ML20027E317

Contentions Re Steam Generator Replacement,Supplementing 820810 Petition to Intervene

ML20027E317
Person / Time
Site:	Point Beach
Issue date:	11/05/1982
From:	Patricia Anderson WISCONSIN'S ENVIRONMENTAL DECADE
To:	Atomic Safety and Licensing Board Panel
Shared Package
ML20027E316	List: ML20027E315 ML20027E317
References
NUDOCS 8211150002
Download: ML20027E317 (25)

Text

t DOCKETED USfRC UNITED STATES OF AMERICA

'82 N]V 10 60:57 NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board Wisconsin Electric Power Company '

POINT BEACH NUCLEAR PLANT UNIT 1 DOCKET NO. 50-266 Operating License Amendment 2 (Steam Generator Replacement Proceeding)

DECADE'S CONTENTIONS CONCERNING STEAM GENERATOR REPLACEMENT s

Pursuant to 10 C.F.R. S2.714 (b), and the criteria set forth in Re Cleveland Electric Illuminating.CDEoany, 14 N.R.C. 175, 184(1981), Wisconsin's. Environmental Decade, Inc. (" Decade") ,

herpby supplements its Pet'ition to Intervene and Petition f or Hearing, dated August 10, 1982, by listing the contentions which, at tnis time, it seeks to have litigated.

The Decade makes this' filing without waiving its previously stated objec' tion to the scheduling order of the Atomic Safety ana l

Licensing Board (" Board") that req tires an impecunious party to make a detailed filing in this proceeding at the same time it is required to prepare direct testimony in a companion proceeding, Docket 50-266 OLA-1. In order to avert legal waiver of the right to participate in this proceeding, it has been necessary f or us to involuntarily relinquish the opportunity to secure af firmative testimony in the companion proceedings, because both tasks could not be accomplished simultaneously. The objectionable nature of the scheduling order is compounded f urther by the f act tnat the B211150002 821105 PDR ADOCK 05000266 g PDR WED-PA-13/05/82-P:50266NRC.P53-5

. 5 Licensee has made no showing, nor has it been required to make a showing as we requested, to demonstrate the need for such an expeditious schedule that has as its intended impact the f atal impairment of the interests of the opposing citizenry.

Although listed separatel,y for ease of reference, the following contentions are all interrelated with each other and shouldf be considered as such.

T FIRST CONTENTION Tube Failures Under LOCA Accident Conditions Contention Degradation of as f ew as one to ten steam generator tubes in either the existing or the proposed steam generators at Point Beach Nuclear Plant Unit 1 (" Point Beach") could induce essentially uncoolable conditions in the course of a loss-of-coolant-accident ("LOCA") , a condition which was not considered by the Nuclear Regulatory Commission (" Commission") with regard to the existing generators prior to licensing the f acility, in the Final Saf ety Analysis- Report _or in any subsequent license amenament proceeding, nor which is addressed in the application for the propo' sed generators.

These factors from secondary-to-primary leakage through degraded steam generator tubes act to lower the thresho,ld for admitt3ng contention.s such as to make a matter with a low probability justiciable,, even if it might otherwise not be so, due to the large c,onsequences from its occurrence. Further, inasmuch as these factors were not evaluated as part of the original operating license, it is necessary that they be evaluated in this proceeding to amend the operating license prior WED-PA-ll/05/82-P:50266NRC.PS.5-5

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to its being approved.

Basis "The basis for our concern about the present course of actions being pursued by the task force * *

• lies in the indeterminancy of the adequacy of the present code formulations. ** * [A] clear demonstration of coolability by wide margins is necessary to satisfy this uncertainties [ sic] regarding the ECCS capability; that is, cooling by narrow margins would have to be regarded by him as an essentially uncoolable situation. * *

• Some of the essential areas of uncertainty in predicting ECCS performance are reflooding and steam binding. * *

• Of paramount concern in this area, however, is the possible ef f ect of steam generator tube f ailures on the ECCS." REG ECCS Task Force, Memorandum to ECCS Task Force Members, dated June 16, 1972.

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"[I]t was the consensus of the [hmerican Physical Society] group that steam generater tube f ailure during a severe LOCA could occur f requently. Moreover, it appears that rupture of a few tubes (on the order of one to ten) dumping secondary steam into the depressurized primary side of th reactor system could exacerbate steam binding problems ano induce essentially uncoolable conditions in the course of a LOCA * * "." Report to the American Physical Society b by the Study Group on Light-Water Reactor Safety, 47 Review d Bodern Physics (Summer 1975), at p. S85.

"Furthermore, se'rious weakening of these tubes f rom similar causes (of tube degradation) could, in the event of a loss-of-coolant-accident (LOCA), result. in tube failures

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i that would release the energy- of the secondary system into the containment." Regulatory Guide 1.83 (Rev.1), at p. 1.

"If the shock loads imposed by the LOCA cause A critical number of tubes to fail, say by a double ended (guillotine) break, the inflow f rom the secondary side can cause choking of flow during ECC preventing adequate cooling of the core. The critical number of tubes is relatively small." Office of Nuclear Reactor Regulation, EEC Procram 191 .thR Resolution af Generis Issues Related .ta Nuclear 22HAI Plants, NUREG-0 410 (197 8 ) , at p. C-29.

"The failure of a number of steam generator tubes as a result of the pressure transients during a loss of coolant accident could. render the emergency core cooling system iner f ec tive." Risk Assessment Review Group, Report .ta .thR L L Nuclear Reaulatory Eggmission, NUREG/CR-0 4 0 0 (197 8) , at

p. 48.

