ML20044B728
| ML20044B728 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/30/1980 |
| From: | Malsch M NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| To: | |
| Shared Package | |
| ML19290F683 | List: |
| References | |
| FOIA-92-436 SECY-A-80-112, NUDOCS 9303030253 | |
| Download: ML20044B728 (47) | |
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o UNITED STATES NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20555 ADJUDICATORY SECY-A-80-112 au,y 30, uso COMMISSIONER ACTION For:
The Commissioners Frcm:
Martin G. Malsch, Deputy General Counsel
Subject:
Director's Denial of 2.206 Relief (In the Matter of Wisconsin Electric Power Co.)
Facility:
Point Beach, Unit 2 Furpose:
To inform the Commission of the denial of a 2 which, Aon regarding act request for enforcement g/
Point Beach, Unit h = %, -..
.v-Review Time Expires:
August 14, 1980 1
~Discussien:
Ey petition dated March 12, 1980, Wisconsin's Environmental Decade, Inc. (Decade) requested the Commission to issue an order to show cause to prevent Point Beach, Unit 2 from returning to operation due to recent steam generator tube degradation.
Decade also requested that an adjudicatory hearing be held prior to returning Unit 2 to service.
As the basis for its request, Decade cites two facts indicating recent problems at Unit 2:
(1) on February 28, 1980 one steam genera-tor tube leaked at a rate of 1400 gallons per day (gpd); and (2) during a subsequent eddy inf amauan c this rewrd as i:eted current test (ECT) inspection 35 tubes showed in acccrdan c udh the f:Wom c!!nimm3 tion defects above the top of the tubesheet.
Att',cnmphons
- - - Decade asserts that these problems, coupled I
with Decade's earlier concerns expressed in q "'
g qyg regard to Unit 1, invalidate the staff's basis for permitting operation of Unit 2.
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CONTACT:
3 L
Marian E. Moe, CGC F
634-3214 7
GILINSK92-436 PDR
2 The Commission referred the petition to the Director of NAR who reviewed the points raised by petitioners and denied their re-quest for enforcement action on July 10, 1980
.: 00-8 0- 2 5 ).
Regarding the 1400 gpd primary to secondary leak, the Director explained in his denial that such a leak is considered to be rela-tively small and similar to leaks which have developed at other FWRs. 1/
The licensee carried out the shut-dcun and eddy current testing required by Unit 2's technical specifications when leakage exceeds 500 gpd. 2/
The 500 gpd limit is used to assure tnat a crack in any one tube is less than the critical size which cculd burst during a LOCA, and is viewed by the staff as an extremely conservative limit.
See Safety Evaluation Report Related to Point Beach Unit 1 Steam Generator Tube Degradation Due to Deep Crevice Ccrrosicn, p.
19 The leak-ing tube, which was mechanically pulled for r 22 tubes which laboratory testing, was one o were plugged dur_ing shu:d:wn [7
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Although Decade referred to the leak as a tube ":..pture",
the staff uses the term to refer to a " sudden and violent cpening of the tube generally accompanied by large plastic deformation and high leakage For example, the staff considered the sudden leak of 180,000 spd, which occurred at Point Beach, Unit 1, in 1975 a tube rupture.
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Sy comparison, the restrictive conditions imposed on the more troublesome Unit 1 by the Confirmatory Order of Novem-ber 30, 1979 required the reactor to be shut down for tube plugging upon any primary to secondary leakage in excess of 250 spd or sudden leakage of 150 gpd.
Eddy current testing is required by the order only if the leakage is in excess of 500 gpd.
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Eddy Current Test Inspecticns Decade's second basis for requesting the order to shcw cause was that the March,1980 eddy current test (ECT) inspections identified l
defects above the tubesheet in 35 tubes.
Decade believes that the occurrence of defects in tubes above the support plate invalidates one of the staff's bases for allowing plant operation, i.e.,
that the tube sheet con-l straint will reduce the possibility of the collapse of tubes sufficiently so that there.
is no hazard presented by sudden tube bursts.
l The staff's ECT inspection in March 1980 found 500 indications (defects) at or near the top of the tubesheet (within ene inch above or below).
Of these 500, 32 showed indications of 39% or more thrcugh-wall corrosion.
30 were plugged after the March j
inspection, and the remaining two were plugged after the April outage. 1/
The eddy current testing during the March 1980-inspec-tion was the first. tim.e. _the multifrequency technique was used at Unit 2.
This technique l
1s much more sensitive to. defect indications t
than the former single frequency technique.
In comparing the results of the latest inspec-i tion with previous tests the' licensee fcund that the majority of the defects for tubes which had previously been tested had shown up i
at an earlier point, albeit not as clearly as l
with the ' new technique, and had not significantly l changed in magnitude.
WEPCO's inspection also shcwed that for those tubes which showed l
defects during previous inspections at or near the top of the tubesheet, the March 1980 inspection showed that the indications (defects)-
remained essentially unchanged.
The staff attributed the detection of the new defects to the ability of the multifrequency tech-t nique to better discriminate signals from the tubes rather than to recent deterioration.
This hypothesis, together with the results of b
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Unit 2's Technical Specifications require the licensee to j
plug tubes with indications of through-wall degradation at 40% or greater, j
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4 metallurgical laboratcry tests, supported the staff's conclusion that deep crevice cracking at Point Beach Unit 2 does not yet present a major safety problem.
The staff is ccmmitted to continuing its review of further test results on Unit 2.
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Recommendation:
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1 Martin G. Malsch Deputy General Counsel Attachments:
1.
Director's Denial 2.
Decade's Petition i
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Conmissioners'commentsshouldbeprovideddirectlytotheOfficeoftheSecretary!
by c.c.b. Aucust 14, 1980.
P t
Commission Staff Office comments, if any, should be submitted to the Commissioners NLT August 7,1980, with an information copy to the Office of the Secretary.
If the paper is of such a nature that it requires additional time for analytical-review and comment, the Commissioners and the Secretariat should be apprised of when conrents r.ay be expected.
DISTRIBUTION Commissioners Commission Staff Offices Secretariat
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UNITED sTATEa '
NUCt. EAR REGULATORY COMMISSION 1
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WAsmwotew. o. c. ::sss i,yG.- l
- MI July 10,198D Kathleen Falk, Esquire
.i General Counsel Wisensin's Environmental i
Decade, Inc.
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302 East Washingten Avenue Macisen, Wisconsin 53703
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l Cear Ms. Falk:
l rce the reascns stated in the attached " Directer's Cecisien Uncer 10 CFR 2.206", the petitten of Wisconsin's Environmental Decade Inc. cf 'darch 12, 1980, is denied.
i A ccpy of this determination will be placed in the Cc Tnission's public document rooms at 1717 H Street, N.W., Washington, D. C. 20555 and at i
the Occument Departrent, University of Wisconsin - Stevens Point 1.ibrary.
Stevens Point, Wisconsin 54t81. A ccpy will also be filed with the Secretary of the Comission fer review by the Cc:mtission in acccedance with 10 CFR 2.206(c) of the Connission's regulations.
Attached also is a copy of the notice that is being filed with the Office h
of the Federal Register for publicaticn.
7 Sincerely, j
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Gr W k n
i n -
Harold R. Denten Director' i
Office of Nuclear Reacter Regulation Attachments:
- 1. Director's Decision
- 2. Federal Register Notice l
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7590-01 Nuclear Regulatory Comission
{ Docket No. 50-301)
Wisconsin Electric Fewer Co.
(Pcint Beach Unit 2)
Issuance of Director's Decision Under 10 CFR 2.206 On April 9,1950, notice was published in the Federal Register (45 Fed.
Reg. 24293) that Wisconsin's Environmental Decade, Inc. (DECADE) had requested that the Nuclear Regulatory Co nission issue an order to shew cause and an order enjoining operation of Foint Beach Nuclear Plant, Unit 2 because of steaci generator tube degrudation at the facility. The petiti:n was referred to the Office of Nuclear Reactor Regulation to be treated as a request for action under 10 CFR 2.206 of the Comission's regulations. Af ter a review of the relevant information, the Director has deter =ined that the Foint Seach Unit 2 facility can continue to operate without undue risk to the public health and safety.
