ML20072L975

Present Status & Projected Future Progression of Steam Generator Tube Corrosion/Degradation at Point Beach Unit 1, Wisconsin Electric Power Co,Two Creeks,Wi

ML20072L975
Person / Time
Site:	Point Beach
Issue date:	06/08/1983
From:	Myers J JRM ASSOCIATES
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ML20072L965	List: ML20072L962 ML20072L975
References
TAC-48752, NUDOCS 8307140268
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Text

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, o ENCLOSURE 2 I l

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PRESE::7 STATUS A::D PROJECTED FUTURE PR03P2SSION OF STEAM-GENE?.ATOR TUEE CORR 3SION/DEORADATION AT POINT BEACH UNIT NO.1, NISCO:: SIN ELECTRIC PO'4ER CO::PANY, T..'O CREEKS, WISCONSIN

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~JRM Associates Consulting Corrosion /:.'etallurgical Engineers 4198 Xerlyn Drive Franklin, Ohio 45005 Prepared for Public Service Co.nission State of Wisconsin 4802 Sheboygan Avenue

.Fadison, Wisconsin 53707

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', a i 6 Tabic of Contents Page Introduction 1 Potential for Steam Generator Tube Corrosion /

Degradation at Point Ecach Unit No.1 3 General 3 Denting 5 U-Eend Cracking 9 Wastage / Thinning 10 Pitting 11 Intergranular/ Stress Corrosion 12 Beneficial Effects of Thema11y-Treated Inconel-600 15

( Secondary-Side Water Chemistry at Point Beach Unit No. 1 20 Anticipated Tube Plugging and Associated Power-Output Reductions for the Existing Steam Generators at Point Beach Unit No. 1 24 Necessity for Ecplacing Equipment at Point Beach Unit No.1 Other Than the Stean Generators 26 Necessity for Condensate Polishing at Point Beach Unit No. 1 28 Conclusions 28 Referunces 31- .

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APPENDIX Replacement Steam Generator Equipment Technical Description l Unit 1 Steam Generator Tube Plugging History a

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PRESENT STATUS AND PROJECTED FUIUP5 PR00?2SSION OF STEAL GENE?lTOR TUBE COP 30] ION /DESPADATIO:: AT POI"T EEACH UNIT UO. 1, ' DISCO:: SIN ELECTRIC PO'.GR C0:.*PANY, T40 CP.EEKS, WISCONSIN Introduction Numerous incidents of mill-annaaled, Inconel-600 steam generator (SG) tube corrosion / degradation have occurred at Point Beach Unit No. 1. The most important of these are intergranular/ stress corrosion cracking above the tubcsheet/ plate (TS), wastage / thinning above the tubesheet, denting in the tube support plates (TSP), and intergranular/ stress corrosion in the tube-tubesheet crevices. These phenomena as they relate to Point Beach Unit No. 1 and other operating steam generators through December 23, 1980 were described and discussed in an earlier report. (I) Tube corrosion /

degradation during the first 9.9 years of operation resulted in the plugging of many tubes in both steam generators. .

Since December 23, 1930, an additional 63 tubes have been plugged in SG-1 A; 31 tubes have been plugged in SG-13. At the present time, the per-centage of tubes plugged in SG-1 A and SG-1B are, respectively,14.4 and 13.2. (2) This represents a significant number of plugged tubes, especially e.

since the Nuclear Regulatory Commission typically prefers to limit the per- -

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cen', age of plu'g ged tubes in a steam generator to about 12.

In order to minimite future corrosion-induced steam generator tube plugging, Wisconsin Electric Power Company (WEP00) presently operates l Point Ecach Unit No.1 at about 77,5 of full power (thereby reducing the hot-leg temperature from 597 to 557 F). The corrosion-related benefits of operating at 77% of full power were identified in the earlier report. ( }

The benefits in this reduction in the hot-leg temperature were further emphasized when the unit was temporarily operated at 90;5 of full power (i.e., with a hot-leg temperature of 575 F); an unacceptably large number of tubes were subsequently plugged because of intergranular/ stress corrosion in the tube-tubesheet crevices of both steam generators.

Because it is conceivable that additional steam generator tubos will require plugging in the future, possibly necessitating further reductions

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from 77% of full-power operation, WEPC0 has proposed to replace the two steam generators at Point Peach Unit No.1. ("} In many respects, the Model 44F replacement steam generators are identical to the original M> del 44 units. The replacement steam generators, however, include numerous design improvements to preclude corrosion-related steam generator tubo degradation. (5,0 Further, the Inconel-600 tubes in the replacer.ent steam generators will be thermally treated to further mitigate the potential for corrosion. A tech-nical description summary of the steam generator equipment replacement pro-

gram for Point Beach Unit No.1 is included in the Appendix. (7)

The purpose of this report is to present the results of an in-depth study regarding the present status of steam generator tube corrosion / degradation at Point Beach Unit No.1. Concurrently, an attempt is nade to' predict future tube corrosion / degradation with and without steam generator re-p-*

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< placement. Specifically, the study was designed to ovaluate the follouing:

1. The anticipated fature performance of the replacement steam gen-erators, with regards to:

a. Full-power operation using the current secondary-side water treatment and water chenistry control systems.

b. Tube degradation by wastage, pitting, denting, U-bend cracking, intergranular/ stress cerrosion, fretting, fctigue, and erosion. ,

c. Anticipated tube plugging. -

d. The necessity for condensate polishers and condenser re-tubing.

2. The anticipated future performance of the existing steam generators with regards to:

a. The possibility of continued tube degradation at 77% of full power operation.

b.1hc possibility of further reductions in power output because of additional tube plugging.

3. The potential need to replace major portions of the total cystem (e.g., feeduater heaters) in order to mitigate:

a. Corrosion of equipment other than the steam generators,

b. Corrosion of the steam generators.

In part, the information required to make these evaluations was ob-tained from State of h'isconsin/Public Service Comnission (PSCW) hearing transcripts and exhibits. Applicable data vere also obtained from the U.S.

Nuclear Regulatogr Conmission (NRC); Westinghouse Electric Corporation (WEC),

PSCW, and WEPCO personnel; and the technical literature. '

Pctential for Steam Generator Tube Corrosion /

Degradation at Point Beach Unit h'o, 1 General A number of nill-annealed, Inconel-600, steam generator tube corrosion /

degradation problems (of varying magnitude and concern) have been identified s ..

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(~ in operating pressurized-water-reactor (PdR) power production systems.

These are:

1. Uastage/ Thinning

2. Pitting

3. Denting

4. U-Bend Cracking

5. Intergranular/ stress corrosion

6. Fretting

7. Fatigue

8. Erosion several of these phenomena, houever, are of little or no concern to steam generator operation at Point Beach Unit No. 1 For example, erosion and fatigue are once-through steam generator (OT33) problems.( } It is further k" understandable that thermal fatigue of the steam generator tubes would not be expected to be a significant problem in recirculating PdRs because of their inherent design (i.e. , the tharx.1 stresses / strains are inherently re-laxed by the U-bends).