"Recent studies have shown that as few as ten tubes l

would need to have ruptured during a LOCA (assuming a l

leakage rate of 130 gal / min per ruptured tube) before the l

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cladding temperature would be significantly affected (i.e.

peak cladding temperature (PCT) [ greater than] 2 20 0

• F) ."

Evaluation nf Steam Generator Tube Rupture Events, NUREG-0651(1978), at p. I-2 (" Event Evaluation Repo r t") .

"One area [of research] that has not been considered sufficiently using recent accident analysis codes is estimation of tne consequences of a transient or some otner failure that might lead-in turn to the failure of a significant number of tubes. Such f ailures could lead to the degradation of 'ECCS f unction." Office of Reactor Safety Research Group, Report in tJul President's Nuclear Saf ety oversight fnamittee(1981), at p. I-2.

1 "The consequences of multiple tube f ailure, excess ef the design base, have not yet been rigorously studied. ***

In tne event of a LOCA, the core reflood rate could be retarded by steam binding. * *

• S[ team] G[enerator] tube f ailures would create a secondary to primary leak patn wnich aggravates the steam binding effect and could lead to inertective retlooding of the co r e." Nuclear Reactor Research, Steam Generator Status Report (Feb.1982), at p. 2 to 3 (" Status Report") .

"At the times Point Beach Unit 1, Surry Unit 2, and Prairie Island Unit 1 were licensed, there were no specific analysis requirements for S[ team] G[enerator] T[ube] rupture event s. * * *

"The starf does not require licensees to' analyze loss-of-coolant accidents (LOCAs) concurrent with an SGT break, but does require all LOCA analyses to include the effects of the pl.ugged tubes'on_ reduced _ RCS flow." Event Evaluation Report, at p.1-2.

l l "The purpose of tnis section is to evaluate the impact, if any, of the repaired [ sic] steam generators on the l

' accident analysis transients for Point Beach Unit 1. Under the guidelines specified in 10 CFR 50.59 such an evaluation is required to verify that no unreviewed safety concerns or changes to the Technical Specifications occur. This section provides a qualitative discussion of the effect on the

' accident analysis of steam generator parameter changes from steam generator repair. Conclusions are made concerning the applicability of the original F[inal) S[afety] A[nalyses]

R[eport] to the repaired [ sic] unit. Consistent witn the requirements of 10 CFR 5 0.5 9, licensing regulations and guidelines of the original licensing of the Point Beach Unit I are assumed to apply, and Dnig changes in the safety analysis due to equipment changes are conside r ed."

Wisconsin Electric Power Company, Steam Generator Report 12I Point Beach Nuclear Plant Unit 1, Docket 50-266, dated August 1982 (" Steam Generator Replacement Report"), at p. 5-1.

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• SECOND LITIGABLE ISSUE Tube Failures Under Normal Operation Conditions Contention Rupture of steam generator tubes during normal operation may release radiation to the environment from the pl an t's secondary side in excess of maximum permissible doses to the extent that:

(a) Iodine. The iodine levels in the primary coolant exceed presently effective Westinghouse Standard Technical Specifica* ons for reactor coolant iodine activity.

3 (b) Unconsidered Leakace. The primary-to-secondary leakage is greater than bounded in the Final Safety Evaluation Report f or Point Beach ("FSAR") or in the Steam Generator Replacement Report due to such things as multiple b tube f ailures or single tube ruptures greater than assumed in tne design basis analysis.

(c) Safety Valve. The secondary side safety valve set point is exceeded and does_not properly. reseat for an extended period.

(d) Main Steam Ling Break. Primary leakage through a ruptured tube overfills the steam generator and floods the main steam line with water that causes a main line steam break. .

(e) Condensor. ,

The condenser is removed from service during a tube rupture accident due to mechanical or economic reasons and the iodine partitioning function is lost.

These factors from primary-to-secondary leakage though degraded steam generator tubes act to lower the threshold f or WED-PA-ll/ 05/ 82- P : 5026 6 NRC . P53-5

s admitting contentions such as to make a matter with a low probability justicible, even if it might otherwise not be so, due to the large consequences from its occurrence. Further, inasmuch as tnese f actors were not, in large part, evaluated as part of the original operating license, it is necessary that they be evaluated in tnis proceeding to amend the operating license prior to its being approved.

't Basis Iodine "A steam generator tube rupture occuring while the reactor coolant iodine concentration is above the level Standard Technical allowed by the Westinghouse Specifications could cause of f site doses exceeding 10 CFR Part 100 guidelines." Saf ety Evaluation Report Relating to Full Scale Steam Generator Tube Sleeving at Point Beach Nuclear Plants Unit 1 and 2, Dockets 50-266 and 50-301 OLA-2, undated but presumed issued July 8, 1982 ("Sle eving S ER") , at p. 43.

s Unconsidered Leakage "The consequences of multiple tube f ailues, in excess or tne design base, have not yet been rigorously studied.

Rapid degradation- between inspections of a 'large number of tunes' could create the potential for multiple tube f ailures in the event of a plant transient or failure of a single tune and tne accompanying jet impingement and the tube whip could cause f ailure of additional tubes. Furthermore, the potential for complicating circumstances involving multiple equipment f ailures such as the stuck open PORV during the Ginna incident and possible steam bubble f ormation _in the primary system have not been evaluated. Another concern is ruptures in multiple S[ team] G[enerators]. In this event, unless the plant can be rapidly depressurized and' brought onto Residual . Beat Removal, there is the potential to continuously lose emergency core cooling water outside of containment." Status Report, at p. 2.

