radiologicaldose during transients and postulated accident conditions.C. Evaluation of Steam Generator Tube Structural IntegrityEvaluation of licensees' and CE's analysis of structural integ-rity of tubes under normal operating and accident conditions(LOCA, SSE and MSLB). Information developed in this task willprovide input for establishing a generic tube plugging criteriaand recommendations for the revision of Regulatory Guide 1.121.D. Establish Tube Plugging CriteriaEstablish a generic tube plugging criteria that is consistentwith the determined allowable leak rate, tube structural integ-rity and degradation rates. These results will allow assess-ment of the adequacy of the requirements defined in RegulatoryGuide 1.21.

Task A-4Rev. No. 1May 1978E. Secondary Coolant Chemistry RequirementsEvaluate the mechanism of tube degradation. The results willbe used to define the requirements for secondary coolantchemistry control including considerations for condenser in-leakage.F. Evaluation of ISI MethodsReview the development of improved eddycurrent probes, coilsand multi-frequency techniques to better quantify dents andgrowth of dents and increase sensitivity for detecting cracksin dented regions.G. Establish Criteria for Revision of Regulatory Guide 1.83Integrate experience from inservice inspection results, theresults from the evaluation of various ISI improvements andthe plugging and secondary water chemistry requirements intocriterion for possible revision of Regulatory Guide 1.83.H. Review Design Criteria for Plants Not Yet LicensedReview and evaluate design modifications proposed by applicantsand CE to prevent denting in plants not yet licensed foroperation.3. BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLE-TION OF TASKThe safety issue addressed by this Task Action Plan is applicableto Pressurized Water Reactors with Combustion Engineering (CE)steam generators.For CE PWRs currently licensed for operation, we have concludedthat, pending completion of this TAP, continued operation does notconstitute an undue risk to the health and safety of the public forthe following reasons:Primary to secondary leakage rate limits, and associatedsurveillance requirements, have been established to provideassurance that the occurrence of tube cracking during operationwill be detected and appropriate corrective action will betaken such that no individual crack will become unstable undernormal operating, transient or accident conditions.,-71 -

Task A-4Rev. No. 1May 1978* Augmented inservice inspection requirements and preventativetube plugging criteria have been established to provide assurancethat the great majority of degraded tubes will be identified andremoved from service before leakage develops.Steam generator water chemistry control requirements are beingconsidered to provide additional assurance that the potentialfor tube degradation during operation is minimized. On acase-by-case basis, additional measures have been taken to (1)minimize contamination of the secondary coolant by in-leakage ofcondenser cooling water.(e.g., condenser tubes with improvedcorrosion resistance have be installed) and (2) minimize buildupin the steam generators of corrosion products generated in thesecondary system (e.g., full flow condensate demineralizers havebeen installed).Tube denting at tube/support plate intersections in CE designedsteam generators has not been severe enough to result in through-wall cracks at dented locations. However, if tube cracking wereto occur in the dented region, it would be constrained by thesupport plates which woul-d control crack stability and preventtube failure (bursting) during postulated accidents.Even if a LOCA or a MSLB were to occur during operation and sometubes were in a state of incipient failure, the radiologicalconsequences of such an event would not be severe.Continuous feedback from operating experience and the TAPefforts will be utilized to update interim criteria andrequirements.For plants experiencing severe degradation, the following additionalinterim bases were also considered:The probability of the design basis accident during normaloperation is small and the probability that the accident wouldoccur during the short period of time between the leak detectionand the plant shutdown is even smaller.Even if an accident occurs when there are cracked tubes, theconservatively calculated consequences are still acceptablysmall until plant shutdown.A small amount of leakage (e.g., less than the TechnicalSpecification limit) can be tolerated during normal operationwithout exceeding the offsite dosage limits of 10 CFR Part 20.-72 -

Task A-4Rev. No. 1May 1978The above-mentioned rationale which constitutes the basis forcontinued operation of licensed CE PWR facilities also supportcontinued licensing of new facilities. Further, to the extent thatis practicable, depending on the status of the design, fabricationand installation of the steam generators for facilities not yetlicensed for operation, "state-of-the-art" design improvements andoperating procedures which eliminate or at last minimize the poten-tial for steam generator tube degradation are required by thestaff. The following design and operational factors are consideredby the staff in the conduct of its reviews:Designs to provide improved circulation to eliminate low flowareas, and to facilitate sludge removal.Designs to minimize flow induced vibration and cavitation.Designs to provide increased flow around the tubes at thesupport plate.Selection of material for tube support plates with improvedcorrosion resistance.Material selection (chemistry), processing and heat treatmentof minimize the susceptibility of tubes to stress corrosioncracking.Secondary system water chemistry control.Secondary side material selection (condensers, feedwater,heaters turbine discs and blades, elbows, etc.), and watercleanup system to minimize erosion and the resulting sludgeand corrosion product buildup in the steam generators.Designs to allow for installation of an ion exchanger (conden-sate demineralizer) in the secondary system to minimize feed-water contamination.Condenser leakage detection systems.In view of the above, we conclude that issuance of ConstructionPermits and Operating Licenses, pending completion of this TAP, cancontinue with reasonable assurance that operation will not presentan undue risk to the health and safety of the public.-73 -