The potential for tube degradation by fretting is equally considered to be of insignificant concern to steam generator operation at Point Beach Unit No.1. This belief is supported, in part, by the observaiion that there has been no indication of significant fretting danage to any of the tubes in any of the operating Westinghouse Model 44 Series steam generators. Model 44-Series steam generators have operated up to 13 years to date without fretting- type tube degradation; there is no reason to believe that fretting will be a problem in the future.

Based upon these and other considerations, it can be concluded that

[bx the existing (and the replaccmant) stean generators at Point Beach Unit'No.

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!;.. 1 can be expcoted to operate until at least the year 2008 without major j concern for tuba degradation by either erosion, fatigue, or fretting. The other five forms of tube degradation, however, are potential problems and j deserve seriouc consideration.

Denting A history of Inconel-600 tube plugging in the steam generators at Point i

Beach Unit No.1 is included in the Appendix. Examination of these data reveals that denting has not been a significant problem. A total of only 11 tubes have been plugged because of denting; none have been plugged since i

September 1978. This is understandable because denting is basically an acid-chloride-related phenomenon. (1'9) Denting is generally associated with FWR units where the condenser-cooling waters contain appreciable amounts of

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chloride (i.e., where sea or brackish water is used for cooling) and there is significant ingress of cooling water (along with dissolved oxygen) into the PdR secondary-side water because of condenser leakage.

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) Condenser-cooling water at Point Ecach Unit No. 1 contains only a I

i small amount of chloride (about 3 parts per million by weight, ppm). (1)

Equally important, c ndenser tube leakage has not been a serious problem at this operating unit since the fatigue-related, condenser-tube degradation

problem was solved in the early 1970s. WEPCO also has ,a continuous, water-l chemistry monitoring program (i.e., cation and/or total conductivity) for

! the four condenser hotwells, the feedwater, and the steam generator blow-down. Condenser leaks can be rapidly detected and located; the affected condenser can be isolated from the system before any appreciable amounts

. of relatively-high quality Lake Michigan water can enter the secondary-side

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,1 Basically, there is no reason to believe that denting-related steam generator tube degradation will be a future problem at Point Ecach Unit No.1. The existing steam generators should operate at least an additional 25 years without significant concern for denting.

Concern for the potential denting problem would be even further re-1 duced if the stean generators were replaced. For example, the tube support 4

plates in the Model 44F steam generators will be fabricated from Type 405 ,

stainless steel instead of carbon steel. Compared to carbon steel, Type i i 405 stainless steel has significantly improved resistance to the chloride-induced, occluded-cell corrosion which causes denting (Figure 1). (10,11) f f (.' ISOTHERM AL AUTOCLAVE EXPOSURE t...rown u.. . ... ,

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does not steel in cause appreciable laboratory environments. corrosg)to ?fpe 405 stainless i

The possibility of having chloride-related, occluded-cell tube corrosion j{ in the replacement stean generators will be further reduced by the qua tre- .

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( foil tube-paEsage design for the tube support plates (Figure 2).(5)

, Improved pcondan-side water flow at the tube-TSP interfaces will i .

!' essentially eliminate the occluded cells (i.e., prevent the locali::ed concentrating of chloride ions) which are required for denting.

Based upon these considerations, there should be no concern for tube denting in the replacement steam generators.

There should equally be no najor concern regarding the stress-corrosion cracking of the Type 405 stainless steel tube support plates. The naterial from which the TSPs will be fabricated are delivered to Westinghouse  !

Electric Corporation in the tempered-martensite metallurgical condition. ( }

Any prior residual stresses introduced into the material during its pro-duction would be eliminated by the 1325 to 1375 F tempering. Further, the IEC quality assurance program precludes any thermal cutting or welding of

! b. the TSPs during their manufacture. Pasidual stresses introduced into the tube support plates during broaching of the quatrefoil openings will be minimal becau:e of the technique used; broaching is accomplished at ever 1

decreasing amounts of metal removal. In addition, stress-concentrating sharp corners do not exist in the TSP fN slits.

i The stress-corrosion cracking of Type 405 stainless steel would be of .

I' significant concern only if the yield strength of the material exceeded l about 160,000 pounds per square inch (psi), the applied / residual tensile

l stresses in the material were an appreciable percentage of the yield strength, and the alloy was exposed to an aggressive environment. The tempered-martensite metallurgical structure of the tube support plates for the replacement steam generators at Point Ecach Unit No.1 uill not have l

i a yield strength of 160,000 psi, the in-service TSPs will not be subjected to high tensile' stresses, and the secondan-side water is not considered to e

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405 stainless steel TSPs by the secondary-side water is not a problem of concern has apparently been confir .ed by extensive testing at Westinghouse Electric Corporation.

U-Ihnd Crackind Two possible stress-related causes are believed to be associated with the prinary-side water U-bend cracking uhich has been obset ted in the mill-annealed Inconel-600 tubes of certcin recirculating PNR steam generators.

One of these is denting-related tube support corrosion which causes the TSP flow slots to " hourglass".(9} "Hourglassing" of the flow slots forces an in-ward displacement of the legs of the tubes at these locations. When this inwani novement of the legs of the tubes l occurs at the upper tube support

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w plate, it has been shown to cause an increase in the tensile stress / strain at the U-bend apex (i.e., through tube ovalization). This additional in-

' crease in the stress / strain at the apex of t.he U-bend is probably the additional factor required to initiate the cracking. ( -

The other possible factor in the U-Send cracking is residual tensile stresses on the inside tube surfaces which ucre not eliminated after the bending operation by stress-relief annealing.

Ironically, there have been no known incidents of U-bend cracking in any Westinghouse Xodel 44-Series steam generators. There should be no major ~ concern regarding this tubo degradation phenonenon for the existing stean generators at Point Beach Unit No.1 during the next 25 years.

'The potential for U-bend cracking in the replacenent steam generators

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resistant quctrefoil tu'be-passage design and the use of highly-corrosion-resistant Type 405 stainicss steel for the tube support plates. Further, the U-bends associated with the innernost rows of tubes in the replacement stean generators will be stress-relieved after bending.

Wastage /Thinnin::

Wastage (tube thinning) of Inconel-600 stean generator tubes is basically a problem associated with ?dR units which operate / operated with coordinated-phosphate secondary-side uater treatnent. The cause of wastage is local concentrations of rssidual acid phosphates (e.g., phosphoric acid). Since the establishment of all-volatile-treatnent (AVT) chemistry control for the secondary-side water, both the evidence and the extent of wastage have been diminished and no further substantial tube degradation due to this

' mechanisn is expected to occur. (9)

No significant number of steam generator tubes have been plugged at Point Beach Unit No.1 since h'ovember 1975 because of wastage. only 14 tubes have been plugged since Novenber 1975 because of either " wastage or cracking" (see Stean Ocnerator Tube Plugging History included in the Appendix).

Regardless of whether the reason for plugging these 14 tubes over a 7.5-year period was cracking above the tubesh ets or wastage, the' number of tubes involved is considered insignificant; further, the danage was un-doubtedly associated with phosphate treatment of the secondary-side water during the early years of Point Ecach Unit No.1 cperation.