"The FDSA [for Ginna Nuclear Plant] predicted that with a douole ended guillotine break of a single steam generator tube, the primary to secondary leak rate would be about 843 gpm. The initial leak rate at Ginna was calculated to be about 760 gpm, even though the break was not a double ended guillotine break.

"During the January 25, 1982 incident at Ginna, the total amount of primary-to-secondary leakage and the total WED-PA-ll/05/82-P:50266NRC.P5.$-5

I amount of water and steam released to the environment were larger than would normally be predicted, because of valve malfunctions and operator actions (see Chapters 3, 4, and 5 of NUREG-0909). A comparison with a previous safety evaluation report input on the radiological consequences of a steam generator tube rupture accident (SGTR) (ref 8.3) shows that the potential exists for doses exceeding Part 100 Guidelines from a design-basis SGTR accident. * *

• Although a more serious event was avoided and the radioactivity releases were not excessive, the staff concluded that additional measur'es must be taken to prevent potential accidents in the future from having similarly large leakages and releases that could cause more severe radiological consequences. " Ginna SER, at p. 3-7.

Safety Valve "In the actual event [on January 25, 1982 at Ginna Nuclear Plant], the damaged B steam gerlerator safety valve lifted five times. After the last operation, the safety valve apparently f ailed to fully close. This f ailure was not explicitly considered in the FDSA, nor was its effect of continued primary-to-secondary leakage." Ginna SER, at p.

3-7.

"However," a slow RCS pressure reduction may, without b careful operator attention, result in opening of the damaged SG ADV or safety valve (SV). The resulting offsite doses could be significantly greater than those experienced [in prior tube rupture events]." Event Evaluation Report, at p.

2-4.

Fiain. Steam Line Break "The PDSA assumed reducing safety injection flow to drop the RCS pressure to below 1100 psi at either one hour or four and one-half hours. In the actual event, plant pressure fluctuated above 1100 psia until two hours and 10 minutes af ter the tube rupture when it was bought [ sic] to 1100 psig. This resulted in lifting the damaged steam

• generator saf ety valves a number of times. Even af ter RCS pressure was bought [ sic] to below 1100 psia, leakage continued into the steam generator since the damaged steam generator pressure also dropped, probably the result of the leaking steam generator safety valve. Leakage into B S[ team] G[enerator], resulted in excessive level and concern regarding the steam line ability to withstand the extra weight of being filled with water.

"Since the FDSA assumed time for primary and secondary pressure equilization did not result in S[ team] G[enerator]

overfill, and the FDSA did not analyze the resulting system (as well a of fsite consequences - Section 8.1.3) effects, the PDSA does not bound the actual event in this respect."

Ginna SER, at p. 3-6.

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Condenser "The staf f's design-basis S[ team] G[enerator] T[ube]

R[upture] analyses presently include the assumption that the concenser is unavailable, since there are a number of reasons it may be unavailable after an accident; loss of ,

of fsite power; loss of instrument air; loss of service water resulting in loss of instrument air; the turbine bypass valves being out of service; and, as during the Ginna accident, a decision to remove the condenser from use.

Witnour the condenser as a heat sink, there will be steam or speam/ water release f rom the unaff ected steam generator (al w ay s) , and from the affected steam generator (for essentially all design basis accidents). In this event

' considerably more radioiodine is released and the dose consequences are dominated by the iodine releases. The

[Ginna] licensee analyzed this type of accident, but the description of the analysis is too vague. The FDSA ref ers to 'all tne primary iodine activity expected from 1% failed fuel,' without specifiying how many curies would be expected to be released to the coolant. The description says this release, via the atmospheric relief or code saf ety valves,

'could only occur if all offsite power to run the main condenser was lost and...' Since there were, in fact, considerable releases from the B steam generator code safety valve during the ginna accident, even though off site power was available, the statf concludes that the licensee underestimated the probability of releases from naths other tnan tne concenser. Secondary leakage is r. __anged beyond the time typically assumed by the licensee and by the staff.

Because tnis prolonged leakage both increases the secondary side fission product (mostly iodine) concentration, and l decreases tne iod ine partitioning, it is important to consider the effects of continued leakage into the S[ team]

G[enerator) and overrilling with liquid rease out the S[af ety] V[alve]." Ginna SER, at p. 3-9.

THIRD CONTENTION '

Elimination of Crevice

! fo_ntention l

• l The proposed steam generator will eliminate the tubesheet crevice where corrosive impurities have concentrated in the past by hyraulically expanding the new tubes to the full depth of the tubesheet holes. At the same time as the crevice is eliminated, l

however, this process will shitt the roll stressed transition zone (between the expanded and unexpanded part .of the tube) from near the bottom ot the tubesheet hole to a point level with and WED-PA-11/05/82-P:50266NRC.P53-5

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above the upper surf ace of the tubesheet. Egmoare Diagram C to Diagram B in Attachment 1. This will create four interrelated problems:

(a) Egsidugl Silasses. The newly situated roll stressed transition zone will be subject to stress assisted cracking due to residual stresses from the hydraulic expansion process. ,

(b) flydsg Dapps112 The zone will subjected to extensive corrosive attack, in addition to and compounded by 3

stress assisted cracking, because it ih located directly under deposits from impurities in the bulk secondary water tnat cannot be entirely eliminated in a pressurized steam generator of the' existing or proposed design operating with h' an all volatile water chemistry treatment and also is in an deposition area subject to alternatc wetting and drying.

