Text

Informs Commission of Staff Actions Re Steam Generator Tube Integrity at Plant
	ML20126E786
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	12/16/1992
From:	Taylor J; NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	;
References
	SECY-92-412, NUDOCS 9212290328
	Download: ML20126E786 (200)
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VQLICY ISSUE December 16. 1992

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SECY-92-412 Lor:

The Commissioners from:

James M. Taylor Executive Director for Operations 54 - 3 Y[

Satdeci:

TROJAtl STEAM GENERATOR ISSUES Pmpms:

To inform the Commission of staff actions related to steam generator tube integrity at Trojan fluclear Power Plant.

bckotaund:

following an extended outage for inspection and repair of the steam generators, the licensee restarted Trojan on February 12, 1992.

To restart Trojan, NRC authorized the licensee to use " interim plugging criteria (IPC)," which enhanced primary-to-secondary leakage monitoring, and reduced allowable primary-to-secondary leakage limits. On November 9,1992, Trojan was shut down as a result of excessive steam generator leakage.

This leakage was detected and mitigated by the enhancements made as a result of the IPC imposed by the restart amendment.

Upon inspection, it was determined that the leak was due to an improperly sleeved tube in the free span region below the first tube support plate.

A document entitled " Differing Professional Opinion" (Enclosure 1) was submitted by an NRC staff member in the Office of Nuclear Regulatory Research (RES). Although it

Contact:

L. Kokajko NOTE:

TO DE MADE PUBLICLY AVAILABLE NRR/DRPW/PDV IN 10 WORKING DAYS FROM TIIE 50bl369 DATE OF TilIS PAPER g

The Commissioners was not a differing professional opinion (DPO), it has been released to the public and intervenors who label it a DPO.

This document will be called a DP0 for purposes of further discussion in this report.

The DP0 was based on concerns that eddy current testing is inadequate to accurately determine the size of flaws in steam generator tubes, and that core safety is at risk, given a main steam line break outside a containment which could result in multiple steam generator tube ruptures.

The DP0 is further explained in a memorandum from T.P. Speis to J. Hopenfeld dated February 19, 1992 (Enclosure 2).

The NRC staff issued an analysis that supports initiating a new generic issue on March 27, 1992 (Enclosure 3). Additional internal staff correspondence on this issue is provided in a memorandum from Muscara to Serpan dated March 16, 1992 (Enclosure 11), and in a memorandum from Beckjord to Murley dated June 16, 1992 (Enclosure 12).

RES identified the new generic issue (GI-163), with a proposed prelimincry priority ranking of HIGH, and sent it_

to the Of fice of Nuclear Reactor Regulation (NRR) on September 28, 1992 (Enclosure 4) for comment. NRR responded to RES on November 24, 1992 (Enclosure 5), giving its review of the proposed GI-163 priority and recommending that the priority be ranked to LOW based on full consideration of the analysis that sup)orted the Trojan technical specification (TS) change in Fe3ruary 1992.

RES agreed on November 30, 1992 (Enclosure 6) to review the relevant facts and appropriately revise its analysis and priority ranking.

Following the November 9, 1992, Trojan shutdown, Chairman Selin received a November 23, 1992, letter from Mr. Robert Pollard of the Union of Concerned Scientists.(UCS)

(Enclosure 7) regarding concerns related to multiple steam generator tube rupture accidents, with references to the DPO, proposed Gl-163, and the Trojan event.

In this letter, Mr. Pollard requested that_the Commission --

1.

release a number of internal documents;-

2.

direct the staff to _ notify him of any meetings. among-Portland General Electric, Westinghouse, or any other licensees concerning the subjects of steam generator-tube leakage limits and repair limits so that UCS can

. arrange to attend; 3.

direct the staff to explain how a new high-priority, generic safety issue concerning multiple steam generator tube leakage will be considered in a Trojan restart decision; and

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-o The Commissioners 4.

direct the staff to explain why, in view of the new generic safety issue on multiple steam generator tube leakage, the other Westinghouse plants that have been granted license amendments allowing operation with flawed steam q,enerator tubes should not be ordered to shut down until the issue is resolved.

On December 4,1992, Mr. pollard sent a second letter (Enclosure 8) to the Commission.

This letter outlined deficiencies in the staff's response to the generic issue.

[Liscussion:

The licensee determined that the cause of the leak at Tro,ian on November 9, 1992, was an improper stress relief on a steam generator tube sleeve weld.

The T ojar. TS require unscheduled inservice inspections following events that could potentially place unusual stresses on the steam generator tubes (e.g., loss-of-coolant accidents, main steam line or main feedwater-line breaks) or following unanticipated tube degradation indicated by primary-to-secondary tube leakage in excess of the TS limits, in the November 9,1992, event at Trojan, the cause of the tube leak was evident and the potential for further tube degradation was limited to a defined set of tubes and a specific tube region (i.e., tube sleeve locations). The licensee believes that an unscheduled inservice inspection under such circumstances is not warranted because:

1) the specific inspection scope required by the TS would not be necessary to address the identified cause and would result in unnecessary radiation exposure of personnel, prolonged operations at raduced reactor coolant inventory, and unwarranted down time for the plant, 2) the potential for other similar failures was limited to the remaining 1095 tube _ sleeves in service, and 3) the examination of these sleeves following the November 9,1992, event, combined with the review of installation records and additional testing and analyses have demonstrated proper installation of the other tube sleeves.- An emergency TS change was submitted by _ portland General Electric to add a footnote that allows the deferment of the required unscheduled steam generator inservice inspection-until the next refueling outage scheduled-for the Spring of 1993. The licensee plans'to provide the staff a revised TS request with additional technical information in about two weeks.

The documents that Mr. Pollard requested to have placed in the public Document Room (PDR). are predecisional document _s and represent internal staff.

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The Conunissioners deliberations about assigning a priority to a multiple steam generator tube rupture generic safety issue, GI--

163.

Internal and peer review of the proposed GI-163 are continuing, and a final staff position on proposed GI-163 has not been established.

EtAff_Atl.ons:

1.

Despite the predecisional nature of the documents cited by Mr. Pollard, they and other related documents (excluding any proprietary information) have been placed in the PDR to ensure full public understanding of the currert staff postilons on the steam generator tube rupture issue.

2.

A public meeting was held on December 1, 1992, at the Trojan site to discuss the licensee's investigation, root-cause analysis, and proposed emergency TS change and corrective actions for restarting Trojan.

Mr. Pollard was invited to the meeting but did not attend.

The staff will invite Mr. pollard to all public meetings related to restarting of Trojan, in addition, the staff will inform Mr. Pollard through the normal public notification process of public meetings related to steam generator tube issues held with any licensees or vendors.

3.

RES's Division of Engineering presented its views on the interim plugging criteria as applied at Trojan Nuclear Plant in a memorandum dated December 9, 1992 (Enclosure 9).

RES's Division of Engineering concluded that

... operation of the Trojan plant with steam generator tube IPC (interim plugging criteria) for one fuel cycle does not constitute a significant threat to public health and safety.*

in adition J. llopenfeld's :omments on this document are found in a note from Hopenfeld to Burdick dated December 9, 1992 (Enclosure 13).

4.

The staff responded to Mr. Pollard's letter on December 10, 1992 (Enclosu'c 10), informing him of the actions outlined in this paper.

In addition, the letter states that we will provide the final determination of the proposed GI-163 priority. The staff is rosionding to the technical issues discuneJ riy Mr. Pollard in his second letter.

1 1

O O

i The Commissioners.

5.

The licensee withdrew its first TS change-request for a limited inspection program, which-stated emergency circumstances were present and requested expedited processing.

The staff will-review any requested TS change. A TS change may be required before the licensee changes the plant condition to Mode 4 (hot shutdown) if the licensee intends to complete a limited inspection )rogram..The staff will complete its review of tie requested TS change, and, if we determine the change is a)propriate and safe, we may issue the requested clange on an exigent basis.

This issuance would include a no significant hazards consideration determination.

6.

'Following full consideration-of the corrective actions for the November 9, 1992, event,- the staff will decide whether to approve the restart of Trojan.

b

o The Commissioners 7.

The staff is continuing the review of the issues of proposed GI-163 and will inform the Commission of the final staff position.

The staff will inform the Commission of any subsequent actions.

/

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/

JamesM.Tahlor E/ecutive' Director

/ for Operations

Enclosures:

1.

Differing Professional Opinion dated December 23, 1991 (Hopenfeld) 2.

Memorandum dated February 19, 1992 (Speis to Hopenfeld) 3.

Memorandum dated March 27, 1992 (Hopenfeld to Beckjord) 4.

Memorandum dated September 28, 1992 (Heltomes to Gillespie) 5.

Memorandum dated November 24, 1992 (Gillespie to Heltemes) 6.

Memorandum dated November 30, 1992 (Heltemes to Gillespie) 7.

Letter dated November 23, 1992 (Pollard to Selin) 8.

Letter dated December 4, 1992 (Pollard to Selin) 9.

Memorandum dated December 9, 1992 (Shao to Beckjord)

10. -Letter dated December 10, 1992 (Murley to Pollard) 11.

Memorandum dated March 16, 1992 (Muscara to Serpan)

12. Memorandum dated June 16, 1992 (Beckjord to Murley)

13. Note dated December 9, 1992 (Hopenfeld to Burdick)

DISTRIBUTION:

Commissioners OGC OCAA OIG OPP REGION V EDO ACRS SECY

DIFFERING PROF.ESSIONAL OPINION Recent experience at the Trojan plant indicates tnat present inspection techniques are not sufficiently sensitive to detect steam generator tube degradation. The problem is i

i l inherent in the oddy current probe design and its use it is essentially mposs b e to 1

detect tight through the wall cracks, especially at the tube support plate regions.

The plants were not designed to operate continuously with a large number of tubes containing through the wall cracks.

My concern is that a Main Steam Line Break (MSLB) outside containment could trigger a mult;nle steam generator tube failure which would than result in a core melt because of depietion in coolant inventory.

NRC is currently addressing the uncertainties in the in service inspection procedures by considering the possibility of allowing affected utilities to operate with tube imperfections beyond the 40% tech specs through the wall plugging limit.

White, the above action is useful for the long term, I believe it is not focused on the main issue. The main issue is whether the core can be maintained intact and radoactivity release prevented with a MSLB outside containment and multiple steam p:enerator tube rupture. While considerable research will be required to define a new plugging limit and change the SRP, the result will not increase plant safety. The basic problem is with the NDE procedures and their inability to predict tube degradation and leakage.

Rather than concentrating efforts on alternate plugging limits, the NRC should request all affected licensees to provide warranties that they have the capability to keep the core intact and prevent allowable dose releases with a MSLB and a multiple tube rupture of no less than 80% of all tubes.

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ENCLOSURE 1

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FEB 1 9 1992 HEMORANDllH FOR:

Joram Hopenfeld Reactor and Plant safety issues Branch Division of Safety Issue Resolution Offica nf Nuclear Reaulatory Research IROM:

1hemis P. Spets Deputy Director for Research Office of Nuclear Regulatory Research

SUBJECT:

YDUR DIFFFRIHr, PROFESSIONAL OPINION (0p0) DATED 12/23/91 On January 24, 1992, a number of RES staffers', including ntyself, met witii you to discuss and explore the various issues / concerns associated with your DP0 related to multiple steam generator tube failures ([nclosure 1). Based on my reading of your DPO, it was not clear to me with what staff position you differed; or whether based on recent experience with detection of steam generator tube degradation (e.g., as you indicated at the Trojan plant), you were recommending the fomutation of a "new" staff position; the objective of the meeting was then to informally explore _and gain a more precise understanding of your concerns and recommendations, including their bases.

Based on your DP0 as described in your 12/23/91 memo, and the discussions with you and the other RFS staffers at the subject meeting, yuu are concerned that "a Main Steam Line Break (MSLB) outside containment could trigger a multiple steam generator tube failure which would result in a core melt because of a depletion of coolant inventory" and as a result, you recommend that "the NRC should request the affected licensees to provide warranties that they have the capability to keep the core intact and prevent allowable dose releases with a MSLD and a multiple tube rupture of no less than 00% of all tubes ' _This is not within the design basis and interpretation of GDC 31 as currently applied by the staff.

In sumary then, based on the discussions we had at the subject meeting, it seems that you are recomending the development of a new staff position.

As you know, we have in place at NRC a process-(described in NUREG-0933 and managed by the Office of Research) whereupon when an issue is identified, it goes through a prioritization process (it involves other Of ficos plus the ACRS) to determine whether the identified issue merits further consideration, including its designation as either a USI or a generic issue. Once it has been categorized as either a U51 or a generic issue, a program plan is developed, including a schedule for its resolution _[ including All the appropriate interactions with the ACRS, CRGR, and Commission) and, depending on the resolution / findings, appropriate staff positions are developed and recommended for implementation, normally via a rule, regulatory guide, changes to SRP, etc.

R, Bosnak, J. Huscara, B. Sheron, P. Norian and L. Gallagher.

ENCLOSURE 2 t

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J. HnPnfeld 2

N O I S 1932 Therefore, based on our discussions and in order to facilitate the process described above, we agreed at the subject meeting that you should provide more detailed information (in fact you told us at the meeting that you were in the process of doing just that) about the basis for your concurn so that we can proceed to prioritize this issue, One of the reasons for the need of this information is to estimate to the extent possible the probability of the event which you have postulated, the associated consequences and the cost of potential backfits which (guld prevent the event from taking place or giltjgt(p its consequences.

Therefore, it will be useful to define the issue as precisely as possibic and provide any further information (and its basis) which you have concerning the probability of a large number of tubes with sufficient degradation which could then fail coherently as a result of a MSLB, the probability of periodic eddy current testing not finding the degradation, the location and the associated probability of a HSLB the probability that the forces associated with a HSLB are oeyond the capacity of the degraded tubes which would then lead to coherent f ailure of large numbers of tubes; the probability that this event would go undetected long enough (i.e., before isolation of the f aulted SG) and that the coolant inventory will be ' lost' outside containment and thus not availablo for the ' recirculation phase

of core cooling.

for more guidance and details on the procedures to be followed and the information to be provided so that a draft prioritization of the issue can be prepared, see RES Office letter No. 1, Revision 2, ' Procedure for Identification, prioritiration, and Tracking of the Resolution of Generic Issues,* dated July 12, 1991 (Enclosure 2).

Especially valuable to you will be Attachment 1 to the RES Office Letter which lists the type of information needed in order to allow RLS to make an assessment of the importance and priority of the issue, ongi.,ig in ThemYP hpYis Deputy Director for Research Office of Nuclear Regulatory Research inclosures:

1.

DP0 frm. J. Hopenfeld dtd.

12/23/91 2.

Memo to Office Dir. and Reg.

Dir frm. E. Beckjord dtd. 7/12/91 cc:

LBeckjord JHeltemes WHinners RBosnak USheron A0urda Distritqt em subj-cir chron TSpeis Offc NQ R Name@4 y

Date: 2/ /92 0FFICAL RECORD COPY

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nocent experience at the Trojan plant indicates that present inspection techniquos are not suthciently sonsitrve to detect steam generator tubo degrodation. The problem is inhofent in the oddy Current probe design and hs use. it is essentially impossibic to detect tight through the wall cracLa, especially at the tube suppon plato regions.

The plants woro not designed to operato continuously with a largo number of tubos containing throu0h the wall cracks.

My concem is that a Main Steam Une Break (MSLB) outsido containment could tri90er a muttepte stoam generator tube failure which would than result in a core melt because of depletion in coolant inventory.

NHC is currently addressing the uncertainties in the in service inspection proceduren by considering the possibihty of allowing affected utsittics to operate with tube imperfections boyond the 40% tech specs through the wall plugging limit.

While the above action is usolul for the long term, I believe it is not focused on the main issue. The main issue is whether the coro can be maintainod intact and radioactMty release prevented with a MSLB outside containtnent and multiple steam Ococrator tubo rupture. While considorable research will be required to defino a new plugging limh and chanGo the SRP, the resutt will not increaso plant safoty. The bdsic problem is with the NDE procedures and their inability to predict tubo degradation and leaka01 Rather than concontratin0 offorts on attemate plugging limits, the NRC should request att attected licenscos to provide warrantics that they have the capability to keep the core intact and prevent allowablo dose releases with n MSLO and a multiple tube rupture of no less than 80% of all tubes, J.

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FROM:

Eric 5. Beckjord, Director Office of Nuclear. Regulatory Research

SUBJECT:

RES orFICE LETTER NO. 1, REVISION 2, " PROCEDURE FOR IDENTIFICATION, PRIORITIZATION, AND TRACKING OF THE RESOLU-TioW 0~ EENERIC-ISSUES" 4

As a result of the NRC reorganization in April 1987,'the functional respon-sibility for the early stages 'of generic issue management was transferred to the Office of Nuclear Regulatory Research (RES). -RES.fice Letter No. 1 (OL-1) was published on December 3,1987, to replace the guidance previously.

provided by NRR Office Letter No. 40,. and Revision 1 was published on March 22, 1989. The purpose of this revision to RES OL-1 is to reflect recent

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changes in the generic issues prioritization procedure brought about by the recent transfer of the' function from the Division of Regulatory Analysis to-i the Division of Safety Issue Resolution. One major change is that NRR is asked to participate in the peer review process of those prioritizations for.

which RES recomends a low or Drop priority, but need not concur for GSis prioritized.riigh or Medium. = NRR will continue to participate in.the peer review of the-final-resolution packages of all such issues.

The generic issue process consists of six phases:

Identification, Prioritiza-tion, Resolution, imposition, Implementation, and Verification. The enclosure to this letter specifies the-procedure to be followed for the management of generic issues through the first two stages (Identification and Prioritiza-tion) as well as the tracking of those issues through their resolution. ' The procedures for managing generic issues through the Resolution stage (RES Office letter No. 3) and the imposition, Implementation, and~ Verification stages (NRR Office Letter No. 25) arp provided separately. This-procedure was-developed to provide a mechanism to document new safety concerns with existing and future reactors and to.have the RES staff formally evaluate these concerns for. safety significance and appropriate action.- Since potential generic.

-issues may arise from different offices within Headquarters or from sources 4

outside of the H '.iquarters staff such as the' ACRS, Regional Offices or the-public and since the prioritization of these issues may involve review by.

other offices, this pcocedure is being provided outside RES for information

__ _ - __ a

'I Multiple Addresses

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and use, as appropriate. RES intends to accept potential generic issues for evaluation from any source provided they generally follow the attached procedure and provide adequate information regarding the concern (as prescribed in Attachment I to the Procedure) to allow RES to make an assessment of their priority.

This procedure is not intended to be applied to screen all current or future work going on within NRC (which may be initiated through various mechanisms) but should be used for new potential generic issues associated with nuc; ear power plants.