Based, in part, upon the rigorous tubesheet and crevice cleaning pro-grams which NZPCO personnel have conducted since the conversion to all-volatile l secondary-side water treatment in September 1974, there is no reason to e

i t 11-( believe that future vastage will be a serious problem in the existing steam generators. This bflief is supported by the results of extensive eddy-current testing (ECT); there has been no indication that significant additional wastage has occurred,in recent years. ( )

With regards to the replacenent stean generators, the secondary-side unter chemistry will not involve caustic-producing phosphates. It is in-conceivable that tube unstage would occur providing reasonable attention is given to control of the secondary-side water chemistry.

Pitting Only three recirculating-type PdP.s have apparently experienced any significant pitting attack on the outside surfaces of the stean generator tubes. ( 3} A cold-leg phenomenon, pitting should be of concern only if it

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is anticipated that there uill be appre.ciable ingress of chloride and oxygen-containing, condenser-cooling water into the secondary circuit as a result of condenser leakage. (I ) Reportedly, pitting can possibly be facilitated by the presence of sludge and scale containing copper or copper oxide (s),(IO)

The results of research by Hickling and Wieling( ) provide significant insight regarding the possibility of Inconel-600 pitting attack taking place in an operating stean generator. Eased upon pitting potential / voltage data, it was shown that Inconel-600 nicht pit over the 300 to 480 F range providing the secondary-side water contained nore than one ppm oxygen in conjunction with a chloride-ion concentration of about 20 ppn.

Since the c'henical conposition of the secondary-side water (i.e., the condensate, the makeup, and the feedwater) at Point Beach Unit No.1 are continuously monitored, especially with respect to condenser-cooling E-r

. (~ uator inleakage, the sufficiently aggressive conditions required for

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pitting should never exist. The inleakage would be detected and its cause corrected long before the conditicns apparently required for pitting were created. Further, time is required for pits to initiate and subsequently propagate. The times required for pit initiation are normally reported in units of days or longer. .

Concern for steam generator tube pitting at Point Ecach Unit No. 1 can be farther nininized by appreciating that Lake Iichigan uater typically contains only about 3 ppm chloride. One-hundred percent Lake Michigan water would have to be concentrated by a factor of about six before it reached 20 ppn chloride; according to Fackling and Wieling's data, this concentrated lake Michigan water nicht pit Inconel-600 if the water also contained about one ppm dissolved oxygen. With regards to an operating stean generator, one

- ppm dissolved oxygen in the secondary-side rater would exceed the " free-world's supply. " The dissolved oxygen content of the feedwater at Point Beach Unit No.1 is typically less. than 5 parts per billion by weight (ppb).

. Based upon the.se considerations, it can be concluded that pitting should not occur on the outside surfaces of the Inconel-600 steam generator tubes at Point Ecach Unit Ho 1. The unit should be capable of operating at least an additional 25 years without concern for Inconel-600 ' pitting attack, either with or without steam generator replacement.

Intergranular/ Stress Cerrosion Two, perhaps similar or even identical, intergranular/ stress corrosion phenomena have baen observed in PdR, mill-annealed, Inconel-600 steam gen-

. erator tubes. These are intergranular/ stress corrosion above the tubesheet 1 k.

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( and intergranular/ stress corrosion in the deep tube-tubesheet crevices ,

{ which exist in certain older stean generators. Both phononena aro un-doubtedly caustic related and associated with units which operated orig-inally with phosphate treatnent of the secondary-side water. 0,3,15)

Intergranular/ stress corrosien of stean generator tubes above the tube-1 sheet at Point Beach Unit No.1 is of no major concern at the present time because the deep piles of caustic-containing sludge on the tubesheets have j

been clininated by the rigorous sludge cleaning / lancing programs conducted by EPCO personnel since September 1974. Further, deep sludge piles will  ;

not form in the future because condenser-cooling water inleakago into the j secondary-sit a water has been significantly reduced. Sludge-producing i

phosphates are not used to " correct" inleakage. Inleakage is corrected j . at its source (i.e., even small amounts of inleakage are readily detected

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by the continuous chemistry-monitoring systen and the cause such as a leaking condenser tube is corrected before any deleterious amounts of Lake Michigan water enter the secondary-side water).

i Intergranular/ stress corrosion of steam generator tubes above the tubesheet should not be a limiting factor in achieving an additional 25 t years of operation at Point Beach Unit No. I regart Vas of the decision to replace or not to replace the stean Generators.

Intergranular/ stress corrosion of stean generator tubes in the 0.007-j inch vide,19-inch deep crevice which exists at each tube-tubesheet intersection apparently can be controlled by c.aintaining the hot-leg tenperature at a value which does not exceed about 557 ? (i.e. , the hot-leg temperature associated with the present 77) of full-power operation). It is obvious that intergranular/ stress corrosion is strongly temperature dependent; raising l(

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( the hot-leg tenperature in July 1981 only 18 ? (i.e., to 575 F) at Point Beach Unit ::o.1 subsequently required an unacceptably large number of steam r,cncrator tabes to be plugged about eight months later (see Stean Generator Tube Plugging History included in the .tppendix).

Since NEP00's ( as cell as other.9) vigorous attempts to completely renove the aggressive (caustic) species fron the deep, narrow crevice have not been successful, the stean generator operation at Point Beach Unit No.

1 is destined / limited to a hot-leg maxinun temperature of about 557 F unless the stcan generators are replaced. Even at a hot-led tenperature of 557 F there is no conplete assurance that unacceptable intergranular/

stress corrosion will not develop uithin the next 25 years - - although the results of recent ECT could be interpreted to suggest that it may not.

Basically, there are too little long-tena data and too many variables to

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accurately predict the life expectancy for the existing stcan generator tubes with regards to intergranular/ stress corrosion in the crevices.

The only practical solution to the deep-crevice problem and its inherent intergranular/ stress corrosien susceptibility is to climinate the crevices. This will be accenplished in the replacement stean generators by hydraulically expanding the . tubes over the entire tubesheet length. This design change, and others such as flow-distribution baffles above the tube-sheet which will direct any secondary-side sludge / scale to an inproved blevdown (see Replacement Stean Generator Equipment Technical Description included in the Appendix), in conjunction with the installation of thernally-treatpd' Inconel-600 tubes should clininate the possibility of any caustic-  !

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intergranular/ stress corrosion in the replacement stcan generators. l I

.r The t*nefits of thernally-treated Inconel-600 steam generator tubes deserve

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special consideration and are presented in the follo:ing section of this report. Briefly, it will be shown that ther= ally-treated Inconel-600 stcan generator tubes have a predicted life expcetancy of at least 28 years even if they are exposed to an aggressive caustic-containing on-vironnent.