Accordingly, the request by DECADE has been denied.
Copies of the Director's decision are available for inspection in the Comission's Public Docu ent Room,1717 H Street, N.W., Washington, D. C. 20555 and at the Occument Department, University of Wisconsin - Stevens Point Library, Stevens Point, Wisconsin 54481. A copy of this decision will also be filed with the Secretary of the Commission for review by the Comission in accordance with 10 CFR 2.2C5(c) of the Comission's regulations.
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Dated at Bethesda, F.aryland this 10th day of June, 1980.
FOR THE NUCLEAR REGULATORY COMMISSION j
+.WS$
. e Harold R. Denton, Director Office of Nuclear Reactor Regulation
00-80-25 UNITED STATES OF AMERICA NUCLEAR REGULATORY e,0W.ISSION In the Matter of
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WISCONSIN ELECTRIC POWER CCMPANY
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Docket No. 50-301 (Point Beach Nuclear Plant,
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Unit 2)
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DIRECTOR'S DECISION UNDER 10 CFR 2.206 By petition dated March 12, 1980, Wisconsin's Environmental Decade, Inc.
(CECADE) requested that the Cornission enter an order to show cause and an order l
enjoining operation of Point Beach Nuclear Plant Unit 2, because of steam generator tube degradation at the facility. The petition was referred to the Office of Nuclear Reactor Regulation to be treated as a request for action under 10 CFR 2.206 of the Ccruission's regulations. Notice of receipt of the petition was published in the Federal Recister en April 9, 1980 (45 F.R. 24293).
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CECADE cites as the basis for its request previous filings datec Nove= er 14, 1979, Novecer 26, 1979, Cece ter 17, 1979, January 8, 1950, and February 8,1980. These earlier filincs address the Petitiorer's concerns regarding the consecuences of a LOCA coincident with steam generator tube ruptures in light of significant tube degradation which has occurred at Point Beach Unit 1 within the tubesheet crevices and the more recent finding of defects l
at or slightly above the top of the tubesheet. E DECADE centencs that while it had previously been believed that no significant tube problem existed at Point Beach Unit 2, the experience at Unit 2
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The substantive issues raised in these filings have previously been addressed in Staff Safety Evaluations dated November 30, 1979 and April 4,1980 for Point Beacn Unit 1. See Attachments A&B.
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on February 28,1980 ("one tube ruptured with a leak rate reported at 1400 gpd")
and subsequent eddy current test (ECT) inspections which identified 35 other tubes with defects. "all of which were above the tubesheet", invalidates any basis for continued operation of that facility.
The Staff has evaluated the steam generator tube leak which occurred at Point Beach Unit 2 on February 28, 1980, and the results of the subsequent steam generator inspection conducted at the facility during March 1980.
For the reascns l
set forth in the attached Safety Evaluation Report, (Attachment C), and sumarized I
i below, I find that the Unit 2 steam goerators have been adequately inspected and I
that the condition of the steam generators is adequate to assure continued safe c:eration of Pcint Beach Uni: 2.
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Unit 2 has previously experienced wastage and stress corrosion cracking at and above the tubesheet affecting in excess of 200 tutes, 2/ of which 36 had been plugged.
As discussed in the attached SER, the Staff believes these defects to be in a generally stable condition, i.e., they are not devel: ping at a significant rate.
No special operating restrictions, such as those imposed at Point Beach Unit 1, have been recuired or imposed.
The 1400 gpd (gallons per day) primary to secondary leak which occurred recently at Unit 2 was a relatively small leak similar to those which have occurred at other PVR units as a result of through wall cracks. The term " rupture" is generally reserved for tube failures involving a sudden and violent opening of i
the tube generally accompanied by large plastic defont.ation and high leakage (e.g.,
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The point Beach Unit 2 steam generators each contain 3250 U-tubes.
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3-No tube ruptures heve occurred at the Point Beach Unit 2 fishmouth tube burst).
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facility. L The findings of the March 1930 steam generator inspecti:n at Unit 2 are f
addresse'd in the attached Safety Evaluation. Deep crevice cracking at Point Beach Unit 2 is clearly at an early stage compared to the situation at Point Beach Unit 1 (and other units), and continued operation of Point Beach Unit 2 is supported by the evaluation and conclusions previously set forth for Unit 1 (see Attachment A).
Should significant deeo crevice cracking activity develop scretire in the future, the Staff would net expect this activity to occur abcVe the top of the tubesheet.
This is supported by results of laboratory examinations of five tube samples removed from Point Beach Unit 1 and one semple (containing the deep crevice indication) removed fran Unit 2 indicating that the general intergranular attack occurring The need for
.itnin :ne tubesheet does not extend outside of the tubesheet.
i accitional tube renovals for laboratory examination will be considered by the 5:a'f snould the deep cresice cracking phenorenon continue to deveico at Unit 2.
As ciscussed in the attached Safety Evaluation, the Staff has concluced that tne fincing of approximately 500 indications at the top of the tubesneet, including approximately 250 indications of 20t or greater, and 32 indicatiens of 39" or greater, is not indicative of a new or highly active corrosien mechanism occurring This is supported by a reevaluaticn of eddy current at cr accve the tubesheet.
indicaticns or tapes from previous inspections for these tubes containinc 39:
1 c,reater indicating that the majority of these indications nave been cresent 1
l 2/ A tuce rupture (1S0,000 gpd) did occur at the Foint Beach Uni + 1 f February 26, 1975.
That tube failure was determined *o be ~h
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ano stress corrosion cracking above tne tubesnect.
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in previous inspections dating back to the period 1974 to 1977.
Previous inspections dating cack to this period have identified the region within a few inches of the tucesneet to be the scene of wastage tninning and/or stress corrosion cracking degracatien wnich recent data indicates is not developing at a significant rate.
The staff attributes the fincing of the top of the tubesheet indications in March 1950 to the ennanced capability of multifrequency ECT to discriminate relatively small amolituce defect signals from the tubesheet entry signal, relative to t
v previously ecoloyed single frequency ECT.
i III Based on the foregoing, I have determined that there is reasonable assurance that the Point Beach Unit 2 facility can continue to operate without undue risk to the public health and safety.
Consecuently, DECADE's request for an order to shew cause and an order enjoining operation of the Point Beach Unit 2 facility is denied.
A cocy of this decision will be placed in the Commission's Public Document Room at 1717 H Street, N.W., Washington, D. C. 20555 and in the Local Public Document Room at the Library of the University of Wisconsin, Stevens Point, Wisconsin 54481.
Additionally, a copy of this decision will be filed with the Secretary of the Comm-ission for review by the Conmission in accordance with 10 CFR 2.206(c) of the Conmission's regulations.
As :rovided in 10 CFR 2.206(c) of the Commission's regulations, this decision will constitute tne final action of the Commission 20 days after the date of issuance, unless the Commissien on its own motion institutes the review of this decision within that time.
Harold R. Denton, Director Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland this 10th day of June, 1980.
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Attach ents:
A - Safety Evaluation Report Related to Foint Beach Unit 1 Steam Generator Tube Degradation Due to Deep Crevice corrosion - 11/30/79 B - Same Subject Report as Attactr.ent A - 4/4/80 C - Same Subject Report as Attachments A & B
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ATTACHMENT A
- "%,o, UMlTED STATES f,$^e( f 1 NUCLEAR REGULATORY COMMISSION gy 3 O Eg 2
E numscie w. o. c. :ms 5A E
SAFETf EVALUATIO.? REPORT RELATED TO POINT EEACH UNIT 1 STE;.M GENERATCR TUEE i
OEGRADATION DUE TO DEE? CREVICE CCRROSION INTRODUCTION Inservice ins;ections of the Point Beach Unit i steam generators perfomed during the August 1979 and October 1979 outages indicate extensive general intergranular attack (IGA) and caustic stress corresion cracking on the external surfaces of the steam generator tubes within the thickness of the tube sheet.
