ML20046D392

Informs That Util Has Reviewed Draft NUREG-1477,dtd 930601, Noticed in Fr on 930702 & in Agreement W/Comments Provided by Epri/Sg Mgt Program Provided to NRC

ML20046D392
Person / Time
Site:	Farley
Issue date:	08/13/1993
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF ADMINISTRATION (ADM)
References
RTR-NUREG-1477 NUDOCS 9308190102
Download: ML20046D392 (45)

Text

s Southem N,3ciear Operatin0 company Post C"ce Box 1295 Birmingham. Alabama 3s201 I

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Te'ephone (205) 86B-5131 i

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L Southern Nuclear Operating Company Dave Morey i

Vee Prescent Farley Project the SC%them electnc Sistem AUGUST 13, 1993 Docket Nos 50-348 50-364 Rules and Review Directives Branch Division of Freedom of Information and Publication Services Office of Administration U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Joseph M. Farley Nuclear Plant Comments on Draft NUREG-1477 Gentlemen:

Southern Nuclear Operating Company has reviewed draft NUREG-1477 dated June 1, 1993 which was noticed in the Federal Register on July 2, 1993.

Southern Nuclear Operating Company is in agreement with the comments provided by the Electric Power Research Institute / Steam Generator Management Program (SGMP), which have been provided to the NRC.

If there are any questions, please advise.

Respectfully submitted, Y!

)R<w Dave Morey REM:maf SGMP1477. REM cc:

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Bethesda,MD 20814 EPRUSteam Generator Management Program Comments on Draft Subject-NUREG 1477 (Announced in Federal Register on July 2,1993)

Reference:

(1) Memorandum for T.J. Murley and E.S. Beckford, from W.T. Russell and T.P. Speis, Draft, NRC NUREG-1477, ' Voltage-Based Interim

-Plugging, Griteria for Cleana C==~no,w Tulwc. Tad..Gmup Reoart".

Availability for Public Comment, June 2,1993.

(2) " Steam Generator Degradation Specific Management (SGDSM), An induitry-im'rt.stwer" ynwaza-ty-rhe-rr-> ~~~ e~ ~~ mua.inm Project at the USNRC Offices, April 15,1993, (also reference 26 in Draft NUKCG M7M.

The purpow of this letter is to comment oztrefarance I which was announced in the Federal Registcr/Vot 58, No.126/Priday, July 2,1993 (1)/Noticos. Thace mmments are being presented by the EPRI/ Steam Generator Management Program (SCMP1).

If con

• procanted i.n thic letter deal _ with Information associated with the sub}ect 50cenents riatuenced proprietarj nwterial, thect-thew enmments must N considered prehminary and sub}ect to revision if and when the proprietary information becomes available for industry wide review and comment.

The SGMP has solicited comments on tha div-ument.from its utility members and

'thbse non-members who are-c.llowedDgespate in the CCMP on subjects dealing with safety related issues. The SGMr he reviewed these commento and i

j consolidated them in this letter.

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g Since the SGMP is working on a generic approach of steam generator degradation management encompassing alternate tube repair criteria (see reference 2), this letter presents comments motivated by (Draft) NUREG 1477s assumptions, analysis, requirements, recommendations and conclusions as they compare to this generic effort.

Additionally, the subject of (Draft) NUREG 1477 is interim plugging criteria for steam generator tubes. Utilities have docketed requests for interim plugging criteda and have had considerable communi :ation with the NRC on this issue. The SGMP has not been involved in these discussions. Therefore, the majority of this letter's comments are divorced from these communications and any agreements reached between these utilities and the NRC on the subject of Interim plugging criteria, although the SGMP recognizes that some of its comments have bearing on IPC related issues and they should also be considered in that context. Additionally, to help reduce the number of submitted utility comment letters to the NRC, attached are comments provided by utilities which deal with IPC issues. These utilities' comments are in addition to those comments provided by the SGMP and intended for the convenience of the regulatory review process associated with (Draft) NUREG 1477.

Comments on (Draft) NUREG 1477 are proportioned into four attachments. identifies areas of agreement and attachment 211sts areas of concern which are titled " imperatives". Imperatives are areas requiring accommodation by NRC and industry to allow successful regulatory action on the industry's steam generator degradation specific management program (SGDSM) initiative. lists suggested improvements to the draft document and Attachment 4 provides utility comments mentioned earlier.

Sincerely, David A. Steininger Technical Advisor Steam Generator Project Office Approved for Distribution by:

Robert E. Smith Chairman, Executive Group Utility Steering Committee Steam Generator 6trategic Management Program i

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j Significant Areas of Agreement with (Draft) ?mREG 1477 (1) As noted on page 3-351, "the regression fit of the burst pressure data as a function of voltage is valid."

The utility industry believes NRC recognition that a NDE measured parameter such as voltage can be related to structural margin through standard regression techniques supports the generic approach of SGDSM. SGDSM maintains that an industry data base is required to correlate the rupture strength of degraded tubes with a NDE measured parameter.

(2) As noted on page 3-36,"the method for estimating the distnbution of voltage changes during the next operating cycle is appropriate for purposes of accounting for crack growth during the cycle in assessing burst and leakage integrity, assuming that crack growth and voltage growth rates are within the bounds of previous experience." This methodology is described on page 3-12 of the document and involves using plant-specific average voltage change observed during previous operating cycles.

Although the SGMP is in general agreement with this position, it should be noted that specific plants may not have sufficiently developed historical degradation growth data to apply this methodology. An alternate approach should be the use of an appropriate growth rate value obtained from data gathered from other operating plants where similar degradation has been observed and for which acmptable degradation growth rate data can be extracted. A recommended growth rate will be provided in the SGDSM approach for use where plant specific degradation growth rate information is unavailable.

(3) Page 3-27 indicated that the " proposed 150 gpd leak rate limit when combined with an effective leak rate monitoring program will provide reasonable assurance that should a significant leak be experienced in service, it will be detected and the plant shut down in a timely manner before rupture occurs and with no undue risk to public health or safety."

It is acknowledged that this is an ac:eptable " defense-in-depth" strategy. It is suggested that an effective leak rate monitoring technique use an ac ptable detection threshold to avoid a spurious plant shutdown unrelated to significant steam generater leakage.

(4) Page 3-37 indicated that " implementation of the Westinghouse (or s~mtilar) guidelines for NDE data acquisition and analysis is essential for ensuring the Page numbers refer to (Draft) NUREG 1477, dated lune 1,1993 (see reference 1).

1 C00*390d 9G: PI CS. 21 900

4 reliable detection of low voltage signals and for ensuring that variabilities in the voltage responses and voltage measurements are minim!wl."

It should be recognized that the generic SGDSM initiative will establish an industry standard of data acquisition and analysis for each degradation that has associated with it an alternate repair criteria which satisfies the protocol established by SGDSM. 'Ihis initiative will satisfy the NRC request that "similar" guideline implementation is essential for reasons noted in the NUREG comment.