3.hkd Eric 5. Beckjord,lJ)irector Office of Nuclear Regulatory Research

Enclosure:

Procedure for identification, Prioritization and Tracking of the Resolution of Generic issues cc:

J, laylor, EDO RFS Staff s

DCLostLRf.

P.80CEDVRLf.0R..!Dmilf1W10L_fil08111ZATION AND TPACKlHG OLTHLRES0WJ10tLOLEGGJL11M5 JEIE00VC110tf A generic issue is an issue that is applicable to all, several, or a class of reacturs or reactor related facilities. Generic issues can arise from various concerns and, accordingly, are classified into one of the following four l

categories. A Generic Safety issue (GSI) is a generic issue that involves a-safety concern that may affect the design, construction, operation or decom-missioning of all, several, or a class of reactors or facilities and may have a potential to require iicensees to make safety improvements and/or require the promulgation of new or revised requirements or guidance. A Regulatory Impact Issue (RI) is a generic issue not related to in) roving safety, but to modifying current NRC requirements or guidance, with the primary purpose of reducing the regulatory impact. usually cost, of requirements on licensees or applicants. An Environmental Issue (EI) is a generic issue involving istpacts on those items protected by the National Environmental Policy Act (NEPA). A Licensing Issus-(LI) is a generic issue related to actions the NRC staff should take to increase knowledge, certainty, and/or understanding in order to increase confidence in assessing levels of safety; improve or maintain the NRC capability to make independent assessments of safety; establish, revise, and carry out programs to identify and resolve safety issues; document, clarify, or correct current requirements and guidance; or improve the offectiveness er efficiency of the review of applications.

lhe generic issue management program is divided into six distinct stages:

identification, prioritization, resolution, imposition, implementation, and verification.

The procedure described herein is to be utilized only for the identification and prioritization stages of generic issue manages >ent, includ-ing their tracking through resolution.

Procedures for the management of generic issues during resolution, imposition, implementation, and verification are not covered by this procedure.

In the identification stage, potential generic issues may be suggested by organizations or individuals within the NRC, the Advisory Comittee on Reactor Safeguards (ACRS), the nuclear power industry, or the public. Generic issues may also be suggested as an outcome of reactor research programs. Potential risk to the public is the principal consideration in suggesting a potential generic issue. Once suggested, potential generic issues are screened for duplication, overlap, or integration with existing generic issues. When a potential-generic issue is accepted as a new generic issue it is assigned a number and title, the scope of the issue is defined, classified by type (GSI, RI, El or LI), and catalogued with all other generic issues #,u NUREG-0933, "A Prioritization of Generic issues."

Each new GSI is normally prioritized by developing a quantitative assessment of safety benefits (risk reduction) and NRC and utility impacts (cost) as described in NUREG-0933. Based on the extent of potential risk r M uation tn the public and the value/ impact ratio developed from this assessoc.t. ed at i

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2 further adjusted by qualitative judgments and other considerations, a priority is assigned to each GSI.

Ris, Lis, and Els are evaluated and quantitative and qualitative estimates of their merits are described. The preliminary priority assesstnents for each issue are sent for peer review and coment. These peer review coments are then addressed, the original preliminary assessment revised, as necessarv, and a final priority recomanded, as appropriate.

Issues which are assigned a HIQi or MEDIUM priority move on to the resolution state. All HIGH priority Generic Safety issues are screened for additional designatto.n as an Unresolved Safety issue (USI). An issue assigned a LOW or DROP priority by nature of the rating standard is of so low a public risk reduction potential that resolution of the issue is not pursued. All issues-are documented in the catalogue of generic issues maintained in NUREG-0933.

Generic issues not initially assigned staff resources for resolution (LOW or DROP) may be reactivated in the future if new information becomes available which may change the original priority assignment or change their classifica-tion. NUREG-0933 serves as the repository for the priority assignment and resolution for all identified generic issues. NUREG-0933 is updated semiannually.

Ris. Lls, Els, and HIGH and HEDIUM priority GSis are assigned by the Office Director, Office of Nuclear Regulatory Research (RES), to the approariate NRC Office for resolution. Most GS!s are usually resolved within RE5; aowever, from time to time other offices are assigned G51s for resolu-tion.

Lis are assigned to NRR for disposition. RIs and Els are assigned within RES, and decisions to work on the resolution of these issues are made by qualitative judgment and the availability of staff resources.

In the resolution stage, an in-depth technical evaluation of the issue is performed by the office assigned the task of resolution.

The status of issue prioritization and resolution is tracked in the Safety issues Management System (S!HS). For each issue, the SIMS includes a synopsis of the issue, work scope, work status, and program milestones.

PROCEDURE:

1.

IDENTIf! CATION:

a.

Anyone inside or outside NRC can identify a proposed generic issue. A generic issue may be proposed by an individual or by an organization unit. However, when proposed, an attempt should be made to include all the information specified in Attachment I so that there is a clear understanding of the issue and its safety significance.

b.

Proposed generic issues submitted to NRC's Office of Nuclear Regulatory Research (RCS) should be addressed to the Director, RES.

c.

RES/DSIR will screen all proposed generic issues for duplication or overlap with previously identified generic issues and to see if, in fact, they are generic, not plant specific issues.

proposed issues that may be plant specific will be sent to NRR for rN iew and appropriate action.

a 3-d.

for' each generic issue accepted, RES/DSIR will assign it a'nutber and will maintain a log of its status and disposition, RES/DSIR will promptly advise the originator of the receipt and initial disposition of the issue. This dis >osition may include a deter-mination that the issue is covered >y another existing issue or Multi-Plant Action (MPA), that it has been accepted for prioritiz-i ation, or that additional information is needed, e.

After their acceptance, generic issues that originate from outside NRR or from an individual within NRR (i.e., not sent through NRR management) will be transmitted by RES to NRR/PtRS for an imediate action determination and screening for identification of overlap or duplication with already imposed or completed Multi-Plant Actions (MPAs).

If HRR cannot complete the inmediate action determination and HPA screening within 15 days, RES should be informod when the NRR review will'be completed.

2 PRIORITIZATION a.

RES/DSIR will classify each accepted generic issue as a GSI, RI.

Li, or EI. Based upon its classification, one of the following actions will be taken:

1.

Issues classified as RIs, L!s, and Els will be evaluated and their merits quantitatively and qualitatively estimated and described in a preliminary assessment report.

2.

Issues classified as GSis will be evaluated by RES/DSIR and assigned a preliminary priority estimate based on engineering judgment and/or very rough quantitative risk calculations.

The preliminary estimate (i.e., safety significance) will be used by DSIR to establish an order for entering GSis into the prioritization process; i.e., those with preliminary priority estimates of HIGH will be entered into the process before GSis with preliminary priority estimates of HEDIUM, LOW and DROP, MEDIUM before LOW and DROP, and LOW before DROP.

b.

DSIR will prepare a draft prioritization write-up for each generic issue using the methodology described in NUREG-0933. GSIs will be assigned a priority ranking of HIGH, MEDIUM, LOW, or DROP based on their estimated public risk reduction potential, their value/ impact ratio, and other considerations, RES/DSIR will send the draft prioritization write-up of each c.

generic issue to appropriate NRC personnel.for peer review prior to finalizing their priority. Those involved in peer review will include the DSIR Division Cirector, DSIR Sranch Chiefs, NRR/PfMS (for distribution to cognizant NRR management and Staff), as appropriate, and the originator.. NRR will be requested to participate in the peer review process only for GSIs with a priority recomendation of Iow or Drop. NRR/PHAS will be provided L

o s

4 information copies of the priority write-up for GSIs with a High or Medium priority recomendation and issues classified as RI. El or Lt. The offices designated for peer review for an individual issue are to provide coments on the draft prioritization and the priority ranking to RES/DSIR within 15 work days of receipt. NRR will perform a screening review of the prioritization and will perform a full staff review only if the result appears questionable.

In that case, a review schedule will_ be developed by NRR and pFS will be informed of that schedule, d.

Based upon the results of the peer review, RES/DSIR will revise the draft prioritization to address the coments or identify the differences, include a recomended final priority ranking, as appronriate, and submit it to the Director, prs, for approval, e.

For GSis' approved as HIGH or HEDIUM priority the Director, RES, will assign them to the appropriate RES Division for resolution or request an other NRC Office to resolve them.

For issues approved as Lis, the Director, RES will send the final assessments to RES or NRR for disposition, as appropriate, based on the nature of the issue. For issues approved as Ris or Els, the Director, RES, will assign the issues within RES based upon a decision to work the issues made by qualitative judgment and the availability of staff resources.

In addition, the Director, RES, will send the prioritization assessments of all approved generic issues to other Offices, the ACRS, and the Public Document Room for infomation and coment, f.

RES/DSIR will review those prioritized as HIGH and provide a recomendation to the Director. RES as to whether they should be designated as candidate Unresolved Safety issues. Criteria to be used are documented in NUREG-0705.

3.

TRACKING a.

Each NRC Office assigned the task of resolving one or more generic issues should prepare and implement a plan defining the respon-sib (11tles, process, and schedule for resolution. The Office Director assigned the task of resolution of a newly approved issue will submit a copy of the work plan, including a detailed schedule and the plan for the regulatory analysis, to the Director, RES, within 6 weeks of being assigned the issue. This submittal shall contain the information listed in Attachment 2.

b.

The Chief of the cognizant NRC branch will submit status reports for each approved work plan to the Director, DSIR, quarterly _ or as requested. When a status report indicates slippage of the' estimated 'resolutton completion date, the revised work plan must be approved by the cognizant Office Director and by the Deputy Director for Generic Issues and Rulemaking, RES.

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A 5-c.

RES'/DSIR will provide-the-approved work plans for.all generic issues as input to SIMS, RES/DSIR will provide updates to'SIMS quarterly :to-incorporate approved work plans for new generic -

issues and incorporate modification to-and/or changes in scheduler information for cristing work plans.

d.

RES/DSIR will issue a quarterly status summary of Generic issue resolutio n progress through the Generic issue Management Control Systees (GIMCS), by the second week of each quarter. Quarterly reports will also be providea by the-Director, RES, to the EDO highlighting progress, problem areas, and schedule changes.

. c.

RES/051R will update NUREC-0933 semiannually to catalog new..

generic issues, document the progress of staff efforts during-the-prioritiration process, and document the priority assignments made for peneric safety-issues.

-Attachment 1 4

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M1EhmtaL1 GENERIC ISSUE INFORMA110N To the extent practical, the following information should be provided in suffletent detail to permit the proposed issue to be analyzed and prioritized with a minimum of additional information gathering.

1.

A title for the proposed generic issue should be suggested. While brief, the suggested title should attempt to define the specific nature and scope of the proposed issue.

2.

Potential, suggested, or known deficiencies in the technical bases of existing staff guides or requirenents should be identified (i.e.,

Regulatory Guides, Standard Review Plan Sections, Rule, etc.).

For proposed issues suggested by examination of LERs a complete listing of applicable LERs and/or a complete set of copies of the applicable LERs should be provided.

3.

A description of the proposed issue should be provided which discusses the background (bases) and perceived safety significance of the issue (i.e., contribution to risk, core melt frequency or public dose). The issue should be scoped to identify those individual plants or classes of plants affected by the proposed issue.

4.

Sufficient attention should be devoted to the proposed issue to suggest a potential solution and/or alternative solution (i.e., design and hardware changes and/or additions; procedural changes; changes in plant staffing and/or management; accident management chanoes, etc.).

5.

The suggested solution and/or alternative solutions should be evaluated in sufficient detail to determine whether the solution (s) would be expected to result in:

a) the need for additional research, staff studies, testing, new procedures, rulemaking, etc.;

b) increase or decrease in operational exposure of the plant operat-ing staff; and c) a plant shutdown or extension of a refueling outage to implement the potential solution (s) for the proposed issuo.

6.

A preliminary value/ impact assessment should be provided for the potential solution (s) for the proposed issue. The reference documents listed below provide methodologies for both risk and cost analysis, and illustrative examples.

7.

Name(s) and organization (s) of all persons currently working on this issue should be provided.

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The name 'of 'the. person supplying preliminary value/ impact assessment

)

.information should be provided.

i 9.-

Appropriate references (memorandum, NUREGs, SRPs, etc.)-should be provided.

10.

The transmittal memorandum should reflect th'e concurrence of.the office -

of the originator, if possible; however. this is not mandatory.

1 Rtf m nee Documents HUREG/BR-0058, Revision 1. " Regulatory Analysis Guidelines for the U.S.

Nuclear Regulatory Comission,' May 1984

]

NUREG/CR-3568, *A Handbook for Value-Impact Assessment,' December 1983, 4

NUREG-0933, "A Prioritization of. Generic Safety Issues,* Decembe'r 1983, b

NUREG/CR-2800, Guidelines for Nuclear Power Plant Safety l Issue'Prioritization Information Development,' February 1983 and Supplements: 1,t 2, 3, and_4..

NUREG/CR-3971, 'A Handbook for Cost-Estimating,' October 1984-

[

NUREG/CR-4568, "A Handbook for Quick Cost Estimates," April 1986.

l l-AE00 Procedure 3. " Application of Risk' Perspectives: A: Procedures Guide,"

[

Peter Lam, U.S. NRC, October 15 1984 4

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At1LChattnL2 Generic Issue _ Ban 190 ment Control Information 11em_Humber (Generic Issue Number) 11_tle (Cenaric Issue Title)

Lead Office /Div/Br As appropriate Other Qfitce/D1yffr As appropriate It5LBAAAptr (Name) lac _MEstr (As tssigned) hr_LAuthor_ization (if different from Parts A, B, and C of Appendix F from Operating Plan)

(ontract Title Provide Contract Title (if contract issued)

Contractor _Name/

Identify Contractor Name and FIN Number FIN NO.

(asapprop,riate) york Scone Describe briefly the work scope for completing the issues Affected Dpigmenti Issue NUREG-Revise and issue Regulatory Guide 1.xx: Revise and issue SRP Section x.x.x., Revise and process STS change Jr.chnical Resolution Select milestones from the initial date Division Director was requested for infonnation through issuance of revised.SRP change.

For the most part the selected allestone dates will vary from issue to issue. Typical milestones should include but are not limited to those on the following page.

Status Describe current status of work.

Problem /Res olution Include potential problems and actions being taken to resolve them.

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Milestone Examples Driainal -furrent' agiy31' o Date information:recuested from division j

o -Date received from Division

-o Proposal-Solicited o'_ Proposal Evaluated and Accepted-o -Contract: Scheduli, if appilcable o Testing Schedule, if applicable o Draft'NUREG/CR report-from-

. i contractor / consultant o Staff revfew of draft NURfG/CR

-report o - Value Impact-Statement _ prepared o. Final. report prepared by Olvision-4 o Final: report forwarded to RES for:

processing o RES Director Review. completed o-Review Package to CRGR:-

o -CRGR review completed o EDO~ approval o Federal Register Notice of -

Issuance of SRP for Public Coement-o:- 0MB. Clearance if' applicable o : Division review of public comment completed '

o' -RES Director review completed o-CRGR review completed.

.o EDO. approval-o_ Federal Register Notice of issuancel of SRP o

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i laAR E l 1932 MEMORANDUM FOR:

E. Beckjord, Director Office of Nuclear Regulatory Research FROM:

J. Hopenfeld Reactor & Plant Safety Issues Branch Division of Safety Issue Resolution, RES

SUBJECT:

A NEW GENERIC ISSUE: HULT!PLE STEAM GENERATOR LEAKAGE The enclosed analysis ' Safety issue Relating to Continuous Operation With Degraded Steam Generator Tubes" indicates a core melt probability frecuency of 10"/Ry with containment bypass.

The analysis is submitted for your evaluation and action as appropriate in accordance with'RES office letter fl.

/6/

J. Hopenfeld Reactor & Plant Safety Issues Branch Division of Safety issue Resolution, RES

Enclosure:

As stated DIST:

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MAR z 71992 NEMORANDUM FOR:

E. Beckjord, Director.

Office 'of Nuclear. Regulatory Research FROM:

J. Hopenfeld Reactor & Plant-Safety _ Issues Branch Division of Safety-Issue Resolution, RES y.

SUBJECT:

A NEW GENERIC ISSUE: MULTIPLE STEAM GENERATOR LEAKAGE 1 The enclosed analysis " Safety Issue Relating to Continuous-Operation With Deg,raded. Steam Generator Tubes" indicates a core melt probability frequency of-10' /Ry with containment bypass.'

The analysis is submitted for your evaluation land action as appropriate in accordance with RES office letter fl.'

/GI J. Hopenfeld 1

Reactor & Plant Safety. Issues Branch Division of Safety issue Resolution,= RES

- Er. closure:

As stated DIST:

RPSIB r/f' DSIR c/f Hopenfeld:CY King:CY Minners:CY-Kniel:CY

~RPSIB HOPENFELD 3/V/92

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r SAFETY ISSUE RELATING TO CONTINUOUS OPERATION WITH DEGRADED STEAM GENERATORS-IN PWR's i

INTRODUCTION Recent operating experience in PWR plants, Reference 1, indicates that tight narrow cracks in steam generator tubes are not all being detected with Eddy Current (EC) probes. While such cracks may not leak under normal operating conditions they could leak when subjected to sudden high stresses. The-pressure differential across the tube wall which would follow a, steam line break accident,SLB, might provide such stresses.

Based on burst tests of sample tubes which were pulled out of _ service, laboratory data, and analysis, the NRC believes that plant safety is not compromised by the degraded tubes, (Reference 2, 3). The Trojan Plant is allowed to operate with more than 600 hundred defective tubes. The NRC-adopted this~ position on the premise that defective tubes will leak-prior to t

rupture and the leaks will be detected in a timely manner.

This writer believes that the above information,1 however,:is not a sufficient -

safety basis for continued operation with defective tubes. This concern was documented in a DPO, Reference 4_which in-response,Jthe Office of RES request-

.ed, Reference 5, that additional information be submitted in accordance with RES Office Letter No. 1.

This document is a response to the'above request.- Its main purpose is_to_-

conduct a preliminary-evaluation to show that continuous operation:with -

4 degraded tubes con ~stitutessa safsty risk. This-risk, however, can be~

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- mitigated by insuring that a sufficient reserve of borated water is' available

'for ECC injection-~at each plant site.

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5 ANALYSIS The f act that degraded tubes neither leak, at normal pressures, nor burst under SLB pressures is not an indication that they will not leak following a SLB accident.

The attached RELAP sample calculations (ATTACHMENT 1) show that the total leakage, not leak origin, is the determining factor whether the plant can be brought to a safe shut down.

It makes no difference whether the leak origin was from one ruptured tube or many pin hole leaks.

In the example calculations, when the leakage is above 650 gpm, Table 1 and Figures 1-5 show that the leak flow must be terminated in less than eight hours to prevent depletion of the refueling storage tank.