Beneficial Effects of Thernally--Treated Inconol-600 The results of research reported to the technical co.mnanity in 1973 provided significant insight regarding the beneficial effects of the2nally treating Incenel-600 Briefly, Blanchet and his co-workers observed a reversal in the usual sensitizing effect in the intergranular/ stress corrosion of Inconel-600 exposed to high-temperature deaerated nator. Cracking did not occur in material which had been heat treated to precipitate carbides at the grain boundaries uhereas high-tenperature annealing (i.e. , mill annealing) lead to cracking in the sane laboratory environments. Subsequently, considerable research was conducted to more completely understand this natallurgical phenonana and how it night be used advantageoasly in nitigating interdranular/ stress corrosion of Inconel-600 steam generator tubes in pressu-rized-water reactors. (17 99) ~~

Domian and his co-investigators 7 conducted experiments wherein highly-stressed U-bend specimens of Inconel-600 were exposed to 650 7, ansonia-hydrazine-treated (AVT), circulating, high-purity water for over 18,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. The results of this research revealed:

1. Intergranular/ stress corrosion of Inconel-600 can occur when the grain boundaries are free of carbides.

2. Cracking is predoninant in natorial annealed at tenperatures above 1600 ?. Crack frequency increases with increasing grain size.

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( 3. Cracking is reduced t;y theraal treatnents uhich produce grain bound-ary carbide precipitation.

4. Cracking did not occur in 18,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> for uterial hich had been thernally treated at 1400 and 1500 F.

5. There was no c.utallegraphic evidence of grain boundary carbide precipitation af ter 13,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> exposure at 650 F.

6. A long incubation period (greater than 13,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) is required for cracking to initiate in highly-stressed caterial.

Later, Airey's research provided further evidence that thermally-treated Inconel-600 would be advantageous for FJR stcan generator tube applications. (18 '90) In part, Airey reported the following :

1. For specimens stressed in tension at 50,000 psi and exposed to deaerated, 10,1 sodium hydroxide (caustic soda) solutions at 600 F, Inconel-600 which had been ther.ully treated in the 1200 to 1300 F range had superior resistance to in'ervra nular/ stress corrosion compared to mill-annealed natorial, bN

2. Maxinun resistance to caustic-induced intergranular/ stress corrosion is associated uith Inconel-500 which has c boundary carbide precipitate (Figure 3) (a18)enicontinuous, grain

3. For plastica 11y-deforned C-ring specinens exposed to deacrated,10j sodiun hydroxide solutions at 600 F, thernally-treated (15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> at 1300 F) Inconel-600 has resistance to intergranular/ stress corrosion uhich can be as nuch as ten tines (or nore) greatop than that exhibited by the nill-annealed product.(Figure 4).(20 f

4. For specinens stressed in tension at values up to 40,000 psi and exposed to deaerated,101 sodium hydroxide solutions at 600 "F for 4,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, the cracks in thernally-treated (15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> at 1300 F) Incenel-600 had propagated only one or two grain depths into the alloy whereas relat4 ve annealed product (Figure 5).120}'y deep cracks existed in the nill-

5. For thernally-treated (15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> at 1300 F) Inconel-600 specinens stressed in tension at 17,000 to 30,000 psi and exposed to deaerated,

10) sodium hydroxide solutions at 600 *F, the crack propagation rate for intergranular/ stress corrosion is so slow that it cannot be meaningfully neasured. For mill-annealed specimens stressed at 20,000 psi and exposed to the sane environnent, the crack pro-pagation rate is approximately 0.050 inch per year (Figure 6). (20)

The early work showing the beneficial effects of ther . ally treating Inconcl-600 has been verified by others. (21) Testing in high-purity water s

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Figure 4 - Crack depth as a function of exposure time for mill-annealed

- and thenr. ally-treated Incone - 0 exposed to deacrated, 10%

N caustic soul'tions at 600 F. 20

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O P M M n 4, . i d Figure 5 - crack depth as a function of stress and natorial conditien (i.e. , mill-annealed and thernally-treated 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> at 1300 F) for Inconel-600 ex solutions at 600

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. l (gesure flee thrst Figure 6 - Crack depth as a function of time, strcss level, and natorial condition (i.e., nill-annealed and thermally treated) for Inconel-600 exposed to deaerated,10,6 sodium hydroxide solutions at 600 F.

containing controlled amounts of dissolved oxygen (0.05 and 8 ppm) at 593

• F, De and Ghosal reported:

l

1. I!ighly-stressed, nill-annealed material was cracked intergranularly i

.. after 1,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> of exposure. '

s l

. 2. Highly-stressed' thernally-treated (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 1100 F) Inconel-600 did not experience interdranular/ stress corrosion during the 1,600-hour exposure.

3. The beneficial effects of the thermal treatnent can be attributed to a semicontinuous-type precipitate in the grain boundaries of the Inconel-600.

The relevance of the 1973-1981 research to the stean generator tuba degradation concern can be appreciated from the 1983 results of Airey and Pement who investigated Inconel--600 tube specimens which had been removed from the hot legs of two operating steam generators. ( They concluded that the most likely aggressive species which causes intercranular/ stress corrosion in operating stean generator tubes is caustic.

Analyses of these( -21) and other data (23) reveal that nill-annealed ,,

Inconel-600 can be expected to experience intergranular/ stress corrosien in high-purity water, high-purity uater containing anmonia and hydrazine, and high-purity water containing sodium hydroxide, providing the environ-ments are sufficiently hot and the specimens are highly-stressed in tension.

It can also be seen that highly-stressed but thernally-treated Inconel-600 has significantly improved resistance to intergranular/ stress corrosion in these clevated-temperature environments.

The relative improvenent in the intergranular/ stress corrosion resistance of ther-ally-treated Inconel-600 is evident from the data presented in Figures 5 and 6. Assuning thermally-treated Inconel-600 is stressed in tensAon at 20,000 psi and continuously exposed to deacrated, 10.5 sodiun hydroxide solution at 600 F, the cracks would be expected to propagate no nore than one or two grain diameters / depths into the alloy (i.e. , about 0.002 inch) in 8,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. For comparison, the cracks in mill-annealed Inconel-500 exposed to the same tensile stress and environ-s.

e

, {' nontal conditions uould be expected to propagate over 0.030 inch in the same time period. Pased upon this comparison, ther.aally-treated Inconel-600 has intergranular/ stress corrosion resistance which is at least 10 times that of mill-annealed material.

Since the Point Eeach Unit Mo.1 stean generators operated approx-inately seven years before any nill-annealed tubes were plugged because of intergranular/ stress corrosion in the tube-tubesheet crevices, the data by Airey suggest that thermally-treated Inconel-600 tubes in the replacenent steam generators could have a life expectancy of 70 years at fu31-power operation (i.e., a hot leg temperature of 597 F) without concern for inter-granula ./ stress corrosion. Although a 70-year life expectancy for thermally-treated Inconel-60O tubes may appear to te overly optinistic, it

,. nust be renembered that it was assu ed that'the tubes uill be ex.oosed to b relatively-concentrated sodiun hydroxide solution for the entire tin.c period and the cracks do in fact propagate beyond one or two grain depths.

Assuming a safety factor of 2.5 for thermally-treated Inconel-600 stressed in tension at 20,000 psi and continuously exposed to deaerated, 10% sodium hydroxide colution at 600 *F, the life expectancy for the tubes in a recirculating P'clR steam generator would be approximately 28 years at full-power operation with regards to intergranular/ stress coiresion degradation.

Secondarv-Side Water Chenistry at Point Fench Unit No. 1 It is uell established that secondary-side water chemistry nonitoring/

control is at least a major factor in nitigating unacceptable steam generator

k. tube corrosion / degradation. Personnel at both the Electric Power Research Institute (EPRI)( ) and IGP00( } fully appreciate this. For example,

j i .