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This ccnditicn a;; ears to have develep-1 rapidly during the last twelve (12) renths as evidenced by small primary to secondary generater tube leaks occurring en Se;; ember 20, 1975 and F. arch 1, August 5, and August 29, 1979.
Ninety-seven (97) tubes were plugged as a result of the August 1979 inspection, and 145 tubes were plugged as a result of the Octcber 1979 inspection.
Of the 145 tubes pluscedi in the Octccer inspecticn, IM : bes were dee ed cefective due to the crevice corrcsi cn phenomenen.
4 Foll: wing the October 1979 inspecticn, the NRC staff met with representatives of Wiscensin Electric Pcwer Company (the licensee) and their Westinghouse con-sultants on Noveder 5, and again on November 20, 1979 to discuss the opera-ticnal experience at Point Beach Unit 1 and the present condition cf the steam This included a discussion of the Poin: Eeach Unit 1 operating cenera crs.
his: cry, results of the August and October 1979 steam generator inspections, results of labora: cry examinations cf tubes pulled during the October 1979 eu age, laboratory tests and calculations to de:nonstrate tube integrity, and
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1 the plans for remedial actions.
Information provided by the licensee and i
Westingnouse at these meetings has forrally been dccumented by letter dated l
Ncvember 23, 1979, from S. Burstein to H. R. Centen.
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At the request of the ARC, Point Beach Unit 1 has not been returned to power r
pending a thorough safety evaluation by the NRC staff.
A safety evaluation was deemed appropriate in view of the degradation which presently exists within the tubesheet crevices and because of the likelihood fer continued tube degradati l
and new leaks, unless reredial reasures are taken to retard the progress j
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Of the steam generater tube degradation.
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CPERATIONAL HISTCRY Water Chemistry Point Beach Unit 1 began comtercial cperation in December 197s using phosphate f
secondary chemistry control.
- n addition to continucus feed, phosphates were batch-f ed to the steam generators.
Steam generater blowdown was perfer:ed I
i nt e rmi tt ently.
Numerous cendenser leaks were exherienced until modifications were made to the condensers in 1971.
Sodium to pnesphate (Na/PD ) levels were 4
c generally high and free caustic was present in the secondary coolant l
through January 1972.
i From January 1972 to Septencer 1972, phesphate cencentratiens were increaset using tr.e same batch and continueus feeding methods.
The ;eriods of c;eration with free caustic were reduced and the Na/PO ratios were generally controlled I
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between 2.0 and 2.6.
The unit was shut down for its first refueling in, 4
Septem:er 1972.
Fei10 ing c:meletion cf the refueling and maintenance outage in March 1973,
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the pnosphate feed system was mocified to allow better control and cen.
e tinucus elescown was initiated.
In early 1974, the Na/PO control ratios 4
i were acjustec to de ween 2.3 and 2.5.
During the April 1974 refueling shut-I Tube lane blocking cevices
- .n, the steam genera: Ors.ere slucge lanced.
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were installed in June 1974 to impreve circulation and sludge ree: val.
l an onlir.e conversion to all volatile treat. ment ( AYT) was In Se::em er 1974, perf:rmec ey discontinuing phosphate feec anc initia:1ng maxt um staim generater Online conversicn was marginally suc:essful, however, anc free D13::wn.
l caustic was confirmec in Novem er 1974.
During hovemoer, a 48-hour shutc0wn anc 5:ak.e e perf orce: f or phespnate removal.
ine unit was sluege lan:ec i
1 anc returned to power with AYT treatment.
Curing 1975, operation wita AYT chemistry inci:atec free causti: was presen curing ::eration and socium phosphate hicecut return was present curing uni During 1976 anc 1977, levels of free caustic generally decreasec i
shutc0*n.
Socium and phosphate continuec i
altneugn free caustic was frequently cetectec.
- te cete::ec during unit shute:wns.
In 197 anc 1979, free caustic was r
cetection limits and socium and phesphate, although still pre-ner:211y tel:%
The ::ntinuing impr vement in sen:,.ere rucn icwer curing unit snuidewn.
AY7 :nemistry control from 1975 :: present has Oeen cue in large par:
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cf increased attention to concenser leatage anc the continuing develcpmen 3
lentage cete::icn capacility.
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i The c:ntr:lling parameter for the varicus c:rrosion mechanisms that lead t:
tube cegradaticn appears to be re',,.n to steam generator setCndary water chemistry control. The predcminar.*
i cf chemistry control at reint Beach l
Unit 1, ;rier to 1975, was coordinated ;M-;hosphate c:ntrol.
In late 1974, l
1 Point Beach Unit 1 ccnverted from phosphate control to all-volatile treatment (AVT). Wiscensin Electric sludge lanced the steam generators and made minor l
changes to improve circulation.
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i The purpose of the chemistry c5a,ngeover from phosphate chemistry control to
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AVT was principally to arrest tube thinning (westage) that pett.arily occurred The Unit 1 steam generators experienced-near but always above the tubesheet.
significantly reduced rates of wall thinning follcwing the chemistry ccnversien, f
The objective cf using phosphate control was to buffer inleakage of ic:urities j
from the concenser and to prevent formatien f boiler scale on the steam gener-ator tubes. Control cf caustic level was also of concern.
In fact, improper I
9 use cf phes; hate eften leads to caustic stress corresion. With the change.
over to AVT centrol, caustic stress corresion has remained a concern. Stress l
i cerrosien cracking after c:nversion from phosphate to AVT contrcl is related 1
to previcus phosphate cencentration and pessibly to makeup water centamination, j
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Plants with only short periods of phosphate control have not experienced opera-i ti nal pr:blens due tc aail ininning er clustic stress c:rresien cracking.
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A seccnd significant effect en the ccnversicn to AVT upon wastage has occurred j
With a due to a change in the character of steam generator sludge dep: sits.
phosphate feed, the sludge is coarse, granular material that forms a cohesive j
mass en the tubesheet.
Operation with AVT af ter a period of phosphate treatment j
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results in a finelv divided sludge of dense particles that are mare easily l
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removec by water-lancing procedures. This sludge is similar in metal compositicn l v
to the pn s;natet slucge because the iren ic:urities in the feedwater are e
r unchanged. The improved ability to remove the "AVT si~ Je" minimizes wastage 3
Cf steam genera :t tubes.
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Tute Inte:rity - Fiu:;in; ".istory The early nistory of tube cegraca-Wastage anc Caustic Stress Corrisien:
ti:n a: Fein: 5eecn Unit i, since beginning c:crercial cperation in Deceraer 197U witn a pnosphate sec:ncary water che:istry, was highlighted by the accumulation of a suestantial amount of slucge ceposits on the tubesheets,-
and the occurrence of wastage and caustic stress corrosion located for the
-::s: par: just adeve the tubesheet in both sten: generators.
By September 1972, a total of 175 tubes in both steam generatcrs had been plugged.
- Hewever, only two tubes recuired plugging in the subse:uent April 1974 inspection, apparg reflecting impreved control of sedium to phosphate (Na/PO ) ratios and free caug 4
in the secondary sa:er.
The changeover to AYT se::ncary water chemistry in Septemaer 1974 was per-f ormec en-line and witncut an intermeciate slu:ge lancing, so that the siccge ce:csits rerained essentially in place curing the first few months of AVT cceration until ine first sluege lancing in A: vender 1974.
On February 25,1975, fc11 ewing the change ever :: AVT, a tute rupture occurrec resulting in a 125 g;m primary-to-seconcary leak.
Subsequent inspection indica.
ted that a ceasination of wastage anc caustic stress corrosion cracking (SCC) hac occurrec resulting in tne tube f ailure locatec a few inches above the ta:e-A total of 157 tutes were pluggec as a result of tne inspection per-sneet.
f or:ec fclic.ing tne tute rupture incicent.
Su: sequent operating experience at Point Beach Unit 1 (since Feeruary 1975) incicates that the wastage anc caustic SCC pnencrenen above the tucesheet have essentially been arres ec.
