ML20127F330
| ML20127F330 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 01/11/1993 |
| From: | Perkins K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Marbet L AFFILIATION NOT ASSIGNED |
| References | |
| NUDOCS 9301200206 | |
| Download: ML20127F330 (13) | |
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WALNUT CnEEK, CAUFORNIA 94596 5368 JAN 11 1993 Mr. Lloyd K. Harbet 19142 SE Bakers ferry Road i Boring, Oregon 97009
Dear Mr. Marbet:
This is in response to your letter dated December 28, 1992, which requested a copy of one of the enclosures transmitted with the Dectmber 10 letter you received from Mr. Lawrence Kokajko.
Enclosures 1 and 2 to Mr. Kokajko's letter are transmitted herewith.
Please advise me if you need additional information.
Sincerely, K. E. Perkins, Jr., Director ,
Division of Reactor Safety and Projects
Enclosures:
As stated t
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( Q MEMOPJdlDUM FOR: ERIC S. DECYJORD, DIRECTOR, FROM: OFFICE OF NUCLEAR REGULATORY RESEARCN LAWRDICE C. SHAO, DIRECTOR, DIVISION OF DIGINEERIllG, RES
SUBJECT:
RES POSITION INTEGRITY 011 STEAM GENERATOR TUDE The Division of Engineering has provided a discu alternate plugging critoria (APCaspects of the rationale r tube used to conclusions on its viability as...a)p, and to provide independent critorion.
outer diameter intergranular stress attack (IGA)'a't . corrosionJqracking and case (ODSCC of intersections in the steam 'ge)ne[Aibe rator.
support plate (TSP) possible on data and analyses elsewhere in the rtechnical literature, available from our res s extent arch and technical experience and opinions. The report endbut also represent staff maintain a clear distinction between staff opinion data. eavorsandto published Lawrence C. Shao, Director,
Enclosure:
Division of Engineering Technology .
As stated ,
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,V # 7, Enclosure Discussion of Technical Rationale for Interim (One Cycle) Alternate Plugg CriteriaforSteamf.eneratorTubes The purpose of this report is to provide a discussion of the key technical aspects of the rationale used to support steam generator tube alternate plugging criteria (APC) and to provide independent conclusions on its viability as an interim (one fuel cycle) criterion. A chronology of events relative to this issue is provided as attachment 1.
The interim APC applies and the steam intergranular generator. attack (IGA) at tube support plate (T The technical rationale presented in this report are based to the greatest extent possible on data and analyses available from N researchexperience technical and elsewhere in the technical literature but also represent staff and opinions.
distinction between staff opinion and published data.The report endeavors to main The rationale presented in this report are based on technical considerations which we believe are adequate to justify APC for one fuel cycle.Longer term technical considerations, such as reliatility and sensitivity of NDE techniques for steam generator tube inspection, are the subjects of HRC research which is being coordinated with the Office of Nuclear Reactor Regulation (NRR) as part of an overall APC action plan.
(1)
Background:
Regulatory Guide 1.121 has generally translated into a 40% t
" plugging technical limit" for flaws in steam generator tubes as part of the plant specifications.
However,ieddy current inspection techniques highly reliable for detecting and sizing the cracks un beyond 40% through-wall, further, it has been argued by the industry that the 40% plugging limit is overly conservative, at least for the case of axial ODSCC/lGA cracks confined to tube support plate intersections.
generators, e.g., pulled tubes.over-conservatism is basedThe onclaim burstfortests of cracke revealed short cracks which,Detailed when examinations tested, of these produced tubes have high b correspondingly pressures even for up to through wall cracks.'
with idealized flaws (electrical discharge machined slots) support theNRC r to through wall cracks provided that the cracks are sh "short" can be defined as less than 0.5 inches, which is the length ,of through-wall crack that would be predicted to result in a burst for 7/8 inch diameter,pressure differential 0.050* inch wall thickness (see attachment 2). tubing under main steam line brea analyses, the industry has proposed an alternative to the Regulatory G 1.121 guidance, generator tubes. the so-called alternate plugging criteria (APC), for steam The APC are based on correlations between the voltage
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. . ;b-i amplitude recorded during eddy current tube inspections with a bobbin coil and subsequent measurements of the tube burst pressures and leak rates. The APC i are also currently restricted to 0050C at tube support plate intersections. l 1he Office of Nuclear Reactor Regulati kRR) has approved several licensee i requests for steam generator APC o tlyim(onefuelcycle) basis.