(5) Page 4-6 indicates that changes to the reactor coolant specific activity technical spedfications would permit a licensee to operate with predictions of primary to secondary tenhge during acddent conditions higher than that which would be allowed using Standard Technical Specifications. This would still ensure the same level of protection to the public.

This operational requirement is intended in part to minimize both simple and compound iodine spiking. Although the SGMP is in general agreement with this position, it should be recognized, as indicated in NUREG-0933, "A Prioritization of Generic Safety Issues," July 1991, item B-65, "that iodine spiking is a significant effect in only non-core melt accident consequences, which are not major contributors to nudear plant risk." It is suggested that NUREG 1477 reaffirm this position of spiking not being major contributors to nuclear plant risk for plants operating under steam generator alternate tube repair criteria.

(6) Page 4-39 indicates that the NRC concluded that because calculations using leak rates which are several orders of magnitude above the expected leak rates from a plant using IPC, and because the fraction of the initial RWST inventory calculated to remain was so large, no plant spedfic calculations are required to demonstrate adequate RWST inventory.

The SGMP believes that SGDSM will allow higher than IPC allowable leak rates, but lower than assumed values used in these NRC calculations for the reason of adequate RWST inventory. Analysis by the SGMP was previously presented to the NRC (see reference 2) substantiating this conclusion.

Therefore, the NRC conclusion of no need for plant specific calculations should remain valid for alternate repair criteria under SGDSM and should be so indicated in the NUREG.

1 P00*390d LG: P1 EB. 21 900

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Imperatives

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(1) Pages 3-14 and 3-17 identify a requirement to pull tubes to meet specded objectives. These objectives are:

1. Enhance and vahdate the empirical burst and leakage correlations,

2. Confirm that axial ODSCC continues to be the dominant degradation mecharism at the TSP intersections, and

3. Provide additional data for assessing the reliability of the inspection methods (e.g., provide verification of "non detectable indication" as indicated by NDE interrogation of the degradation).

The objectives are intended to be met by pulling a minimum of nine tube, tube support plate (TSP) interactions exhibiting voltages ranging from less than 1

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volt to the maximum observed per plant outage.

Within the framework of the SGDSM approach, the above note.1 objectives and method of attaining these objectives are not completely valid ot warranted.

Limited tube pulls across the industry (not on individual plant-by-plant per i

outage basis) should be performed to verify degradation's burst and leak correlations maintained in the industry's SGDSM program. The number of tube pulls over time should not be open ended. Engineering criteria will be established identifying tube pull needs. Obtaining useful tube pull data exhibiting a high benefit / cost ratio should be the objective. Practical tube pull requirements are required. For example, desired tube intersections for ODSCC examination should be taken from the lower bundle elevation to maximize a successful tube pull effort. Also verification of cnrrelations should be done for the range of independent variables which have the largest safety significance and/or exhibits unacceptable dependent variable uncertainty. Areas of safety significance are located at the safety citeria's limit value; not at values far j

removed from the limit.

Plants that haven't pulled tubes establishing degradation morphology may need to do so to a ilmited but practical extent. But sufficient NDE data may be available from the plant to establish the degradation medariism by comparing fully characterized NDE signatures available for example, in.an industry SGDSM data base or acceptable alternative. To date approximately 200 tube i

intersections from field pulled tubes exldbit consistent eddy current signatures for ODSCC at tube supports.

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~. s Memo To:

David Steininger From:

Rick Mullin g

Date:

August 11,1993

Subject:

Comments on EPRI Draft Letter on Draft NUREG-1477 I have reviewed the draft EPRI letter forwarded by your letter dated August 9,1993.

My comments are provided below.

1.

The cover letter indicates that Attachment 1,2, and 3 are comments on the SGMP developed DSM approach. While that is true, several of the comments apply to a

the use of an IPC, e.g.,9 tube pulls per outage. To say that these only relate to DSM or that Attachment 4 contains the IPC issues is incorrect.

I recommend that a statement be added to the following issues that the issue pertains both to the DSM and IPC:

Issue (1), Page 5 - Tube pulls Issue (3), Page 6 - POD t

issue (8), Page 11 - Probe size and probe variability Issue (2), Page 12 - In-line probe standard measurements for each tube issue (6), Page 13 - POD Issue (9), Page 14 - 100 TSP inspection with RPC 2.

In that issue (8) on page 11 and issue (4) on page 18 are the same issue, I would combine the two and indicate it applies to DSM and IPC.

3.

Issue (2) on page 12 requires testing probe wear "using a delivery method that would reproduce the vertical position of the probe in a smooth standard / conduit environment." l'm not sure why checking wear requires a vertical position. If it meets a 15% wear limit lying horizontally, I believe it would meet the limit vertically.

4.

Although I do no have a problem with identifying SNC as the author of the f comments, I believe the way it is presented implies that SNC was the only utility with the comment. On the basis of there being no objections voiced at the July 22-23 meeting to the issues I raised, I believe all utilities interested in an iPC would support those comments. That would include AEP, WPS, Duke, and Duquesne, in addition to SNC.

5.

No mention is made of the 20% analyst cutoff which Westinghouse has used in the past. This cutoff is based on the requirement to resolve voltage differences if the difference is > 20% and over i volt. To ignore this approach seems like leaving a benefit out of the ARC and we may be bit by the " tails of the distribution."

6.

No mention is made of the requirement to review RPC mix channels. I thought the result of the Chicago meeting was that the RPC mix channel was an option to the analyst; however, no requirement to review the RPC mix channel for every indication was required.

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Comments on EPRI Draft Letter on Draft NUREG-1477 08/11/93Paga 2

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The draft NUREG requires the use of 100% through-wa!! holes for calibration. I thought 'he recommendation of the NDE subcommittee at the Chicago meeting was that no benefit was gained by using through-wall holes. Therefore, we would keep the current cal standard unless some benefit was shown by using the 100% holes.

8.

On the basis of Information Notice 93-52, the letteris addressed to the wrong location. The address should be:

U.S. Nuclear Regulatory Commission ATTN.: Document Control Desk Washington, DC 20555 A marked up copy of the draft letter containing minor comments is attached.

If you want to discuss these comments, please call me at (205)868-5502.

CC: B.D. McKinney, Ei.E Moore, D.N. Morey, J.D. Woodard

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Subject:

EPRI/ Steam Generator Management Program Comments on Draft l

NUREG 1477 (Announced in Federal Register on July 2,1993) d

Reference:

'1).\\1emorandum.'or T.J. Aiurley and E.5. Beckford, from W.T. Russell and T.P. 5:eis. Drzit, NRC NUREG-1477. " Voltage-Based Interim Plugging Criteria,;or Steam Generator Tubes-Task Group Repcrt"-

1 Availability for Public Comment June 2,1993.

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" Steam Generator Degradation Spectfic Afanagement (SGDSM). An l

Industry i=:itiative." presented by the EPRI Steam Generator Reliability Project at ::e USNRC Offices. April 15.1993. (also reference 26 in Draft i

1 NUREG ~ '77L 1

l The curpose of this let:er is to comment c. reference I which was announced in the j

Federal Reg:sterz Vol. 55. No. ".25/ Friday..luly 2.1993 i;)/ Notices. These comments are being presen:ed by -he EPRl/ Steam Generator Management Program (SGMP1).