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Denting was first detected in Neverber 1975, at Point Beech Unit i
Centin;:
1, art currently affects the tube to tube su;;crt plate intersections of f
I apprcxicately 100 tubes. Of these, ten tuces were plusced in November 1977, j
and cne in September 197E.
The criteria for plugging includes the plugging cf all tubes restricting the passage of a.540" eddy current probe, and the surrcunding tubes.
i The degree of denting at Point Beach Unit 1 is considered to be only moderate, t
and flow slot hourglassing has not been cbserved to date.
The eddy current l
inspe:tions performed in August and Cctocer 1979 indicated no prcgression in denting since the September 1975 inspection.
i The ecst recent ccncerns regarding the integrity l
- eec Crevice Crackin;:
1 of steam generatcr tubes at Point Beach Unit 1 involve ccrresion darage _to i
t;ces within the thickness of the tubesheet.
This phencmenon, known as
'desc crevice cracking", affects early generation cf Westinghcuse designed scsar generators in which the tubes were not fully expancec in the ;Lcesheet.
l This " deep crevice cracking" involves both caustic intergranular attack and t-i j
cracking nithin the tubesheet crevice.
This phenomenen can affect steam genera.
4 ters wnich have ccnverted from phosphate to AYT secondary water chemistry, such as Point Beach Unit 1, or have operated exclusively on AVT.
'lthcugn we are aware that the " deep crevice ccrresion" phenceenen has been t
ctserved in at least seven cther Westinghcuse designed plants (San Cncfre i
Unit 1, H. 3. Robinson Unit 2, R. E. Ginna Unit 1, and Prairie Island Uni l
2, and three foreign units) the Pein; Beach Unit I situation'is unicue in terrs of the extent and the rapid progression in the last twelve (12) months.
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5-The " deep crevice cracking" phenomenon at Point Beacn Unit I was first dete:ted in Ncvember 1977, and has caused several small tube leaks ( l.5
- m) in the last twelve acnths.
prior to the August 1979 inspection, a r
total of 22 tubes were :lu;;ec because of deep crevi:e cracking.
}
DISC'J5SION Aucust 1979 Steam Generater Inscection i
Felloning a return to power on.lugust 5,1979 after ceing shutdown to repair a high cressure turbine steam leak and to make a tencorary repair to the auxiliary feedeater line, a 1.45 gpm (2CSE gpd) leak develcped in steam generator A.
This leakage exceeded the Technical Specification limit of
.35 gpm (500 spd), and the plant was shutdown.
Subsequent hydrostatic leak testing and eddy current testing (ECT) revealed three leaking tubes which failed within the thickress of the tubesheet (tubesheet crevice region).
In addition, the 100 eddy current inspection of the net leg tubes in both steam geraratcrs
- anc i incicated 52 tubes and 45 tubes, rescectively, with deec crevice crt: king ind' cations in excess cf the a0: clugging 1.imit.
These i
crack indications generally eccurred in the " kidney" shapec regien of low fl os velocity.
No cole leg indications were found.
Six percent of the het i
leg tubes were probed around the U-bend to the ccid leg side.
The rer.aining tubes were probed through the first su: pert plate.
The eddy current testing was perferred with a 40C KHz prebe.
The edcy curren: testing in August did not indicate any progression in wastage or caustic stress corrcsion cracking above the tubesheet since the previous ir. service ins;ection in Se:tember 1978, nor did the six percent tube sample indicate progression of tube denting.
a
-e
. The unit was re urned to p>er en Augus: 19, 1979, fellcwing the steam cene-rator repai r c;t age.
Hewever, en August 29, 1979 the Unit was again shutdcun due te a 324 gpd leak which is less than the Technical Specificatien limit of 500 spd.
Tnis leak had existed since the August 19 restart, increasing i
at the rate of apprcxir.ately a0 gpd per day.
Subsequent inspecticn and a j
review of the tape record of the 100% eddy current inspection perfomed during the earlier outage shewed that 9.e leaking tube was One of two tubes with 1
eddy current indications exceeding the plugging limit but inadvertently left ur.cle;ged.
-hese indica:icns (ECT indication for leaker was ES") were apparen-
- 1y everleckec during the licensee's cata e valuatien effer and were net f
identified as pluggable.
1, The leaking tube was plugged and the unit was returned to service en September 2, 1979.
h Oc cter 1979 Steam Generator Inscection i
Fcr the Oc:cter 1979 refueling cu: age, the licensee nac criginally scheduled 75 tu:es in each steam generater fer eddy current inspection as part of a centinuing monitcring program.
These tubes were located in the kidney shaped
- ene of the het leg side where the deep crevice cracking phenomenon had been 1
l ebsened previcusly.
Based upon the nt;mber of tubes with pluggable indication DC of the tube wall :nickness removed), the sa=le size was first T
h incress?d to 200 tubes, i
[
F i
t P
l
-c-9 and finally to 100t cf the tubes in both steam generators per Technical Specific ca:icn re;uirerents.
In all, 77 t'.bes in steam ger.erator A and SE tubes in steam ger. era:ct 3 were plugget.
This included, for steam generater A, two (2) tthes with no indications that were pulled for labora: cry analysis, three (2) tubes with defects less than the plugging limit (all tubes with detectable indicatiens were plugged), and two (2) tubes which were plugged by mistake.
The 53 tubes plugged in steam 'ger.erator C included three (2) tubes with defects I less than the plugging limit and one plugged by mistake.
To assure that all tubes containing detectable ECT indice: ions were actually l
plugged, the eddy current tapes were reviewed by two qualified engineers or technicians.
In addition, tubesheet phc:cgrapns were taken and hydretesting l
was perfor ed to detect mis-iccated plugs.
As a result of the Oct:ber 1979 steam generator inspection, a total of 10.lt of the tubes in steam generator A and 9.St cf the tubes in steam generator 3 have been plugged.
The previcusly NRC-appreved LOCA-ECCS analysis was only,k for tu:e ;1ugging up to 1Ct in eacn steam generater.
Therefere, the licensee su:mitted for NRC approval a revised analysis to demonstrate acceptable ECCS perfermance during LOCA for tube plugging up to IEt.
The acceptability of thisl re;cr is addressed in Appendix 5 cf this SER.
Whe eas et:y current testir.; during the previous steam genera:cr inspection in,
i Augus: 1979 had been performed fer the most pare using the single frequency (ACD KM:) prete, all eddy current testing in Oc cber 1979 was perfor ed using a cultifrecuency (10,100, and 4C0 X:-::) prebe.
Basically, the naltifrequency tecnnique provides enhanced capability for the discrimination of defects agains
-9 ncise er interference effects (e.g. ; support plates).
However, since both j
- echniques involve the cellection of data at 400 KH:, it was pcssible to
- erfern a direc: cc parisen cf the August and Oc cher 1979 eddy curren:
test results. A reevaluation cf the eddy curren: : apes frc the August i
1979 inscecticn revealed the fellowing distribution of degradation for tubes i
identified as having greater than 200 indications in October 1979:
l 1
No. of Tubes Category 400 KM Data-Aucust 1979 SG A SG B i
12 12 1.
No detectable. indications e
'2.
Noisy signals - no estimate possible 37 29-l 6
5 3.
Tace EeCord unavailable for review 17 13 4
Eddy current signals >40%
0 1
+
5.
Eddy current signals <40%
r I
1 7
6.
Tubes not compared t
No, tubes plugged per Oct. 79 ECT data 73 67 i
Thus, the nuncer of tubes plugged in October 1979 is not necessarily wnclely in:icative cf acciticnal tube degradation cccurr'ng since August 1979, but cculd l reflec: enhanced cacability to detect tube defects using the rulti-f recuency eddy current technique.
i i
The average eddy current indications obtained fcr each of the above five tube l
i categories are as follcws:
Average t ECT Indicatien (10/791 i
Category SG A 53 5 l
1.
73 BD 84 54 2.
78 75 3.
36*
75*
4.
= *e-e dai ter 5/G-A and 59% for S/3-3, res ectively, in August,1979 i
E
l I.