Because secondary pressure is near atmospheric following an unisolated SLB, it is difficult to reduce primary pressure below that of the secondary side in a timely manner.

Consequently, the 1~-k flow could continue for extended period of time causing

~

the eventual depletion of the RWST. Operator action to m'inimize the primary pressure can delay the time to exhaust the RWST, however unless the break flow is terminated, a means for replenishing the RWST appears to be the only viable solution.

The frequency of SLB accident outside containment without the ability to isolate the affected steam generator is postulated (Reference 6) to occur at a 10-4 /RY.

If a high leakage was to follow it could lead to a core melt because the RWST will be depleted in a period of five to eight hours, as shewn in the attachment and discussed above The NRC also assumes (Reference 6) with a 99.99 certainty that the operator will be able to depressurize the secondary

6 in the above time. The RELAP results, with no credit for operator action, are used in this report. What the operators could do to mitigate the accident will depend on the leak flow rate and is beyond the scope of this study.

The determination of core melt frequency can be obtained by multiplying the probability of leakage following a SLB by the probability of-10-4. The detennination of the leak flow from all the degraded tubes requires knowledge of the leakage from each degraded tube and the ~ total number of affected tubes.

From Laboratory data of precrack specimens (Reference 7) one can only conclude-that leakage under SLB loads is higher than under normal operating conditions.

The cracks in the above specimen were generated in a non prototypical environ-ment and the leak tests were of short duration, therefore,-the data cannot be used for leakage estimates in an actual plant, (see Attachment 11 for addi-tional discussion).

Plant data is not available on leakage of tubes with through the wall cracks at SLB pressures.

However, the available plant data suggests that there is 'a high probability that-a leakage will occur but the. data is too meager to allow meaningful leak flow estimates. Twenty one specimen which were removed from the Trojan plant, Reference 8, show that the depth of penetration will determine whether the a tube will leak when subjected to high pressure differentials. With the exception of two specimen all the other failed without leakage, on ascent to burst pressures. The two specimen that leaked

- prior to rupture, however, also exhibited very deep cracks (98% max.). The leak occurred at high than SLB pressure but below the burst pressure. The

7 only conclusion that can be drawn from these twenty one tests is that crack morphology will determine whether a tube will or will not leak at certain applied pressure.

Tube R4-C73 and Tube R21-C22 were pulled from other U.S. plants (Reference 10). Under steady state (delta P) the aoove tubes leaked between 0

.3 ml/hr and 0 - u 7 ml/hr.

In contrast, under SLB delta P the tubes leaked at a rate of 174 ml/hr and 108 ml/hr.

In another case, a tube at a Belgian react $r (R19 - C35) was leak tested in a laboratory and~ found to leak at a rate of.07 gpm at normal operating pressure and.53 gpm at SLB pressure.

The above plant data indicates a high probability of leakage __with through the wall cracks and a significant increase in leakage when SLB-loads are applied instead of normal operating loads. One may conclude that the probability is one that some leakage will follow a SLB if the defects have penetrated the tube walls.

EC analysis of degraded tubes is more an art than a science and, therefore, a proper evaluation of probe signals require a good knowledge of stress corro-sion.

In spite of considerable research in this a'rea for the last thirty years the ability to predict crack propagation in the field is still very.

. limited. No practical methods are available to predict probability of leakage from periodic tube inspection. Also,-current practice is to shut down the plant when leakage occurs rather than conduct inspection on the predication of a leakage probability.

Current RES-aging research does not appear to be designed to provide practical information to reactor operators in this regard.

I

4 8

Data on leakage as a function of crack morphology will be required to determine how degraded tubes will behave during_the accident. Since experience shows that crack morphology varies not only with location within the tube bundle but also from one reactor to another and from one operating cycle to another the generation of such data-is not a practical solution for leakage probability determination.

In conclusion there is no way:of predict-ing how many tubes will develop deep micro cracks, how many of them will' leak and how much will they leak during an accident.

, provides additional-examples of why present data is not usable for leak flow estimates fullowing SLB. The main conclusion drawn-from these examples is that the laboratory data used by Westinghouse does not support Westinghouse conclusion that the leak flow following.SLB is very small_ as long-as the bobbing coil probe voltage is below 2 volts.

Even if one, for an instant, ignores the question of prototypicality and-

- accepts Westinghouse contention that =theirs were valid tests, statistical analysis of the Westinghouse: data, Attachment 4,=shows that the leak flow rate at the 95% confidence limit and 0 voltage'could be.07 gpm. Using the Westinghouse estimate that 680 defective tubes'will remain in service, the total leakage-per-steam generator L(50 gpm) is significantly larger than the 0.16 gpm leak rate estimate tar Westinghouse.-- It-is-not the purpose here to l

-question Westinghouse analysis, but rather to point out that leak rate-l calculations are very' sensitive to model assumptions..

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' Based on the above discussion a core melt frequency of 10-4/RY may be the best that can be estimated.

Plant operators, therefore, must provide assurances that sufficient water to prevent core melt is available to'them to avoid RWST depletion.

Practically speaking, if a supply of water is available for several days there will be sufficient time to define a solution.

Five to eight hours, on the other hand, may not be sufficient.

The first step towards the resolution of this issue is to document the amount of borated water reserves presently available at each plant.

I~t is estimated that one week (NRC+ PLANT) time would be required for this activity for each plant. The corresponding total cost is estimated at $160K ($100K man-yr

80 plants /50 weeks).

Multiplication of the 10-4 /RY by a dose of 2.710+6 ( PWR-4 seq) gives 2.7 10+2 man-rem /ry.

Assuming 30 yrs remaining life and 80 reactors,.we get 50*10+4 man-rem.

CONCLUSIONS:

The present analyses shows that continuous operation with degraded tubes could lead to a core melt due to simultaneous le:kage from many tubes following an.

unisolated steam line break.

The risk for such an event can not reliably be estimated because of lack of data. Although a design basis multiple tube l

rupture could bound the above leakage it is not practical at-this time to request the industry to modify present plan + 6esigns. The available data does-not support NRC position that operation with degraded tube is safe. That

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4 10 position is based on ' leak before break' consideration which is acceptable for normal operation but is not applicable to the SLB accident.

Public safety will be served by requiring plants to have sufficient borated water reserves on hand. The first step towards this solution is to document present water availability at each plant.

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11 REFERENCES 3.

Subconmittee one Material and Metallurgy ACRST-1872, ' November.6,1991.

2.

Nucleonics week, February 13, 1992.

3.

Personal Communication Emmet Murphy (NRR), December 22, 1991.

4.

J. Hopenfeld, DP0 December 23, 1991.

5.

T. Speis to Hopenfeld, memo "Your DP0 Dated December 23 1991',

February 19, 1992.

6.

HRC Integrated Program for Unresolved Safety Issues A-3, A-4, and A-5, Regarding Steam Generator Tube Integrity,2NUREG-0844, Apr.il 1985.

- 7.

Steam Generator Tube-Integrity Program NUREG/CR-2336; also, WCAPil313.

8.

T. P.,Hagee et al, Trojan Steam Generator 1 Tubing Destructive Examination-Interim Report TR-MCC-186.

P 9.

Steam Generator Tube Integrity Safety Analysis Report 1for Tube-Support Plate Intersections,- PGE, December 1991.

10.

Trojan Nuclear Plant Steam Generator Tube Repair Criteria, December 1991 WCAP 13130.

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REFERENCES (Con't.)

11.

R. N. Parking, Sress Corrosion Crack' Coalescence, Corrosion, June 1990, pp. 485.

12.

G. Ishack " Steam Generator Tube Degradation - Is it a Safety Concern:

Nuclear Eng. International, Jan.1992.

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13 ATTACHMENT 2 EVALUATIONOfTHETROJANPLANTANALYSISOFLEAKAGEFOLLOWINGSLB In Reference 9 PGE requested the NRC to review their justification for restart of the Trojan plant.

PGE concluded that the recently discovered through-wall steam generator tube degradations does not involve unreviewed safety questions (USQ) as described in 10 CFR 50.59 (a) (2).

PGE reached these conclusions in reliance, on a Westinghouse study (Referen'ce 10) which examined the consequence of operating steam generators with defective tubes.

Westinghouse concluded that primary to secondary leakage under Steam Line Break (SLB) condition will not constitute a safety problem.

The following analysis shows that primary to secondary leakage during a SLB accident is very sensitive to model assumptions and data source.

Different assumptions and different laboratory data leads to drastically different conclusions than those reached by Westinghouse.

2.

Westinghouse Analysis, (Reference 10).

The Westinghouse work consists essentially of three parts:

1.

Experimental and analytical study of primary to secondary leak rates through cracked tubes under SLB conditions.

2.

Determination of crack growth rates and the resultant increase in tube. degradation for the next operating cycle.

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14 3.

Prediction that'a SLB accident at any time during cycle 14 will not exceed 0.12 gpm.

Implicit in the analysis are the following assumptions:

ASSUMPTIONS:

1.

SLB leak rates with bobbin voltage indications in the 6 to 50 volts range can be extrapolated to SLB leak rates in the O to l'.4 volts range.

Figure 8-3, page 8-13, of Reference 10.

2.

The laboratory leak rates of Figure 8 are prototypic of SLB leak rates.

3.

Successive average voltage changes between three operating cycles determines crack growth rates.

DISCUSSION:

ASSUMPTION 1 The bobbin coil probe voltage depends not only on the total volume of SCC voids in the wall but also on void geometry. Voltage amplitude is not a unique, linear indicator of the propensity of the tube to leak under SLB' loads as assumed by-Westinghouse. Several, partially through the wall IGA craclis, may give e larger voltage signal than a single deep crack yet the' deeper crack, will more likely leak during an SLB accident. The data from Plant A-2

i -

s 15 (page 10-11 of Reference 10) indicates that the depth of penetration and not the voltage amplitude determines whether degraded tubes will or will not leak.

Tube R4-C73 with, a through the wall crack and a 2.8 voit indication leaked under SLB conditions while another tube with a 2.31 indication but shallower cracks did not leak. (Differences in above voltage; are not considered significant w'ithin the 40% NDE uncertainty). An indirect indication that crack geometry, not merely crack void affects leakage is provided by the dependence of the burst pressure (Reference 7).on both the depth and the

~

length of the crack.

PGE (Reference 9) stated that the Bobbin probe cannot reliably characterize the depth of penetration of micro flaws. Statistical analysis of laboratory data (Attachment 4 shows that even if the coil probe voltage is zero a finite leakage is possible.

In conclusion, leakage under SLB loads from samples with large voltage signals, is not sufficient to show that through the wall, undetected cracks with indication < 2 volts will not leak.

ASSUMPTION 2 The main variables affecting stress corrosion cracking, SCC, are material condition, temperature, exposure time, environment chemistry and local stresses. While the first three parameters can be simulated in a laboratory it is not practical to simulate water chemistry and local stresses as claimed by Westinghouse.

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l 16 Experimental data, Reference 11, demonstrate that under cyclic loadings, even small f1 actuation in the applied stress rapidly increases the coalesscene of cracks. The Westinghouse specimens were not exposed to the same type of stresses that exist in an operating unit. The interface between the subcooled and saturated region and the foam region are sources for thermal stress fluctuations.

Support plates and regions near the tubesheet are subject to local stress fluctuations due to flow induced vibration. The recent SGTR accident at Mihma, is an example (where anti vibration bars which were not

~

installed in'the proper locations) of wear tube movement at a support plate.

Chemistry excursions from condenser leakr, and primary coolant leaks which occur in operating steam generators are other examples of parameters which cannot be properly duplicated in a laboratory.

Besides the lack of proper environmental simulation, the leakage tests were terminated by Westinghouse after 30 minutes.

Considerable industrial experi-ence indicate that high velocity two phase flow can cause material erosion by droplet impingement.

The time scale for a jet, emerging into an empty space, from a leaking tube to penetrate a 0.040 inch wall of an adjacent tube is on the order of several hours.

(Attachment 3) s The RELAp calculations (Attachment I) indicate that SLB accident with primary to secondary leakage will not be terminated within half an hour as assumed by Westinghouse especially if the leak rate is large.

If the accident proceed for several hours, leakage may be increased due to tube to tube damage propagation.

a 17 ASSUMPTION 3 As already mentioned above the Eddy Current probo can not characterize the depth of micro cracks in the tube wall. Yet it is the growth and coalescence of these micro cracks that will result in leakage. The average growth rate as obtained by Westinghouse may be related in some manner to these micro cracks but without knowing what that relation is the average growth rate of 45% is simply a guess.

By a way of co,mment, even if growth rate data was available, the use of average value for leak before break predictions is highly questionable. The' proper procedure would be: first establish that cracks grow in a random manner, second obtain a distribution function to the largest crack in a randomly selected tube samples, fourth, use this distribution to predict the time for the largest crack to penetrate the wall, The Westinghouse assumption that cracks grow at constant rates is in disagree-ment with the conclusion reached at a recent AEA conference (Reference 12).

The. applicable conclusions from that conference were that: (1) Inconel 600 is not a stable alloy when used in a steam generator tube material.

(2) crack growth rate in tube roll transitions is not constant and is dependent on water chemistry and the particular point in time of the units' cycle.

In conclusion, leakage will be determined by the fastest growing crack in a given tube sample and not by the average crack growth rate as assumed by Westinghouse,

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IB DISCUSSION Westinghouse efforts in support of PGE request for approval' of Trojan restart is focused on showing that primary to secondary leakage following SLB will be within allowable limits.

The accomplishment of this task require data which is not available.

Of course the ability to predict leakage per crack is only part of the problem, the'oiber part is the predictions of how many of the c' racked tubes will leak when subject to SLB loads. Westinghouse calculates a total of 680 defects which could be left in service.

If we accept this number and apply the CE leak data (Ref. 7, pg. 27) to a 0.5 inch long crack, we obtain a total leakage under SLB loads of 6800 gpm instead the.12 gpm per steam generator calculated by Westinghouse.

It is obvious that the cor. sequence of SLB is very sensitive to the leakage data employed, r

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March 11. 1992 Dr. Joram Hopenfeld U.S. Nuclear Regulatory Comission Washington, DC 20555 DPAFT RESULTS OF STUJi GENEPATOR TUBE RUPTURES CONCURRENT WITH STUJi LINE BREAK OUTSIDE CONTAINMENT CALCULATIONS - LW 05 92

Dear Dr. Hopenfe1d:

The attached report prepared by C. Heath sumarizes the results of the calculations performed as you requested to detemine the expected behavior of a Westinghouse RESAR III plant after a steam line break concurrent with a steam generator tube rupture.

The calculations perfomed led to prediction of refueling water storage tank desletion (RWST) bes ruptured.in a period of three to e and a half hours depending on tie number of tu It should be emphasized that the time to exhaust tt.e RWST could vary substantially due to operator action, thus the predicted times are not absolute and are useful as scoping calculations only.

Please note an NPA mask was developed as part of the analysis should you desire to see the results displayed on the DEC 5000. Also 1 have included, as a second attachment, a copy of the critical flow equations we discussed.

if you have any additional questions or comments please call me at 492-3688 or Chris Heath at 492 3691.

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Leonard L. Ward

(/INEL Program Manager for NRR Projects l

Enclosures:

As Stated cc:

P. Norian G. Berna (EG&GIdaho,Inc.)

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o STEAM GENERATOR TUBE RUPTURE CONCURRENT WITH STEAM LINE BREAK OUTSIDE CONTA Prepared by:

C. Heath I!GED M llGU At the request of Dr. Joram Hopenfeld of the USNRC, Office of Research, scoping calculations were performed for a double ended rupture of a main steam line, outside of the containment, concurrent with multiple failures of steam generator The f ailed steam generator tube break areas evaluated in this study tubes.

included sizes equivalent to 1, 2.5, and 5 double ended guillotine ruptures. A RLSAR 111 Nuclear Steam Supply System model was used for the evaluation.

The results of these calculations show that without operator U.tervention, a steam line break, outside of the containment, concurrent with the double ended rupture of a single steam generator tube in the failed generator results in The double-depletion of the refueling water storage tank (RWST) in 8.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

ended rupture of five steam generator tube results in exhaustion of the RWST inventory in about three hours. With operator action to throttle Emergency Core Cooling Systen (ECCS) injection flow, exhaustion of the RWST with five failed tubes is delayed to 7.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . While operator actions can significantly delay exhaustion of the RWST, timely accident management strategies such as those to replenish the RWST with borated water would be needed to prevent the accident from progressing to a core melt. Because the sec.ondary pressure of the failed steam generator decreases to near atmospheric conditions due to the large ' steam line rupture, operator actions to reduce reactor coolant system (RCS) pressure to a value below that of the failed steam generator secondary (to teminate the RCS break flow) may not be timely enough to prevent exhaustion of the RWST.

The results of the scoping calculations are discussed below.

EllWilLQU The The SCDAp5/RELAP5/M003 code, version 7(o), was used in the calculations.

calculations were performed on a DEC 5000 computer for a four loop RESAR III PWR at a themal power of 3400 MW,.

The RELAP5/H003 nodalization diagram is r, resented in Figure 1.

The model consists of two separate loops.

The single ioop contains the f ailed steam generator with the broken steam line and failed steam generator tubes while the other loo) combines the three remaining loops.

The calculations were carried out to one leur into the event at which time the primary and secondary pressure responses achieved a near quasi-steady state condition.

Three steam generator tube failure cases were evaluated consisting of break areas equivalent of 1, 2.5, and 5 double erded guillotine ruptures. The main steam line break size included a double ended guillotine failure, outside of the containment, with an area of 4.9 ft'.

With a steam line break outside of the

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containment concurrent with a multiple f ailure of the steam generator tubes, exhaustion of the RWST inventory can potentially occur which could lead to a possible core melt.

With the break located outside of the containment, exhaustion of the RWS1 cannot be followed by a switch in ECC alignment to the recirculation mode of cooling. From an accident management perspective, the time to exhaust the RWST inventory is therefore of particular interest since in the event of ro additional actions, core uncovery and melt could occur.

Table 1 presents a sumary of the results of the scoping calculations. The time to exhaust the RWST inventory for the three steam generator tube rupture sizes varies from 8.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for one failed tube to 3.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for five failed tubes. For illustrative purposes, no operator actions were assumed for these first three Cases.

In estimating the time to exhaust the RWST, the capacity of the tank was assumed to be 350,000 gallons, which is approximately the minimum allowable technical specification value. Clearly, any additional borated water would lengthen the amount of time to drain the RWST. Also, the time to exhaust the RWST is based on the injection flow at one hour into the event, which consisted of high pressure safety injection and charging flow. Low pressure safety injection was never initiated in our calculations and the safety injection tank (SIT) contributions were insignificant by this time for all cases.

While RCS and secondary pressure has stabilized at this time, use of the injection or break flow at one hour results in minimizing the drain time for the RWST since break flow is expected to decrease during the latter portion of the events.