(c) Detectability.

It will be more difficult for eddy current testing' to detect stress-assisted def ects or corrosion in the transition zone than in the unexpanded portion of the sleeve.

(d) Unconstrained Leakaae. Through-wall defects in the stressed and corroding transition zone of the proposed steam generators, unl-iKe defects in the transition zone of the will be unconstrained by the existing generators, surrouncing wall of the tubesheet, and the resulting secondary-to-primary in-leakage will lead to the safety concerns discussed in the First Contention and primary-to-secondary leakage, to the concerns in the Second Contention.

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These problems with eliminating the crevice create a justiciable controversy as to whether the proposed steam generators, by their design, will suf fer tube degradation, and do so in more ominous locations, and thereby f ail to comply with applicable Commission regulat'lons, 10 C.F.R. 5 5 0.4 0 (a) ("the health and safety of the public will not be endangered"), and 10 C. F.R. [ Pa r t 50 App. A Crit. 14 (" pressure boundary shall * *

• 1 have an extremely low probability of abnormal leakage, of rapidly propagating f ailure, and of gross rupture"), such as to mandate denial of an operating license amendment.

Basis Introduction "Following insertion into the tubesheet hole, tack rolling, welding and gas leak testing, the tubes are hyrau11cally expanded to the full depth,of the tubesheet holes. Full-depth closes e]iminates the tube sheet crevice in which concentration of impurities has occured in the original steam generator." Steam Generator Replacement Report, at p. 2-8.

' Residual Stresses

" Westinghouse now, on new V[ertical] S[ team]

G[enerators], is applying a full roll to the [ tubes at the

, bottom of the steam generator] to roll out the long l

crevices. When asked [by the utilities] if they had done research on the top of the rolled zone for possibilities of stress-assisted cracking, Westinghouse replied tnat they had l not. The Westinghouse position is that the crevices are l

okay, but tney are yielding to customer pressure t'o roll them out. That being the case, then Westinghouse should do research on the top of the roll-stressed area to keep f rom jumping f rom the f rying pan in to the fire."

"Wesi:inghouse is rolling out the crevices, but is not concerned about the residual stress left in the Inconel 600 by rolling. Testing of rolled out specimens should be done under realistic environmental conditions." Ad Hoc Committee on Steam Generators, final Reoort in the Edison Electric Institute Nuclar Plant Design and Operations Task Force an Pressurized Water Reactor Steam Generators, August 1, 1974, at Part VII, p. 2 15 and p.12 132f.

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"Wh11e tne actual f ailure mode of tne two leaking tuces cannot be known without removal _and additional inspection, tuce leaks identified on the inner row [ot North Anna Power Station Unit 1) have been attributed to residual manuf acturing stresses created during tube bending. * * *"

Letter f rom C. M. Stallings (VEPCO) to H. R. Denton (NRC),

Docket 50-338, dated December 10, 1979, at Attachment p. 4.

Sludge Deposits "The use of 'zero solids treatment' is specifically not advised, owing to the lack of counteracting chemical treatment to absorb impurities introduced.by condenser leakage.

"The alternate water chemistry is 'zero solids treatment in which hydrazine is used for oxygen control, and in some cases ammonia or a volat 11e amine is added f or pH contro1. SM2h IIAA132D1 13 D21 1222mmandAIL. It is considered risky because in the practi' cal case the steam generator water chemistry will not be maintained truly free ot solids. * *

• All of the steam genrators have areas where the thermal and hyraulic conditions are such as to cause tne ' drop out' of solids. Without the influence of proper phosphate adjustment of the boiler water chemistry, experience at Eeznau I indicates that the resuling deposits can create an environment which is detremental to the tubing b material." Westinghouse Electric Corporation, Hummary Paper DD Beznau I Steam Generator Tube Leakace Problem, May 10, 1972, at pp. I and 15.

Detectability

- "The only pla'ce ton the sleeves where cracks would be expected to have a circumferential orientation (it they were to occur) would be at the expansion transitions of the joints. Routine inspections with bobbin probes generally have not been capable of detecting circumferential flaws at similar joint transitions which already exist on tne unsleeved tubes. Should such cracks occur, it will likely be necesary to employ a non-standard probe such as the pancake probe to detect these cracks. Circumferential cracks at expansion transitions have not generally been of concern since (1) such cracks typically involve only a small fraction of tne " tube circumf erence bef ore resulting in a detectable leak and (2) even if complete severance of the tuce occurred during accidents, the resulting leakage would be severely limited by the tubesheet crevice. For sleeves, tne resulting leakage would be expected to be severely limited by the narrow sleeve to tube g ap. " Prefiled Testimony of Commission Statf Witness, Emmett L. Murphy, in the Full Scale Sleeving Proceeding, Docket 50-266 and 50-3 01, a t pp. 9 to 10.

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FOURTH CONTENTION Balance of Plant Contention The replacement of the lower assemblies and moisture separators or tne Point Beach st.eam generators will not serve to repair or substitute for. other interrelated structural weaknesses in tne/ balance ot tne plant, including the following:

1 (a) Condensers. The major source of corrodents in the steam generators in tne past has been from leaks through

, failing condensor tubes. The condensers at Point Beach will not be replaced even though they do not meet present construction standards and remain a continuing source of

. tertiary-to-secondary in-leakage.