Task A-4Rev. No. 1May 19784. NRR TECHNICAL ORGANIZATIONS INVOLVEDA. Engineering Branch, Division of Operating Reactors, has theprimary lead responsibility for the overall review and evalu-ation of steam generator tube integrity in operating plants.This includes operational experiences, tube failure mechanismsand potential repairs, plugging criteria, ISI requirements,tube failure probability studies, leakage rate limits, andsecondary coolant system control. This also includes the leadresponsibility for determining the probability of LOCA andMSLB initiating events and the probability of tube failuresduring these events and responsibility for determining thenumber of tubes assumed to fail in LOCA and MSLB analyses.Manpower Estimates: 0.1 man-year FY 1977; 0.5 man-year FY1978;0.5 man-year FY 1979.B. Environmental Evaluation Branch, Division of Operating Reactors,has the lead responsibility for the review and evaluation ofthe offsite dosage related to the consequence or probabilityof a Main Steam Line Break (MSLB) accident or a LOCA shouldsuch evaluation become necessary. EEB will also consult with EBand provide support for the probabilistic evaluation of MSLB andLOCA initiating events and the probability of tube failuresduring these postulated events.Manpower Estimates: 0.1 man-year FY 1977; 0.2 man-year FY 1978;0.2 man-year FY 1979.C. Reactor Safety Branch, Division of Operating Reactors, has thelead responsibility for the review and evaluation of: (1) theECCS performance related to secondary to primary leakage as aconsequence of a LOCA, and (2) the effect of primary to second-ary leakage during a MSLB accident on fuel failures should suchevaluation prove necessary.Manpower Estimates: 0.1 man-year FY 1977; 0.13 man-year FY 1978;0.13 man-year FY 1979.D. Mechanical Engineering Branch/Materials Engineering Branch,Division of Systems Safety, has responsibility in factoringall steam generator operating experience into the review ofnew design/material concepts and new system component require-ments. This will apply to PWR facilities not yet licensed foroperation.-74 -

Task A-4Rev. No. 1May 1978The activities involved will include the review and evaluationof the applicant's and the NSSS's proposed improvements on thedesign and/or operation of the steam generators; for items suchas secondary coolant chemistry, design modifications to avoiddenting, ISI requirements, recommendations for revision ofRegulatory Guides, condenser design to avoid in-leakage andprovisions for access opening and space in the containment tofacilitate steam generator inspections.Manp~ower Estimates: 0.1 man-year FY 1977; 0.5 man-year FY 1978;0.5 man-year FY 1979.E. Analysis Branch, Division of Systems Safety, has the leadresponsibility in developing analytical capabilities (computercodes, etc.) to evaluate the effects of steam generator tuberupture(s) concurrent with various reactor transients thatinclude MSLB and LOCA accidents. The purpose is to determinethe equivalent number of tube failures that can be toleratedduring transient events. This information will then be factoredinto the overall program of determining an adequate sample planfor tube inspections.Manpower Estimates: 0.2 man-year FY 1978; 0.2 man-year FY 1979.F. Reactor Systems Branch, Division of Systems Safety, has theresponsibility of evaluating the design and performance ofnew associated auxiliary systems for CP/OL plants yet to belicensed, should any be required as the result of this tech-nical activity; e.g., full flow condensate demineralization,etc., for PWR secondary coolant.Manpower Estimate: 0.15 man-year FY 1979.5. TECHNICAL ASSISTANCEA. Contractor: Brookhaven National Laboratory (BNL) -DOR, OSSFunds Required: $98K FY 1977; $125K FY 1978; $225K FY 1979.This effort is funded as part of an overall program at BNLapplicable to the three Category A Technical Activities (A-3,A-4, and A-5) related to PWR steam generators. Funding valuesunder DORSAT are not included.This program is needed to obtain technical consultation andassistance to review information in areas of water chemistryand corrosion analysis, monitored jointly by EB/DOR and MTEB/DSS.-75 -