7he fellcwing tables des: ribe the distribution of eddy current indi:atica by their depth cf penetra-i:n and their Iccaticn er elevation within the tube-
]
i sheet crevice for both the August and October 1979 f nspections.
i DISTRIBUTION OF INDICATIONS (per:ent) l By Depth of k'all Penetration l
40 49:
50-595 50-69 70-79: 80-89% 90-1005 i
SG'A 1
2.5 3.4 12.1 40.5 40.6 i
-a me
<a..
0*.!
C.O
,e u "
.g
.z-4..
".C 1-6 By Lc:atien in Tube Crevice ("from tube end) 0 4" 5-9" 10-14" 15-19" 20"-too cf tube sheet t
SG A 0
15.2 15.0 26.4 42.4 55 5 3.6 23.2 27.7 21.4 24.1 j
No crevice indicati:ns extending above the tubesheet have been observed to da:e.
Laterat ory Examiratiens Segments of three tubes were pulled during the Oct:ber 1979 cutage for further
-l study of the deep crevice cracking phenomenon.
One tube (R15-C45;' i.e., Row 15, Column 45) was taken from the " kidney" shaped :ene of previous deep crevice i
corr:sion activity and centained an 59: eddy current indication.
The secen:
tu:e (R22-C37) ras also taken frca " kidney" shaped :ene, but containec no field eddy :urrent indicati:n.
The third tube (R20-C73) was taken outside the zone of previously observed ectivity, and which also did not exhibit eddy current A
indications during field examinations.
. Tube R15-C 5 was cut just bel:w the first suppert plate, but brcke at the location of the indicated defect during removal.
The break cccurred under a pulling force of abcut 25,600 pcunds withcut significant ;1astic elcoga-tion.
Tubes R22-C37 and R20-C73 were cut belcw ti.e first and second suppc'rt plates, respectively, and recuired fcrce applications of 25,400 and 13,000 pcunds f or remcval.
These pulling leads induced in excess of 101 elcngation cf the tubes.
These tubes were delivered to the Westingheuse R&D Center where they were subjected to intensive ar.alyses including laboratcry ECT, radiegraphy, metallegraphy, micreanalysis, and testing fer mechanical prccerties and i nt egri ty.
Metallegra; hic examination revealed, a general etallurgical Examination:
v ccndition cf (uniform) intergranular attack (ICA) within the c
- ce regions x
of each cf the three tubes examined.
Varicus micreanalytical.
niques indicate this condition to be a result of a residual caustic materials remain-ing from phesphate chemical treatment and pcssibly from earlier conder.ser tube leakage, No intergranular attack or cracks were fcund in the tube s eci ens abcve the tube sheet.
Tube R15-C45, which shcwed an 59' in-plant ed'dy current indication, exhibiter.
uniform 11A 40% through weil and cracks about 90% thrcugh wall adjacent to the lccation of the in-plant eddy current indication, thereby confirming the in-plant eddy current signal.
This was also confirred by SEM fractography
)
I of the fracture surface,
1 a
i
[
t Tubes R22-C37 and R2C-C73 showed uniform IGA abcut '0 and deer.er crack
- era
- rati:n :c 5D: and 33 of the wall, respectively.
For tube R20-C73, f
a re:all:gra; hic sample was taken which extended 3/4' above and below the
{
- cp cf :ne tubesheet, with no IGA cbserved along the entire length.
l The results for Tubes R22-C37 and R20-C73 are of particalar interest since i
they exhibiter no eddy current indications during in-plant inspection.
H ow-t ever, the local crack penetration was detected during laboratory eddy current l
examination, and it is therefore likely that these cracks were devel0 ped under the hich tensile leads during the tube removal process.
j i
t The licensee and Westinghouse conclude that the eddy current testing is f
currently nc able to detect intergranular corrcsion within the tubesheet.
Significan: (>20 percent through wall) cracks or tube wall penetrations in
- ne :uteshee; area are, hosever, detectable by eddy current testing.
The con-j ditions within the tubesheet crevices are such that the tubing ma:erial affect-ed :y intergranular corrosion is held in place by the tubesheet itself and the crevice c:nditien shess a minimum of grain dislocations er material loss.
As i
a result, the grains in the suspect region remain in physical and electrical contac: providing a continuous path for eddy currents induced in the tubewall when the eddy current test is performed. The caterial, therefore, may show no ed:y curren; indication of the corr:sien within the tubesheet crevices unless i
- here is cracking through a portion cf the tube wall.
Me:nzrical Testing: - Specimens were removed from R15-C45 and R22-C27 fer i
me:hanical tests.
Tensile tests, lead plug tests, and burs; tests were ;er-4 fCTOed.
'I t
t t
6
i The tensile test results indicated that the properties of the base core material i
whi:." has r.:: seen ceneral intergranular attack is similar to virgin material j
in terrs cf streng:h and duc ility.
Lead plug burst tests were run on s;ecimens frem tube R15-C45 for the-purpose of determining the diametral expansion of the tube prior to failure.
The resuits
,-r dem0r. strate that the tube has sufficient ductility to expand into contact with 1
the ;besheet within the crevice'.
S u rs: tests were perforred with specimens from tubes R15-C45 and R22-C37 with the i
foll >in; results:
1.
Sar:les removed f rom the intergranularly attacked region of each tube i
(at leas: 5 inches belou top of utesheet) exnibited burst pressure of 1
1 EiC0 psi and 6800 psi fer tubes R15-C45 anc R22-C37, respec*ively.
The l
.2xinum ;ressure applied during a F.SL5 will, be approximately 2000 psi..
)
2.
- 5 in:h tube san:le fron Tube.22-C27 at a 10:ation extencing down ::
l 21/2 inches below the top cf the tubesneet exhibited a burst streng;h l
in excess of 11700 ;si.
Thus, there was no cegradation in burst l
strength relative to that for a virgin tube.
This corresponds with i
i the result of the retallurgical examination which indicated no in-tergranular attack above the tubesheet.
i 7.te :- e;-ity On :ne tesis of test results and analysis, the licensee concludes:
l 1
r
'rsi:e One Tubesnea+
1.
';; re aining wall thickness (i.e. no :enetrated by IGA or cracks) i t
is reccired tc ensure tnat a dcubie ended tube failure will not i
1
. l' t'
occur during a pestulated ezin steam line break (MSLB), and is I
indicated to exis ty the concition of the tubes examined and i
the test results.
Ductility will alloa a degraded tube to expand to contact tube-1 2.
sheet (i.e. no tube burst during MSLB).
7 Tute cellapse during LOCA is highly unlikely since tube ovalizaticn 3.
during collapse would be constrained by the tubesheet.
The naximum leak rate as a result of a crack within the tubesheet 4
is governed by the annular gap.
Tube breaks 0.15 inch or more belos the tcp of the tubesheet will not pull cut of the tubesheet For j
during MSLB because of the restraint of the tube buncle.
rates l
breaks within 0.15 inches of the top of the tubesheet, le t
will be large enough to allcw detection during-normal operation.
Cutsice the Tucesheet i
The concitions cf the tubes examined and the test results incicate 1.
that intergranular attack does not occur outside the tubesheet l
l c revi ces.
400 remaining wall is required to. resist pressure leading curing-2.
a LC;A, and is indicate to exist.
i Test results show that the leak-befcre-break criteria is valid and 3.
will require timely. shutdown and corrective actions.
1
' Remedial A:; ions
.e licensee plans to. Or already has, implemented the foll wing interi:
r
.tasures to provide additional assurance of c ntinued safety:
1.
A hydrostatic test has been successfully performed at 303 psi secondary to primary pressure.
Such test pressure exceeds the pressure which might be imposed on the steam generat:r tubes in the event of a loss-of-coola.nt accident.
2.
A primary :: secondary hydr: static test has been suc:essfully performed at 2000 psi.
This test pressure exceeds that which c:ald develop during a steamline or feedwater line break and.
thus, demonstrates the tubes' ability to maintain their integrity during such events.