21- -

i J

( . IEPCO Chemistry Standing Order No. C50-8 establishes operational and action levels for the feedwater at Point Eeach Unit No.1.(25) easically, this standing order sets the feedwater pH range at 9.2 to 9 3; the ammonia i

concentration (as NH ) range at 0.50 to 0.70 ppm; and the total conductivity range 3

j at 4.2 to 5.2 mi' 'romhos/en c 3 ; the feedwater is to contain .a minimum hydrazine concentration of 7 ppb. The standing order provides positive direction

if any of these limits are exceeded. In addition, IEPCO has (since about 1978) a secondary-side water chemistry monitoring program which is designed l

to nitigate steam generator tube corrosior4/ degradation and other system 4

corrosion. (26) It is understood that the existing secondary-side water chem-s j istry monitoring progran will be refined in the near future, especially if 4

the steam generators are replaced. (2)

Exact 1v, hou well '.!EPCO personnel have controlled the feedwater chem-i\'

, istry at Point Beach Unit No. 1 can be scen by exanining the data in 1

j Figures 7 and 8 which are reproductions of actual IEPCO logs for 1983.

1 i Excluding the to-be-expected transient conditions, the feedwater during

) 1983 contained less than 5 ppb (the louer limit of the detection equipment) dissolved oxygen, 7 to 12 ppb residual hydrazine, and 0 5 to 0.7 ppm 4

l .anmonia; it had a pH of 9 2 to 9.4 and a conductivity of 4.2 to 5.4 micro-3 t

mhos/cm . Examination of the 1982 data revealed that the feedwater normally '

I contained less than 5 ppb dissolved oxygen, 0 5 to 0.7 ppm ammonia, and 7 to i

15 ppb hydrazine; it had a pH of 9.2 to 9.3 and a conductivity of 4.2 to 5.2 3

micrombos/cm . Similar feeduater data ranges are available for the 1978-1981 l time period.

l WEPC0 personnel have naintained reasonably good control of the AVT -

f secondary-side water chemistry. Secondarj' -side water chemistry control 4

4

---

: ,w ,wm---- ,-~.a ~c, -e - , m,n-.~--ev,, e, ,-----,.ww. .,..we.wm~,..,.w.eg,r,w.w ee, ,, , m e ,m e - r ~~ r-e ,,r.r e w e r s w w.m r.- e

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( IEPCO Chemistry Standing On!cr I:o. CSO-8 establishes operational and action levels for the feedwater at Point Ecach Unit !<o.1. (25) Basically, this standing order sets the feedwater pH range at 9.2 to 9 3; the annonia concentration (as NH ) range at 0.50 to 0.70 ppn: and the total conductivity range 3

at 4.2. to 5 2 micronhos/cn3 ; the feedwater is to contain a ninimun hydracine concentration of 7 ppb. The standing order provides positive direction if any of these linits are exceeded. In addition, IEPC0 has (since about 1978) a secondary-side water chemistry nonitoring program which is designed to nitigate stean generator tute cerrosion/ degradation and oth;r system corrosion. (26) It is understood that the existing secondary -side water chen-istry nonitoring progran will be refined in the near future, especially if the steam generators are replaced. (2)

Exactly how well ',GPCC personnel have controlled the feedwater chen-i istry at Point Beach Unit !!o.1 can be seen by exanining the data in Figures 7 and 8.which are reproductions of actual UEPCO legs for 1933.

Excluding the to-te-expected transient conditions, the feedwater during 1983 contained less than 5 ppb (the louer limit of the detection equipment) dissolved oxygen, 7 to 12 ppb residual hydrazine, and 0.5 to 0.7 ppm annonia; it had a pH of 9.2 to 9.4 and a conductivity of 4.2 to 5.4 nicro-3 nhos/cm . Examination of the 1982 data revealed that the feedwater normally

• contained less than 5 ppb dissolved oxygen, 0.5 to o.7 ppn am=onia, and 7 to 15 ppb hydrazine; it had a pH of 9.2 to 9.3 and a conductivity of 4.2 to 5.2 micronhos/cm 3 . Similar feeduater data ranges are available for the 1978-1981 time period.

WEPC0 personnel have naintained reasonably good control of the AVT secondary-side water chemistrj. Secondary-side uater chenistry control

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Figure 8 - Condnetivity (C), residual hydrazine (H), and a.conia (A) data for

~

the feedwater at Point Beach Unit No.1 during 1933.

O e

, _2!4 .

( has undoubtedly been sonewhat handicapped by the phosphate residues which still exist in the systen and a lack of continuous monitoring sites.

Very likely, further inprovenents in the water chenistry nonitoring progran (e.g., additional continuous monitoring at locations throughout the secondary-side of the systen) will be established if the stean generators are replaced.

Anticinated "'ube Plu; rine and Associated Pouer-Outout Reductions for the c.xistine Stcan Gen-orators at Point Ecach Unit :;o. 1 F.xanin2 tion of the Stean Generator Tube Plugging History for Point Ecach Unit No.1 (see Appendix) reveals that the only anticipated, corrosion-related reason for the future plurging of tubes in the existing stean generators uculd be intergranular/ stress corrosion in the tube-tubesheet i -

crevices. The other corrosion-related tute degradation phenomena .

have been effectively citigated (Figure 9). Analysis of the intergranular/

stress corrosion (crevice) data in Figure 9 between February 1930 and Oct--

ober 1932 provides the only neans a L the present tino of predicting future tube plugging at Point Beach Unit ::o.1. Assuning a linear extrapolation of these data on a senilogarith~.ic plot and allowing a correction factor for the temporary operation at a hot-leg tenperature of 575 F, it can be roughly estinated that an additional 80 to 85 tubes uill require plugging during the next 2.5 years of operation (i.e. , between June 1983 and December 1985); poss-ibly another 110 to 120 tubes will require plur;ging betueen Decenber 1985 and Decenter 1990. It must be emphasized that these predictions are based upon continued operation at 77% of full power with a hot-leg tenperature of 557  ?.

t, ' ! ~.

CUKGLATIVE NUMBER OF TUBES PLUG 3ED FOP, CORROSION PHZ'40.'ENA ,

2 -

g il - g 8 ~8 2 o I I IIi w .

l 1 1 I I Iiij l i I i 1.1 I l l l 1 1

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7 ooo E c: F

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-*--- Ma r 82 i c Det182 I I I IIII F i I i I I I III I I I I I III to '

i I

26-i

)( . The need to further reduce the power output (i.e. , from 77% of full I

l power) at Point Beach Unit No.1 because of additional steam Cencrator j tube plugging cannot be accurately estimated. Based upon experience, it is reasonable to assune that (after a reasonably large nunber of tubes have ,

f aircady, been plugged) there vill be a 0 5 to 0.7 j reduction power output for each additional one percent of tubes plugged.1 Based upon this hypothesis, i'

the power output by December 1985 could be reduced to about 75% of the full-power rating; by December 1990, the unit could be operating at about l

731 of tha full pouer rating.  :

Recessity for Fanlacine Eaui?nent at Point j

Beach Unit l'o. 1 Other Than the Stean Gen-erators 1

K Deterioration of the 90Cu-10Ni (Copper Alloy 70600) tube interiors in 1

the No. 4 feeduater heaters by general corrosion (undoubtedly ammonia re-J

lated) has resulted in a large number of tubes being plugged in these units at Point Beach Unit No.1.(2,27) The tube plugging has resulted in higher i

, vator velocities in the renaining tubes and crosion-corrosion damage. l i Similar deterioration has not been observed in the other feedwater heaters (according to the results of recent tube uall thickness measurements). (2)

Very likely, the No. 4 feedwater heater corrosion is a major source of the copper contamination which is occasionally observed in the secondary-side water / sludge.