If ccmmen:s presented :n tais :etter deal with informa: ion associated with :he i

subject document s referenced proprietary material, then these comments must be '

considered preliminary and subject to rev:sion if and when the proprietary information becomes a cailable for industr wide review and comment.

The 5GMP '.as solicited comrnents on the document from its utility members and those non-members who are allowed to participate in :he SGMP on subjects dealing

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with safetv related issues. The SGMP has reviewed these comments and consolidated them in :his letter.

Since the SGMP is working on a generic a:proach of s:eam generator degradation managemer: encompassing a;:ernate tube repair criter:a (see reference 2), *his ietter 1

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Presents comments motivated by (Draft) NUREG 1477s assumptions, analysis, requirements, recommendations and conclusions as they compare to this genene effort.

t Additionally, the subject of (Draft) NUREG 1477 is interim plug ng criteria for steam generator tubes. Utilities have docketed requests forde.c.aplugging criteria and have had considerable communication with the NRC on this issue. The SGMP has not been involved in these discussions. Therefore, the majority of this 4

letter s comments are divorced from these communications and any agreements reached between these utilities and the NRC on the subject of interim aPem:t*

t plugging criteria. But to help reduce the number of submitted utility comment letters, attached are f pantMtility comments N'!e d b,

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[lPC issues. Inese are m aIdition to the SGMP comm f NUREG 1477.

_D_" convenience of the regulatory review process associated with (Dra t)

Comments on (Draft) NUREG 1477 are proportioned into four attachinents. identifies areas of agreement and attachment 2 lists areas of concern which are titled " imperatives Imperatives are areas requiring accommodation by NRC and industry to allow successful regulatory action on the industrv's steam generator degradation specific management program (SGDSM) initiative. lists suggested improvements to the draft document and[tachmen Provides individual utility comments mentioned earlier.

Sincerely, Robert E. Smith Chairman, Executive Group Utihn steermg Committee Steam Generator Strategic Management Program i

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i Significant Areas of Agreement with (Draft) NUREG 1477 1)

As noted on page 3 332, "the regression fit of the burst pressure data as a function of voltage is valid."

The utility industry believes NRC recognition that a NDE measured parameter such as voltage can be related to structural margin through standard regression techniques supports the generic approach of SGDSM. SGDSM maintains that an experimental data base is required to correlate the rupture strength of degraded tubes with a NDE measured parameter.

As noted on page 3-36,"the method for estimating the distribution of voltage changes during the next operating cycle is appropriate for purposes of accounting for crack growth during the cycle in assessing burst and.eakage mtegrity, assuming that crack growth and voltage growth rates are within the bounds of previous experience." This methodology is described on page 3-12 of the document and involves using plant-specific average voltage change observed during previous operating cycles.

l Although the SGMP is in general agreement with this position,it should be i

i noted that specific plants may not have sufficient or appropriate historical degradation growth data to apply this methodology. An alternate approach t

'lue ob ained frem data should be the use of an appropriate growth ra+

gathered from other operatmg plants wher identical degradation has been g observed and for which acceptable degradation growth rate data can be extracted. A recommended growth rate will be provided in the SGDSM approach for use where plant specific degradation growth rate inicrmation is unavailable.

Page 3-27 indicated that the " proposed 150 gpd leak rate limit whe. combined with an effective leak rate monitoring program will provide reasonable assurance that should a significant !eak be experienced in service. :: will be detected and the plant shut down in a timely manner before rupture occurs and with no undue risk to public health or safety."

It is acknowledged that this is an acceptable " defense-in-depth" stra:egy.

(4)

Page 3-37 indicated that " implementation of the Westinghouse (er similar) i guidelines for NDE data acquisition and analysis is essential for ensuring the reliable detection of low voltage signals and for ensuring that var: abilities m the voltage responses and voltage measurements are minimized.

Page numbers refer to (Drait) NUREG 1477 (see reference '.

2 gj he

4 It should be recognized that the generic SGDSM initiative will establish an industry standard of data acquisition and analysis for each degradation that has associated with it an alternate repair criteria which satisfies the protocol established by SGDSM.

5)

Page 4-6 indicates that changes to the reactor coolant specific activity technical specifications would permit a licensee to operate with predictions of primary to secondary leakage during accident conditions higher than that which would be allowed using Standard Technical Specifications. This would still ensure the same level of protection to the public.

This operational requirement is intended in part to minimize both simple and compound iodine spiking. Although the SGMP is in general agreement with this position, it should be recognized, as indicated in NUREG-0933, ~A Prioritization of Generic Safety Issues," July 1991, item B-65, "that iodine spiking is a sigruficant effect in only non-core melt accident consequences, which are not major contributors to nuclear plant risk." It is suggested that NUREG 1477 reaffirm this position of spiking not being major contributors to nuclear plant risk for plants operating under steam generator alternate tube repair criteria.

) Page 4-39 indicates that the NRC concluded that because calculations using leak rates which are several orders of magnitude above the expected leak rates from a plant using IPC, and because the fraction of the initial RWST inventory calculated to remain was so large, no plant specific calculations are required to demonstrate adequate RWST inventorv.

The SGMP believes that SGDSM will allow higher than IPC allowable leak rates, but iciwer than assumed values used in these NRC calculations for the reason of adequate RWST inventory. Analysis by the SGMP was previously presented to the NRC (see reference 2) substantiating this conclusion.

Therefore, the NRC conclusion of no need for plant specific calculations should remain valid for alternate repair criteria under SGDSM and should be so indicated in the NUREG.

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3 Imperatives j

(1)

Pages 3-14 and 3-17 identify a requirement to pull tubes to meet specified

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objectives. These objectives are:

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1. Enhance and validate the empirical burst and leakage correlations, u*

2. Confirm that axial ODSCC continues to be the dominant degradation p

mechanism at the TSP intersections, and i

3. Provide additional data for assessing the reliability of the inspection methods (e.g., provide verification of "non detectable indication" as indicated by NDE interrogation of the degradation)-

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The objectives are intended to be met by pulling a minimum of nine tube TSP

nteracnons exhibiting voltages ranging from less than 1 volt to the maximum-abserved per plant outage.

M TP4 Within the framework of the SGDSh proacghe above noted objectives and.

method of attaining these objectives are not completely valid or warranted.