Since decay heat generation decreases with time, the operator could continue to throttle ECC flow to minimize RCS pressure and the resulting break flow, while maintaining a minimum of subcooling.

The last case presented in Table 1 shows the effect of the operator actions to delay drainage of the RWS1. These actions included throttling the ECC flow to maintain a minimum of subcooling in the RCS, while cooldown of the RCS by opening the atmospheric dump values (ADVs) in the intact steam generators was also initiated.

As mentioned earlier, with the double ended steam line break, cooldown of the RCS with the objective of reducing RCS pressure below that of the broken steam generator requires many hours since the failed steam generator depressurizes to very low values early in the event.

Table 1 shows that throttling ECC flow to maintain a minimum of subcooling results in delaying exhaustion of the RWST from 3.1 to approximately 7.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months after initiation of the event.

All the cases were run for six seconds at full power to reach equilibrium thre99 out the system and then the breaks were opened and the reactor was h

scramed.

The results from the final case of Table I which included operator action are discussed i" the following paragraphs in detail. The results for the cases involving the rupture of 1,2.5, and 5 tubes are phenomenologically similar and are included in Appendices A, B, and C to this report.

A summary of the assumptions and initial conditions for these scoping calculations are provided ir. Table 2.

Figures 2 through 6 present the calculation results of the main steam line rupture concurrent with five failed steam generator tubes for the operator action Figure 2 presents the RCS and failed secondary steam generator pressure case.

responses.

Becau'se of the large steam line break size, the failed steam l

i generator depre:surires rapidly to near atmospheric conditions. As a consequence of the rapid cooldown of the failed steam generator, the RCS also experiences an initial rapid cooldown, which stabilizes due to the activation of the ECCS early in the event. The sudden decrease in RCS pressure at about 750 seconds in figure 2 is due to the Sli discharge which condensed the steam and collapsed the voids which developed during the initial portion of the transient.- The condensation caused the P.CS to depressurize, increasing the SIT flow and further reducing the saturation temperature and hence RCS pressure. Continued ECC flow then pressurized the RCS to the condition where break flow equaled the ECC injection flow which occurred at about 750 seconds. At about 1000 seconds, operator action was initiated to throttle the ECCS, reducing RCS pressure during the latter portion of the event as shown in Figure 2. Note that without operator action to throttle ECC flow, the RCS pressure will remain at significantly higher pressures as shown in figure C1 of Appendix C.

The ECC injection and rupture steam generator tube mass flow rates are given in Figure 3.

The tnass flow rate through the failed steam 1'ne is given in Figure 4.

Using the. ruptured tube break flow rate of about K i lb/s from Figure 3 at 3600 seconds, the RWST is estimated to drain in about 7.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . The ECC flow, shown in figure 3, temporarily decreased at about 3300 seconds into the transient as a result of the emptying of the SITS. Although the ECC pumped injection finw is lower than the break flow at the end of the transient shown in Figure 3, pumped ECC flow would be increased at this time to maintain RCS-subcooling and an RCS pressure of approximately 160 psia.

Figure 5 presents the primary and intact secondary temperature responses and shows that RCS temperatura has stabilized after one hour into the event.

The failed steam generator temperature transient is given in Figure 6.

It is important to note that there is a flow restrictor in each steam generator at the entrance of the steam line which is designed for a 2.75 psi pressure drop at a flow of 1051 lb/s. This restrictor had little or no impact on limiting the break flow through the broken steam line for the conditions calculated, it should be recognized that other strategies or actions may be successful in further delaying exhaustion of the RWST or terminating the break flow through the failed steam generator tubes.

It should also be emphasized that break flow and hence ECC flow can vary significantly depending on the operator throttling actions to achieve the degree of desired subcooling. As a consequence, the time to exhaust the RWST can also vary significantly.

The significance of the calculations should not emphasize the exact times for exhausting the RWST, but that operator actions can extend the time to drain the RWST. Other strategies that may be considered could include:

1.

Opening the PORVs early in the event to establish sufficient inventory in the sump to initiate ECC recirculation.

2.

Activate Residual Heat Removal and attempt to establish mid-loop operation to terminate the loss of RCS liquid through the break in the steam generator abes.

3.

Replenish the RWST inventory with borated water at a rate greater 4

that the ECC injection rate.

CO'gil!S.LM A double ended steam line break outside of the containment concurrent with five f ailed steam generator tubes results in exhausting the RWST in about three hours without operator action.

With operator action to throttle ECC flow, the exhaustion of the RWST is delayed until about eight hours after opening of the break. Because the break is located outside the containment, the eventual loss of the RWST inventory will lead to a core melt since there will be no coolant in the containment sump to initiate the ECC recirculation mode of cooling.

The importance of these results are that operator actions can successfully delay exhaustion of the RWST.

However, to prevent a core melt additional accident management actions during the long tertn would be needed to terminate the break flow or identify alternate sources of ECC injection water.

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TUBE BREAK M.$.L. BREAK OPERATOR ACTIONS TUBE BREAK ~

HOURS TO FIGURES l

f.. D.E.G FLOW'.(1b/s)

EMPTY RWST 2

S.G. :-TUBES -

AREA (ft')

AREA (f t )

I-0.004 4.9 None 83 8.5 Appendix A

' 2.5 0.010 4.9 None 155 4.1 Appendix B 5

0.020 4.9 None 200 3.1 Appendix C 5

0.020 4.9 Opened intact steas 105 7.7 2-6 generator ADVs, throttled charging pueps, and terminated HPSI and LPSI after 18 minutes.

1 The steam generator tube break flew rate is based on the value at 3600 seconds.

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TABLE 2 CALCULATION INITIAL CONDITIONS AND ASSUMPTIONS 1

Simultaneous break in main steam line aied rupture in steam generator tubes.

2 Instantaneous scram of reactor coincident with break initiation.

3 Intact steam generators isolated.

4 All ECCS consisting of HPSI, LPSI, and SIT, as well as charging pumps actuated.

5 Tjm,e to exhaust RWST based on break flows one hour after break.

6 No operator action (except for last case) e

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7 APPENDIX C 5 Double Ended Guillotine Tube Breaks Calculation,Results e

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[deg F]

i Primary Tilot l

6MM Secondary 640.00 620.00 600.00

580.00 560.00 540.00

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Figure C5. Broken Steam Generator Temperature,5 DEG Tube Breaks

[deg F]

660.00 640.00 I

620.00 600.00

', 580.00 560.00 540.00 520.00 500.00 480.00 460.00 k

440.00

\\

420.00 400.00 380.00 360.00 340.00 320.00 300.00-w

\\

p 280.00 260.00 3

240.M

[s] x 10 0.00 0.50 1.00 1.50 2.00 2.50 3.00 3.50

^

_4

-Notes on critical Flow Rate-Estimation

- The Darcy equation -is applicable -- to incompres_sible. steady-state flow through a constant diameter--straight pipe where the pressure difference is given by:

A P=fE E V (1)' -

D 29, Rearranging a'd expressing the flow as a a mass. flux.

3 G=

- (2).

=

fL v

- p If the flow through is desired through a length-of pipe with a _ flow.

loss coefficient K, the'above equation-becomes:-

G= 2

[P -cP,(T,)

(3)'

e The Zaloudek correlation for subcooled critical flow is-given by

' (P -P e) l( 4 l' G=c u

u t

for the range 400< P,.< 1800 psia and vnere c

= discharge' coefficient P,

= upstream pressure-P, = saturation pressure o

v,

= specific volume of saturated liquid-W e

5 4

...,.ecr

. ~.,

1

--w.

If the Zaloudek correlation is modified to predict a frictionless -

Moody flow at saturation, thus:

[.* [ P,- c P,,e ( T,) )

(5)

G=

where c is a coef ficient that matches the Zaloudek flow rate with frictionless Hoody critical flow at saturation.

If the upstream pressure of Eq. (3) equals the downstream pressure of Eq. (2), and the mass velocities are the equal, Eqs. {2) and (3) can be combined to yield:

2 9,[P.-0. 8 5 P,,e ( T,) ] 14 4 ' i (6) v,(K+1) where the pressure is in psia and c = 0.85.

P,

= system pressure, psia P,,, = saturation pressure of subcooled liquid at temp. T,

'F, pria 3

= specific volume, f t /lb v

g,

= gravitational constant, ft/sec/sec K

= flow loss coefficient, dimensionaless L

= flow length, ft D

= hydraulic diancter, ft f

= friction factor 2

G

= mass flux, lbs/sec-ft AP

= pressure difference Pressure losses due to area changes and bends etc. can be accounted for through changes to K, the flow loss coefficient. The upstream pressure of the Zaloudek correlation (P,) equals the downstream pressure of ths Darcy equation (P,)

so that the flow rates predicted by each formulation are equal.

The Zaloudek estimated 1

- 4j.i

-: t '

flow equals the:frictionaless Moody flow at-saturated conditions.'

Eq.-(6)'is app 1'icable to--subcooled and saturated-fluid discharge.

For:the critical flow of superheated steam, Murdock and-Bowman is--

used where:

G=44.52 (7)

.v.

0 t

9 t

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-EADY... 2250 580

'Y k

LIQUID STAT,L-PROPER 580.0000

-F 225 PSIA T

%.0.000

=

P-

--\\

585.7833 BTU /LBM

<e113 LBM/FT3

'H

'=

RHO ~

=

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=

S

=

0.0000000E-01 0.0000000E-01 ALPHA

=

X --

=

HEAT CAPACITY 0.7248301 BTU /LBM-R 1.335268 BTU /LBM-R CV

=

CP

=

1,842232 CP/CV

=

VISCOSITY 1.812E-06 LBF-SEC/FT2 MU

=

THERMAL CONDUCTIVITY BTU /HR-FT-F 0.320

'K

=

20979.43 LBM/S-FT2 GCRIT

=

tAoY...

lekI

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~

u 6

e 9

uy C 4?

0.3

/

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RT 3

6 BM

-TT2-i

" EADY... : 5 0 0. 3 00 pf isre"-

LIQUID STATE QDES/

300.0000 F-50 PSIA T

=

D.>.0.000 0 LBM/FT3 H

P

=

270.4979 BTU /LBM

=

.40928 RHO-

=

268.8862 BTU /LBM 0.4363126 BTU /LBM-R U

=

-S-

=

0.0000000E 0.0000000E 01 ALPHA

=

X

=

HEAT CAPACITY 0.8490872 BTU /LBM-R 1.026471

-BTU /LBM-R CV

=

CP

=

1.208911 CP/CV

=

TIISCOSITY 3.847E-06 LBF-SEC/TT2 MU

=

THERMAL CONDUCTIVITY 0.396 BTU /HR-FT-F K

=

15072.92 LBM/S-FT2 GCRZT

=

EADY...

}

3 7

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i

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January 29, 1992 TO:

Warren Minners

-THRU:

Jim Glynn FROM:

Les Lancaster j

SUBJECT:

Confidence Lines For the Bobbin /LeakRate data I ran a simple. regression taken from a R

computer package called STATGRAPHICS.

I borrowed the package from-Dick Robinson and quickly learned how to use it and quickly ran the-regression on the-data.

On presenting the results to you a question emerged on the resulting confidence bounds which I shall attempt to answer in this' note.

R STATGRAPH2CS gives two_ limits-which they call confidence limits and

. prediction limits.

It turns out that their ' confidence limits' is the confidence limits on the predicted mean and their ' prediction limits' is the confidence limits on the prediction of a single observation.

The bounds closest to the fitted line is their ' confidence limits'.-

'See attached three pages taken from HUREG/CR-4604.

Using this information,I can answer your original-question, which prompted this exercise, with the following table-(Remember. your original question was:

At_a specified confidence. how big can the. Bobbin be to_ expect a.zero

~

LeakRate?):

Using-Using Confidence

-Prediction Limits -

Limits 50% Level-6.5 11.7 95% Level-9.1' 27,2-R-

From the attached plots, printed from the STATGRAPHICS run, note that your-

' commented observation or question on the number'of points lying outside of-

=the bounds would hold.for the ' prediction limits' if the-fit had'been better.

4 to D l5 nQ h]e s hu Ls us<

y ay e y

)2ro p < /rwy L t j.' v - ~ C L i

.9 lw As sb lc. ib.1 l-r m h

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1

. - + -

9 d

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s us<

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[

.S NUCLEAR REGU TO Y OMMISSION wAswiNoton o.c. anus

%....*/

SEP 2 81992 MEMORANDUM FOR:

Frank P. Gillespie, Director Program Management, Policy Development, and Analysis Branch Office of Nuclear Reactor Regulation FROM:

C. J. Heltemes, Jr., Deputy Director Office of Nuclear. Regulatory. Research

SUBJECT:

GI-163 MULTIPLE STEAM GENERATOR TUBE LEAKAGE

Reference:

NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity, Final Report. NUREG-0844, September 1988 License amendments have recently been granted to Westinghouse plant owners for operation during one refueling cycle with modified technical ssocifications for steam generator tube leakage limits.and steam generator tu>e repair limits.

One such amendment (No. 178,-Facility Operating License No. NPF-1) prompted a DP0 and the opening of the subject new generic issue. The prioritization for this issue is now being peer-reviewed.

The preliminary-priority ranking is HIGH. Although not normally done for_ HIGH priority issues, NRR-is being provided a concurrent copy of this prioritization for review and comment because of recent NRR experience in evaluating licensee submittals with respect to steam generator technical specifications.

'NRR is also invited to assist in defining the scope and priority of issue resolution efforts by identifying the steam generator tube integrity issues which would be most usefui to address on a priority basis to.best-suit the needs_of NRR. The reference contains previous discussion, conclusions, and recommendations with respect to many such issues.

Issues not discussed in the reference might also be important in plant operation with-revised steam generator tube repair and leakage limits (e.g. a PWR ATWS). Severe accident-issues such as steam generator tube heat-up from hot degraded core gases might also need consideration. Finally, a number of. regulations and supporting documents might be impacted and require revision if the technical specification revisions-are to be permanent, rather than limited to the current single-cycle. NRR comments in this regard are also invited.

o i

ENCLOSURE 4-l.-

.. ~.

y

- i SEP !8 890 Please assig'n an NRR contact to work with the task manager in coordinating activities bearing on this issue. The Task Manager for this issue is Dr.' Gary Burdick,- 492-3950.

C.

He eme r., Deputy Director Of e of Nucle r Regulatory Research

Enclosure:

As stated O

b W

O

+

d v

e e

J

e Prioritization Evalu3112D Issue 163: Multiple Steam Generator Tube Leakage O

gg t

4 e

l

V Issue 163: Multible Stram Generator Tube leakaoe DEKRIP_1LQ!i Backoround This issue was opened (Reference 1) in response to a Differing Professional Opinion (DPO) with supplemental information contained in Reference 2.

The NRC is currently considering changes in steam generator technical specifications and monitoring. The changes being considered include utility proposals to permit tubes with up to 100 percent through-wall cracks to remain in service (Ref. 3, Pg. 326).

From the same reference, "The staff has approved higher depth-based limits, up to 64 percent, for specific types of flaws at specific plants.* One utili'y proposal combines eddy current testing augmented by more restrictive leakage limits and inspection programs. The DP0 states that,

'Wnile considerable research will be required to define a new plugging limit and change the SRP, the result will not-increase p. ant safety. The basic problem is with the NDE procedures and their inability to predict tube degradation and leakage." At least one other NRC staff member has also expressed concern with industry proposals for alternate tube plugging criteria allowing operation of steam generators with known through-wall tube cracks (Reference 4). That reference, among other issues, also questions the capability of eddy current testing in steam generator tube flaw detection and sizing.

~

2 License amendments have been_ granted for. plants to operate with eddy current-limits on steam generator tube testing augmented by more-stringent primary to

~

secondary leak criteria. One such plant, the plant specifically mentioned in-the DPO, was selected as the Base Case for this prioritization ~(a four-loopf E _

+

PWR).

The concern stated in the DP0 is "...a Main Steam Line Break outside-containment could trigger multiple steam generator tube failures which would.

- then result in a core melt because of depletion.of coolant inventory-" LThe concern applies to PWR operation with multiple steam generator tube through-wall-cracks or other tube degradations. This statement of concern defines the issue prioritized herein.

Safety Sionificance A PWR main steam line break-(MSLB) concurrent with steam generator tube rupture-(SGTR)- could result in-a containment bypass loss _of _ coolant accident.

If-the SGTR involved enough tubes and if the MSLB was not-isolable, core, damage would ensue upon refueling water storage tank-(RWST) depletion. ~ The ruptured tubesL and open steam line would :then-provide a direct l path.to the,.

atmosphere for fission products from the deteriorating reactor core - For prioritization purposes. the' MSLB-will be assumed to occur in the steam line segment between containment and the first MSIV.

BWRs are not affected by this issue.

N

-e-

_w--

. +.

p

- - -.. -. ~ - -

1.

3 3

Possible Solutions _

One solution is repair of tubes or replacement of.large portions of seriously degraded steam generators.

A second solution is to install a feature to prevent depressurization, given.a.

MSLB; i.e. the steam line between containment and the first main steam isolation valve (MSIV) should be provided with a guard pipe assembly ~to maintain secondary system pressure, given a MS'B.

The MSIVs should-be of L

sufficiently high availability for the case where MSLBs occur downstream of.

the MSIVs.

A third solution is to mitigate the seriousness of loss-of-RCS inventory by providing berated water sources sufficient to maintain core cooling long enough beyond.RWST depletion -to initiate and maintain cooling by the decay heat-removal (DHR) system.

A fourth solution might be :to add MSIVs inside containment.

However, due to the potential _for installation and maintainability. problems for such valves,.

this option was not pursued.

PRIORITY DETERMINATION-Frecuency Estimate Reference 5 estimated the probability of.a MSLB in the segment between.

containment and the MSIV as 1.0E-3 per reactor year (RY) with_an assumed-

- _ 4

.<m.

~.

r er we

6

~

4

- worst-case peak differential of 2600 psid. An unpublished estimate has also been made, using an expert judgement approach, by a team assembled under the

' Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors

program sponsored by the NRC. The team median estimate (to be published) was' I.7E-4 per reactor year per segment, for a l'eak volume i

greater than 50 gallons per minute, which gives 6.BE-4 per reactor year for a four loop plant. Due to lack of data, Reference 6 could only estimate an-upper bound of 7.0E-3 per RY for breaks in secondary system piping greater than three inches in diameter. for this exercise, 6.8E-4 per RY per four loop unit will be assumed as the initiating event frequency.

The Base Case y plant identified 3025 flawed tubes 'of which 2597 were repaired by either plugging or sleeving during 1991. The balance of.428 were approved to be left in service (Ref. 7) under new steam-generator '(SG) tube eddy curren' testing criteria and more stringent primary to secondary leakage limits.