(b) feed ater SysteE. A new source of corrodents in the proposed steam generators may come from other plant components operated unoer the new water chemistry. The AVT water chemistry treatment that will be used may corrode pumps and piping tnat feed water to the steam generator of older plants such as Point Beach using copper based alloys, anc cause degradation of tne tubing from copper oxides.

I t These components with copper alloys which modern standards i

discourage will not be replaced.

(c) Condensate Polishers. Because AVT does not absorb impurities, this wa'ter chemistry treatment is f recuently coupled in new plants with a condensate polisher to remove the inevitable corrodents that will be part of the feedwater. No concensate polisher is proposed for inclusion within the operating license amendment.

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These problems with corrosive impurities f rom other impaired plant components that will not be replaced or f rom the f ailure to install new components as a necessary adjunct to this license amendment, create a justiciable controversy as to whether the proposea steam generators, by tneir limited scope of repair, will continue to suf f er tube degradation and thereby f ai.1 to comply witn applicable Commission regulations, 10 C.F.R. S 50.4 0 (a) ("the health and safety of the public will not be endangered"), and 10 C.F.R. Part 50 App. A Crit. 14 (" pressure boundary shall * *

• have an extremely low probability of abnormal \leakage, of rapidly propagating f ailure, and of gross rupture"), such as to mandate denial of the. operating license amendment.

Basis Condensers 9

" Design improvements (over admiralty based condensors at Point Beach] are. recommended to bring condensers to modern construction standards. Westinghouse has proposed titanium tubes and titanium clad tubesheets (in concensers]." W. D. Fle tche r (}i) , " Operating Experience witn Westinghouse Steam Generators," 28 ' Nuclear Zgghnolooy 357, March 197 6, at p. 36 9.

"Anotner area of improved design concentrates on the selection of more corrosion-resistant materials in the concenser because water leaks tnrough the failed condenser tubing, when combined with air, can contaminate the conuensate, feedwater, steam generator water, and steam.

This contamination in turn degrades the structural integrity or tne steam generator tubes, turbine, and other components in the cooling system. The utilities are eliminating the use or ammonia-sensitive alloys from the condensers and replacing them with more corrosion resistant alloy tubing. *

• The copper alloys are being replaced by materials such as titanium, AL 6X, or stainless steel (f or freshwater se tv1ce) ." SLgam Sgnargtgr Zukg ExpgrigDgg, NUREG-0886 (February 1982)(" Updated Experience Repo r t") , at p. 37.

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Feedwater System "A major disadvantage or AVT is that the boiler water is unbuffered and subject to extensive and rapid pH excursions in the event of feedwater contamination.

Further, excessive hydrazine can decompose producing ammonia; unacceptable amounts of ammonia in the condenser or f eedwater train can corrode copper-base alloys and allow potentially-deleterious corrosion products to enter the steam generator." Testimony of Public Service Commission Witness, Dr. James R. Meyers, in H2 Wisconsin Electric Power coinpany, Dockets 6630-ER-10 and 6630-UI-2, at F.xhibit 68, p.

1/g "Until better control of the feedwater chemistry is established, we suggest that samples from the discharge of tne high pressure feedwater heaters be monitored frequently for iron and copper. A review of steam generator crud analysis f rom Units 1 and 2 sampled February,1973 shows tne presence of Cu in steam generator crud at levels suostantially higher than six otner operating plants reviewed.

"As discussed in the previous chemistry review continued operation with elevated oxygen and ammonia concentr ationr in tne feedwater presents a potentially deleterious chemical environment to this copper and f erritic system. Accelerated corrosion can occur under these conditions to the feedwater reheat system. These corrosion procucts would be transported to the steam generator thus contributing to the formation of sludge on the tube sheet."

Letter f rom G. W. Hood (R) to C A. Reed (WE), dated February 4, 1974, re Point Beach Chemit.,yry Review, at p. 2.

Condensate Polishers "The incorporation of f ull-flow condensate polishing l

(FFCP) into tne feedwater cycle is viewed by many as a l necessary adjunct to AVT. " W. D. Fletcher (R), " Operating Experience with Westinghouse Steam Generators," 28 Nuclear Technoloay 357, March 1976, at p. 370.

FIFTH CONTENTION All Volatile Treatment

,f.0Atsntion The water chemistry treatment intended for use in the proposed steam generators is an all volatile treatment ("AVT"),

instead or a congruent phosphate treatment that had originally been used in the existing generators. This creates four new problems:

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(a) Solids Removal. AVT fails to perform the function of removing impurities f rom the bulk secondary water that had been perrormed by phosphates and which may otnerwise lead to corrosive conditions in the steam generators.

(b) D212stign. Thls problem of unprecipitateo impurities with AVT is compounded in a pressurized steam generator such as Point Beach because detection is done largely in the bulk water and not in localized areas where corrocents concentrate and deposit.

(c) Feedwater Train. Excessive hyd'razine with AVT can decompose producing ammonia which, in the feedwater train, can corrode copper-based alloys and allow corrosion products to enter the steam generator.

b These problems withi AVT create a justiciable controversy as to whether the proposed steam generators, in operation, will suffer from corroding tubes and thereby fail to comply witn applicable Commission regulations, 10 C.F.R. 55 0.4 0 (a) ("the l health ano sarety of tne public will not be endangered"), and 10 C.F.R. Part 50 App. A Crit. 14 (" pressure boundary shall have an extremely low probability of abnormal leakage, of rapidly propagating f ailure, and of gross rupture"), such as to mandate denial ot tne operating license amendment.

Basis Solids Removal "The use of 'zero solids treatment' is specifically not advised, owing to the lack of counteracting chemical treatment to absorb impurities introduced by condenser leakage.