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Limited tube pulls across the industry (not on individual plant-by-plant per

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outage basis) should be performed to verify degradation's burst and leak l

g correlanons maintained in the industry's SGDSM program. The number of p

tube puils over time should not be open ended. Engineering criteria will re

[h established identifying tube pull needs. Obtaining useful tube pull data

-:xhibit=g a high benefit / cost ratio should be the objective. Practical tube pull 3

requiremems are required. For exampigpesired tube intersections for ODECC -

examination should be taken from the Tower bundle elevation to maximize a uccessful tube pull effort. Also verification of correlations should be dor.e for the range of independent variables which have the largest safety. significance and/or exhibits unacceptable dependent variable uncertainty. Areas of safety i

significance is located at the safety criteria's limit value; not at values far.

removed from the limit.

Plants that haven t pulled tubes establishing degradation morphology may need to do so to a limited but practical extent. But sufficient NDE data may be available from the plant to establish the degradation mechanism by comparing to fully characterized NDE signatures availabgr example,in an industrv EGDSM data base or acceptable alternative. To date approximately 200 tube

.ntersecnons from field rulled tubes exhibit consistent eddy current signa-ures

'I for ODSCC at tube supports.

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Cost / benefit associated with pulling tubes to assess the adequacy of ISI techniaues used in SGDSM is unacceptably high and may not enhance safety.

ISI procedures should be adequately evaluated and qualified by reliability assessment programs before plant application. Such a program is exhibited by SCDSM with its requirement to implement the EPRI PWR Steam Generator Examination Guidelines document which contain protocol requirements for data analyst and ISI equipment qualification. It is emphasized that this reliability assessment program is in addition to and does not substitute for a plant's own site specific ISI qualification program which can be satisfactory.

An example of an unacceptable cost / benefit ratio for a tube pull requirement is that associated with verifying "no detectable defect (NDD)." If an acceptable ISI reliability assessment program is mandated under SGDSM, justification requiring verification of in-field NDD calls can not be substantiated.

Additionally, an infinite amount of data is required to prove a " negative", that is, to verify NDD. resulting in an infinite cost / benefit ratio.

2)

Page 3-20 indicated a lack of a proven correlation between leak rate and voltage and page 3-21 recommends a leak rate relation which is not a function of voltage.

U Standard statistical tests should be used establishing the very low probability of

' /y a zero slope linear relationship (i.e., no functional relationship between leak I

rate and voltage).

Althouzh one could arrive at a relativelv low correlation coefficient for the subject iiata, this does not necessarily mean a correlation does not exist. Data scatter causing a relatively low correlation coefficient can be directly related to some physical phenomenen. In the case of ODSCC at tube suppor:

mterse:nonghe phenomenon is norphology variations ii.e., number of ligame-s) associa:ed with the cracking. But uncertainties associa:ed with the regress:en model and supportmg data can be applied to account for this data scatter.

Cogen: statistical arguments conceming these issues were presented at a previous utility industrv NRC meeting (see reference 2).

Similar statistical techniques and not arbitrary qualitative judgments should be used es:ablishing correlations for other degradation mechanisms. Correlations under EGDSM should not be precluded by NUREG 1477 assumptions, such as the dis:nbution of leak ra:e not depending on voltage, unless adequately suppor:ed by standard sta:istical arguments.

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Page.M3 provides a probability of detection (POD) value oi o.6 for ODSCC at tube su: port p!ates. As noted th;s value was the average POD for 20-percent

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through-wall to 100-percent through wall cracks. Page 3-29 recommends consideration of multiple ISI techniques as a means for enhancing POD.

It is SGMP's understanding this average POD value was based on the resuits of the Surry project as summarized in NUREG/CR-5117, PNL-6446, S:eam Generator Tube Integrity Program / Steam Generator Group Project, Final Project Summarv Report, May 1990. The SGMP believes this is an inappropriate use of the data. The Surry data was taken without following.

O[f pe,p. ( specific ISI enmininn tmidelines. This is not t proposed SGDSM prograrrf Specific inspection guidelines will be followed which have supporting POD information considerably higher than the 0.6 value recommended in the draft NUREG document. Justification for a higher value is provided in the SGDSM program. Specifically, for ODSCC at tube support plate intersections, the applicable value vs. voltage is greater than 95%

at a 90% confidence for the 1.0 to 3.0 voltage range (the range of most interest).

A lower probability value will be allowed for lower voltages that are not j

structurally significant. Additionally, structural margin is not compromised even when growth rate is evaluated with these initiallower voltage values for the next operational cycle of the plant.

Although the industrv provided POD is acceptable, the actual value is higher because of independent dual analyst interrogation of the inspection data as required under SGDSM. For example,if the probability of detecting degradation of a certain extent is 0.9, then the probability of a single miss (POMi is 0.1, the probability of two misses is 0.1 x 0.1 = 0.01, and so the POD for two inspections is 1.0 - 0.01 = 0.99, assuming independence between inspections.

Credit for dual, independent analyst inspection of the data should be allowed.

Other ISI techniques to develop an enhanced effective POD appears i

unnecessarv. Since the POD, for example that associated with eddy curren:

bobbin coil probes used to interrogate ODSCC at tube support plates, is acceptable, use of complementary devices is superfluous and only results m additional occupational radiation exposure and O & M expense.

/(4)

Page 3-17 indicates that all input parameters for setting the voltage limit shouic be evaluated at consistent 95-percent confidence values. Additionally,it is stated,"the task group concluded that the deterministic methods used to establish the IPC, generally, satisfy the requirements of Regulatory Guide 1221.

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even when consistent 95-percent confidence values of the input parameters (i.e., voltage variability, voltage growth, and burst pressure) are used.

It is inappropriate to require high probability values of for example. 95% 9Pb on all input values to ensure that certain limits specified in Regulatory Guide 1.121 such as 3 times normal operating pressure for tube burst, no: be exceeded.

This suggested methodology produces an6necessarili3high probabilin Hat a safety limit will not be exceeded.

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It should be noted that the margin to burst factor of 3 specified in Regulatorv l

Guide 1.121 is an engineering safety factor intended to bound uncertainties which have not been quantified. This conservative, deterministic methodology of setting limits is analogous to the ASME code requirement of fatigue where fatigue test data is arbitrarily modified by factors of 20 on stress or 2 on cycles, whichever is limiting, to produce a conservative design fatigue curve.

The 95%/95% or acceptable probability value on for example, tube burst, derived with uncertainties accounted using appropriate values, should be specified at actual normal operating pressure or main steam-line break l

conditions.

If an engineering safety factor of 3 is mandated, per Regulatory Guide 1.121.

then credit should be allowed for example. for the constraint provided by the tube support plate preventing tube burst under normal operating pressure using an appropriate definition of " burst".

i If the above described probabilistic methodology, demonstrating a "high level

{

of probability" exists that safety is maintained, is not allowed, then the ASME code requirement of 1.4 times main steam line break pressure conditions as i

dictated by Regulatorv Guide 1.121 should be retained. It should be met l

without additional conservative bounding confidence limits on the test data.