S (2) Technical Precedent: The NR actions accepting industry requests for APC are not the first time th NRC has allowed specific exceptions to the 40% plugging limit for flaws in steam generator tubes.' Plants with steam generator tubes pitted beyond the 40% limit have been allowed to continue operation based on burst test results of pulled tubes. Exceptions to the tube plugging criteria have also been based on the f* criterion. "The f* criterion exempts a utility from the 40% plugging limit for that portion of the defective tubing within the tubesheet below a certain dimension (the F*
distance) from the top of the tubesheet or from the top of the hardroll, whichever is lower."' The purpose of the F* exemptions was to provide a distance within the tubesheet beyond which it could be reasonably assured that a double-ended break in a tube would not result in the tube leaving the tubesheet.
(3) Support Plate Considerations: While there is a small gap between the TSP and the tube, the TSP should provide restraint against tube rupture for cracks contained within the TSP thickness (0.75 inches for the Trojan TSP's). The restraint would be provided by the rigid TSP as the tube expands against it under the influence of a differential pressure such as that caused by a MSLB.
(4) Degradation Mechanisms: While all of the initiation, growth, environmental sensitivity, and synergistic effects of the various mechanisms of steam generator tube cracking are not completely understood, growth of 005CC cracks in TSP regions is not expected to be significant during one fuel cycle. For purposes of this report, significant can be defined as a through- <
wall crack growing on the order of 0.5 inches beyond the TSP. Evidence from the Trojan' pulled tube examinations has shown that the OD lengths of the cracks ranged almost up to the TSP thickness. For Trojan, the cracks were generally confined within the TSP thickness. Thus, for the Trojan cracks to become significant, they would have to extend on the order of 0.5 inches beyond the TSP intersection during one fuel cycle. Upper bound laboratory ODSCC growth rate data indicate that crack growth of this magnitude would not be expected to occur during one fuel cycle, f (5) Probability of Initiating Event (s): The key initiating event for SGTR is considered to be a main steam line break (MSLB). The MSLB would cause approximately a 2600 psi pressure differential across the steam generator tubes. An MSLB has never occurred in a U.S. plant. Quoting from the l
enclosure to Reference 5, "Under the Evaluation and ImDr.ovementJof NDE l Reliability for Inservice Inspection of Licht Water Reactors Proaram sponsored l by the NRC, a team of experts estimated the median frequency of an MSLB to be l 1. 7 x 10 per reactor year . . . . . This extrapolates to a frequency estimate of 6.8 x 10 per reactor year for a four-loop plant.
(6) Steam Generator Tube Rupture Experience: U.S. industry experience since 1975 has documented only 6 SGIR's and only two of those were due to 0050C cracking. The more recent of the two was the SGTR at McGuire Unit 1 in 1989.
The McGuire SGTR was initiated by ODSCC at a manufacturers burnish marking.
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.. c;;a The cracking progressed under the combined influence of residual stresses from the burnish marking and a " local metallurgical contaminant." failure was
' attributed to linking of several axial ODSCC cracks to fom a 3.5 inch OD macrocrack nn the cold leg side of the o m generator. The ODSCC cracking was in the free span away from TSP in
' ions. None of the documented SCTR's were associated with ODSC nt rsections.
(7) Pulled Tube Burst Test Results. >ressure test results from the pulled 1rojan steam generator tubes have shown no leakage under nomal operating or d' MSLB pressure conditions.' When pressurized to failure, the burst pressures measured for the pulled tubes were in excess of the MSLB pressure by at least a factor of two.'
(8) Steam Generator Tube Rupture Tests and Research - Research results from burst tests of tubes using idealized flaws (unifom thinning and EDM slots) has provided a significant body of data on tube integrity. Equations which have been fitted to the data generated for EDH slot specimens have shown that 0.5 inches would be the length of through-wall flaw that would be expected to 9 result in burst at MSLB pressures for 7/8-inch diameter, 0.050 inch wall thickness tubing' (see attachment 2). The equation for the through-wall EDM slot represents an extrapolation from data measured on up to 90% through wall slots. NRC sponsored research has shown that 'the empirical equation developed from EDM notches provides a realistic estimate of remaining margin to failure for tubes with stress corrosion cracking when bounding flaw dimensions are used." An empirical aquation fitted to data from burst tests of uniformly thinned steam generator tubes has also been developed.' This equation is contrasted with the EOM equation in attachment 2. It can be seen that the two equations are of similar fom but that the unifom thinning equation provides more conservative estimates of tube burst pressures for flaw depths greater than a/t of 0.8, where a - flaw depth and t tube wall thickness. Use of either equation to bound degradation below a/t - 0.8 should yield similar results in terms of burst pressure. However, the EOM equation provides a more accurate representation of stress corrosion cracking and should be used for flaw depths greater than a/t - 0.8. j (9) Summary and conclusions: Based on a review of steam generator tube operating experience, on destructive examinations of tubes removed from the Trojan plant, and on expert opinion concerning the frequency of main steam line break, it is concluded that it is reasonable to continue operation of the Trojan plant for one fuel cycle with flaws greater than 40% through-wall at TSP intersections. Q conclusion is supported by the information provided in the preceding paragraphs, which is summarized below. Subsequent operation Thi require acoiuunai TWiew eiter completion of one cycle and will include consideration of information developed at that time.