"The alternate water chemistry is 'zero solids treatment in which hydrazine is used for oxygen control, and WED-PA-ll/05/82-P:50266NRC.P53-5

in some cases ammonia or a volatile amine is added for pH contro1. SMgh 11AA1EAni 111 n21 12E22EAndAIL. It is considereo risxy because in the practical case the steam generator water chemistry will not be maintained truly free or solids. * *

• All of tne steam genrators have areas where the thermal and hyraulic conditions are such as to cause the ' drop out' of solids. Without the influence or proper phosphate adjustment of the boiler water chemistry, experience at Beznau I indicates that the resuling deposits can create an environment which is detremental to the tubing material." Westinghouse Electric Corporation, Engmary Paper AD Beznau .I Steam Generator Tube Leakace Problem, May 10, 1972, at pp. 1 and 15. [ Emphasis in original.]

1 "The primary advantage of AVT is that no dissolved solid additives are used (such as phosphates) which can concentrate in the steam generators to induce corrosion, sucn as phosphate wastage of Inconel-600 tubing. The disadvantage of AVTis that it provides no buf f ering capacity to mitigate the effects of impurities in the cooling water through the condenser or corrosion products. Thus, when conuenser leakage occurs, the resultant impurities can enter the steam generators and cause severe changes in the pH, wirn resultant increases in corrosion rates. Although tne three PWR vendors. currently recommend AVT, both Westinghouse ano CE had in the past recommended tne use of phosphates to buffer impu r i' tie s in their recirculating U-tube steam generators. ** *" Updated Experience Report, at p. 3/.

Detection "THE WITNESS: You see, the problem we have, sir, in this b.usiness of water chemistry, is even if your bulk water is perrect[,] on that localized basis where' deposits come out, the whole thing could be blown right there. Because tne bulk water could be perrect, immaculate, but on the localized basis ycur water chemistry could be totally dif ferent than it is in tne bulk water.~ Ana tnerein lies the problem.

" EXAMINER WOLTER: By localized basis I assume you mean at the point where the corrosion is taking place.

"THE WITNESS: Or under a sludge deposit tnat sets on the bottom of the tube sheet.

" EXAMINER WOLTER: So really looking at this water chemisty, I take you answer at face value-- it doesn't really mean much. -

"THE WITNESS: Not if you got crevices and deposits.

This only works for'the bulk water.

" EXAMINER WOLTER: And tnat's, of course, the important part, isn't it?

"THE WITNESS: The bulk --

" EXAMINER WOLTER: No, not the bulk, but where the corrosion takes place or the crevice is or.where the deposit is, isn't tnat the critical area?

"THE WITNESS: That's the critical area.

W ED-PA-ll/ 05/ 82- P : 50 266 NRC . P53 -5

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t * " EXAMINER WOLTER: And you are telling us that there's

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no way to measure that water chemistry at that location?

"THE WITNESS: That's exactly -- in the crevices and under the deposits, that's correct.

" EXAMINER WOLTER: And there's no direct relationship or even indirect relationship, from your testimony, if I unoerstano it correctly, between the bulk water and the water under thoe deposits or in those --

"THE WITNESS: I can haye anything in the water crevice area. The chemistry -- that can be totally different from tnis. And tnerein.11es the problem.

" EXAMINER WOLTER: So what you are saying is that tnere's just no way we can tell whether the water chemisty at the point where ,it did the damage was similar c;r dissimilar to tne chemistry tnat you have described in your Exhibit 68.

"THE WITNESS: In Exhibit 68 I addressed tne bulk water chemistry

" EXAMINER WOLTER: That's right. 3 "THE WITNESS: And explained ho'w the bulk water chemistry can go astray. But in Exhibit 68 I also explained how the localized water chemistry can be significantly dif f erent in the crevices and under deposits than it is in the bulk water.

' EXAMINER WOLTER: But tnere is no correlation necessarily between the bulk water chemistry that you have described in detail and the chemistry -- the water chemistry h at the location where the corrosion takes place. There is no correlation between those two.

"THE WITNESS: No correlation between the two. Except we do know that in those deposit areas that the causticity or tne environment has to be very, very high.

" EXAMINER WOLTER: In order for corrosion to take place.

"THE WITNESS: That's correct. See, this is part or tne problem that we're into here today. As the steam generator tuoes were tested certainly the Inco people and tne Westinghouse people tested that material and tested it and experimented it in a wide variety of waters, but they certainly never expected to have a concentrated caustic solution exposed to the metal wnich in fact did occur.under deposits and in crevices and in places of this nature.

t

" EXAMINER WOLTER: So am I correct, then, in assuming l

that the only way that you can tell whether the chemistry is bad at tne point..where the corrosion takes place is to wait until the corrosion actually takes place and then physically inspect periocically,to see if corrosion is taking place?

i "THE WITNESS: Basically, that absolutely correct. If l I'm pulling a water chemistry oft of the continuous l blowdown, which WEPCO did, for example, very carefully and very continucusly, and they may be reading along and they say: Well, find, I have no f ree caustic present in the bulk

. water. But they may very well have f ree caustic present under the deposits.

" EXAMINER WOLTER: And there's no way they can tell that.

WED-PA-11/ 0 5/ 82- P : 50 26 6 NRC. P53 -5

"THE WITNESS: There is no way to tell that. Because '

there's no way to get a water sample out of that a r e a."