/(5)

Page 5-1 indicates that "the task group concluded that evaluation of offsite and control room doses can be used in conjunction with the recommended me: hod for predicting primarv-to-secondarv leakage rates to provide adequate assurance that offsite doses would not exceed establishe

'mits (i.e.,10CFR Part 100 and GDC 19) under postulated accident condition. /

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The SGMP be eves that an acceptable probability value be specified for the 6

safety criteri= of interest only, which in the present case is the radiological dose

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limit at th A as dictated by 10CFR100. An appropriate probability value should be established for meeting less that 10CFR100 limits.

It is unreasonable to dictate a high probability value (e.g.,95%/95%) on all variables to meet a small fraction of or 10% of 10CFR100 limits. Quantification of uncertainties should remove the need for applying engineering saferv factors i

such as one tenth of 10CFR100 limits. Such safety factors were originally J

l incorporated to compensate for nonquannfiable uncertainties. Additionaiiv, not allinput variable uncertainties need be specified at high probability values j

in setting an acceptable high level of probability that 10CFR100 limits are not exceeded. A modern approach to such analysis is the so-called bes:-estima:e BE) methodology, in which one predic:s realistic results with explicit wnsideranon of and accounting for uncertainties that are introduced at even i

1 9 l j

i stage of the calculation. One would then need to demonstrate to a high level of j

probability that safety limits of 10CFR100 are met. For example, such an approach in the acceptance criteria for emergency core cooling systems (ECCs) in light water reactors is now allowed per a NRC rule change in 1988.

l Unfortunately, in this case a high but unspecified probability (although the Standard Review Plan repeatedly calls for a 95% probability at a 95% confidence Icvel that certain limits should not be exceeded) is dictated for limits (e.g., peak s

I clad temperature and minimum DNBR) which themselves have incorporated conservative, possibly 2 sigma tolerance bounds on safety limit distributions.

In any event, the SGMP suggests that similar analysis, with appropriate modification defining acceptable safety margin, be allowed to establish the allowable steam generator primarv to secondary leakage under faulted load conditions.

i The SGMP favorably recognizes that somewhat similar analysis is suggested for investigation on page 5-2 of (Draft) NUREG 1477. But the SGMP recommends analysis more closely aligned with the 19SS ECCS methodology with appropriate improvements in actually establishing quantifiable, probabilistically defined safety margin under SGDSM.

(6)

Page 4-2 indicates that in order to place the Standard Review Plan (SRP) radiological dose analysis in perspective, a realistic analysis was performed.

Specifically, it is stated that, "this analysis calculated offsite doses using assumptions and data closer to the expected conditions, that is, without as p,

much margin as is included in the licensing (SRP) calculations?

The above statement does not appear to be consistent with the so-called rea:istic analysis presented in the draft NUREG. Specifically, the realistic analysis uses a

ompound iodine spike for the faulted load event of a main steam line break xhich is 500 times the release rate from fuel which gives a coolant activity ci 11pCiigm.

The SGMP agrees that 0.1pCi/gm is a more realistic value of the equilibrium 1131 coolant activity. But the compound spiking value of 500 is not. NRC contractor data indicates that the compound iodine spike is one to perhaps two 3

orders of magnitude less than 500. This information was previously presented by the SGMP to the NRC on April 15,1993 (see reference 2).

It should be noted that NUREG 0844,"NRC Integrated Program for the 4

8 Resolution of Unresolved Safety issues A-3, A-4. and A-5 Regarding Steam Genera:or Tube Integrity, September 1955. "srecifies a best estimate value of 10-i 20% oi the 500 value.

Finallv Supplement 13 to NUREG-0933. 'A Prioritization of Generic Safety

ssues. December 1991, Item e7.5.1 refers to a reassessment of radiological

10 consequences associated with SGTR accidents and reevaluation of the applicability of the assumptions in the Standard Review Plan, Section 15.6.3.

De conclusion listed in this item indicates in part that the, ~ DST agreed that a

~best estimate" analysis modeled after plant experiences, like Ginna. could be beneficial in more realistically determining the nsk and conservatisms inherent in the current SRP requirements." It further goes on to say, if this limited scope comparison of the SRP model with best estimate analysis is followed, this issue could be considered as an imp ovement tu current licensing positions (a licensing issue)." Since this SRP reevaluation would 131 spike of 500, the SGMP obviously include a reassessment of the compound 1 recommends that th NRC initiate reassessment of the SRP, Section 15.6.3, as MEommended b, AE in item 67.5.1 of NUREG-0933 (Supplement 13). Such a l

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reassessment shouso use NRC contractor data gathered to support Generic Issue 67.5.1. Such a reassessment and its conclusions would also apply as it relates to 131 spiking models used in analysis of the main steam line break event with 1

concurrent steam generator leakage.

(7)

Page 3-11 provides discussion on a NRC model which assesses crack growth, fracture behavior, leakage rates, and is a traditional mechanics based approach.

It concludes that sensitivity studies using the model indicated the difficulty in reliability modeling complex phenomena like IGA /ODSCC. It goes on to further state,"the model can use, as input data, the flaw lengths and depths (j

derived from inservice inspection results of the flaws in tubes that are left in l

service,' and " improvements in the ECT technology should enable the use of l

1ength - and depth - based criteria for plugging and repairing, thereby providing

(

the reauireo link to inservice inspection data for the NRC modelf Additionally page 3-39 suggests that,"over the long term, the use of length -

and/or depth-based criteria for plugging and repair would be a preferable i

a?Froach, and the mdusry should pursue improved nondestructive testing technoiogy to support such criteria.

I Re SGMP takes issue w th the suggestion that the above described approach j

can become a viable option to interrogate the structural integrity ei steam generator tubing experiencing ODSCC at tube support intersections. The SGMP I

believes that pursing this approach for this type of degradation would be a wasteful utilization of finite resources with a low probability of success in the foreseeable future. But of course advances in NDE technology is desirable and being pursued by the unlity industry. The questions which arise on the present subject are issues of practicality and timeliness.

As noted on page 3-29 of the (Draft) NUREG,"the most desirable situation is to define a plugging limit that (1) is reliability enforceable and (2) provides reasonable assurance that tubes that satisfy the limit will continue to have i

structural and leak-tight :ntegrity/ The SGMP believes that a voltage based repair limit for ODSCC 2: tube supports satisfies this desired objec :ve of the NRC.

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NUREG 1477 clearly indicates, as noted above, "the difficulty in reliability modeling complex phenomena like IGA /ODSCC." In addition to -his problem.

-l state-of-the-art NDE technology available today and in the foreseeable future

.[

should not and can not provide the necessary, highly accurate inputs to l

compensate for the inadequacies of a recognized, unreliable mechanistic model j

of IGA /ODSCC.

I The SGMP believes that such an approach will ultimately result in more uncertainty in its application than a voltage based criteria.

q (8) Page 3-33 indicates that IPC is not applicable at any TSP intersections where the l

intersection cannot be inspected with the full-sized bobbin probe specified in j

the Westinghouse guidelines. Additionally, page 3-31 indicates that the task l

group recommends that bobbin coil probe variability be limited to 5 percent.