o Examination of tubes removed from the Trojan steam generator has revealed cracks which are generally confined to the TSP thickness.
Pressure tests on these tubes showed no leakage under normal operating or MSLB conditions. The burst pressures were in excess of the MSLB pressure by at least a factor of two. Further, these tests did not account for the additional restraint that could be anticipated from the ISP which should further increase the burst pressure for the service geometry.
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o Based on empirical correlations between burst pressure and crack size, validated by both laboratory data and by burst tests of tubes removed from service, for cracks to be significant, they must extend outside of the TSP intersection on the order of 0.5 inches. However, based on service experience and examination of tubes removed an steam generators, the existence of cracks extendin preciably be the TSP intersections is unlikely at this timaf _Sase on conserva,ond tive laboratory test data on stress corrosion crack growth rates, it is unlikely g xisting cracks within the TS would extend to a critical length during one fuel cycle. .
Th'us, cracks that are within the TSP can tand the HSLB pressure, and it is unlikely that such cracks would extend to a critical length during one fuel cycle. On this basis, it is concluded that operation for one fuel cycle is justified.
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References s
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1- ' Trojan Nuclear Plant Steam Generator Tu Repair Criteria for Indications at Tube Support Plates,' Westinghouse Energy Systems, WCAP-13129, Revision 1, December, 1991, WESTINGHOUSE PROPRIETARY CLASS 2 2- HUREG/CR-0718, Steam Generator Tube Integrity Program Phase 1 Report, USNRC, September, 1979.
3- itVREG/CR-5150, Steam Generator Operating Experience, Update for 1984-1986, USNRC, June, 1988 4- NUREG CR/5117, Steam Generator Tube Integrity Program / Steam Generator Group Project, final Project Summary Report, USNRC, May 1990.
5- Hemorandum, From C.J. Heltemos to F.P. Gillespie, Gl-163, Multiple Steam Generator Tube Leakage, September 28, 1992.
6- NUREG CR/5796, Steam Generator Operating Experience, Update for 1989-1990, USt1RC, Occcmber, 1991.
7- NUREG CR/2336, Steam Generator Tube Integrity Program, Phase 11 Final Report, USNRC, August, 1988.
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\ Attachment 1 Chronology of staff related events relative to Steam Generator Tube Integrity O The NRR staff had been reviewing Alternate Plugging Criteria (APC' during 1991, and in mid-December 1991, roccived a
' series af requests from Portland General Electric relative to operstian of the steam generators of the Trojan Nuclear Plant, wm rein use of the industry APC was an important issut On December 23, 1991, Mr. J. Hopenfeld filed a Differing O
Professi:nal Opinion (DPO), noting that "Recent experience at the Trojan plant indicates that the present inspection techniques are not sufficiently sensitive to detect steam generator degradation." Mr. Hopenfeld's DPO went on to note his concern "that a Main Steam Line Dreak (MSLB) outside containment could trigger a multiple steam generator tube failure which would than (sic) result in a core melt because of depletion of coolant inventory."
O A Safety Evaluation Report by HRR was published on February S, 1992, wnich allowed operation of the Trojan Nuclear Plant tor one additional cycle (cycle 14) under 3duced leakage allowances, and under repair criteria t. Anelude application of APC.
O On March 16, 1992, Dr. Joseph Muscara published a Memorandum on " Steam Generator Tube Inspection, Integrity and Plugging Issues" wherein he stated his " concerns with generic acceptvice of industry proposals for alternate tube plugging critt ta which would allow operation of steam l genclators with known through-wall cracked (leaking) tuben."
i O on March 27, 1992, J. Hopenfeld filed another paper titled "A New Generic Issue: Multiple Steam Generator Leakage."
! This document contained an analysis titled " Safety Issue f Relating to Continuous Operation with Degraded steam l
Generator Tubes" which indicated "a core melt probability I frequency of 10 E-4/Ry with containment bypass."