In the following evaluation, the 428 tubes are assumed to have flaw depth greater than 40% through-wall based on evidence from pulled tubes (Ref. 7).

from Reference 8, there are 87, 67,101, and 173 flawed tubes in the Base Case SGs A, B, C, and D respectively.

It is assumed herein that flaw depth:is uniformly distributed through the tube walls. Under.this assumption, there -

are approximately-15,11,17, and 29 flaws in the respe:tive SGs with depths between N and N+10 percent through wall for N=40, 50, --, 90 assuming only one flaw per tube.

[There are probably several flaws per tube, particularly at the support plates.

It should also be recognized that many more flawed-

d o

5 tubes could exist at the Base Case Plant than the 428 in question. These conditions are both due to the inherent weaknesses in eddy current testing.

Reference 4 describes an exercise where several different techniques were used to detect laboratory produced flaws. Detection probability ranged from 0.2 to 0.75 with an average of 0.5.

Flaw sizing ability was poor.)

Figure 1 is a plot of curves illustrating normalized burst pressure vs. f1aw length for a family of flaw penetration curves (cf. Reference 9, Uniform Thinning Equation). Po is the burst pressure of a virgin (unflawed) tube, P is the burst pressure of a flawed tube, t is virgin tube thickness, and a is flaw depth.

Po is approximately 10E+4 psid, thus 0.26 on the P/Po scale corresponds to approximately 2600 psid.

From the figure, tubes with an a/t ratio of 0.75 or more would leak or break at a pressure of 2600 psid or less depending upon flaw length.

Again assuming uniform distribution of flaw depth, there are about 36, 28, 42, and 72 flaws, in the respective SGs with depths 0.75 or more. No attempt will be made to estimate flaw length.

It is, however, noted from the figure that flaw length can be quite small for flaws greater than 0.9 through-wall to exhibit leakage at pressure much less than 2600 psid. Actual rupture of ten or more tubes at 2600 psid appears to be a reasonabic assumption. A leak-before-break (LBB) concept is inherent in the Safety Evaluation (Ref. 7) for the Base Case plant licensing amendment which is the subject of the DPO. Reference 11 cautions that, 'It is not recommended that the LBB concept be applied to piping systems that are susceptible to IGSCC [intergranular stress corrosion cracking) or to water hammer and to piping subject to erosion, such as portions of the steam extraction systems."

O UNIFORM THINNING EQUATION 1-i O.9 - g--

(

)r.

08 -

i

_o Q$

0.7 -

a gj

0. 6 -

e o_

t;;

0.5 -

E u

0.4 -

8 e

=j 0.3 -

- ~

g = o.73

0. 2 -

all = 0.9

^

M = 0.95 0

M = 0.99.

0 0.4 0.8 1.2 '

1.6 2

2.4

.2.8 F1aw Length (inches)

Figure 1.

8' W

e e

4 7

For a large main steam line break with more than 10 steam generator tube ruptures, Sequence Bc of Reference 5 has a.n assigned probability of 0.5 for failure to depressurize the RCS to atmospheric pressure before RWST depletion.

The estimated core damage frequency for the case under consideration is then F(MSLB) X P(more than'10 tube ruptures) X P(failure to depressurize) or (6.BE-

4) X (1.0) X,(0.5) 3.4E-4 events per reactor year. The 0.5 for failure-to depressurize is not elaborated upon (other than being time-based) in Reference 5 and could be conservative.

Reference 12 states that, " Initiatives that are directed to prevention of core damage should also assess the potential for early failure or bypass of the containment in order not to exceed the large release frequency of 1.0E-6/ reactor year."

Although Reference 12 is not yet implemented, with a frequency goal of 1,0E-6 per reactor year for this containment-bypass sequence, virtually all of the 3.4E-4 events per reactor year constitutes the frequency reduction potential ((3.4E-4) - (1.0E-6)

-3.39E-4).

Lo_01ecuence Estimate Reference 13 (cf. Table-7, Page 3-24) assigns a PWR-1 release category to an HSLB with multiple (>10) SGTRs. The whole body dose per PWR-1 event is 5.4E+6 person-rem.

The offsite public dose consequence is thus (3.4E-4) x-(5.4E+6)

-1.8E+3 person-rem per reactor year. Although the license amendment granted--

for the Base Case unit was for cycle 14, it is assumed that extensions will be granted through the remaining four years of unit operation (not yet approved by the NRC) currently desired by the owners. With an assumed 4 years of

_ -. =. --

6 2

f 8

remaining unit-operation, the total risk reduction per four loop unit _is 7.2E+3' person-rem, with 5.4E+3 and 3.6E+3 person-rem for three and two loop units respectively. Currently, two four loop units and one three loop unit have been granted license amendments allowing-operation with flawed SG tubes.

The total risk reduction potential from these plants is 2.0E+4 person-rem, assum_ing 4 years remaining unit life for each.

Assuming all currently operating PWRs will be in similar circumstcnces-to_the case considered herein, and using number of loops per unit information from Reference 14, the total risk reduction available is: (31) (7.2E+3) + (14)

(5.4E+3) + - (30) (3.6E+3) -4.1E+5 person-rem.

Cost Estimate Industry Cost:

From information in Reference 15, replacing the steam generator would cost 3150M with other estimates exceeding 5200M.

Plugging and sleeving the 428-flawed tubes at the same ratios experienced at the Base Case' plant up to

-Refueling cycle 14 (71% plugging, 29% sleeving) would cost (Ref.15) about.

52M. This action would probably cause the plant power rating to be lowered L

over the remaining plant-life. Assuming ~a-10% reduction in power over four l:

years.with a 0.65 industry average plant availability. and approximate per' day Base Case plant replacement power cost of $774K (Information from Ref.16)'the-net cost, including plugging and sleeving, is about $76M. Thus, neither of-these options appear to be economically attractive.

(

s a

9 Another possible solution, involves installing main steam line guard pipes.

This appears to be economically more attractive due to short main steam line segments between the containment wall and the first MSIV.

This is also a preventive action. Which would meet the Reference 12 C0F goal _ for a containment bypass sequence since the calculation of f(MSLB) x P(>10 tube ruptures) x P (failure to depressurize) yields (6.8E-4)' x (1.0) x (0.5) -

2.3E-7 < l.0E-6 events per reactor year.

It was assumed that four MSIVs would require replacement in addition to guard pipe installation because of new large flanges required on each MSIV.

Cost estimates were taken from Tables 4.1 and 4.2, Abstract 2.1.9, Reference 16, after noting from Reference 17 that guard pipes need not be designed for greater pressure and temperatur6 than the enclosed pipe.

Allowing room for the enclosed pipe and insulation, 36 inch diameter carbon steel pipe was selected. The nearest satisfactory pipe thickness tabulated was 2 inch. A 4% inflation factor of 1.22 was used for the years 1988 - 1992 inclusive.

o Carbon Steel Pipe (36 in diam., 2 inch thick)

Niamber of Linear feet:

44 Unit Installation Labor:

316 MH/LF (man-hours / linear feet)

Labor Rate X Overhead Factor:-

$24.90/MH X 1.59 -$39.60/MH Factory Cost:

$2914/LF Site Materials:

3818/LF Carbon Steel Pipe Cost:

1.22 X 44 X [(316 X $39.60) 4 $818 + $2914) -$872,064

6 10 o

Carbon Steel Stop Valves (30 inch)

Number of Valves:

4 Unit Installation Labor 333MH/ Valve Labor Rate X Overhead Factor

$29.90/MH X 1,59 =$39.60/MH Factory Cost:

$34897/ Valve

Carbon Steel Stop Valves Cost

1.22X4[(333X$39.60)+$34897)=$234,649 Total Guard Pipe Installation Costs =$1,106,713 per four loop plant.

ML Call Development of a solution is estimated to rost three and one half. person-years of contractor labor at a cost of $125K/porson-year to investigate solutions to this issue and related problems [cf. Other Considerations). NRC Project Manager cost would be about $41K or 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months (Ref.16). Assuming the results indicate a rulemaking activity for operation with degreded SG tubes, an additional 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months of staff time would be required. Tech Spec, changes, supporting Reg. guic'es, and other regulatory costs for the 75 reactors involved total $1,6M (also Ref.16). Total NRC cost is estimated to be $2.1H.

hitLCall Assuming the current population of 75 PWRs for eventual operation similar to the Base Case unit with modified SG tube plugging requirements and l

1

'] y '-

4 0

11 primary / secondary leakage criteria, total industry and NRC costs are about

$65M for the total number of steam lines involved. That figure is obtained by scaling the $1.llM per plant appropriately for 2 and 3-loop plants, totalling over the numbers of plants, and adding the NRC costs.

Ytlye/Inoact Asstmeni Based on an estinated total indus*ry risk reduction of 4.1E+5 person-rem and total cost of $65M for the solution considered, the value/ impact score is: S .4.1E+5 person. rem 365M = 6.3E43 person. rem /5H Q1htr Considerations The PWR issue considered herein was a core damage scenario initiated by a main steam line break (MSLB) in the pipe segment between containment and the MSIV.- Loss of pressure in the secondary side resulted in a primary to secondary side ~ pressure differential sufficient to rupture flawed steam generator tubes. It must be recognized that a MSLB is not the only initiating event for a high primary / secondary pressure differential. Stuck open relief valve, break in the main feed 'line, or MSLB downstream of a stuck open MSIY should also be considered as initiators of similar scenarios. Other issues may be affected-by multiple steam generator tube ruptures, such as Pressurized Thermal Shock. From figure 1 3 note that, depending upon flaw length and depth, such SGTR initiators can occur at pressures less than or equal to PWR operating- =

12 pressures. It should also be noted that if operating years for the Base Case were more than the tour assumed, e.g. 24 years (to achieve a 40 year plant life), the individual plant and total plant risks would be correspondingly larger. Severe accident considerations should also be made during resolution-activities. A PWR ATWS under degraded SG tube conditions, for example, would likely result in rupture of significant numbers of SG tubes with possible MSLB due to overpressure of the secondary system. In other severe accident scenarios, hot gases from a damaged core could heat-degraded tubes, including tubes with through-wall cracks, causing tubes to slump, resulting in open pathways, to the secondary system and environment, for radionuclides. Consideration should also be given to any new permanent leak rate criteria vis-a-vis existing regulations such as 10 CFR 50.34a (ALARA) and Part 50, Appendix 1, as well as GDCs 14, 15, 30, 31, and 32. {0NCLU510N Based on the potential public risk reduction associated with-this issue and additional potential concerns under Other Considerations, this issue is assigned a HIGH priority ranking. )

13 Beferences 1. E. S. Beckjord memorandum to Thomas E. Hurley, 'A New Generic Issue: Multiple Steam Generator Tube Leakage,' June 16, 1992. 2. J. Hopenfeld memorandum t. E. Beckjord,,'A new Generic Issue: Multiple Steam Generator leakage,' March 27, 1992. 3. W. 1. Russell and J. T. Wiggins, ' Staff Initiatives in Steam Generator Integrity,' presented at the 1992 NRC Regulatory Information Conference, Washington, D.C., July 22, 1992. 4. Joseph Huscara memorandum to Charles Z. Serpan, " Steam generator Tube Inspection, Integrity and Plugging,' March 16, 1992. 5. NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity, Final Report, NUREG-0844, September 1988. 6. ' Reactor Safety Study, an Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants,' WASH-1400/NUREG-75/014, October.1975. 7. Letter from Lawrence E. Kokajko, NRC to James E. Cross, Portland General Electric Company, February 5. 1992. Docket No. 50-344. r C =

e 6 l 3-1 14 f 8. J. E. Cross, PGE, letter to U.S. Nuclear Regulatory Commission, January 16, 1992, Docket 50-344._ 9. Alzheimer, J. M. et al., *$ team Generator Tube Integrity Program - I Phase ! Report,' NUREG/CR-0718, Pacific Northwest Laboratory, Richland, Washington, 1979. 10. E. C. Rodabough, ' Comments on Leak-Before-Break Concept for Nuclear Power Plant Piping System,' NUREG/CR-4305, Oak Ridge National Laboratory, Oak Ridge, TN, August 1985. 11. Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, NUREG-106), Vol. 5, P-17, April 1985. i 12. SECY-91-270, ' Interim Guidance on Staff Implementation of the-Commission's Safety Goal Policy,' August 27, 1991, 13. - W. B. Andrews, et al., ' Guidelines for Nuclear Power Plant Safety issue Prioritization - Information Development," NUREG/CR-2800, Pacific Northwest Laboratory, Richland, WA, February 1983. 14.- Information _ Digest, NUREG-1350, ').S. Nuclear Regulatory Commission,: March 1992. 1 + 'T w s -er t wn n+ vr,'m-r-v r.4-w.s v g w -rx< ~ -+-e,-emw w w s--n

w w s--

-,----e- --,--r~-w- +-=ew,: m,sm -,e m s a---o--s m e4<mr g sw --

15 15. Wayne L. Miller and Lloyd C. Brown, ' Generic Cost Analysis for steam Generator Repairs and Replact>ents,' EGG-PE-6670, EG1G Idaho, Inc., Idaho Falls, ID, August 1984. 1 16. E. Claiborne, et al., " Generic Cost Estimates,' NOREG/CR-4627, Science I and Engineering Associates, Inc., Albuquerque, NM, February 1989. 17. Section 3.6.2, ' Determination of Rupture Locations and Dynamic Effects Associated with Postulated Rupture of Piping. " Standard Review Plan, NUREG-0800, USNRC, Rev. 1, July 1981. L c v

ljs esog%, UNITED STATES e's e ! (f i NUCLEAR REGULATORY COMMISSION (., ' N [pf WASHINoToN o.C. 20EM v s,, ....+ November 24, 1992 MEMORANDUM FOR: Clemens J. Heltemes, Jr., Deputy Director for Generic Issues and Rulemaking Office of Nuclear Regulatory Research FROM: Frank P. Gillespie, Director Program Management, Policy Development and Analysis Branch Office of Nuclear Reactor Regulation

SUBJECT:

GENERIC ISSUE 163, 'HOLTIPLE STEAM GENERATOR TUBE LEAKAGE" This. memorandum is in response to your memorandum of September 28, 1992 requesting comments on the prioritization of Generic Issut 163, "Hultiple Steam Generator Tube Leakage". Our review of the prioritization analysis has identified several aspects of the analysis that are incorrect and do not accurately reflect operating experience or current licensing positions. These aspects of the analysis were discussed with members of your staff in a meeting on November 13, 1992 and are summarized in the attachment. We recommend that your staff review and update their prioritization analysis considering the comments we have provided. NRR staff is available to provide input for this activity as required. We appreciate the opportunity to comment on the prioritization analysis and look' forward to continued cooperation. o Frank P. Gill.s ie, ector Program Management, Policy Development and Analysis Branch Office of Nuclear Reactor Regulation

Enclosure:

3 As Stated cc: W.T. Russell W. Minners J. H. Sniezek ENCLOSURE 5

e l ENCLOSURE COMMENTS ON PRIORITIZATION ANALYSIS Of GENERIC ISSUE 163, "HULTIPLE STEAM GENERATOR TUBE LEAKAGE" Generic issue 163 was opened in response to a Differing Professional Opinion (DPO) filed in response to NRR's approval, in January 1992, of an interim alternate plugging limit for Trojan. The interim plugging limit is applicable to flaws which can be characterized as dominantly axially oriented, Outer diameter stress corrosion cracks (005CC) at tubc locations within the thickness of the tube support plates. The interim plugging limit permits tubes with low voltage indications (<1 volt) to remain in service subject to certain restrictions, implementation of this interim limit at Trojan in January 1992 permitted 428 tubes with indications to be left in service. As r.sted in the staff's SER for Trojan, many of the indications left in service may reasonably be expected to grow entirely (100%)luded on the basis of-through-wall during th next eperating cycle. However, the NRR staff conc extensive analysis and testing that the proposed interim limit would provide adequate assurance of steam generator tube structural integrity and that leakage would be within allowable regulatory limits. The generic issue prioritization analysis concludes that "... actual rupture of ten or more tubes at 2600 psid (for steamline break) appears to be a reasonable assumption..." [for Trojan due to implementation of the interim plugging limit). Tnis conclusion follows directly from the assumption in the prioritization analysis that the flaws at Trojan can be idealized as uniform thinning flaws. (With this assumption, the tube burst strength will degrade to less than 2600 psid before the flaw becomes 100% through-wall.) The generic issue proposes a number of potential solutions including-plugging or sleeving tubes with the subject low voltage indications or, alternatively, various actions to mitigate a multiple tube rupture event. We believe the assumption of uniform thinning to be totally inconsistent with the actual degradation mechanism at Trojan. The assumption ignores the considerable body of evidence described in the staff's SER that the degradation mechanism at Trojan is dominantly axially oriented 0050C with minor general intergranular attack (IGA) involvement. In contrast to uniform thinning, it was shown in the Trojan submittal that _ axial cracks may penetrate up to 100% through-wall over lengths ranging to a maximum of 0.84 inch without degrading tube burst strength to less than 2600 psid. There is no I

j 2 evidence of any uniform thinning at Trojan, either from the pulled tube data or from eddy current rotating pancake coil data. IGA involvement observed on the pulled tube specimens was not sufficient to affect the burst strengths of the tubing; the observed burst strengths for the pulled tube specimens were found to be consistent with expected burst strengths for the limiting axial cracks based on measurements of the crack geometries performed subsequent to the burst tests. Even foi tubes that exhibit significant IGA, the licensee for Trojan has demonstrated that such tubes would be expected to conform to the same burst preswre/ bobbin voltage relationship as that developed from the pulled tube' data and laboratory OD5CC specimens. In our opinion, the prioritization analysis dun not provide a uedible basis f or concluding that implementation of the interim plugging limit at Trojan has created the potential for rupturing 10 or more tubes during a postulated steamline break (SLB) accident. This conclusion ignores the considerable evidence, discussed in the staff's SER, that all tubes can be expected to retain adequate integrity, consistent with the criteria of Regulatory Guide 1.121. This conclusion also fails to consider the probabilistic assessment of the potential for tube rupture during postulated SLB accidents that was included as part of the Trojan submittal. This assessment concludes that the conditional probability of one,or more tube ruptures, given an SLB, will reach a maximun value of only 3.4X10' during the current operating cycle at Trojan. This is an extremely low value compared to estimates employed in NUREG-0844 wnich were found to lead to low risk, finally, we note that the development of the interim plugging limit has conservatively taken no credit for the potential strengthening effect provided by the tube support plates on the portions of the tubes affected by ODSCC. Other comments we have concerning the prioritization analysis include the following: 1 Page 5 of the analysis refers to " inherent weaknesses in addy current testing" and that a RES sponsored study showed an average detection procability of 0.5 for liberatory produced flaws. We agrt,t that the probability of detecting short cracks at the support plate locations is relatively low until the cracks have penetrated beyond 70X through-wall, even with a fully implemented state-of-the-art inspection. This is because such flaws typically produce signals exhibiting low signal-to-noise ratios. In other words, the standard depth-based 40% plugging limit is not enforceable for short cracks at the support plate locations. We believe, on the basis of considerable field evidence, that cracks are detectable by eddy current inspection before ehey become suf ficiently large (in terms of both length and depth) to potantially impair tube integrity, assuming that appropriate test procedu4es and equipment are employed and assuming that the test personnel hhve )een adequately trained, 11 is our experience that such flaws will prtduce suf ficiently large amplitude signals so as to be detectable against the background noise. Based on our review of the inspection progrAn implemented at Trojan, we believe there is high assurance that all significant cracks which could potentially impair tube integrity were

identified and that the affected tubes were plugged or repaired. As discussed earlier, the NRR staff believes that the interim 1 volt limit is adequate to ensure tube integrity. Furthermore, we believe the 1 volt limit to be an enforceable limit (i.e.,1 volt indications can generally be discriminated from the background noise). 2. Page 5 of the prioritization analysis states that "a leak-before-break i (LBB) concept is inherent in the (staff's) safety evaluation" (for Trojan). The prioritization analysis states that this is contrary to the recommendations of the NRC Pioing Review Committee in NUREG-1061 which included a recommendation tlat the LBB concept not be applied to piping systems susceptible to IGSCC. Actually, the LBB concept in the context of GDC-4 played no role in the development of the interim ! volt limit. Furthermore, the development of the interim limit takes no credit for leak-before-break behavior of cracked tubes. The limit was developed to ensure no ruptures by enforcing the structural margin criteria of Regulatory Guide 1.121, in addition the limit is intended to ensure th:t any it k:ge which occurs during normal operating, transient, or postulated accident conditions will be within regulatory

limits, for the reasons given above, we continue to believe that the interim plugging limit approved for Trojan provides adequate assurance of tube integrity.