Task A-4Rev. No. 1May 1978Stress and/or burst strength calculations are funded in partunder DORSAT contract on an as-needed basis. This program willprovide assistance in accomplishing Tasks 2C, 2E, and 2G.B. Contractor: Idaho National Engineering Laboratory (INEL) -DSSFunds Required: $75K FY 1977; $100K FY 1978.This effort is generic in nature and will be applicable to thethree Category A Technical Activities (A-3, A-4, and A-5)related to PWR steam generators.The purpose of this program is to determine the effect of steamgenerator tube plugging on the predicted peak clad temperaturesfollowing a postulated LOCA. The primary activity is to producea reliable computer code to aid the evaluation of the effects oftube plugging on the ECCS performance. An addition to the pro-gram will be needed to consider steam generator tube failuresconcurrent with MSLB or a LOCA. This program will provideassistance in accomplishing Tasks 2B and 2D.C. Contractor: Sandia Laboratories -DORFunds Required: $50K FY 1977; $10OK FY 1978; $150K FY 1979.This work is of generic nature, and will be applicable to allPWR steam generators.The purpose of this program is to perform a statistical analysisof steam generator tube failures in operating reactors in orderto establish the bases for the sampling plan for inservice inspection.This is a new program to augment staff effort in steam generatorsafety reviews and will assist in addressing Tasks 2A, 2F, and2G.6. ASSISTANCE REQUIREMENTS FROM OTHER NRC OFFICESA. Office of Nuclear Regulatory Research, Division of ReactorSafety Research, Metallurgy and Materials Branch and Proba-bilistic Analysis Branch.RES has funded, at the request of NRR, a major confirmatoryexperimental program at Pacific Northwest Laboratory. Theactivity of this program consists of a series of tests to verifythe burst and cyclic strengths of degraded steam generator tubesand the leakage rate data. This program is managed by Metallurgyand Materials Branch, (Task 2C).-76 -

Task A-4Rev. No. 1May 1978RES has funded, at the request of NRR, a program to address thefactors which determine Inconel 600 susceptibility to stresscorrosion cracking in primary water. Metallurgical condition,chemistry, temperature, stress and environment will be considered,(Task 2E).The Probabilistic Analysis Branch funded the program to assistEEB in probabilistic analyses, (Task 2B).B. Office of Standards Development, Division of Engineering Stand-ards, Structures and Components Standards Branch.OSD has funded a confirmatory research program at BattelleColumbus Laboratory to evaluate eddy current methods forinspecting steam generator tubes as a subcontract toBrookhaven National Laboratory, (Part of Task 2F).C. Office of the Executive Director for Operations, AppliedStatistics Branch.Provide assistance to EB/DOR for statistical assessment of steamgenerator tube integrity, (Part of Tasks 2A, 2F, and 2G).D. ACRSThis task is closely related to one of the generic items iden-tified by the ACRS and, accordingly, will be coordinated withthe committee as the task progresses.7. INTERACTIONS WITH OUTSIDE ORGANIZATIONSA. Licensee(s) of Combustion Engineering Nuclear FacilitiesAt present all CE plants experiencing tube denting will bemonitored to evaluate the progress of denting. Each licenseewill submit an analysis of the consequences of tube dentingon tube integrity and demonstrate that adequate safety marginsexist for continued safe operation.B. Combustion EngineeringThe primary interactions with CE has been and continues to berelated to their investigation program for the resolution ofthe tube denting problem at CE designed plants, and its genericimplications, such as the licensing bases or justificationsfor continued operation of CE plants with known tube degrada-tions. For interim periods of operation until the cause of tube-77 -

Task A-4Rev. No. 1May 1978denting is identified and corrective measures(s) implemented,this interaction will be needed to ensure that CE developscapabilities for the evaluation of ECCS performance for postu-lated accidents concurrent with tube failures, should such alicensing basis become necessary. In conjunction with licensees,CE will be requested to submit a test program and correctiveaction plan for Maine Yankee and Millstone Unit 2 and an analysisof the structural integrity of degraded tubes under normal oper-ating and accident conditions (LOCA, SSE and MSLB).In addition, CE will be requested to keep NRC informed of steamgenerator design changes and modifications in secondary watertreatment systems to alleviate tube degradation in future CEplants. This information will be incorporated into all Tasks ofthe program.C. EPRI, PWR Owner Group etc.Interactions with other organizations such as the Electric PowerResearch Institute (EPRI) and the "ad hoc" organization of PWRowners may also be required because of the mutual interests inthe safe operation of steam generators in general and, in par-ticular, the various problems associated with the operation ofsteam generators. Current programs sponsored by EPRI includethe CE model boiler studies and the round robin program for ISItechniques.8. POTENTIAL PROBLEMSIt should be anticipated that required feedback from related programsfunded by outside organizations may delay the timely completion ofcertain subtasks. However, it is hoped that effective participationof NRC representatives at "ad hoc" organizational meetings willimprove mutual interests in NRC goals. Any delays in submittalsrequired by licensees and NSSS vendors would certainly delay thereview and evaluation of tasks defined in the program. Timely inputis required from all technical organizations involved.-78 -