3.
Upon NRC approval.of the Technical Specif.ication change requested ty letter dated November 2,1979, the reat: Or c::lant system would be cparated at the n:minal pressure of 2000 ;sia rather than 2250 psia.
This would reduce internal pressure stresses during operation approximately 15t.
a.
(a) Within 30 effective full power days, a 2,000 psi primary to secondary hydrostatic test and a 800 psi se:endary to primary hydrostatic test i
will be performed.
Should any significant leakage develop as a resul of eitner test, the leaking tubes will be identified and plugged.
?
I E
E r
f
1 I
lea -
1 i
(t) Witnin 60 effective full pcwer days, the same primary to secondary and secondary to primary hydrostatic tests will be repeated, and an
[
eddy current examination of the steam generator tubes will be per-formed.
This eddy current program will be submitted to the NRC for Staff review.
j i
5 Primary coolant activity for Foint Beach Nuclear Plant Unit I will be limited in accordance with the provisions of Sections 3.4.8 and 4.4.8 cf the Standard Technical Specifications for Westinghouse Pressurized t
Water Reactors, Revision 2, July 1979, rather than Technical Specifica-tien 15.3.1.C.
The acceptability of this action is addressed in Appendix C cf this SER.
t S.
Close surveillance of primary to secondary leakage will be continued and
.l l
- .,e react:r will be shut dcwn for tute plugging cn detection and confir-mati:n of ary of the follcwing ccnditions:
i Sudden primary to secondary leakage of 150 spd (0.1 gpm) in either l
a.
t steam generator; 5.
Any primary to secondary leakage in excess of 250 gpd (0.17 gpm) l in either steam generator; or i
An upward trend in primary to secondary leakage in excess of 15 gpd c.
(0.01 gpa) per day, when measured prirary to secondary leakage is above 150 gpd.
i j
b The rea:c:r wili be shutd:wn, and leaking steam generator tubes plugged, 7.
and an eddy current examinatien performed if any of the follcwing c:nditions l
are present:
3 1
C:nfirmati:n of primary to secondary leakage in either steam generator
=
a.
I in excess of 500 gpd (0.35 gpm); or l
Two leaking tubes are identified within a 20-day period.
b.
This eddy current program will be submitted to the NRC for Staff review.
The. RC Staff will be provided with a scatary of the results of the eddy N
5.
current examinations, including a description of the quality assurance This summary will include pr: gram c:vering tube examination and plugging.
i a photograph of the tubesheet of each steam generator which will verify j
the iccaticn cf tubes which have been plugged.
t g.
T:e licensee has completed a review of Emergency Operating precedure 2A, Revisien 9, dated March 29, 1975, and has confirmed that this procedure j
is appr:;riate for use in the case of a steam generator tube rupture.
This procedure has been reviewed and found acceptable by HRC.
The licensee will ccmplete a retraining program for all licensed reactor 10 crerat:rs and senier react:r operators in the conduct of E0p-3A, the steam generator tube rupture procedure, before return to pcwer operation.
- lea -
l i
In addition, the licensee plans to also implement the follosing measures in an
~
attem:t te retard further tu e degradation:
1.
The reacter c:clant system het leg temperature will be reduced r
to approximately 557 F.
This will result in lower secondary steam pressure and, hence, loner power output due to limited flos capability cf rain turbine control valves.
Maxicum cut-put under these conditions is expected to be 83% full power i
or "13 "We ne:.
Testing will be perfcrmec :: assure main steam meisture carryever does not exceed design value of 0.25t.
~
)
?
r l
1 h
2
[
i b
I f
i.
i f i.
I i
2.
Close surveillar.:e cf feedsater chemistry c:n-l i
ditions and c ndenser tube leakage will c:ntinue, j
L 3.
Sludge lancing will be performed within 12 months of return to power.
With regard to Item 1 above, deep crevice corresien has not been observed to i
date on the cold leg side.
The primary coolant temperature on the cold leg side is 542 F at 10:t.
It is he;ed, theref re, that reduced temperature operati:n will be effe:tive in retarding the rate cf deep crevice corrcsien i
on the het leg.
L I
EVALUATION i
The staff has met on two se;arate o::asions, November 5 and 20.1979, with r
the licensee and their ::nsultants :: review the inspection results (bcth
'ugust and Oct:ber,197? inspections) and :: discuss the conhition of the Point Beach Unit 1 steam genereters and the measures which have been taken
- assure their safe c;eration.
In addition, the staff has also reviewed information submitted by the licensee on November 2,1979 in response to Our concerns regarding the apparent increase in the rate of deep crevice ccrreston at Point Beach Unit 1.
This information includes results of two su::essive 100t inspections of all the steam generat:r tubes using both single frequency and multi-frequency eddy current echniques, results of lateratory examinations of tube spe:imens that were removed fr:m the tube-sheet crevice regions, and results of analysis of degraded tube behavier under normal cperating and ;cstula ed ac:icen; conditions.
i I ~
Several levels of defense are generally relied upon to ensure steam generator tube integrity.
These are inservice inspection, preventive i
tube plugging, and a primary to secendary leak rate lirnt.*
The fellowing evaluation addresses the areas of inservice inspection, pre-ventive tube plugging, prirary-to-secondary leak rate limit, mechanical tube integrity and the licensee'.s prc;csed remedial actions.
s Inservice Inscection and Preventative Tube Pluggina The licensee has perforced a 100% inspection of the steam generator tubes using multi-frequency ECT, and all tubes with ECT indications in the crevice 20ne have been plugged.
The 1000 irspection of the steam generator tubes re;.esents a total inspection of the tube crevice regions, and the culti-frequency l ECT is the test sensitive technique currently available 'for this type of ins;e'c-The plugging criteria were censervative in that it included tubes with tion.
Furthermore, improved QA and QC ar(y E;T indication in the crevice region.
i prececures have been implecented to assure that all the tubes containing pluggable indications are indeed plugged.
However, the accuracy of the ECT j
technique is somewhat diminished in the tubesheet region and cannot be fully
~
1 P
This relied upon to detect every tube degraded by deep crevice corrosion.
appears to be particularly true for tubes, subject to general intergranular attack, but which do not contain cracks.
Partially through wall cracks t
of significance are generally detectable, even in the tubesheet region, t
1
'As c:scussec in NUP.iG-0523, " Summary of Operating Experience with Hesever, the bases for continued Recirculating Steam Generaters".
cperation presented in this document did not consider deep crevice corrosion as it was not identified as a significant mode of tube degradation i
prier to publication.
, l i
with the i creved sensitivity asscciated with the multi-frecuency ECT prebe.
It is unlikely that the general attack (IC-A) will penetrace ccm-plately through the tube wall withcut a loss of scme wall material, and where i
significant loss of wall material occurs, it is generally detectable by tha t
ECT.
i F rina ry-to-Seconda rv teak F. ate. Limits i
The third level of protection involves limits on the primary to secondary leakage rate.
These limits are established to assure that (1) the occurrence cf leaks during normal operatien will be detected and (2) corrective action I
will be taken before any individual thrcugh-wall crack becomes large enougn to open up during postulated accident conditicns and affect the ability of i
the ECCS system to ecol the core during a LOCA or result in u. acceptable radie-i activity releases during a MSLE.
For straight section tubes wih no radial res t ra i nt, a C.35 ;;r (500 spd) limit assures ina; ary incivicual thrcush-wall crack is less than :ne critical flaw si:e which c:uld burst under leacs associated with pcstul ated accident.
1 For through-wall cracks which may exist within the tubesheet, leakage during
- ostulated accidents will be severely restricted by the tight annular region te
- ween the tube and tubesheet.
j v :nanical Tube Incegritv e
Ee:ause of the nature of deep crevice cracking, the mechanical integrity of l
l the degraded tubes offers an addicional level of crotection.
Because the ceeo crevice cracking is peculiar to the local chemistry conditions in the i
i
?
i
I.
i tube to tubesheet crevice, the phen menon will be limited to that area. This is ::nrred by the 10:ation of all the defects wnich have been cbserved i
durir.g inservice ECT inspecti:n, and by the laborat:ry examinations and mechanical testing of tube samples removed from Point Beacn Unit 1.