Replaconent of the No. 4 feedwater heaters with units, tubed with Tjpe 304 stainless vill eliminate both the antonia-induced general corrosion and the Each of the Westinghouse yodel 44 steam generators at Point Ecach Unit

No. I'contains 3,260 Inconel-600 tubes. ,

- _ . , _ _ _._.. . _ , __ _ _ _ _ . _ . _ . _ . - - . _ . _ _ . _ . _ . _ _ . ~ . _ . _ . _ . - _ _ . - , . . . - ,

l

(' high-velocity-water induced erosion corrosion.

x -

Very likely, the l'o. 5 feedwater heaters do not exhibit general corrosion because they are tuted with 80Cu-20 :1 naterial (Copper Alloy 71000). The other feedwater heater tubes probably have not correded to any appreciable extent because they operate at lower tenperatures.

There is no obvious reason to repitee any feedwater heaters other than those proposed.

It is also known that thernal fatigue has been a problem of concern in the Coppar Alloy 70600 noisture separator reheater (MSR) tubes.( ' The problen is apparently associated with temperature differentials, uneven tube expansion / contraction, and tube binding.

Replacing the copper-nickel tubes with Type 439 stainless steel, en-lard ing the tube-support holes, and reducin'g the tube-support spacing to r

'V 25 inches should effectively nitigate the MSR themal-fatigue problen.

There is no reason to believe that the main condensers should be retubed at the present time. Only about 10 to 12 condenser tubes (out of a total 24,000 condenser tubes) have been plugged each year since the fatigue-related condenser tube problen was corrected in the early 1970s. (28) Con-sideration should be given to retubing the nain condenser only if an un-anticipated, unacceptably large number of tubes nust be plugged in futurayears or condenser-tube corrosion is found to be a source of deleterious copper ingress into the secondary-side water. 'dith regards to the latter, the data at Point Beach Unit No. 1 do not suggest that condenser-tube corrosion is a najor source of copper contanination. Furthce, it has not been fimly established that copper is in fact an actual factor in Inconel-600 tube corrosion / degradation,

, f Necessit" for gondensate _ Poli shinc

\- -

at Point 5'ach Unit No. 1 Fixed success has been achieved with condensate polishing (cendensate denincralization) at the operating units which have installed them. There is no reason to believe that they should be routinely installed without question.

For example, there were 11 PJR units which had operated over 1,000 effective full-power days (EF?Ds) that did not experience a single steam generator tube problen/ defect during 1973 and 1979; of these 61+j uere on AVT,18% uere on phosphate treatment, and 18) were on AVT uith condensate demineralitation for the secondarf-side water treatm.ent. (7) AVT can be a viable chenistry control program for the secondary-side water uithout condensate' polishing. (7'9}

Condensate polishers are ba,1cally a desirable option for plants uhich v.

{ use sea or brh l:ish water for condenser cooling and/or have a high incident rate for condenser inleakage. Neither of these conditions exists at Point Beach Unit No.1. Further, there is aluays concern from resin carryover which could possibly introduce an accressive species into the secondary-side water when condensate polishers are included in the secondary-side uater system.

There is no obvious reason to install condensate polishers at Point Beach Unit No.1 under the present or anticipated operating conditions.

Conclusions Eased upon the results of this comprehensive, updated study of steam generator tubo degradation / corrosion at Point Beach Unit No.1, it can be concluded:

- (_ ' 1. There should be no major future concern for stean generator tube corrosion / degradation by either fretting, fatigue, crosion, pitting, r

denting, wastage, U-bend cracking, or intergranular/ stress corrosion

.{ ,

above the tubesheet with either the existing or replacement stean gene rators. 2

2. Continued operation at 77,5 of full power requires that the hot-leg temperature for the existing stean generators never exceeds about 557 F.

3. There is reason to believe that some additional stean generator tubes uill require plugging because of intergranular/ stress corrosion in the tube-tubcsheet crevices associated with the existing units.

It is believed that future plugging could require the existing units to be operated at about 73 I of the full-pouer rating by 1990.

4. There should be no significant intergranular/ stress corrosion of the thernally-treated Incorel-600 tubes in the replacement steam generators for at lesst 23 years even if they were continuously exposed to a hot caustic environment.

5. WEPC0 personnel have established a reasonably viable secondary-side water chemistry program. They have achieved good success in maintaining the chemistry limits set for the feeduater in 1978

6. Considerable attention has been given by WEC engineers in the design and naterials selection for the reolacement stcan generators with

( regards to mitigating the known corrosion problems. Tneir materials selections is supported by data contained in the technical literature.

7. Replacing the No. 4 feedwater heaters and the XSRs should significantly reduce the amount of copper ingress into the secondary-side water. The stainless stecls selected for use in these two systens are con-sidered acceptable for the intended applications.

8. The design changes proposed by the WEC engineers should eliminate the fatigue problem in the existing MSRs.

9. There does not appear to be any obvious reason to replace the No. 5, No. 3, No. 2, or No. 1 feedwater heaters at the present tine.

Similarly, there is no reason to retube the condenser.

10. It has not been established that small amounts of copper in the secondary-side water have an adverse effect en steam generator tube corrosior/ degradation.

11. There is no obvious reason to install condensate polishers (con-densate denincralizers).

12. The improved-design replacement steam generators should have at least 2

This conclusion assumes 773 full-power operation for the existing steam

_(. genera tors, 100% full-power operation for the replacement steam generators, and proper control of the secondary-side water chenistry.

s

, I'

• a 90 to 953 probability of achieving a nininum 25 years of oporation at full power providing the AVT seco:dary-side water chenisty progran is refined and rigorously controlled.

13. There should be no significant tube plugging in the replacencnt stean generators providing the AVI secondary-side water chanisty progran is refined and ri;orourly t controlled.

14. At least ninor consideration should be given to stean generator re-placement in order to preclude the re.T.ote possibility of continued corrosion object danage of plugged (F03) totubes and nearby the associated unplugged tubes. p(ot.qntial 13/ for foreign

'k i

'~,

I f

t . ,,

-31 .

i Pa ferences 1

- 1. J. R. !@crs, "Stoan Generator Tube Corrosion, Point Beach Unit lio.1, Wisconsin Electric Power Conpany, Two Creeks, Wisconsin," report pre-pared for the Public Service Conmission, Stato of Wisconsin, March 22, i 1981 j

i 2. D. K. Porter (UEPCO), private conmunication to J. R. %ers and T.-L. ,

j Poon, June 1, 1983.