Testimony Under Cross-Examination of Public Service Commission Witness, Dr. James R. Meyers, in Ae Wisconsin El e c t r ic EQEf.I Ennoan y, Docke ts 66 30-ER-10 and 66 3 0-UI-2, Transcript pp. 1775 to 1779 Feedwater Train

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"A major disadvantage of AVT is that the boiler water is unbuffered and subjec._ to extensive and rapid pH excursions in the event of feedwater contamination.

Furtner, excessive hydrazine can decompose producing ammonia; unacceptable amounts of ammonia in the condenser or f e'eawater train can corrode copper-base alloys and allow potentially-deleterious corrosion products to enter the steam generator." Testimony of Public Service Commission Witness, Dr. James R. Meyers, in Eg Wisconsin Electric Power

- Comoany, Dockets 6630-ER-10 and 6630-UI-2, at Exhibit 68, p.

17.

"Until better control of the feedwater chemistry is established, we suggest that samples from the discharge of tne high pressure feedwater heaters be monitored frequently for iron and copper. A review of steam generator crud analysis f rom Units 1 and 2 sampled February,1973 shows tne presence of . Cu in steam generator crud at levels suostantially higher than six other operating plants reviewed.

"As discussed in the previous chemistry review continued operation ~ with elevated oxygen and ammonia concentrations in the feedwater presents a potentially deleterious chemical' environment to this copper and ferritic system. Accelerated corrosion can occur under these conditions to the feedwater reheat system. These corrosion products would be transported to the steam generator thus contributing to the formation of sludge on the tube sheet."

. Letter f rom G. W. Hood (E) to G. A. Reed (WE), dated February 4, 1974, re Point Beach Chemistry Review, at p. 2.

SIXTH CONTENTION Operator Performance ,

. Contention An extremely high degree of operator performance is required both to properly maintain the proposed steam generators to prevent new corrosion and to respond to tube rupture accidents.

Operator performance at Point Beach has seriously eroded in the past two years and no longer provides that necessary margin of WED-PA-ll/ 05/ 82-P : 50 26 6 NRC . P53-5

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l _19_

safety.

These problems with operator perrormance create a justiciable controversy as to whether the maintenance of the proposeo steam generators will lead to continued tube degradation and as to whether operator response to tube rupture accidents w111 be adequate and tnereby fail to comply with applicable Co m m,is sio n regulations, 10 C.F.R. S 5 0.4 0 (a) ("the health and sarety of the puolic will hiot be endangered"), and 10 C.F.R. Part 50 App. A Crit. 14 (" pressure boundary shall * *

• have an extremely low probability of abnormal 14akage, of rapidly propagating f ailure, and of gross rupture"), such as to mandate denial or tne operating license amendment.

Basis p'

"There has bden a discernable decline in the higher than average perf ormance thet had come to be expected of tnis utillcy * * * [and] there was a significant increase in the number of items of noncompliance. The increase is attributable to tne lack of management attention and involvement necessary to maintain the discipline that has characterized tne~per,tormance of Point Beach during previous years, and apparently stems f rom the loss of experienced

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personnel. This loss combined with the increased regulatory cequirements and the more extensive maintenance caused by tne steam generator tune corrosion have strained the licensee's resources. * *

• During this evaluation period eight experienced people terminated inciding the Maintenance Supertendant, Superintendant of Chemistry and Health Physics, the Health Physicist, an Operations Supervisor, a Shif t Supervisor and three experienced operators. In the prior evaluation perico, eleven experienced management people or engineers had left; thus, the only. management posicion not to be recently vacated is tnat of tne Plant Manager. The personnel lost were replaced by promotion, depleting the overall experence level." Systematic Assessment of Licensee Performance, Point Beach Nuclear Plant, Units 1 and 2, Docket Nos. 50-266 and 50-301, dateo June 19 82, at Transmittal Letter p. 2 and Enclosure pp. v.

ana 5.

WED-PA-ll/05/82-P:50266NRC.P53-5

N SEVENTH CONTENTION Unspecified Problems with Proposed Steam Generators ContentiDD The proposed Model F Westinghouse steam generators may be expectea to experience new forms of tube degradation of an undefined nature that cannot be specifically anticipated at this time,fjust as first tne Model 51 and later the Model D steam g e n e r a't o r s , which succeeded the existing Model 44 steam generators, experienced new and unanticipated forms of degradation.

This inability to anticipate the entire scope of potential problems creates a justiciable controversy to support any inquiry reasonably related to whether the proposed steam generators will cohtinue to suf fer f rom tube degradation, and tnereby f all to comply with applicable Commission regulations, 10 C.F.R. S5 0.4 0 (a) ("the health and saf ety of the public will not be enaangerea"), and 10. C.P.R. Pa r t 5 0 App. A Crit.14 ("pr essur e

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boundary shall * *

• have an extremely low probability of abnormai leakage, ot rapidly propagating f ailure, and of gross 1

rupture"), such as to mandate denial of the operating license amenument.

Basis "The small radius U bends in the inner (or first) row of tubing in [model 51] Westinghouse steam generators [which succeeced tne model '44 generators at Point Beach] have been

, subjected to primary-side-initiated stress corrosion cracking. Thes'e cracks have occurred eitner at the apex of the U-bends or at the tangent point transition between the 1 U-bena and tne straight span portion of tne tubing. At domestic plants such as Surry Units 1 and 2 and Turkey Point Unit 4, apex cracks have occured as a result of service-induced ovality of the tube as a result of the denting proce ss. * * *

" Apex cracks have also been observed in at least two UED-PA-ll/05/82-P:50266NRC.P53-5

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• Westinghouse-designed foreign facilities. Doel, in Belgium, experienced a large leak at the apex of an inner row U-bend.