The mechanical tests demonstrate that caterial beneath the depth of general grain bcundary attack and crack penetration exhibit similar mechanical r
pr::ert'.es as the virgin material and is sufficiently ductile to allow the j
tube to expand to c:ntact the tubesheet.
Therefore, under MSLS, the con-st raf.: ;-:vided by the tubesheet eliminates the :ctential for tube burst.
Regarding pcstulated LOCA conditions, it is t isee's conclusion that the tubesheet c:nstraint against tute ovali:ation c ompanying collapse i
redu:es the possioility cf collapse within the tubesheet.
I nde pendently siculated collapse tests were conducted on unrestrained tubes ith defects as large as 75% :o 50; wall thinning and 1.5 inches in len;:h. The lowest c:1'a:se ;ressure cbsened in these tests is 1760 :si which is well in e
excess f the pressure differential expected during a LOCA. These tests in-l dica:e inat tube collapse wculd not be expectec' during a LOCA.
These incepen-dent test results can be extrapolated to envelop the conditions within the i
tucesneet region.
e
~$e inde:endently simulated collapse tests resulted in small openings in the tu:es wnich wculd have correspending leak rates much smaller nan would be ex:e::ed from a burs; tube.
The secondary to primary in-leakage rate wcult be further limited by the restricted flow through the tube to tube-i e
i e
I I
I t !
An faC staff evaiva-icn indicates that critical everheating sheet crevice.
of the fuel durir.; a LC A c:uld cnly :::ur for leakage rates in exce:s Of 1300 ;;m. A large num:er cf tube failures (collapses) would therefore he necessary before the secondary 10 primary leak rate would result in steam binding and adversely affect the ability of the ECCS to cool the c:re.'
l t
t j
Licensee's proccsed Eases For Continued Oceration
^
The licensee nas pr: posed a ;rcgram to pr: vide additional assurance of l
continued safety. This program incluces (1) performing ;eriodic ;rimary to sec:r:ary hydr: static tests to m:ni;cr the tubes' ability to maintain their t
integrity under varicus differential pressure loadings, (2) impesing a primary f
to secondary leak ra e limit that is m:re restrictive than the current Technical i
r Specifi:a: ion limit, (3) increasing the frequency of ECT inspe:: ions beyend that re:uired by the Technical Specifications, and (4) adepting more restrictive i
reac er c clant activity limits.
The staff agrees tha hydrestatic pressurt tests prior :: returning :: ;oaer at:
peri:dically during operation will provide a positive incication and increase:
confi:ence in steam generator tube integrity.
These tests are concucted in a quasi-static mode that ade~quately models postulated accident conditions.
I I
Sirilarly, the proposed decrease in the primary-to-secondary leat rate limit wil' ;rovide conservative limits which will re:uire timely plant shutdown and c rri::ive actions.
Inservice inspection by ECT techniques is intenced :: identi-f uzes wnich re uire plugging er are ex:ected to require plugging prior :: the nex:
i ns:ecti on.
Therefore, the inservice inspection and tube plugging criteria are i
1 in more detail in A;;endix A of this SEE.
l
- inis as ciscusse:
-c2-i m
I, 1
tied ::gether by the..argin left for continued degradaticn and by the rate cf ce;racatien.
The de:reased effectiveness cf ECT ia the tubesheet region, and i
tne limited data base f:r defining the rate cf c:rr:sion indicate that more i
\\
f re;uent ins;ecti:ns are necessary.
Therefore, the staff 1s in agreement t
i wi-h the licensees prepesal that ECT inspe::icn cf the steam generators sh uld be c:nducted more frequer.tly as described on page 15 of this SER.
The f
in:rcased frequency Of ins;ection will ensure tha: tubes with large defects t
f wi".1 be detected and rec:ved from service and that the rate of degradation sili te carefull.
nit: red.
t I
Measures for Recucine The Rate of Degradation l
J he licensee has also ;r :csed varicus measures to be implemented in an a tter:: to retard further tube degradation.
These reasures include 1) f a crevice flusnir; pr:;rar to remove harmful chemi:als from tubeshee; 4
crevices, and 2) reduced reactor coolant system cperating, pressure and f
- er;erature to reduce
- e stresses and tenperature.
6 Eegarding the crevice flushing program, residual se:ium and phosphate in :ne
{
i tubeshee crevice regien will be removed by crevice-flushing techniques (i.e.,
t a stean flashing technicue to dissolve material in the crevice).
This should l
t hel;.inini:e further tube degradation in the deep crevice of the tubesheet.
5::iu, an alkaline ferting species in bciler cooling water, is a principal f
i elire n: that causes intergranular corr:sive attack Of :nconel 500 alley wnten I
lea: to caustic stress ccrresion cracking.
Caustic stress corrosion cracking l
is de:endent en tem:e-a ure, hydroxyl-icn cencentra icn and stress.
Labcra-
- ry tests in NaOH solutions have shown that the time for stress corresion cracking to occur in Incenel 500 alley increases at temperatures below 550*F.
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The license 2 prc;cses to operate the Point Beach Unit 1 with a reactor c clant inle: terperature to the steam generaters cf 557 F with apprcxi-mately a 10% reduction in pressure differential to reduce the stress level.
The loser cperating temperature will reduce the rate cens: ant for intergranular corrosi en.
Also the icwer stress levels will reduce the rate of crack growth.
1 Operation of loser reacter ccolant pressure (2000 psia) is currently under review by the NRC staff.
The a,cceptability of this propesal will be addressed l
separately.
t Regarding reduced reacter c00lant system c;erating pressure and temperature, the only ef fect of the reduced temperature cperation en the integrity of the major ca:penents will be a siight increase in the rate of radiation damage
- the ceit-line of the reactor vessel.
This is expected to be a minor effect, but shculd be taken into consideration when evaluating the pre sure-temperature limits in the technical specifications.
On the basis of informatien available, ;
t t
it is estimated that the acditicnal shift in RT wculd amount to about 50 F, NDT if the loser tec;erature o;eratien is continuec to the end of life.
Additional assurance that this effect will be pr:perly evaluated will be obtained from the reacter vessel material surveillance program.
The loser operating para-d meters are net expected to affect the design limit for steam quality;.i.e.
noisture carry over.
1 C SCLUS I O!:5_
i i
Sasad en the abeve evaluation, tne staff has reached the follcwing cenclusiens:
1.
Eddy-current-testing cannot be relied upon to cetect all deep-creYiCe corresien degradation but the majority of the defects, pirticularly these that are significant, will be ce:ected.
a t 2.
Mycrestatic ressure tests performed prior te and during c;e i
ration will identify any significant remaining defects.
j m
3.
Cctservative primary to secendary leak rate limits will previde a rance that-in the event that large defects go undetected, er the corresien rate accelerates, timely plant shutdonn and corrective actions can b 4
The ecnstraint previded by-the tubesheet and the rechanical 4
r es of the tubes greatly cecrease the probabilty cf gress tube failu r
nermal caerating er pcstulated accident ccnditions.
5.
A maximum 50 effective full poser day cperating pericd pri er tc the next ECT inspecticn will previde adequate assurance that a large number s
u es will net simultanecuily reach a peint of incipient failure, 6.
Remedial actions propcsed by the licensee will sitigate the eff ects cf pes-tulated accidents anc retart the rate of cerrosion 7.
Tne ecnditicn that the plant will be shut denn for ECT examinati on when two leaks are experienced in any 20-day period will previde an early cation-cf any accelerated degradation.
This will add further confidence of steam generater tube integrity.
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Firally, even if a few tubes went undetected by ECT and 5ydretests, became
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pl 1%f*p severely degraded without leaking, and collapsed during a pcstulated 1.0CA, the l
f aw resulting in-leakage would be tolerable because of the collapse failure nede and l
the large leak rates required to adversely affect ECCS performance.