Furthermore, a criterion that we will continue to apply in our assessment of this issue is that there is no substantial increase in the risk of multiple tube ruptures as a result of implementation of the interim plugging limit. Thus, we conclude that Generic issue 163 should have a low priority and be dropped from consideration. NRR is currently reviewing permanent voltage-based alternate olugging limits (typically 4 volts), which have been proposed for a number of plants. The staff's lead-plant review is being conducted for Farley Units 1 and 2. It should be noted that a number of tough technical issues remain to be resolved before we will be in a position to approve these limits and to write an SER. We share some of the concerns expressed in background documentation for Generic Issue 163 (memorandum to C. Serpan from J. Muscara dated March 16, 1992, and memorandum to E. Beckjord from J. Hopenfeld dated March 27, 1992), and these are among the issues the licensee will be expected to resolve before we issue the SER. The lead-plant SER will set a precedent for how the staff deals with similar proposals for other plants. For this reason we plan to submit the SER for CRGR review prior to issuance, it is our plan to provide the aforementioned DP0 and the other background documents for Generic issue 163 as part of the

e . CRGR package. The CRGR package will address all concerns identified in the DP0 and other background documents. RES will be invited to comment on the CRGR package prior to its being transmitted to CRGR. Finally, we plan to work with RES to incorporate review guidelines for voltage-based plugging limits i into Regulatory Guides 1.83 and 1.121. T I l >mwe=,m=,y 3 g._,,.

now p , y,

'n UNITf D F#TATES i NUCLEAR REGULAT0fiY COMMISSION sd.r g'..v /.f ""i WASHINGTON. D C. 20%6 '/ ..+ WOV 3 01922 MEMORANDUM FOR: Frank Gillespie, Director Program Management, Policy Development, and Analysis Staff Office of Nu, clear Reactor Regulation FROM: Clemens J. Heltemes, Deputy Director for Generic issues and Rulemaking Office of Nuclear Regulatory Research

SUBJECT:

G1 163 ' MULTIPLE STEAM GENERATOR TUBE LEAKAGE" In response to your November 24, 1992, memorandum and the November 25th meeting with HRR representatives, we agree that the preliminary prioritization of this issue that we sent to you for comment on September 28, 1992, may have not accurately reflected the current licensing position that has been applied to Trojan and other plants. Based on your memorandum and the further information provided in the meeting, we will review the relevant facts and make appropriate revisions in our analysis. The revised prioritization will be forwarded to you for your further comments. k ok h Clemens J. Heltemes, Deputy Director V for Generic issues and Rulemaking Office of Nuclear Regulatory Research ENCLOSURE.6'

)

To

3cwwu s

np exstrc m UNION OF CONCERNED SCIENTISTS 4 November 23, 1992 Ivan salin, Chairman James R. Curtiss E. Gail de Planque Torrest J. Remick Kenneth c. Rogers U.S. Nuclear Regulatory Commission Washington, DC 20555 Sub$ects Multiple Steam Generator Tube Rupture Accidents Refereneet Memorandum for Frank P. Gillespie, Director, Program Management, Policy Development, and Analysis Branch, office of Nuclear Reactor Regulation f rom C.J. Heltenes, Jr., Deputy Director, Office of Nuclear Regulatory Research, September 28, 1992. As you are aware, in February 1992, the NRC etaff approved an aceniment to the operating 11 cense for the Trojan nuclear plant which permits Portland General Electric to operato the plant with flaws in the steam generator tubes ' deeper than 40-percent of the tube wall thickness. As described in the ref erence citod above, this prompted a nonbar of the NRC staf f to file a difforing proicssional opinion (DPo)r other members of the staff have also expressed concern about the risks to the public pesod by plant operation with known cracks in the staan generator tubes. You.nay also be aware that the Trojan plant is currently shutdovn because of a recent. steam generator tube leak in excess of the 130 gallons per day permitted under the February 1992 license amendment. As a result of the safety issues raised by some members of the i NRc staf f, s. new, high-priority-generio safety-issue -- The nultiple steam generator tube leakage -- has been opened.. analysis.by the of fice of Huclear Regulatory Research shows that operation with known flaws in the steam generator tubes can result in an accident in which savaral-steam gonorator tubos rupture, leading to molting of the reactor core and the release.of radioactive material directly to the. environment outside the reactor containment building. The staff has concluded that the probability of such an accident is over 300-tinos nore likely than your safety goal policy permits. FAX: 101438 09e8 =1614 P 8treet, NW Sulle ate Washingtsm, Do-2003e 102432 Otoe Camoroge Hascovaners: - 26 Churen Sveet teneridge MA CP738 - 617 547 5552 - fax:c17 664 540 ENCLOSURE 7

In accordance With the Chairman's policy of an open regulatory I am writing to make the following roquests of the

process, Consissisn.

1. Direct the staff to place the following documents in the Public Document Room promptly a) The Differing Professional opinion that led to the establishmont of Generic Issue 163, Multiple StJan Generator Tube Leakage. b) E.S. Beckjord nemorandum to Thomas E. Hurley, "A How Generic Issual Hultiple Steam Generator Tube Leakage,# June 16, 1992. c)

3. Hopenfold nonorandum to E. Beckjord, "A new Generic Issue Hultiple steam Generator Leakage," Harch 27, 1992.

d) Joseph Huscara memorandum to Charles 2. Serpan, " Steam Generator Tube Inspection, Integrity and Plugging," March 16, 1992. (Note Docunants b, c, and d are, respectively, references 1, 2, and 4 in the enclosure to C.J. Heltemos, Jr. namorandum to Frank P. Gillespie, "G1-163, Multiple steam Generator Tube Leakage,n september 28, 1992.) Direct the staff to notify me in advance of any meetings 2. with Portland General Electric, Westinghouse, or any other licenseos concerning the subjects of steam generator tube leakage limits and repair limits so that UCS can make arrange ents to attend. 3. Direct the staff to. explain to the public how the new, high-priority, generic safety issue concerning nultiple steam generator tube leakage will be considered in deciding whether the Trojan nuclear plant should be permitted to resume operation prior to rosolution of this generic safety issue. .4. Direct the staff to explain to the public why, in view of the new generic safety issue on multiple steam generator tuba leakage, the other Westinghouse plants which have been granted license anendments allowing operation with-flawed steam generator tubes should not be ordered to shut down until this generic safety issue is resolved. Sincerely, f Robert D. Pollard Nuclear Safety Engineer

'D' t UNION OF CONCERNED SCIENTISTS December 4,1992 Ivan Selin, Chairman James R. Curtiss 1 E. Oail de Plangue Forrest J. Remick Kenneth C Rogers U.S. Nuclear Regulatory Commission .m Washington, DC 20555

Subject:

01163, Multiple Steam Generator Tube 12akage I am writing to express my concerns about the adequacy of the NRC's responses to the requests mnde in my letter to you dated November 23,1992. \\ Thank you for the prompt Hetion in placing the documents I requested in the p'.tblic document room and for advance notification of the agenda for the meeting held at the Trojan plant on December 1,1992. I did not attend beenuse of the location, the short notice and the fact that the agenda was not focused on the stuffs assessment of the risks posed by multiple steam genertinr tube leakage. As far as I am aware, the stnff has not yet provided an explanation of why Westinghouse plants, other than Trojan, are being allowed to continue operation with flawed steam generator tubes before the subject generleissue is resobed. As for the staffs explanation, thus far, concerning the Trojan plant (l.c., a November 24,1992 memo io C). Heltemes, Jr. from F. Ollicsple and a November 30,1992 memo s to F. Olliespie from CJ, Heltemes, Jr.),I consider Jt to be wholly inadequate. Two major deficiencies in the staff's response, which are discussed in more detail below, are: 1) Addressing only the September 28,1992 memo from CJ. Heltemes, Jr. and generally ignoring the well documented technical issues raised in J. Hopenfeld's memos dated March 27,1992 and September 11,1992 and in J. Muscara's memo dated March 10,1992. 1eie P street. NW autte a10 washitoton, DC 2003s 20E 333 0900 - FAX: 301333 0D05 Camtydge Hencouarters: 26 Church Stres Camtydge.MA o7738 617 547 5552 FAX: 617 864 9406 - awee=co m -ENCLOSUllE 8.

(RJN IUOS/DO TO 13ClE042275 1992 12*04 11862CN 01M P.e2/EN 2) Falling to explain why the Trojan plant may be permitted to operate prior to the resolution of a generle safety issue that was opened as a direct result of the Trojan lleense amendment permitting operation with hundreds of flawed steam generator tubes. The following technicalissues are among those that the staff has generally Ignored in i r response.

An necident invoMng core mehing and radioactive releases outside the containment can necur even without the rupture of any stenm generator tubes. lenkuge of many Unwed stentn generator tubes could huve the same result as the rupture of fewer tubes.

The reason is that Dawed steam generator tubes which are not leaking during normal operation, could begin leaking under the higher differential pressure caused by n main steam line break (MSLD). Simi!ntly, the leakage through finwed tubes during normal operation at a rate less thnn 130 gpd could inerense signl0cnntly as the result of the higher differential pressure of a MSLB. The statements supporting the Differing Professional Opinion (DPO) are more succinct: 'The fact that degraded tubes neither leck, at normal pressures, nor burst under SLB pressures is not an indication that they will not leak following a SLB necident. * *

lt makes no difference whether the lenk origin was from one ruptured tube or many pin hole leaks." [ Memo from J. Hopenfeld, March 27,1992, Enclnsure, p.N

' The staffs response asserts that the Office of Nuclear Regulatory Research made assumptions that ignored the information obtained from tests and inspections of the Trojan steam generator tubes described in the stuffs safety evaluntion report (SER) for the Trojnn license amendment no.178. Even a cursory examination of the documents cited above would have shown that this assertion is false. The September 28,1992 memo from C Heltemes, Jr. and the March 27,1992 memo from J. Hopenfeld explicitly took into consideration the Trojan SER and the tests and inspections conducted on the Trojan steam generator tubes. l

The staff's response falls to even acknowledge the existence of, much less address, the large number of uncertainties involved in assessing the risk of pihnt operation with hundreds or even thousands of cracked tubes. For example, there is a low probability that flaws will be detected and that, even if detected,it is difficult to determine the length and depth of the cracks. There also appents to be insufficient data to be confident that the estimates of crack growth during operation are conservative or that l

leak rate monitoring during operation con provide an adequate basis for evaluating crack growth during operation und during accidents such as MSLH. Although the staff may have attempted to be conservative, there is insufficient data to make a compelling case that an ndequate safety margin remains, i -- _= 4 -e,m e r- ,-eww+w,-.--e~.-. -y-w...-- y g. p. ,wy-y., ,.-:c- -a,--e,- y-,w--- mg,r-. mm 7m -.-.mm.

, rrm u.cserse to isessuarts sce2 12.s 11 scam esco n.c: e '. The staffs response claims that the Office of Nuclear Regulatory Research erred stating that a leak.before brenk concept is inherent in the staffs safety evnluation report for Trojan. [ Memo hem F. Gillespie, November 24,1992, Enclosure, p. 3.} if that were a correct claim, the staff would have to acknowledge that the reduction in the allowable leak rate to 130 gallons per day provides no improvement in safe the tube leak rate previously applienble to the Trojan steam generntors. This is so because, unless the tube leak rate inerenses slowly concept is inherent in the staffs safety evaluation re(f.e., unless the leak before break wrt for Trojan), the tube leakage rate will go from negligible to a rate in excess of bot,i the new 130 gpd limit and the previous higher limit. In fact, this is essentially what occurred on with the tube leak at Trojan on November 9,1992, although I recognir.e that the licensee believes the ca was Imptoper hent treatment of a sleeve wcld rathe' r than a finwed tube. Never the same behavior of a rapidly increasing leak rate or tube rupture could oc stenm line break accident. As a final example,it should be noted that the staffs response does not uddress the point that relaxation of the steam generator tube repair criteria provides no sa benefit to the public. The sole benefit is to the economics of continued operation of the Trojan plant because, absent NRC approval of the more lenient repair crite:In plant probably could not r,perate at 100 percent of rated power. Furthermore, the approval of the Trojan license r.mendment violated the much touted princ defense in depth. What remains in terms of protection for public may be httle more than a Maginot Line. With regard the second major deficiency in the staffs response to date, t that the new generic safety issue established ni directly resuh of the Trojan license amendment need not be reschcd in order to permit operation of the Trojan plant. This ignores the critielsm of NRC's handling of generic safety issues express President's Commission on the Accident at Three Mile Island (the Kemeny Commission) and the U.S. General Accounting Office (OAO). The staff of the Kemeny Commission found that defining an issue as generic was i mechanism the NRC used to " insure the granting of an operating license for constructed plant." [ Staff Report to the President's Commission on the Accident at Three Mile Island, The Nuclear Regulatory Commissinn, October 1979, p. 43 The Kemeny Commission itself pointed to NRC's handling of generic safety issu "an important exumple" of how "NRC's primary focus is on licensing and insufficient attention has been paid to the ongoing process n assuring nuclear snfety." The r Kemeny Commbsion concluded that "the evidene indicates that labeling of a pro as ' generic' may provide a convenient way of postponing decision on a difficult question." [ Report of the president's Commission on the Accident at Three Mile Island, October 1979, p. 20.) t r eryw- -- -v 'vvT v w,ere-h=y we w -sar "T-"Ww-7 V T H r*'%

to isesse422vs sesa.sa-e4 s t i esen sis ne w.e4ee4 ,, m n,ves re More recently, the O AO performed an assessment of NRC's safety standards, enforcement activitics and inspection efforts. [U.S. General Accounting Office,

Efforts to Ensure Nuclent Plant Safety Can Be Strengthened," 0AO/RCED.87141, August 1987.) The following are some of the issues raised by the GAO:
The lack of guidelines to identify safety violations severe enough tu require a nuclear plant to shut down;
The slow corrective actions on the part of the NRC to shut down nuclear plants with records of chronic sufety violations, it was documented that the NRC has taken from several months to up to ten years to resolve generic safety issues including those that the Commission believes pose the highest safety risk.
The backlog of 163 unresolved generic safety issues as oflast December [1986),

including 32 considered to pose a significant risk to public health and safety. I am inclined to conclude that the same problems are affecting the NRC staffs handling of the hsues concerning the flawed s. cam generatur tubes at the Trojan plant,- at least based on the staffs response to date. I remain open to the possibility that there may be an adequate techn :al basis for 8 concluding that operatiun of the Trojan plant given the current condition of the steam generator tubes is neceptable. However,I do not believe thnt the stuff has yet provided such a basis. Thus, any requests for an adjudicatory hearing before Trojnn resumes operation which have been made by the public are understandable and should be given careful attention. The problem fucing Commission is that there is a fundamentnl disagreement between two of your staff offices about the risks posed by flawed steam generator tubes. One office bases its conclusions and recommendation solely on technien1 considerations. The other office feels oblignted to justify its decision to allow plant _ operation despite the unresolved technicalissues. How the Commission resolves this,roblem will depend,in part, on whether higher priority is given to protecting pu:>lic health and safety or the financialinterests of the nuclear industry. Sincerely, E ,($t .f Robert D. Pollard Nuclear Safety Engineer

'o,, UNITED $TATES [' n, NUCLEAR REGULATORY COMMISSION g t W ASHING TON, D. C. 20f.55 k.**, DEC 0 919?2 = MEMORANDUM FOR: Eric S. Beckjord, Director Office of Nuclear Regulatory Research FROM: Lawrence C. Shao, Director Division of Engineering Office of Nuclear Regulatory Research

SUBJECT:

INTERIM PLUGGING CRITERIA FOR TROJAN NUCLEAR PLANT The Division of Engineering has provided a discussion of the key technical aspects of the rationale used to support steam generator tube interim plugging criteria (!PC) for the Trojan nuclear plant and to provide independent conclusions on the viability of IPC for one fuel cycle. The IPC apply only to the specific case of outer diameter stress corrosion cracking (ODSCC) and intergranular attack (IGA) at tube support plate (TSP) intersections in the steam generators. The technical rationale presented in the enclosure are based on data and analyses available from NRC research, Trojan plant operating experience, and the technical literature. The enclosure also reflects staff technical experience and opinions. The Office of Nuclear-Reactor Regulation (NRR) has been consulted on technical details regarding IPC during the preparation of this document. The report endeavors to maintain a distinction between staff opinion and published data. Based on the discussion presented in the enclosure, the Division of Engineering concludes that continued operation of the Trojan plant for one fuel cycle is justified. This justification is based on: (1) Examination of steam generator tubes removed from service at the Trojan plant which has revealed cracks that are generally confined to the tube support plate intersections. (2) Burst test results from cracked tubes removed from service at the Trojan plant which showed burst pressures well in excess of main steam line break (MSLB) pressure. (3) Stress corrosion crack growth rate results which indicate that' incremental growth of the cracks to a critical length beyond the tube support plate during one fuel cycle is unlikely. ENCLOSURE 9

i 4 ENCLOSURE i Discussion of Technical Rationale for Steam Generator Tube Interim Plugging Criteria (IPC) at The Trojan Nuclear Plant i lhe purpose of this report is to provide a discussion of the key technical aspects of the rationale used to support steam generator tube interim plugging criteria (IPC) for the Trojan nuclear plant and to provide independent conclusions on the viability of IPC for one fuel cycle. The IPC apply only to the specific case of outer diameter stress corrosion cracking (ODSCC) and intergranular attack (IGA)hnical rationale presented in this report are based at tube support plate (TSP) intersections in the steam generators. The tec on data and analyses available from NRC research, Trojan plant operating experience, and the technical literature. The report also reflects staff l technical experience and opinions. The Office of Nuclear Reactor Regulation (NRR) has been consulted on technical details regarding IPC during the preparation of this document. The report endeavors to maintain a distinction-between staff opinion and published data. i lhe rationale presented in this report are based on technical. considerations which we believe are adequate to justify IPC for one fuel cycle. Subsequent operation with IPC would require additional review after completion of one cycle and would require consideration of additional information developed at that time. Longer term technical considerations, such as reliability and sensitivity of NDE techniques for steam generator tube inspection, are the subjects of on-going and new NRC research which is being coordinated with NRR as part of an overall steam generator tube alternate plugging criteria (APC) action plan. (1)

Background:

Steam generator tube structural integrity guidance provided in Regulatory Guide 1.121 has generally translated into a 40% through-wall " plugging limit" for flaws in steam generator tubes as part of the plant technical specificatiuns. However, evidence from pulled steam generator tubes at several plants has revealed numerous short cracks at TSP intersections which are greater than 40% through-wall and yet can withstand pressures in 1 excess of three times operating as required by Regulatory Guide 1.121.- it has therefore been argued by the industry that the 40% plugging limit is conservative, at least for the case of short axial ODSCC/lGA confined-to TSP intersections. Burst testing of cracked tubes removed from service at the 1rojan plant has resulted-in burst pressures of at least a factor of two in excess of main steam line break (MSLB) pressure, even for through-wall cracks? NRC research results on tubes with machined and chemically-induced flaws support the contention that the tubes retain significant structural integrity even for up to through-wall cracks, provided that the_ cracks are short. From this research "short" can be defined as less than 0.5 inches, which is the length of a near through-wall crack needed to burst for 7/8-inch diameter, 0.050 inch wall thickness tubing under HSLB ;ifferential pressure' (see figure 1). The burst pressure is defined as the pressure required to_ penetrate the tube wall. Tube burst then, can result in either small or. large leakage. Tube burst results when the-differential pressure acts from the primary-side. Tube rupture relates to a.significant opening under burst 1 ---+-.----.----__--,-._m , _. ~. ~~.-,,,,-_.m.,,,N ,e--r. w,.w,c -.,..ev_v .m.,,

e wall crack extending on the order of 0.5 inches beyond the TSP intersection. As described previously in (2), the Trojan cracks were generally confined to the TSP thickness; hence, growth beyond the TSP on the order of 0.5 inches would be required for these cracks to be considered critical from a MSLB pressure standpoint. Upper bound laboratory ODSCC growth rate data' indicate that crack growth of this magnitude would not be expected to occur during one fuel cycle. While a through-thickness, full TSP length crack would be expected to fail at MSLB pressure, the opening or rupture would be constrained by the tube.;upport plate. True rupture for the portion outside of the TSP would be expected to occur at MSLB pressure only if the crack had grown on the order of 0.5 inches beyond the TSP intersection. Further, little or no movementoftheTSPwhichcouldpotentially' uncover"thecracksispredicted to occur for the HSLB condition. (5) Probability of Main Steam Line Break: The probability of a MSLB, the key initiating event for a steam generator tube rupture, is very low. The HSLB would cause approximately a 2600 psi pressure differential across the steam generator tubes. A MSLB has never occurred in a U.S. nuclear plant. Quoting from reference 5, "Under the Evaluation and Imorovement of NDE for Inservicg inagglion of liaht Water Reactors Prooram sponsored by the NRC, a team of-experts estimated the median frequency of a NSLB to be 1.7 x 10" per reactor year for a volume of 50 gallons per minute. This extrapolates to a frequency estimate of 6.8 x 10" per reactor year for a four loop plant such as Trojan. (6) Summary and

Conclusions:

Based on a review of Trojan steam generator tube operating experience, on destructive examinations of tubes removed from the-Trojan plant, stress corrosion crack growth rates and expert opinion concerning MSLB frequency, it is concluded that-operation of the Trojan plant with steam generator tube IPC for one fuel cycle does not constitute a significant threat to public health and safety. Subsequent operation with IPC would require additional review after completion of one cycle and would include consideration of information developed at that time. - In summary, the above conclusion is based on: (1) Examination of steam generator tubes removed from service at the-Trojan plant which has revealed cracks that are generally confined-to the tube support p' ate intersections. (2) Burst test results.from cracked tubes removed from service-at the Trojan plant which showed burst pressures well in excess of main steam line break (MSLB) pressure. (3) Stress corrosion crack growth rate results indicate that incremental growth of the cracks to a critical length beyond the tube support plate during one fuel cycle is unlikely. (4) The probability of a _ main steam line break, the key initiating-l event for a steam generator tube rupture is very low for one fuel cycle.- i 3

ly.... %, y" UNITED STATES c .i NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666 o %, " " * * / December 10, 1992 t Mr. Robert D. Pollard Nuclear Safety Engineer Union of Concerned Scientists 1616 P Street, NW, Suite 310 Washington, DC 20036

Dear Mr. Pollard:

This is in response to your letter dated November 23, 1992. In your letter you expressed concern regarding the steam generator tubes at the Trojan Nuclear Plant and requested that the Nuclear Regulatory Commission (NRC) place certain documents in the Public Document Room (PDR). You specifically expressed concern about the possibility of a multiple steam generator tube rupture and the licensee's plans for restarting the Trojan plant. The documents we are providing are considered pre-decisional and reflect ongoing staff discussions on a complex technical issue. During development of the resolution to generic technical issues such as the steam generator tube integrity issue, different opinions and analyses are developed by members of the NRC staff as a normal course of agency business. The NRC has a formal internal review process, including peer and management review of all analyses and opinions, to ensure that all technical and safety issues are identified and resolved before requiring industry action. This further ensures that modifications are not imposed on the basis of incomplete information. Nevertheless, we are providing the following documents to the public at this time to make available all current opinions and technical analyses: The differing professional opinion that led to the establishment of Generic Issue 163, Multiple Steam Generator Tube Leakage. E.S. Beckjord memorandum to Thomas E. Hurley, 'A New Generic Issue: Multiple Steam Generator Tube Leakage,' June 16, 1992. J. Hopenfeld memorandum to E. Beckjord, "A New Generic issue: Multiple Steam Generator Leakage,' March 27, 1992. Joseph Huscara memorandum to Charles Z. Serpan, " Steam Generator Tube Inspection, Integrity and Plugging,' March 16, 1992. However, the NRC staff wants to make clear that full review of the steam generator tube integrity proposed generic safety issue has not been completed. We will provide the public with the final NRC staff resolution of this proposed generic safety issue when we have completed all appropriate reviews. W ENCLOSURE 10 r--- -*= -em r +r

l 1 ,l djrn s /G /c,. I 2 /,,, s'o, UN11 ED STAT Es [' h NUCLEAR REGULATORY COMMISSION /f g /7g W ASHING TON, 0. C..*t".$5 p \\...../ maicim .ga c /.33m /u (E l%% (a fd.,. HEMORANDUM FOR: Charles Z. Serpan, Jr., Chief .z Haterials Engineering Branch, RES /Wr7 e i e nf4n,/;.- Joseph Huscara, Senior Metallurgical Eng$e r *b d/4 FRON: Haterials Engineering Branch, Division of Engineering, RES

SUBJECT:

STEAM GENERATOR TUBE INSPECTION, INTEGRITY AND PLUGGING ISSUES f'/)g frj% As you know, I have been the research program manager for NRC's Steam / Generator Tube Integrity Program. This program started in FY 1977 and was completed in FY 1988 with the validation work on the Surry 2A steam generator removed from service. The validation work on Surry was conducted with international cooperation. The total program cost approximately IBM dollars, 4H of which came from EPRI and consortia from France, Italy and Japan. Many results, analyses and insights were derived from this program. Based on the results, several improvements have been incorporated in an update of ASHE Code Sec. XI Appendix IV on Eddy Current (EC) testing of steam generator tubes. Other improvements are being incorporated in revisions to Regulatory Guides 1.83 and 1.121 on steam generator tube inspection and plugging criteria. The results and insights gained also lead to concerns with generic acceptance of industry proposals for alternate tube plugging criteria which would allow operation.of steam generators with known through-wall cracked (leaking) tubes. My concerns with acceptance of this alternate plugging criteria which would permit operation of steam generators with through-wall cracked tubes are on two grounds: Firstly, acceptance would imply violation of the long standing engineering principle of defense-in-depth used by the NRC and of the leak-concept in the spirit of tight reactor coolant pressure boundary (RCPB)d us well for many years. 10CFR50 Appendix A - these concepts have serve Secondly, if we accept violation of the leak-tight integrity of the RCPB, then a strong engineering case must be made and actions taken to ensure maintenance of structural integrity during operation of power plants with leaking steam generator tubes. Based on NRC's independent research results, other. evaluations and experience, I believe a case assuring structural integrity with today's capabilities for detecting and sizing flaws, for estimating crack growth and progression during operation, and for evaluating crack growth and stability during operation through leak rate monitoring cannot be made. The structural integrity of degraded tubes is governed by the flaw length and depth; for tubes with through-wall cracks, the important parameter is crack length. Under HSLB conditions the critical through-wall crack length for axial cracks is about one inch. Tubes with through-wall cracks of the order of one inch and longer would burst or develop large leak rates under HSLB conditions. The comor.1v used eddy current (EC) techniques have poor ENCLOSURE 11

j c. 3 0 1932 Charles 2. Serpan, Jr. 3 Unfortunately, for the types of cracks of through-wall. crack length. interest, correlations of leak rates to crack length and measured leak rate to predicted leak rates, show approximately two orders of magnitude variabilities. Further, in service, cracks tend to be tight, can become fouled with small particles and corrosion products, and may be surrounded and constricted by corrosion products and sup> ort structures. Therefore, through-wall cracked tubes in service tend to leac very little whether the cracks are short or long. Furthermore, approach of cracks to critical sizes cannot be determined since very small changes in leak rates are expected in service. These considerations indicate that critical cracks cannot be distinguished from subcritical ones based on the observed leak rates. Finally, when hundreds or even thousands of tubes may be leaking, how does one determine or distinguish those tubes with cracks that are or may be approaching critical size under MSLB7 How many are there? Small leakage does not necessarily mean short cracks. In recent proposals from industry, the EC voltage has been used to relate an inspection parameter to tube integrity. This parameter does not uniquely measure the crack depth or length. For some crack morphologies of interest the voltage is not expected to relate to tube integrity for the following

1) the voltage saturates at a crack length of approximately 0.5 reasons:

inches. Longer flaws will not produce a larger voltage;

2) the voltage produced is related to the crack tightness, if cracks are tight enough (and conductivity paths exist), the voltage response is low whether the cracks are long or short; 3) voltage produced is insensitive to critical crack morphologies - short, tight cracks axially aligned and separated by short ligaments produce low voltage indicative of the short segments; but from a structural point of view, an effectively long crack exists which is not characterized by the voltage. Plots of voltage vs. leak rate and of voltage vs. burst pressure have been presented. As expected, those plots show little correlation and show two to five orders of magnitude variability in the data.

Further, these plots lack data for long-tight cracks and for short cracks axially aligned with short ligamerts in between which could produce low voltage and also low burst pressure. Also a " voltage growth rate" is proposed as a measure of crack growth rate. Again, since the voltage does not relate uniquely to crack size, for the cracks of interest, all these voltage correlations have little meaning. Based on the abovp discussions, variabilities and uncertainties, I believe that the NRC position of the recent past that through-wall flaws and cracks are not acceptable is still prudent at this time. In recent years NRR staff have-agreed that through-wall flaws in steam generator tubes are not acceptable and have adhered to the concept of maintaining leak-tightness of the RCPB.- Further, NRR staff has agreed that if alternate plugging criteria is considered on a case specific basis that a strong engineering case must be made and actions taken to ensure structural integrity of tubes during operation. Enclosure (1) to this memorandum briefly discusses the issues related to steam generator tube inspection, integrity and plugging, and provides limited examples from the large body of data and y -w sw +- w -er 7 s m= = - g-

1 Enclosure (1) STUJi GENERATOR TUBE INSPECTION, INTEGRITY AND PLUGGING ISSUES The following discussion addresses issues related to operation of steam tubes. Two general areas are generators with through-wall cracked (leaking)d the policy of defense-in-depth discussed; 1) engineering design philosophy an and 2) technical issues related to assurance of maintaining tube integrity of cracked steam generator tubes during reactor operation. Enoineerino Desion Philos.gshy and Defense-in-Depth of Appendix A to 10CTR50 require that the General Design Criteria (GDC)dary (RCPB) have an extremely low probability of reactor coolant pressure boun abnonnal leakage, of rapidly propagating failure and of gross rupture. Further, the RCPB is to be designed to permit periodic inspection and testing to assess the structural and leak-tight integrity. Using materials that exhibit leak-before-break behavior, maintaining leak-tightness of the RCPB and conducting inservice inspection (ISI) to assess structural and leak-tight integrity are important elements of defense-in-depth for maintaining safety and are not meant to allow operation with a leaking RCPB. The GDC indicate and the NRC staff has interpreted that through wall cracks in the RCPB are not acceptable. Several recent actions attest to this interpretation:

1) GDC 4 on exclusion of dynamic effects from ruptured pipes does not apply to materials susceptible to degradation; 2) ASME and NRC rules for evaluation of cracked stainless steel pipe do not allow operation with pipes containing through-wall cracks even though these pipes may exhibit leak-before-break.

Pipes with cracks deeper than 75 percent through-wall must be repaired; 3) NRC guidance for leak monitoring of RCPB allows for a small amount of unidentified leakage, however, if leakage is from a through wall crack, the component must be repaired; 4) HRR coments from review of a proposed revision to Regulatory Guide 1.121 required the guide to state that through-wall flaws of any type and identified cracks of any size are unacceptable. Since the steam generator tubes comprise over 50 percent of the RCPB surface area and hundreds, even thousands of tubes could be leaking with an alternate tube plugging criteria, it is important to adhere to the policy of non-penetration of the RCPB. TECHNICAL ISSUES If it is decided that it is acceptable to operate a nuclear power plant with the primary pressure boundary violated, i.e. with through-wall cracked steam generator tubes for the situation under discussion here, then a strong engineering case needs to be made and actions taken to assure maintenance of structural integrity. The important parameters relating to the structural integrity of steam generator tubes are the crack length for through-wall cracked tubes and the crack length and depth for other cracks. Cracked tubes can exhibit no leakage, small leakage or large leakage and burst behavior - under normal operating and accident conditions. For through-wall cracked tubes, with axial cracks, the crack-length at which large leakage or burst occurs (critical crack length) under MSLB condition is approximately one-inch. Various combinations of crack lengths and depths for part-through-wall flaws can lead to burst under normal operating or HSLB conditions. Therefore, 1 i 4

s phase angle for detection and sizing of flaws. Recently a parameter has been This errerging, the voltage (or amplitude), as a measure of tube integrity. parameter does not uniquely measure the length or depth of flaws, the critical l parameters from a structural integrity point of view. Table I shows data from laboratory produced part-through-wall stress corrosion cracks. For various crack morphologies of interest the voltage is not expected to relate to tube integrity for the following reasons:

1) For flaws of given width and depth a correlation exists of increasing voltage with length up to a flaw length of approximately 0.5 inch.

Longer flaws will not produce a larger voltage than this saturation level; this saturation of voltage for approximately 0.5 inch long flaws and longer is based on the coil design.

2) The voltage produced can be related to the tightness of the cracks; if the cracks are tight enough, and conductivity paths exist, low voltage response is expected whether the cracks are short or long. Of course from a structural point of view the larger flaws are more important and the voltage parameter would not distinguish between them. 3) The voltage produced is insensitive to critical crack morphologies.