Atnough there was no active denting at 'this unit tnat the staff is aware of, there was significant ovality of the tuoing, which was believed to have been introduced during the fabrication process. ***

"Anotner category of U-bend cracks includes stress corrosion cracks located in the transition area between the U-bena ana tne straight portion of the tubing. These cracks have generally been observed at plants which have not experienced denting. This tangent point cracking phenomenon has been responsible #or numerous small leaks over the past tnree years atfecting Westinghouse Model 51 steam genrators, particularly those at. Trojan Unit 1.

"It is believed tnat the ' opposite side' transition geometries [where the cracks have been located] were introoucea during the f abrication process and resulted in increased residual stress at this location. The fabrication procedure includes tne insertion of an internal ball mandrel

? through the U-bend during the bendin process to prevent excessive tuce ovality. Westinghouse has reviewed tne bending techniques used by [its] S[pecialty] M[etals]

D[ivision] during the period in which U-bends exhibiting

. opposite side transitions were fabricated. Westinghouse has been unable to'date, to identify exactly why certain tuces were affected and others were not." Updated Experience h Report, at p. 13.

"Ringhals Unit 3, a tnree-loop Westinghouse plant in

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Sweden [with Model D steam generators which succeeded the Model 44 generators], was shut down on October 21, 1981 because of a 2.6 gpm primary-to-secondary leak. * *

• The steam generators, E preneat type (Figure 2), are similar in l

design (Model D) to those at McGuire Unit 1, the only domestic operating plant with this type of steam generator.

"The leaking tube was located within the preheater section on the cold leg side of tne steam generator. The ECT results revealed numerous tubes with ECT indications localizea witnin tne preneater section at baffle plate loca tions. * *

• u ***

" Westinghouse believes the ECT indications are attributable to excitation of the steam generator tubes f rom high fluid velocities and tnat the tube walls are being worn down from vibrational rubbing against baf fle plates in the preneater sections of tnese steam generators. Westinghouse further belie.ves that a reduction of flow velocity by controlling total feedwater flow should reduce tne potential f or vibration." Updated Experience Report, at p. 16.

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WED-PA-11/05/82-P:50266NRC.P53-5

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, DATED at Madison, Wisconsin, this 5 th day of November, 1982.

WISCONSIN'S ENVIRONMENTAL DECADE, INC.

by .

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PETER ANDERSON Co-Director 114 North Carroll Street Suite /2u 8 Madison, Wisconsin 53703 (608) 251-7020 e.

4 e

e S

G e

a WED-PA-ll/05/82-P:50266NRC.P53-5

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' ATTACHMENT 1 Diagram of Differina Tube Expar.sion Technioues -

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A . a .

ht I l' A welded C completely rolled;n ,d ws:ded

'j S perdy rolle'd in and melded D esplosjon bonded and welded 1

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00LKETED ystM

, UNITED STATES OF AMERICA 82 H0V 10 N0:37 NUCLEAR REGULATORY COMMISSION f IA Before the Atoraic Safety and Licensing BoardFi -}5[sfRVfCk BRANCH Wisconsin Electric Power Company POINT BEACH NUCLEAR PLANT UNIT 1 DOCKET NO. 50-266

/ Operating License Amendment 2 T (Steam Generator Replacement Proceeding)

< AFFluAVIT OF PETER ANDERSON STATE OF WISCONSIN )

) ss.

COUNTY OF DANE )

PETER ANDERSON, being first duly sworn, on oath states:

1. He is Co-D'irector of Wisconsin's Environmental Decade, Inc., a party in tne above-captioned matter.

2. He prepared Decade's Contentions Concerns Steam Generator Replacement,, dated November 5,1982.

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3. The excerpts cited in th'e Basis section of the Decade's Contentions Concerning Steam Generator Replacement are, to his own knowledge and belief, tr e an correct quotations f rom the documents incicated tnerein.

Subscribed and sworn to berore me tnis @ day"or j

Noyember,x 1982.

) l -

li (_i[_ .d ( 4 L i e- / 1_

Notary Pub' lib t ,

State of Wisconsin My commission J f 1AV s 3 E(p WED-PA-ll/05/82-P:50266NRC.P53-5

y t

DOCKETED

. USNRC

' UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 12 NOV 10 40:37 0?f LE C4 SELntIAh.

Wisconsin Electric Power Company 00CKETUG & SERVICE POINT BEACH NUCLEAR PLANT UNIT 1 BRANCH Docket Nos. 50 -266 OIA-2 CERTIFICATE OF SERVICE I certify that true and correct copies of the foregoing document will be served this day by depositing copies of the same in the first class mails, postage pre-paid and correctly addressed, to the following:

Peter B. Bloch, Chairman 3 Atomic Safety & Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Dr. Hugh C. Paxton 1229 - 41st Street Los Alamos, New Mexico 87544 hDr. Jerry R. Kline Atomic Safety & Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Docketing & Service .

U. S. Nuclear Regulatory Commission Washington, D. C. 20555 i Mr. Richard Bachmann l Office of Executive Legal Director l U. S. Nuclear Regulatory Commission Washington, D. C. 20555 l Mr. Bruce W. Churchill Shaw, Pittman & Potts 1800 M. Street N.H.

Washington, D. C. 20056 i

' /l M 6/k-w Carol PfeffePdo @

Date:

l IKTIE: The mailing here is Federal Express l

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