Therefore, the staff has concluded that implementation of the remedial actions
(
prcp: sed by the licensee will assure safe operation of the unit for a cen-At the end of 1
sen'atively established period of 60 effective full ocwer days.
i this pericd an ECT inspection of the steam generatcrs should be perfor ec and results evaluated.
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APPENDIX A j
CALCULATICN OF SECCNCARY-PRIF.ARY LEAKAGE l
First addressing the cut-leakage flow phenonenon the staff assumed a ncminal crevice gap of 0.003 inch, the crack is located mid-depth (about 10 inches) below the top surface of tubesheet and primary-to-secondary ap of 1500 psi, i
f the leakage rate is calculated to be 9.5 gpm.
Within 9 seconds of a LOCA,,
the pressure difference drops to :ero psi primary-to-secondary, the leakage rate would then be zero.
After this time the to reverses and the in-leakage takes place.
l Under LOCA conditions that the in-leakage is of concern, this in-leakage rate l
ds calculated to be 5.5 gpm under the foll: wing assumptions:
P.a s s Fl ux G: 3500 lbm/ft2.see f
Nominal Crevice Gap: 0.003 inch t
Saturation condition of secondary water at the maximum pressure differ-ence of 300 psi.
t I
In addition a conservative calculation was made which assumed gu'~lotine tube rupture at.5 inches below top of tubesheet giving an in-leakage rate of 9.2 gpm.
l Based on the above two calculations, the in-leakage flew rate was estimated to i
be 7 g;m.
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Therefore, a very large number of tubes has to be simultaneously broken in a 1
i guillotine manner to induce a large total in-leakage (>1300 gpm) to be of i
concern regarding the steam binding effect that may slow down the ECCS performance. ;
Thus, the concern in the Wisconsin's Environmental Decade's Petition of November 14 -l l
and 26,1979, regarding the APS study of steam binding is not an applicable cencern in this case.
Fcr:ner, these estimates are conservative in tha: the guillotine break has to be initiated from circumftrential cracks which have not been observed, and the gaps are filled with sludge and not clean as assumed.
L
i i
b APPENDIX B I
ECCS Analysis for 13: Steen Gereretor Tubes Plucced l
Ey letter dated November 19,1979 (Reference 1), as supplemented Novemoer 1
i 26,1979 (Referen:e 2), Wis::nsin Ele::ric Poser Company (the licensee) i submitted an Emergency Core Cooling System (ECCS) reanalysis for Point i
Beach Nuclear Plant, Unit 1.
The analysis was performed assuming 18 i
percent of steam generator tubes plugged.
It supersedes the previous ECCS j
i ana1ysis in which 10 percent of steam generator tubes were assumed to be I
plugged (Reference 3).
Evaluaticn The recently experienced steam generater tube degradation required plugging i
of additional tubes and the level of the tubes plugged is nos at. the 10 percent limit assumed in the current ECCS analysis * (Reference T;.
In order j!
to allow for some additional tube plugging, the licensee has requested that the limi: of steam generator tubes plugged be raised from 10 percent to 15 t
percent.
In support of his request, the licensee has submitted a new LOCA i
analysis based on 18 percent of steam generator tubes plugged (Reference 1).
The analysis was performed with the NRC approved February 1978 version of i
the Westinghouse Evaluation Model (References 4, 5, and 6) and it included I
the following assumptions:
Tc:al Peaking Fact:r: 2.32 Prima ry Coolan: System Pressurt:
2280 psia Core Inle: Temperature:
544*F (nominal value) i Although the submitted analysis was limited to a single break, the DEGCL with C = 0.',the licensee has provided an acceptable justification by referencing D
-10.1; in steam generater A; 9.S" in B.
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I the generic analysis which was performed for the whcie break s:ectrum an:
wnich was previously sucmitted to the NRC (Reference 3).
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The ::nsideratica cf upper pien;= inje: tion (UPI) effect was n :
t incluced in the present analysis.
Henever, it was previcusly demonstrataa (Ref-erences 7 and B) that this effect would cause a 60 F increase in peak i
- lad tem:erature (PCT).
In order to use the preser.t ECCS evaluation i
ncdel to analyze a postulated LOCA in the Feint Eea:n plant and re=ain i
ccc:lian:e with 10 CFR 50.45, a limit c' 214:
s F :n calculated peak clad temperature must be observed.
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The results of the aralysis are provided below:
if i
peak Clad Temperature:
2053*F i
i L :al 2r-Water Reacti:n:
5.11 percent t
i Total Ir-Water Reaction:
less tha'n 0.3 cer:ent h,
I All :r.ese values are below the limits of 1C CFR 50.45 an i
3 a::r::atle.
I'
-Cenclusien i
I, Ease:
en cur review of the submitted dccuments, we conclude froc :ne results i
ef tr.e ECCS analysis perfer ed with the previcusly approved February 1 versica cf the Westinchouse evaluation r.ccel that c: era icn cf F i
i i
Eea:n Unit 1 a: a primary coclant pressure of 2250 ;sia and a peaking facter li.-i; cf 2.32 will be in confermance with the 10 CFR 50.45 criteria i
i 1
We i
- rsicer the ECCS analysis acceptacle for allcwing the plant to te opera:ed i
j vith up ::
a maximum at is percent of stent generator tubes plugged.
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B-3 References 1.
Letter f rom Wisconsin Electric Pcwer Ccmpany (S. Burstein) to NRC l
1 (H. R. Denton), dated Novemoer 19, 1979, transmitting:
ECCS Reanalys for IBt Steam Generator Tube Plugging Licit, Point Beach Nuclear k
Plant, Unit 1.
2.
Letter from Wiscons'in Electric. P>er Company (S. Burstein) to NRC (H. R. Denton), dated November 26, 1979.
3.
Letter frx Wi consin Electric Pcwer Company (S. Eurstein) to NRC (H. R. Denton), dated Maren 20, 1979, transmitting:
LOCA Reanalysis with IC", of Steam Generat:r T:.bes Piugged, Point Beach Nuclear Plant, Units 1 and 2.
1 4.
WCAP-9220-P-A, Westinghous'e ECC5 Evaluation Model, February 1978
- "ersien, Feb ruary 197E.
t 5.
Letter NS-TMA-1951 frx Westinghcuse Electric Corporation (T. M. Andersen) to NRC (J. Stolz), dated November 1,1978.
6.
Letter NS-TMA-2014 from Westinghouse Electric Corporation I!
n (T. M. Anderson) to NRC (R. L. Tecescc), dated December 11, 1978.
li 7.
Letter from Wisconsin Electric Pcwer Com:any (S. Eurstein) to NRC (E. G. Case), dated February 20, 1978.
8.
U.S. t'ucirar Regulatery Comissien, " Safety Evaluation Repcrt on
[
Interim ECCS Evaluation Model for Westinghouse Two-Loop Plants,"
Maren I?7's.
il i
APPEND ~t C REVISE REICTC? COCL*NT ACT!VITv LIMITS _
4, l
We have evaluated the pctential radiolegical consequences of steam line l
1 breaks and steam cenerator tube ruptures for the Point Scach Unit No.1 j
The consecuences of these accidents can be limited to small ol ant.
l fracticns cf the 10 CFR 100 guidelir,es by appropriate limits on the j
i fission product concentrations of the primary coolant.
The present tech".ical specifications do not include a specific limit on iodine I
conctneration n the primary coolant.
In response te the staff's request, j
1 i
the licensee has agreed to operate within the limits of tne staff's
]
Standard Technical Specification en prir.ary coolant activity fcr Point Witn :nis Standard Technical Specification in place, we conclude Se a cn 1.
that the consecueaces of postulated steam line break and steam generater tube rupture accidents would result in deses which would be a small
- raction cf e 10 CF? 100 guiteiires.
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ATTACH"ENT 5 P
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SAFETY EV;tUATION KEPORT RELATED TD i
9
> INT EEACH UNIT 1 STEAM GENERATCR TUEE e
DEGRADATION DUE TO DEEP CREVICE.00RRC$10N i
A:ril 4,
1980 i
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