O RES moved ti decide on prioritization of this issue, via a Memorandum of September 28, 1992 from C. Heltcmes to F.
Gillespic titled "GI-163, Multiple Steam Generator Tube Leakage;" this prioritization evaluation was revised on November 16, 1992.
1 l 0 On November 24, 1992, N R) . replied via a Memorandum from l Gillespic to Heltemos titled Generic Issue 163, " Multiple Steam Generator Tube Leakage."
0 RES submitted a subsequent Memorandum on Hovember 30 from Heltemos to Gillespie , on the same subject, stating "'te
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'f'+' agree _that the prolininary prioritization of this issue that we sent to you for comment on September 28, 1992 may not have accurately reflected the current censing, position that has been applied to Trojan and ,r plants."
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ENCLOSURE 2 NOTE TO: G. Burdick FROM : J. Hopenfeld
SUBJECT:
Reply to your request for comments on Draf t "RES POSITION ON STEAM GENERATOR TUBE INTEGRITY" by L.C. Shao Based on certain data, discussed in Items 1-8, the document concludes fuel cycle that with"it flaws is reasonable to to continue operation for one intersections" the document greater than
- 40) through-wall at TSP further suggests that " subsequent operation will require additional review after completion of one cycletime".
that and will include consideration of information developed at
GENERAL COMMENT
The information provided in Items 1-8, of the subject document does not address the main issue concerning steam generator tube integrity which arose from recent operating experience with ODSCC.
The issue is as follows:
Is it safe to operate plants where an accident such as steam or feed line break may open existing but previously undetected cracks, leakage. which will result in a significant primary to secondary Whether the leakage is significant or not would depend on whether the operator can stop the leak before the RWST is depleted.
The and fact thatthat cracks their lengthwithin will the notTSP can withstand the MSLB pressure become critical during one fuel cycle is not an indication that they also will not leak. The Trojan burst test results show that three out of the 21 test specimen developed with leaks atdepth an average pressures, of 3300 psi, 7500 of penetrations of psi, and 5500 psi,.
respectively. 38%, 58% and 72%
Item (7) under normal operating or MSLB pressure conditions" points out that the abo IT FAILS TO POINT OUT, HOWEVER, THAT THERE IS NO DATA WHICH WOULD ALLOW ONE TO REL?.TE In other words, THE ifABOVE LEAKAGE these specimen hadWITH THE OBSERVED DEGREE OF DE more severe wall penetration would these specimen had leaked under MSLB loads. Considering that the 21 specimen conclusions in (7)represent above area questionable.
sample of a population of 13,000 the The document does not mention the fact that two other tubes which were pulled out of two US plants developed leaks at SLB pressures which delta ps'. wereAat least third an order of magnitude higher than under normal tube, 8 increase in leakagefrom a Belgian under SLB plant ildicate a factor of considerations, on the other hand, indict con lit ions. Theoretical increase in leakage under SLB conditions.
> a factor of 1000 s
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In. conclusion, the absence of a deterministic and empirical models for these newly observed cracks precludes the conclusions reached in the subject document. The document-claims that the conclusions in Item 9 are supported by items 1-8 could be considered valid only if one ignores the theoretical considerations and the available plant data which indicate that substantial LEAKAGE may result at SLB pressures even if the tube does not rupture.
Finally the justifications for any plant operation should not be based on staff opinions or published data on SCC.
SCC is a semi empirical art, in the absence of applicable database other routes of approaching the problem should be considered.
The justification for operating with cracked tubes should be based .
on what procedures would the operate follow given certain primary to secondary leak and a MSLB between the containment and the PSIV.
These justifications should clearly demonstrate compliance with 10CFR100.
I beleive that the staff can more properly judge operator action than predict localized corrosion behavior.
SPECIFIC COMMENTS!
Item 1.
The EDM initiated groves studies provide some measure of the ability imperfactions.
of the tube to resist rupture given certain known wall It bears little relation to how ODSCC form, propagate and leak in steam generator environments.
Item 4 This definition of "significant" is questionable. It makes no difference whether the cracks extend beyond the TSP if they leak at the g:p.
because they It appears lowered that operators rely on such leakage the leakage requirements during normal operations. Unles one can show that the TSP will cause cracks to plug and they will stay plugged under MSLB pressures the above definition may lead to confusion.
The statement that " upper bound laboratory ODSCC ....
would not be expected to occur during one fuel cycle "
is not supported by data,_The document should compare and present plant and laboratory data with regard to stress intensities..and environments before making such claims.
Item 5 e
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