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it must be recognized that smaller type probes are needed to interrogate 6

previously sleeved tubes. Such probes should be allowed as long as they meet

,h acceptable performance requirements (i.e., adequately qualified) wnh required l

calibration. The major concern with the use of smaller probes is ::Mr:bW i

probe centering within the tube. This has not been a problem to date.

j i

j Industry lacks capability to meet a probe variability specification of 5 percent. In any event, this specification is unnecessarv. Realistic probe to probe variability

)l is already taken into account in the correlations database because carious probes where used in its generation. -Additionally, allowance for a maximum of 15% probe wear induced voltage change encompasses this voltage variabilitc j

concern.

i Finally. under the SGDSM initiative equipment cualification recu:remen-c will establish acceptable performance standards on such probes resulting :n weil defined voltage variability uncertainties which can be appropnately l

handled in best estimate analysis establishing well defined safety margins.

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Suggested Improvements (1) Page 2-1 states that the NRC evaluates each applicant's SG design, water chemistry, and inspection program in accordance with the regulaton guidance in the Standard Review Plan (SRP) NUREG-0800 before issuing a license.

It is SGMP's understanding that the SRP was first issued in 1975 (NUREG-75/087). A number of PWR plants were issued their construction permit after 1975, but were not licensed to the SRP. It is suggested that the above noted t

paragraph be clarified to recognize this fact. Additionally, appropriate discussion should be added explaining why plants not licensed to the SRP must now be evaluated with respect to it for application of alternate steam generator tube repair criteria that satisfy safety margins dictated by Regulatorv Guide 1.121 and ASME code requirements.

l V(2)

Page D-1 recommends that a probe wear measurement be made automancally on an in-line standard each time a tube is inspected.

0

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The SGMP believes it is not practical, efficient, or necessary to check the probe h

wear after each tube has been inspected. The guide tube standard is not capable 1

i reproducing the same path for probe travel after each inspection. The guide h6 tube standard also creates abnormal probe movement due to missed alignment

/g/g h) wear at the beginnine and end of a calibration periodfE with the tube end. A more reasonable approach may be to inspect for probe pt

,M that would reproduce the vertical posit:en of the probe in a smooth}

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standard / conduit environment.fProbe wear uncertainty is taken into account establishing the repair limit. This in combination with.e iodic and accurate probe wear measu'rementst-W cf :-pcc. Mcr."rd ould be accep able.

/j If the probe is found to be out-of-spec.. re-examination of previously interrogated tubes with a new probe should be performed.

(3)

Page 3-31 stipulates that noise criteria should be incorporated that would require a certain specified noise level not be exceeded, consistent with the cbjectives of the inspection.

,av It should be noted that a signal / noise (S/N) issue is a generic one irrespective (V

of alternate tube repair criteria. It exists under a plant's present steam generator repair criteria which is 40% depth based. It is difficult to set objective noise critergspecially if required to propornon limits per noise inducing mechanism (e.g., tube noise, electrical noise, etc.). Present piant inspecten procedures account forj$/h] issues and should be followed during application

f alternate tube repair critena. The SGDSM process will provide guidance in 4

-his area. The drait NUREG should be modified b:. removal of tFis noise srecificat:en and reflect present plant procedures for cealine vith S/N issues.

13 f

4) Page 3-16 and 3-36 indicate that data outliers should not be removed from the data set from which the alternate tube repair correlations are developed.

O The SGMP agrees that data outliers should not be arbitrarily removed. Data points should be removed if they are clearly associated with a so called bad test g

3,y or inappropriate test specimen (e.g., damaged by other factors). Additionally, data should be removed if they are associated with degradation morphologies which do not represent the damage mechanism for which the alternate tube repair criteria is developed. This has been done to the SGDSM data set only l

when data removal has a conservative effect on the correlation line. Outlier removal should be allowed under these circumstances.

5)

Page 3 23 discusses the need to test other functional forms of the probability of I

leakage curve.

The SGMP believes this is a reasonable request for the SGDSM approach. But it

[g should be noted : hat SGDSM incorporates an uncertainty distribution with the v

P5 probability of leakage (POL) curve. This compensates for incomplete knowledge of the functional POL relationship. Additionally, as in any data-

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fitting endeave,rghe adecuacy of the model should be examined using standard statistical techrtf6ues. It is suggested that the discussion on page 3-23 reflect this approach for future application of SGDSM.

Page L28 recomr ends that an average POD value be used to determine the l

number of undetected flaws for use m the leakage calculation.

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SGMP believes that the POD in the range of interest is sufficiently high under SGDSM and that mcorocratint so cal!ed undetec:ed indications in the leakage

'y calculation is unwarran:ed. For the coltage range exhibiting low POD values, i

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such indications have been shown to be structurally insignificant.

l Addit:onally, these indications are r.ct expected to grow over the subsequent

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operating cycle to an extent which causes a leakage or burst concern. It is l

recommended that the ciraft NUREG document be modified to support this j

conclusion.

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Page 3-36. fourth bullet appears to imply that probabilistic analyses on a per outage basis should be used to evaluate the conditional probability of tube burst

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during a postula:ed MSLB.

U The SGMP recornmends that any such analysis on an outage basis be avoided.

in this regard, the gener:c SGDSM intiative will provide a methodology that g

avoids this plan: rutage requiremem of a probabilistic calculation ensuring It is margm to burst i:r the f aulted load. main steam line break event.

recommended tha: the craft NCREG be modified to avoid requiring a probabilistic calculation en a rer outage basis if an alternate method is found

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acceptable as presented in the generic SGDSM initiative. Avoiding involved probabilistic analysis on an outage to outage basis is consistent with communicated NRC's desires in previous SGMP/NRC annual review l

meetmgs.

8)

Page 3-19 raises a concern about very large amplitude indications that have recently been identified in the field.

()

The SGMP recommends that the NUREG be rewritten to put these incidents in the proper perspective. High growth rate degradation is not uniquely Mg associated with alternate tube repair criteria. Irrespcctive of the repair criteria,

,0 whether it is 40% depth, crack length, or voltage based, high degradation t

growth must be adequately handled to maintain safety margin as it has been done in the past. He SGMP believes that this process can be enhanced through the SGDSM initiative with its requirement of more extensive, engineering based inspection criteria allowing early identification of steam generator tube degradation.

[9)

Page 3-33 requires that a minimum of 100 TSP intersections below 1 volt be inspected with RPC, regardless of bobbin voltage amplitude.

He SGMP believes that a voltage based repair criteria for ODSCC at tube support plates, such as being proposed under SGDSM, justifies leaving as a minimum indications less that I volt as indicated by bobbin coil in service.

q h p The SGMP can not identiiv any reasonable justification requiring g/ g supplementary RPC inspections of these indications which clearly do not I

compromise safety margm. Such interrogation increases O & M cost with little if any incremental increase in safety.