3. "::cw York Asks !!RO to Ease Rule to Let I:uclear Power Plant Operate," f

The tall Street Journal, yarch 7,1933. -

.i

4. G. Charnoff, letter to 3. E. James, :hy 3,1983, i

l 5. Point Ecach Nuclear Plant No.1, Steam Generator Repair Report, Wis-consin Electric Power Company, Milwaukee, Auguct 1982.

l 6. B. W. Churchill, letter to the Ihclear Regulatory Commission, Washington, D.C. , April 27, 1983.

7. C. W. Fay (WEPCO), letter to H. R. Denton (1:RO), May 27,1982.

( 8. O. S. Tatone and R. S. Pathania, " Steam Generator Tuha Performance:

Experience With Water-Cooled !;uclear Power Reactors During 1979," Atomic Energy of Canada Limited Report No. AECL-7251, Chalk River, Ontario, l March 1981

9. D. G. Eisenhut, B. D. Liau, and J. Strosnider, " Summary of Operating Experience Uith Recirculating Stean Generators," U. S. Nuclear Reg-ulatory Commission Report No. NUREG-0523, Washington, D.C. , January

1979.

I 10. R. T. Begley, " Steam Generator Xstorials Evaluation," paper presented at the Stean Generator Systens Symposium, Tampa, Florida, September 1977.

l

11. A. R. Vaia, G. Economy, M. J. Wootten, and R. G. Aspden, " Corrosion Per-fornance of 12j Chromium Stainless Steels in High-Temperature Chloride

Solutions," paper presented at Corrosion /79, Atlanta, Georgia, March 1979.

12. W. E. Gunson (WEC), letter to J. R. :@crs, June 2,1983.

13. E. M. Blake, "The Effort to Prevent Tube Degradation," !Melear Uews, pp.

49-53, December 1982.

4

14. J. Hickling and H. Wieling, "Electrochenical Investigations of the Re-sistance of Inconel-600, Incoloy-800, and Type 347 Stainless Steel to Pitting Corrosion in Faulted ?da Secondarv Water at 150 to 250 C,"

Corrosion, Vol. 37, pp.147-152 (1981). '

{

l

• *

• i

( 15. J.1hkansi, " Protecting Today's Systens for Long-Tern Reliability,"

Power, Vol. 124, "o. 4, pp. 5-1 to S-24, April 1983.

16. J. Blanchet, H. Coriau, L. Grall, C. Xahicu, C. Otter, and G. Turluer,

" Influence of Various Parameters on Intergranular Crackin.: of Inconel-600 and X-730 in Pure Water at Elevated Tenparatura," proprint G-13 of paper presented at the Stress Corrosion Crceking and Rf d rogen Enbrittle-nont of Iron Base Alloys Synposinn, Firaing, France, June 1973.

17. H. A. Domian, R. H. Ennuelson, L. W. Sarver, G. J. Theus, and L. Katz, "Effect of Microstructure on Stress Corrosion Cracking of Alloy 600 in High Purity Water, " Corrosion, Vol. 33, pp. 26-37 (1977).

18. G. P. Airey, " Microstructural Aspects of Thermal Treatment of Inconci 600,"

Metallor raphy, Vol. 13, pp. 21 -41 (1980).

19. C. P. Airey, "The Effect of Carton Content c.nd Thernal Treatnent on the SCC Schavior of Inconel Alloy t'00 Stean Generator Tubing," Corrosion, Vol.

35, pp.129-136 (1979).

20. G. P. Airey, "Effect of Proccasing Variables on the Caustic Stress Corrosion Resistance of Inconal Alloy 600," Corror, ion, Vol. 36, pp.

9-17 (1980).

21. P. K. De and S. K. Ghosal, "A Comparative Study of Stress Corrosion

(' Cracking of Stean Ganerator Tute .hterials in Water at 315 C," Corrosion, vol. 37, pp. 341-349 (1981).

22. G. P. Airey and F. W. Penent, "A Comparison of Intergranular Attack in Inconel Alloy 600 Observed in the Laboratory and in Operating Stean Generators," Corrocion, Vol. 39, pp. 46-55 (1933).

23. D. Van Rooyen, " Review of the Stress Corrosion Cracking of Inconel-600,"

Corrosion, Vol. 31, pp. 327-337 (1975).

24. P4R Secondary Water Chenistry Guidelines, Electric Fouer Research Institute, Special Report No. HP-2704-SR, Palo Alto, California, October 1932.

25. Fecduater pH, Conductivity, Annonia Routine Operational Guidelines and Action Levels, WIPCO Standing Order Mo. CSC-3, Ravision No. 2, July 2, 1932,

26. Secondary Water Chenistry Monitoring Program, WEPCO Report No. PEMP-8.4.1, Revision No. 2, July 2, 1982.

27. N. A. Ricci (UEPCO), letter to J. K. Reynolds (PSCW), March 15, 1983.

28. D. K. Porter (WEPCO), letter to S. Jenkins (PSCW), Occomber 30, 1982.

REPLACEME!!T STEAM GE!!ERATOR EQUIPME!!T

.TI!CHMICAI7 DESCRIPTION Wisconsin Electric Power Company will repair the Point Beach Nuclear Plant Unit 1 steam generators by replacement of the present louer assemblies with two Westinghouse Model 44F steam generatcr louer assemblies and refurbish _the primary moisture separator equipment of the present steam generators.

The design objective for the replacement steam generator equipnent is to provide the equiv ~ .nt performance iof the equipment being replaced. However, many design improve-

.ments are included that are intended to improve the flow distri-l

'bution, corrosion. improve tube bundle access, and reduce secondary side The replacement 1cwer assembly includes the following features:

1.

A cast channel head will be used'. Improvements incorporated will alter the weld preps on the primary nozzles to f,acilitate inservice inspection.

2. A primary shell drain will be incorporated at the base of the channel head to improve drainage.

3.

The channel head support pads will be identical to the present Point Beach design.

4.

The tubesheet previous Serieswill 44 be the same tubesheet. dimensions as the Plush tube-to-tubenheet welds will be used in conjunction with full depth tube expansion for all tubes to eliminate tubesheet crevices.

5. The secondary shell up to and including the transition cone will incorporate additional shell penetrations.

I a.

Four.six-inch handholes will be placed in the secondary shell just above the tubesheet-to-shell

, weld seam.

b.

Two additional six-inch handholes will be placed in the stud barrel just above the flow distribution baffle. These openings will be 180 i apart and on i the tubelane. -

c.

One three-inch handhole will be located in the shell at the elevation of the top tube support

{' plate.

- __ ..-_____--__u______.__- _ _ _ _ _ _

i

6. A welded tubelane blocking device will be installed to limit tube bundle bypass flow. Its design will be such that it shall not hamper sludge lancing.

7. A flow distribution baffle will be placed approximately 18-20 inches above the tubesheet. 'This baffle will be made of ferritic stainless steel (as will all of the tube support plates). The purpose of this baffic will be to direct the recirculation water across the tubesheet to the center of the bundle. Here any sludge will be deposited in a limited region near the blowdown intake. -

8. The tube support plates will have a broached hole pattern using the quatrefoil design. This design has

,a smaller pressure drop than the most current circular

hole designs.