., I.APPENDIX CCONDENSER TUBE MATERIALSFOR OPERATING PLANTS (PWR)Plant CondenserName Tube MaterialNSSSVendorWCEBeaver Valley 1Cook 1Cook 2Farley IGinnaHaddam NeckIndian Point 2Indian Point 3KewauneeNorth Anna 1 & 2Point Beach 1Point Beach 2Prairie Island 1Prairie Island 2Robinson 2Salem 1San Onofre 1Surry 11Surry 22TrojanTurkey Point 31Turkey Point 41Yankee RoweZion IZion 2Arkansas 2Calvert Cliffs 1Calvert Cliffs 2Fort Calhoun 1Maine YankeeMillstone 22Palisades 13St. Lucie 1Stainless steelCu and stainless steelCu and stainless steelTitaniumAdmiralty and stainles!Admiralty and stainles!AdmiraltyAdmiraltyAdmiralty brassStainless steelAdmiraltyAdmiraltyStainless steelStainless steelAdmiralty brass and sti90-10 CuNi and titaniui90-10 CuNi and titaniw90-10 CuNi90-10 CuNiAdmiralty brass and 70-Al-brassAl-brassAdmiralty and stainlessStainless steelStainless steels steels steelainless steelnn-30 CuNi; steel90-10 CuNi70-30 CuNi70-30 CuNiStainless steelAl-brass and 70-30 CuNi70-30 CuNi90-10 CuNiAl-brass-79 -

I I -APPENDIX C (continued)CondenserTube MaterialNSSSVendorPlantNameB&WArkansas 1Crystal River 3Davis-Besse 1Oconee 1Oconee 2Oconee 3Rancho Seco 1Three Mile Island 1Three Mile Island 2Admiralty70-30 CuNiStainless steelStainless steelStainless steelStainless steelStainless steelStainless steelStainless steelNotes: 123Retubing condenser with titanium.Was Al-brass up to 5/77.Previously Admiralty.-80 -

GLOSSARYextensive denting -(a) presence of tube denting that is widespread through-out whole steam generator in which the average total reduction in tubediameter equals to or exceeds twice the tube wall thickness; (b) measurablesupport pate in-plane deformations, such as hourglassing of flow slots inWestinghouse plants; (c) damage has caused leaking from dents.moderate denting -(a) presence of the tube denting that is widespreadthroughout whole steam generator in which the average total reduction intube diameter exceeds 20 percent of the tube wall thickness; (b) no -measurable support plate in-plane deformation; (c) damage has not causedleaking from dents.minor denting -(a) presence of tube denting isthe average total reduction in tube diameter isthe tube wall thickness; (b) no visible supportdamage has not caused leaking from dents.spotty to widespread, butless than 20 percent ofplate deformation; (c)-81 -

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORTNUMBER i4sidby ODC)BIBLIOGRAPHIC DATA SHEET NUREG-05234. TITLE AND SUBTITLE (Add Volume No., if ppropriers) 2. (Loa blank)Summary of Operating Experience with Recirculation .RECIPIENTS ACCESSION NO.Steam Generators7. AUTHOR(S) 5. DATE REPORT COMPLETEDD. G. Eisenhut, B. D. Liaw, J. R. Strosnider, Jr. January 19799. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (include Zip Code) DATE REPORT ISSUED.~MONTH lYEARU. S. Nuclear Regulatory Commission MONTH ______Division of Operating Reactors 6. (La. blank)Washington, D.C. 205558. (Leave blank)12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (include Zip Code)10. PROJECT/TASKIWORK UNIT NO.Same as 9 above 11. CONTRACT NO.13. TYPE OF REPORT PERIOD COVERED (inclusive dats)Technical Report15. SUPPLEMENTARY NOTES 14. (Leave blank)16. ABSTRACT 200 words or less)Operating problems have occurred in the steam genertors of each of thethree manufacturers of pressurized water reactors (PWP) nuclear steamsupply systems (NSSS): Babcock & Wilcox, Combustion Engineering, andWestinghouse Electric Corporation. This report focuses on the problemsassociated with steam generators of the recirculation type that are designedby Westinghouse and Combustion Engineering, It identifies the operationalproblems observed to date, including the NRC staff's evaluation of suchproblems, and provides a status report summarizing the NSSS, licensee, andstaff programs for the resolution of each problem.17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS17b. IDENTIFIERSIOPEN-ENOED TERMS18. AVAILABILITY STATEMENT 19. SECURITY CLASS (This repont 21.NO. OF PAGESUnlimited 20. SECURITY CLASS (Thispop ]w22. PRICE$NRC FORM 335 17-771