[10) Page 3-33 presents discussion that implies that the industry has not developed an acceptable POD for RPC inspection of ODSCC at tube support plates.

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Although under SGDSM supplementary RPC requirements is not as 3g

,b4 extensively defined as it has become during the IPC initiative, the development I

of an acceptable POD for RPC interrogation of ODSCC at tube support plates will be an objective of the utility industry in support of SGDSM.

11) Page 4-38 indicates that the staff intends to examine in detail the effect of l

primarv-to-secondary steam generator leakage greater than allowable under IPC l

under severe accident conditions.

The SGMP believes that a severe accident is an extremely low probability event.

not part of the plant's design basis, and should not be considered in the g

regulatory review process dealing with SGDSM. Severe accidents is a cornplex issue with system wide implications and therefore SGDSM should not be its focal point of resolution. It should be resolved in severe accident space and not I

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5-in licensing space associated with SGDSM. In spite of this recomrnendation, the SGMP did investigate the effect of significant primary to secondary steam generator leakage on the course of a severe accident. Results were presented to the NRC during an all day meeting at NRCs White Flint facility on December 22,1992. The SGMP conclusion was that SGDSM's anticipated, allowable steam generator. leakage would not alter the conclusion reached in NUREG 1150 for the TMLB' severe accident sequence. NUREG 1477 should be rewr tten to reflect these findings.

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1 Individual Utility IPC Comments 1.

Increased Voltage Interim Plueging Criteria (IPC) i NRC Position ne task group believes that use of a 1-volt limit on an interim basis constitutes an appropriately conservative and cautious approach pending additional data l

and experience and will ensure that mature stress corrosion cracks will not

- remain in service. (Page 3-38)

Respense (Southern Nuclear)

As indicated in draft NUREG-1477,"a strong relationship exists between voltage and burst pressure." As a result, the voltage required to ensure structural integrity during normal operation or during a steam line break is well defined. The primary issue with increasing the IPC voltage concerns calculation of the predicted leak rate in the event of a steam line break.

For any IPC, if the predicted steam line break leak rate is less than the acceptable i

limit; if the leak rate is calculated using an acceptable leak rate model; and if the IPC voltage is a voltage less than the structural limit; the IPC will be safe and acceptable for use. As a result,if a leak rate model acceptable to the NRC is used to predict leakage, an increased voltage IPC should be justifiable.

A 1 volt IPC using the draft NUREG-1477 leak rate modelis extremely conservative. In fact, in the 66 tubes that have been burst tested, no tube has

,i been found to leak at steam line break differential pressures less than 2.8 volts.

Berefore, a 1 volt IPC is approximately 1/3 of the lowest voltage indicatica found to date to leak at steam line break differential pressures.

One possible method for implementing the increased voltage IPC would be to increase the limit in.5 volt increments from I volt to 2 volts. Although i: is not anticipated that any problems would arise, use of the 3 volt incremen:s would allow operating experience to be gained as the voltage limit is increased.

l This approach would allow tubes which are currently in service as a resuit of the implementation of the 1 volt IPC to remain in service as long as they were not a concem from a structural or leakage standpoint.

2.

Approval of IPC for Periods in Excess of One Cycle NRC Position The task group concludes that it is appropriate to limit approval of the imerim 1-volt limit to one operating cycle at a time to ensure that future applications of voltage-based pluggmg criteria properly reDect the la:est data and experience.

IPage 3-38)

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17-Resronse (Southern Nuclear) v As discussed previously, IPCs up to 2 volts are conservative. Reviewing 1 volt t

IPCs on a cycle by cycle basis requires the extensive use of utility and NRC resources which are better spent on the review of the steam generator degradation specific management proposal. Additionally, the IPC and NUREG-1477, when finalized will have received adequate opportunities for public comment. Based on the conservatism contained in the 1 volt IPC, a cycle by cycle approval is not justified.

3.

Continued Use of Rotating Pancake (RPC) Probe NRC Position The task group concluded that all bobbin indications for tubes not plugged or repaired should be included in the assumed BOC distribution, regardless of whether the indications are confirmed by RFC. (Page 3-17)

The task group concluded that additional justification is needed to support

caving tubes in service that exhibit greater than 1 volt bobbin signals not confirmed by RFC at the TSPs. Even if this can be justified, all bobbin indications should be incirded in the BOC voltage distribution used for evaluating the potential for tube rupture and leakage, regardless of whether the indication has been confirmed by RPC. (Pages 3-34,3-38)

Resronse (Southern Nuclear)

The issue is not whether the bobbin indication is a false call, but whether a real flaw not detected by RPC would be too small to contribute significantly to either the leakage of burst probability. As noted in Section 7.3 of WCAP-13692, Response to NRC Questions on NRC Analytical Model for SLB Leakage of ODSCC at the TSPs," the pulled tube database for which field RPC data is available includes 93 specimens. Of these 93 specimens,68 were detected

.ndications and 23 were NDD. The earliest of the tube pulls used in the RPC databases dates to 1986. Prior to that time, reliable RPC systems that could be used at the tube support plate elevations were not available for general use. All tube specirnens in the RPC database are also present in the bobbin database.

Crack morphologies represented are only those that are dominantly axial in orientation, making the RPC database prototypical of axial ODSCC at TSP intersections that is the base for IPC applications.

On the basis of the database, RPC detection probability is greater than 70% at

>40% depth and approaches 100% for depths >70%. RPC detection studies on

~

crack length for PWSCC indicate delectability near 100% for very deep

.ndications >.2 inch length. Therefore. RPC can be expected to detect flaws of significant !ength for depths greater than 50% to 70 'o deep. Potential bobbin

ndications not confirmed by RPC inspection would not be a concern for tube

.ntegnty relative :o burst or leakace during a subsequent operating cycle.

18 Furthermore, in Information Notices 90-49,91-87, and 92-80 the NRC Staff indicated that the RPC probe was acceptable for detecting outside diameter -

corrosion at various plants. At Trojan, the use of RPC only for detection of tube support plate degradation was characterized as conservative.

In addition to these cases, several plants have operated allowing these flaws to remain in service without incident. The practice of allowing bobbin indicanons to remain in service if they are not detected by RPC has been explicitly detailed in previous IPC submittals. Both in the description of the IPC and in the revised technical specification pages, the repair criteria are noted i

to allow flaws to remain in service if an RPC inspection does not detect the flaw. Tnese technical specification amendments were submitted and approved i

in accordance with 10CFR40-92. Both amendments were noticed in the Federai Register and received no comments form the public.

l To revoke this important part of the IPC is unjustified and will result in the removal of tubes from service that are not a concern from a structuralintegrity or leakage standpoint.

4.