9. The Inconel-600 tubes will be thermally treated. The tube dimensions are 7/8-inch O.D. with 50 mil wall thickness.

10. Increased capacity blowdoun will be provided to enlance maintenance of secondary side chemistry. In order to provide this capability, an enlarged blowdown pipe will be provided along with an increased size blowdown nozzle.

11. Tubesheet markings to improve tube identification will be provided on the primary side of the tubesheet

indicating the locations of selected tubes. The tenth tube in every row and column will be marked so as to form a 10x10 array of marked tubes.

12. The secondary surface of the tubesheet will be machined to aid in the accurate and uniform definition of tubesheet thickness.

~

13. The height of the wrapper above the tubesheet will be reduced to improve the flow characteristics near the secondary side of the tubesheet.

14. Downcomer resistance plates will be eliminated to improve the circulation ratio.

15. Drainage. holes will be included at the location of '

the primary manway openings to allow for the drainage of primary water from that area prior to opening of the manways.

., .s o s. .

(

16. All tubes in the innermost rows will be thermally stress relieved after tube bending. -

17. The leak tightness of the tube-to-tubcuheet weld will be confirmed by leak test. ,

18. The heat transfer characteristice for these units will be the same as the original units.

.19. A preservice baseline inspection.in accordance with Section XI of the ASME Code will be accomplished after the steam generator has been installed in the coolant loop and a secondary side hydrostatic test has been performed.

In addition to the replacement of the louer assembly, the following modifications to the upper stcam drum assembly '

are included:

1. The addition of a wet layup nozzle to the upper shell designed for a tvo-inch pipe connection is include 2.

2. In order to increase the moisture separation capdbility of the steam generators, the existing primary moisture separatcrs will be changed out to a more efficient separator design. The replacement primary separator package contains swirl cane assemblies attached to an

' upper and lower dock plche. The primary separator-package vill be suppor ed off the tube bundle Grapper.

The drains from the secondary separator will be enlarged and rerouted through the.-primary separator assembly. The pattern of swirl vand assemblies requires the installation of a replacement feedwater distribution ring with 9 different shape than the existing feeduater ring.' The feedwater distributton ring will be welded to the feedwater nozzle and sGpported off the shell. The n e w - f e e d v a *. e r ring wi'l1 have inverted "J" tubes arranded so that a,large porti5n of the feedwater flow will.be directed toward the hot leg side of the tube bundle, ,

x 4

3. A flow restrictor will be installed in the steam outlet nozzle. ~y

4. In addition, the Peericus' vanes that comprisc,the secondary moisture sdparators will be ~ replac'ed.' .

1

-- 4

/

4

.. . . . . . . . . _ . . . . . . - . . . _ - . ...n - -

,. . s l

1 .

O ,

l &

r ,.

r e

. y

' '/ UNIT 1 STEAM GENERATOR TUBE PLUCGING HISTORY Tubes Plugged DIte of '-

Thinning or Crevice Cumulative Outage - Elapsed Tin,e Denting Cracking Corrosion Othe'r _ Total _rercent (Years) . & E & E h E & E d' E h E 12/21/70 1(1)

~

0 .

1 -

<0.1 0 9/30/72 1.8 - -

. 87 91 - -

14 4(2) 102 95 3.1 2.9 4/6/74 3.3 - -

1 1 - - - -

103 96 3.2 2.9 2/26/75 4.2 - -

59 98 - - - -

162 194 5.0 6.0 11/16/75 4.9 - -

6 4 - - - -

168 198 5.2 6 .1 -

10/1/76 5.8 - - - - - - - -

168 198 5.2 6.1 6/24/77 G.5 - - -

• 1 - - - -

168 199 5.2 6.1 10/4/77 6.9 10 - - -

1 2 - -

179 201 5.5 6.2 2/1/78 7.1 - - - - - -

1(3) -

180 201 5.5 6.2 5/26/78 7.4 - -

1(3) -

'181 201 5.5 6.2 s, 9/20/78 7.7 1 - - -

6 4 - -

188 205 5.7 6.3 3 3/1/79 8.2 - - - -

8 1 - -

196 206 6.0 6.3 8/5/79 8.6 - - - -

52 45 - -

248 251 7.6 7.7 8/29/79 8.7 - - - -

2 -

2 (4 ) -

252 251 7.7 7.7 10/5/79 8.8 - -

2 3(6) 68 61 7 4(5) 329 319 10.3 9.8 12/11/79 9.0 - - - -

19 15 1(7) -

349 334 10.7 10.2 2/28/80 9.2 - - -

1(8) 24 26 -

9(9) 373 370 11.4 11.3 7/26/80 9.6 - - - -

20 22 3(10) -

404 392 12.4 12.0 11/26/80 9.9 - - - -

3 7 - -

407 399 12.5 12.2 7/4/81 10.5 - - -

1(8) 2 2 - -

409 402 12.5 12.3 10/9/81 10.8 - - - -

9 7 - -

412(11)409 12.6 12.5 3/25/82 11.3 - - - -

37 14 2(12) -

453 423 13.8 13.0

'10/22/82 11.9 - -

2 4 4 4 12(13) -

469 431 '14.4 13.2

o e- U,

./ $.

e 1

NOTES (1) Plugged during manufacture.

(2), Fourteen tubes in A were plugged due to gouging during nachining for clad repair. Three tubes in B were removed for analysis and one was plugged by mistake.

(3) Plugged tubes were in periphery and were leaking. During the October 1982 outage, these leaks were found to be due to wear by loose parts.

(4) An audit of tubecheet photographs indicated two tubes which we e plugged but previously not included in inspection reports.

(5) Seven tubes in A included three with defects less than the plugging limit, two tubes which had no indications but which were pulled for analysis, and two tubes plugged by mistake. Four tubes in B included three tubes with indications less than the plugging limit and one tube plugged oy nistake.

(6) Two tubes in A and three tubes in D were plugged due to defects identifiad at or above the tubenheet using reulti-frequency eddy current techniques. These defects are attributed to thinning or cracking in prior years, based upon comparison with sing]e-frequency eddy current results from previous inspections.

(7) One tube plugged by nistake.

(S) One tube in D was plugged due to a def ect above the tubecheet which was identified using multi-f requency techniques.

This defect is attributed to thinning or cracking in prior years, based upon comparison with results frcm previcus {

intpections. 3 (9) Four tubes in B were plugged due to the possibility of being damaged during tube pulling operations and five leaking tubes were plugged without identifying the defect location.

.,(10 ) Three tubes plugged by mistake.

(11) One tube which was in excess of the plugging limit was repaired by sleeving. Plugs were renoved from s3x tubes and the tubes were sleeved and returned to service.

(12) Two tubes were leaking and were plugged. Eddy current inspection revealed no indications. One of the tubes was leaking on the cold leg end from which one explosisc plug was removed during the sleeving demonstration in 1981.

(13) Twelve tubes include eleven plugged for suspected damage from Icase parts in the cold leg and one sleeved tube from which the sleeve vos removed for inspection.

_ . . - . . _ _ _ _ _ _ _ _ _ _