For example a number of short, tight cracks (deep or through wall) axially aligned with short ligaments between them would produce a small voltage indicative of the tight short segments of the cracking. From a structural point of view such cracking would behave like a long crack 1.e. tubes would have low pressure holding capability; under pressure the ligaments i would join to produce critical length cracks and high leak rates. The voltage response from such cracking would not predict the structural integrity. frackina Hechanitms and Growth Rate Variabilities The discussion on crack detection and sizing reliability indicates that important cracks can be easily missed and those that are detected cannot be adequately sized. Even if important flaws were adequately detected and sized, the crack growth rates, both in terms of depth and length, are required in order to estimate the crack sizes at the end of the operating cycle, before the next inspection, to assure that accepted cracks remain below the critical size by a reasonable margin. Research results show that variabilities of one order of magnitude can be easily expected in crack initiation times and growth rates for environmentally assisted cracking even under test conditions where samples of the same material are exposed to the same environment, temperatures, stresses, etc. Much variation in the operating environment of steam generators exists for the power plants in the U.S. Conditions of water chemistries, temperatures and thennal hydraulics can differ from plant-to-plant; geometries, crevice conditions, heat of maarial, temperatures, water chemistries, stresses, etc. can vary even within the same steam generator. As a consequence, many different types _of cracks have been experienced at different U.S. steam generators and even within the same steam generator. Primary and sernndary side cracking has been experienced. Cracks in tubes at-various locations has occurred such as in the tube sheet crevice, at top of tube sheet, in free span zones, within the tube support plate regions, at U-bends etc. Fatigue. cracks, intergranular corrosion crack:, intergranular attack and crevice corrosion cracks have been experienced.- Some of the intergranular cracking is associated with stress such as at dented regions, other intergranular cracking is not associated with any significant stress such as at crevices in undented regions. Several forms of intergranular attack and combinations of intergranular attack and cracking have occurred. 3 .w,-io e ,w-_.. .=... - - -... .y %r+r -r*-- n ~t-- w-.>- -re----- r----ww*rT+- t<-- w ns'- +'-w*T--wv^r--wPt +r T w +M--> w e'wm-- 1 s*~-"- T~P*9

- - ~. - s Leak Rate / Crack Size Predictions and Variabilities Assuming for a moment that a) through wall cracks can be reliably detected and their length accurately sized b) the amount of crack growth, any changes in cracking mechanism and morphology, and growth outside of original zones can be reliably predicted and c) particular cracks will not approach critical sizes during the next operating cycle and are left in service, then these cracks must be monitored during operation to assure that they will not approach critical size. To accomplish this monitoring, leak rate measurements and j specifications are established. Unfortunately, for the types of cracks of interest, correlations of leak ratas to crack sizes and measured leak rates vs. predicted leak rates (from fluid flow and fracture mechanics models) show approximately two orders of magnitude variabilities Figures 14 and 15. PLEASE NOTE THE PROPRIETARY NATURE OF FIGURE 15. The variability is due to several unknown or uncontrollable factors. The length of cracks varies from ~ the inside surface to the outside surface, these' lengths are not always known or easily measured in service; the leak path for IGSCC is variable and highly tortuous; cracks can be very tight and of variable tightness. Further, in service, cracks can become fouled with small particles and/or corrosion products and may be surrounded by support structures and corrosion products. Under these conditions it is difficult to relate leak rate to crack length (to assure it is below critical length). Furthennore, through-wall cracked tubes in steam generators leak very little inservice whether the cracks are short or long because of tightness, fouling, and constriction by corrosion products or support structures. So, critical cracks cannot be distinguished from suberitical ones based on observed leak rates. Furthermore, the approach of cracks. growing to critical sizes cannot be detennined since very small changes in leak rates are expected in service. Finally, when hundreds or even thousands of tubes may be leaking a very small amount, how does one distinguish those tubes that have cracks of, or approaching critical size under MSLB conditions? How many are there? Small leakage does not necessarily mean short cracks. A recent drafi EPRI report attempts to show correlations between leak rate and EC probe voltage, Figure 16. These correlations are used in submittals to support alternate tube plugging criteria. The log - log plot of Figure 16 shows very little correlation of voltage with leak rate. Five orders of magnitude variability is shown for leak rates at a given voltage and one to two orders of magnitude variability in voltage for a given leak rate. The correlation coefficient for this plot is reported as 0.73 which also indicates very poor correlation. This plot also lacks data for cracked SG tubes which produce low voltages. Burst Pressure vs. Deoradation The NRC tube integrity results indicate that tubes with short flaws exhibit more strength than tubes with longer flaws of the same depth, also tubes with shallow flaws can withstand considerably more pressure. Figure 17 shows plots from an empirical equation derived from the data for EDM notches and validated by testing of stress corrosion cracks. It is not surprising that tubes removed from service have exhibited high burst pressures; this can be predicted for short through-wall flaws or for other reasonably deep flaws. 5

TABLE 1. ODSCC Flaw Dimensions and Bobbin Voltages

4 Spodmon Maximum OD Surf acc Haw tsoboln Honormalized Numbar Depth,%

Length. In. Voltage, Volts Voltage Volts B-6348 26 1.41 0.32 1.16 0-46 02 31 1.06 1.00 3.61 F 10 37 0.25 1,62 5.85 F-15 38 0.25 2.38 0.59 B4010 38 0.53 1.37 4.95 B-8147 42 0.25 0.43 1.55 B47 43 0.66 1.48 5.34 B-8341 44 1.14 0.46 1.66 8-62-08 42 1.43 2.04 7.36 B-6103 47 0.69 0.62 2.24 E-1105 50 0.64 1.31 4.73 E-07 07 58 0.45 3.44 12.42 B-62 02 61 0.50 1.71 6.17 B-4604 58 0.70 0.37 1.34 B-5544 59 0.91 1.92 6.93 B-6346 59 1.11 1.82 6.57 B 5907 76 0.81 2.22 8.01 E-1103 86 0.44 4.57 16.50 B-1048 99 1.09 7.24 26.14

Data from NRC Steam Generator Tube Integrity Program-Phase II 7

t. .o I Team V-1.0 Nf O 'S 0.8 - e oo 0.6 - O m x 3_?. 0.4 - p 'A $.0.2-u n. 1 0,0 - i i 20-40 60 80 100 Metallography Wah Loss, % 2fgure 2 - Resu!is from HRC Steam Generator Group Project j R: ported in HUREG/CR-5185 .v.. d f g g 8. '4 " % r

Team V' i.. 100. g 80 - m x o x x .J y x x x 60 - k x gx 7 x x *xx x x xy x x, o 40 - x e x xx (O I: x xx X g .;s 20 4: x x. x m W 0-x =xx x x i i i 4 0 -20 4 0-60 80 .1 0 0 Metallography Wall Loss, % Figure 4 - Results from NRC Steam Ge6erator Group Project Reported in NUREG/CR-5117 and NUREG/CR-5185

C

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E '.lGSCC Round Robin Team MD - Bobbin 100 80-o / 5 60-l o l t S O m 5 40-0 l w I-m 20 - 1 / Os w- =, 0 20 40 60 80 100 Measured Crack Depth, % Figure 6 - Results from NRC Stean Generator Group Project i Reponted in NUREG/CR-2336 and HUREG/CR-5117 i N

IGSCC Round Robin Team MF - Bobbin 100 o O O 80-

R o

D id D c) 60-O t S 0 m .E 40-u m .o 20- -0+ woc - 0 20 40 60 80 100 Measured Crack Depth, % Figure 8 - Results from NRC Steam Generator Group Project Reported in NUREG/CR-2336 and HUREG/CR-5117 e'

4 IGSCC Round Robin Team MH - RPC 100 w 00 D D 80-o D i o. 8 60 - "O oE o ~ E 40-D M W Fm 20-0+ '-c w => i 0 20 40 60 80 100 Measured Crack Depth, % ~ Figure 10 - Results from HRC Steam Generator Group Project g Reponted in NUREG/CR-2336 and NUREG/CR-5117 c

IGSCC Round Robin Team MO - Array Coll 100 / 80-2 E. o 60-O u2 a .E 40-n m 3 o 20-o on 0 0 0S s--m moc r 0 20 40 60 80 100 Measured Crack Depth, % Figure 12 - Results from NRC Steam Generator Group Project Reported in NUREG/CR-2336 and NUREG/CR-5117 ~.. O

.;. A, Comparison of French Leak Test Data with Model Predictions Normal Operating Condition Plastic, Analy,s,1s 16 - +-a -

Ela,stic Analysis

_e. y /.Outside ~~ Average % 's inside 1 Predicted d' ' #N \\/ / French Dato 5 / / / / 5 A ~ 8 -' d_f ^ i o.lo M. 3 ' ya g / i 3 0.01 Range of Crack Length 3 as Measured from ID to OD 1 I I I 3 1 1 I L 0 0.2 0.4 0.6 0.8 1.0-0.001 Crack Length, inch Figure 14 - Data taken from various sources, see NUREG/CR-2336 l-1 and NUREG/CR-5117 / l-21 L-

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y hs1 a,;- 1 W 14 h 4+- 4 t f['[4,.h= hi es ,a q! q .i t .t- ~ J BURST PRESSURE PARAMETER CURVES EDM Slot Specimens (7/8 x 0.050 inch Tubing) t 1 :- 0.9 - 0.8 - 0.7 - O Q. H<d 0.6 - h/t=0,5 Q 0.5 - ,~ c. <F 0.4 - dO 0.3 - h/t=0.80 "^0^^^^^^^^^0 0.2 - h " " " - -. a a /t=Q.90 a - a a a a-an 0.1 - h/t=1 "77---mm m mm o m.. 0 4 a i a a a i i ~ 0 0.4 0.8 1.2 1.6 2 LENGTH Cnches) Figure 17 - Results from NRC Steam Generator Tube Integrity Program - Equations for Curves and Similar Plots Reported in NUREG/CR-0718 3.,., X, /Sc (9 s,n Y M pap,e6 y ik p m ~n a J k., ban cGl.c.G/ p<- M 20 ^' aq'y .h s - 4 Enclosure (2) ISSUES AND QUESTIONS RELATED 10 ASSURANCE OF MAINTAINING STRUCTURAL INTEGRITY OF THROUGH-WALL CRACKED STEAM GENERATOR TUBES Has the Commission decided that it is acceptable to o)erate steam generators, whose tubes comprise over 50 percent of tie reactor coolant pressure boundary (RCPB) surface area, with hundreds or even thousands of through-wall cracked tubes? If it is decided that it is acceptable to operate SGs with through-wall cracked tubing, and therefore will have eliminated the lenktightness of the RCPB as a very innportant element in defense-in-depth for maintaining safety, then a strong engineering case needs to be made to assure maintenance of ctructural integrity during operation. To maintain structural integrity,, flaw length must remain below a critical length. Key issues in assuring structural integrity are knowledge of: a) the through-wall flaws present, b the crack length and accuracy of measurement, c cracking mechanisms and crack growth rate and d knowledge of crack size and progression from leak rate monitoring. Questions and comrnents related to these issues are as follows: Nhat is the probability of detection (POD) for deep and through-wall cracks as a function of crack length? Past experience and results indicate low POD. What is the accuracy of length and depth sizing? If voltage is used as a measure of tube integrity, how is voltage related to length (and depth for deep flaws)? Voltage saturates as a function of length below the critical length. What is the voltage response for tight cracks - even if long? What voltage response and variation is expected for effectively long cracks made up from a series of short cracks axially aligned with small ligaments in between? Tubes with short cracks, even if through-wall, will exhibit high burst pressures. However, tubes with deep part-through wall flaws (85 percent and greater) and through-wall flaws above approximately 0.6-inch long will exhibit burst pressures below HSLB. Again, from the voltage, what is the flaw size for these flaws? 1 E. In the burst pressure vs. voltage correlation, were effectively long, tight cracks (expected of producing low voltage and low burst pressure) considered? Regarding voltage vs. leak rate, considering five orders of magnitude scatter in the data and correlation factors of 0.7 to 0.8, is it considered that a reasonable correlation exists? How well do we understand the various mechanisms of cracking? What are the causative factors and synergisms? What assurance is there that cracking mechanisms will not change during operating cycles? Why wouldn't existing cracks grow beyond the initial locations, i.e. outside of support plate for crevice corrosion cracking? What are the crack growth rates to be applied to estimate crack length (or depth) at the end of operating cycle? What reliable correlations exist between crack length and measured or predicted leak rates?. How do leak rates measured inservice relate to crack length considering corrosion products, fouling, residual stresses, etc. which tend to restrict leakage? What changes in leak rates are expected and can be measured as cracks approach critical lengths? In Monte Carlo evaluations to predict expected leak rates under normal operating, and accident conditions, how are non-detections of through-wall cracked tubes considered? M 2 l 4 HEMORANDUM FOR: Thomas E. Murley, Director Office of Nuclear Reactor Regulation FROM: Eric S. Beckjord, Director Office of Nuclear Regulatory Research

SUBJECT:

A NEW GENERIC ISSUE: MULTIPLE STEAM GENERATOR TUBE LEAKAGE A Differing Professional Opinion (DPO) was recently filed in the Office of Researth regarding steam generator tube integrity during a postulated main steam line break (MSLB). The DP0 specifically addressed the operation of steam generators with microscopic tube wall cracks. This concern deals with the possibility of having multiple steam generator tube leaks during a MSLB that cannot be isolated. This sequence could lead to core melt resulting from the loss of all primary system coolant and safety injection fluid in the refueling water storage tank. The DP0 recomended that plants operating under these conditions have additional sources of borated water readily available to replenish the refueling water storage tank. It was agreed to resolve the DP0 that this concern would be proposed as a new generic issue. This new generic issue is entitled, " Multiple Steam Generator Tube Leakage," the details of which are enclosed. Since your staff is currently evaluating the issue of microscopic cracks in steam generator tubes, we are submitting this information to you for your review. This is also in accordance with RES Office letter No. 1, Revision 3, ' Procedure for Identification, Prioritization, and Tracking of the Resolution of Generic Issues," page 3, second paragraph, which states: "After their acceptance, generic issues that originate from outside NRR or from an individual within NRR (i.e., not sent through NRR management) will be transmitted by RES to NRR/PMAS for an imediate action determination and screening for identification of overlap or duplication with already imposed or completed Multi-Plant Actions (MPAs). If HRR cannot complete this imediate action detennination and MPA screening within 15 days, RES should be informed when the NRR review will be completed." We will also keep you informed of our progress in resolving this issue. oRICINAL SIG10D 3[ y% LO.W Eric S. Beckjord, Director Office of Nuclear Regulatory Research

Enclosure:

As stated cc: F. Gillespie J. Richardson A. Thadani CONCURRENCE: .SEE PREV 1005 CONCURRENCE jf f 0FFICE:RES:DSIR RES:DSIR RJS R DD/RES DD/RES D/l NAME: G.MARIN0* T. KING

W./16NERS T.SPEIS C.HELTEMES E.i

'30RD DATE: 6/11/92 6/11/92 6/(f/92 6/ /92 6/ /92 6/y92 ENCLOSURE 12

l, [ -; t-:2

[ n.te;.

R 'c; o. 01STRIBUTION l Copies w/o'en't. R- =DSIR c/f -T. King RES-Circ./Chron- 'R.Emrit-- .. Beckjord -./ J.Hopenfeld iMinners <- P.Norian

Marino /

K.Kniel 'Marino c/f ~Heltemes v : 1 ' -:q 9 L b 6 9 s. ( s I I i E m:

Agaq, NC0 1992 NOT E TO: G. Burdick FROM :

J. Hopenfeld

SUBJECT:

Reply to your request for comments on Draft 'RES POSITIOt' JN STEAM GENERATOR TUBE INTEGRITY' by LC. Shao Based on certain data, discussed in items 1-8, the document concluder that "it is reasonable to continue operation for one fuel cycle with flaws greater than 40 irough-wall at TSP intersections.' The document further suggests that " subsequent operation will require additional review after completion of one cycle and willinclude consideration of information developed at that time".

GENERAL COMMENT

The information provided in items 1-8, of the subject document does not address the main issue conceming 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, which will result in a significant primany-to-secondary leakage. Whether the leakage is significant or not would depend on whether the operator can stop the laak before the RWST is depleted. Degraded tubes also may cause a significant increase on risk from severe accidents. The fact that cracks within the TSP can withstand the MSLB pressure and that their length will not 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 leaks at pressures, of 3300 psi, 7500 psi, and 5500 psi,. with an average depth of penetrations of 38%,58% and 72% respectively. Item (7) points out that the above specimens "have shosn no leakage under normal operating or MSLB pressure conditions". IT FAILS TO POINT OUT, HOWEVER, THAT THERE IS NO DATA WHICH WOULD ALLOW ONE TO RELATE THE ABOVE LEAKAGE WITH THE OBSERVED DEGREE OF DEGRADATION. In other words, if these specimen had undergone a more severe wall penetration would these specimens have leaked at 2600 psi.7. Considering that the 21 specimen represent a sample of a population of 13,000, the conclusions in (7) above are questionable. The document ignores two other tubes which were pulled out of two US plants and developed leaks at SLB pressures. The leakage was at ! east an order of magnitude higher than under normal detta ps'. A third tube from a Belgian plant indicates a factor of eight increase in leakage under SLB conditions, (see Mar. 23 memo). Theoretical ENCLOSURE 13

i considerations also indicate a factor of 1000 increase in leakage under SLB conditions. In conclusion, the absence of a deterministic and empirical models for these newly observed cracks precludes the conclusions reached in the subject document. The claim that the conclusions in item S are supported by items 1-8 could be considered valid only if one ignores the available data which indicate that higher than normal leakage will occcur at SLB pressures even if the tubes do not rupture. Finally the justificaSons for any plant operation should not be based on staff pinions or published data on SCC, SCC is a semi empirical alt, in the absence of applicable databasu other routes of approaching the problem should be considered. The justification for operating with cracked tubes should be based on what procedures would the operator follow given certain primary to secondary leak and a MSLB between the containment and the MSIV. These justifications should clearly demonstrate compliance with 10CFR100. I beleive that the staff can more prcperty judge operator action than predict localized corrosion behavior, SPECIFIC COMMENTS: ltem 1. The EDM inniated grooves studies provide some measure of the ability of the tube to resist rupture given certain known wall imperfections. It bears little relation to how ODSCC form, propagate and leak in steam generator environments. Item 4 Tr.is definition of 'significant" is questionable, it makes no difference whether the cracks extend beyond the TSP if they leak at the gap. it appears that operators rely on such leakage because they lowered the leakage requirements during normal operations. Unless one can show that the TSP will cause cracks to plug and they will remain plugged under MSLB pressures the above definition may lead to confusion. The statement that ' upper bound laboratory ODSCC....

b ^ 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 The high frequency quoted,6.8x 10-4/RY contradicts the statement that

it is reasonable'

, item 9, because this frequency would result in a core melt probability of 6.8x 10-3/RY with containment bypass as discussed in the March 27 Memo, The above number is considerably higher than present safety goals. The statement that the key initiating event for SGTR is MSLB is incorrect when taken in the context of the entire document. Item 6 cont,adicts this statement. Item 7 Although this item is correct, as stated, it presents only part of the data. As already discussed, three tubes leaked at Trojan. Three tubes from other plants also leaked at MSLB pressures. Rudimentary consideration dictate that leakage increases when delta p across the wall is increased. Item 8 The lengthy discussion of uniform thinning only confuses the main issue. Tnere are several ways that the reduction load bearing capabilities of a component due to corrosion can be accounted for there is nothing special about these equations. The ASME code takes this into account. The main problem here is LOCALIZED corrosion with an UNKNOWN ATTACK RATE. Item 9 A discussion should be added of the type of new information which is required for the ' additional review' to justify subsequent operations. ATTACHMENT 1 i: Second item : Dr. instead of Mr. or just Hopenfeld The following is missing: On Sept 11,1992 J. Hopenfeld filed an addendum to the March 27,1992 concluding that ' strong coupling exists between hot leg mass flow, SG tube leakage and crack

On Sept 11,1992 ~ J. Hopenfeld filed an addendum to the March 27,1992 concluding that

strong coupling exists between hot leg mass flow, SG tube leakage and crack propagation. If confirmed, such a relation between system behavior and undetected tube defects may cause small leaks to quickly enlarge and results in a MULTIPLE TUBE RUPTURE BEFORE THE RCS IS DEPRESSURIZED BY FAILURE OF Tl4E SURGE LINE.

THE RESULTANT CONTAINMENT BYPASS WILL INCREASE T E S URCE TERM." J. H engd / P. Norian, Gm. Mazetis, W. Minne,. Beckjor .:}}

ML20126E786

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Dear Dr. Hopenfe1d:

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Background:

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Dear Mr. Pollard:

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Enclosure:

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GENERAL COMMENT

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ML20126E786

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Contact:

Enclosures:

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SUBJECT:

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Dear Dr. Hopenfe1d:

Enclosures:

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Enclosure:

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Subject:

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Background:

Conclusions:

Dear Mr. Pollard:

SUBJECT:

SUBJECT:

Enclosure:

SUBJECT:

GENERAL COMMENT

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