Use of Smaller Probes Where Necessarv i

NRi ?osition IPC is not applicable at any TSP :ntersections where the intersection'cannot be inspected with the full-sized bobbin probe specified in the Westinghouse guidehnes (0.610-inch-diameter probes for 3/4-inch diameter tubing; 0.720-inch.

diame:er probes fer 7/8-inch-diameter tubing.) (Pages 3-33,3-38)

Resconse Gouthern Nuclear-As dccumented in WCAP-13692. which was fonvarded to the NRC Staff in Westm:; house letter ET-NRC-93-3863 dated April 14,1993, two recent inspecnons have been made using reduced diameter probes centered for 7/8

)

inch OD tubing. At the first plant,46 TSP intersections were examined with both 0.720 inch and either 0.560 or 0.580 inch diameter probes. It has not been confirmed that the small diameter probes had a centering device for the nominal 0.775 inch tube ID. The normalized amplitudes obtained were compared to determine whether non-conservative results might be obtained due to reduced sensitivity associated with lower fill factor probes. In all cases.

the signal observed with the 0.720 inch probe was detected with the smaller probe. Only 4 of ~6 signais had less than 1009b of the 0.720 inch probe amplitude: 1( 1.43 volts 199%); 2) 1.24 volts (92?a; 3) 1.03 volts (91 b); and 4) 1.37 volts G3'M.

At a second plan:,19 in:ersections were similarly compared using 0.7Z inch

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and 0A40 inch prebes. In this trial, three signals observed with the 0443 inch

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probe exhibited less than 100% of the corresponding 0.720 inch probe amplitude: 1) 0.20 volts G8%); 2) 0.26 volts (29%); and 3) 0.5 volts (86%). All 19 specimens produced measurable signals with both probes, and in none was a i

signal greater than 1.0 volt with the 0.720 inch probe undersized with the 0.640 inch probe.

At voltages up to 20 volts, the smaller probe diameters calibrated to the same l

ASME standard voltage tend to consistently overestimate the voltage. In other i

words, use of a smaller probe results in more conservative voltage calls and has not resulted in any flaws not being detected.

As a result, smaller bobbin probes with the use of prescribed calibration procedures, should be used to inspect areas of tubes that are not accessible with the standard bobbin probe.

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tO10RY STORAGE REPORT AUG 1193 10:42AM)-

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Post-it" brand fax transmittal memo 7671 aof pages eft /

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Memo NWM A

To:

David Steininger From:

Rick Mullin p&. et,/

l Date:

August 11,1993 i

Subject:

Comments on EPRI Draft Letter on Draft NUREG 1477

)

I I have reviewed the draft EPRi letter forwarded by your letter dated August 9,1993.

My comments are provided below.

1.

The cover letterindicates that Attachment 1,2, and 3 are comments on the SGMP developed DSM approach. While thatis true, several of the comments apply to the use of an IPC, e.g.,9 tube pulls per outage. To say that these only relate to DSM or that Attachment 4 contains the IPC issues is incorrect.

j i recommend that a statement be added to the following issues that the issue J

pertains both to the DSM and IPC-issue (1). Page 5-Tube pulls l

Issue (3), Page 6 - POD l

Issue (B), Page 11 - Probe size and probe variability issue (2), Page 12 - In-line probe standard measurements for each tube 1

issue (6). Pace 13 - POD 1

/

1 i

Memo To:

David Steininger.

j From:

Rick Mullin AA#g j

Date:

August 11,1993 l

Subject:

Comments on EPRI Draft Letter on Draft NUREG-1477

.j I have reviewed the draft EPRI letter forwarded by your letter dated August 9,1993.

l My comments are provided below.

l 1.

The cover letter indicates that Attachment 1,2, and 3 are comments on the i

SGMP developed DSM approach. While that is true, several of the comments apply to j

the use of an IPC, e.g.,9 tube pulls per outage. To say that these only relate to DSM or that Attachment 4 contains the IPC issues is incorrect.

I recommend that a statement be added to the following issues that the issue pertains both to the DSM and IPC:

l Issue (1), Page 5 - Tube pulls Issue (3), Page 6 - POD issue (8), Page 11 - Probe size and probe variability i

i issue (2), Page 12 - In-line probe standard measurements for each tube 1ssue (6), Page 13 - POD Issue (9), Page 14 - 100 TSP inspection with RPC e

2.

In that issue (8) on page 11 and issue (4) on page 18 are the same issue, I would combine the two and indicate it applies to DSM and IPC.

q 3.

Issue (2) on page 12 requires testing probe wear"using a delivery. method that would reproduce the vertical position of the prolm in a smooth standard / conduit.

-I environment." l'm not sure why checking wear requires'a vertical position. If it meets a l

15% wear limit lying horizontally, I believe it would meet the limit vertically.

l 4.

Although I do no have a problem with identifying SNC as the author of the. comments, I believe the way it is presented implies that SNC was the only utility with the comment. On the basis of there being no objections voiced at the July 22-23 meeting to the issues I raised, I believe all utilities interested in an IPC would support those comments. That would include AEP, WPS, Duke, and Duquesne, in addition to SNC.

]

5.

No mention is made of the 20% analyst cutoff which Westinghouse has used in

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the past. This cutoff is based on the requirement to resolve voltage differences if the difference is > 20% and over 1 volt. To ignore this approach seems like leaving a j

benefit out of the ARC and we may be bit by the " tails of the distribution."

6.

No mention is made of the requirement to review RPC mix channels. I thought the result of the Chicago meeting was that the RPC mix channel was an option to the :

I analyst; however, no requirement to review the RPC mix channel for every indication i

was required.

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~J Comments on EPRI Draft Letter on Draft NUREG-1477.

08/11/93Paga 2 7.

The draft NUREG requires the use of 100% through-wall holes for calibration. I thought the recommendation of the NDE subcommittee at the Chicago meeting was that no benefit was gained by using through-wall holes. Therefore, we would keep the current cal standard unless some benefit was shown by using the 100% holes.

l 8.

On the basis of information Notice 93-52, the letter is addressed to the wrong location. The address should be:

U.S. Nuclear Regulatory Commission

[

ATTN.: Document Control Desk l

Washington, DC 20555 A marked up copy of the draft letter containing minor comments is attached.

If you want to discuss these comments, please call me at (205)868-5502.

CC: B.D. McKinney, B.L. Moore, D.N. Morey, J.D. Woodard t

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%searcn nstnute

.eaaership in Science ana echnorogy.

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August 9,1993 i

hlCC Rules and Review Directives Branch g g)

Division of Freedom of Information and Publication Services i

M Office of Administration y y' U. S. Nuclear Regulatory Commission F

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Washington, D.C.

20555 l

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Subject:

EPRI/ Steam Generator Management Program Comments on Draft NUREG 1477 (Announced in Federal Register on July 2,1993)

Reference:

'1) Memorandum lor T.J. Murley and E.S. Beckford, from W.T. Russell

.md T.P. 5:eis. Draft, NRC NUREG-1477, " Voltage-Based Interim l

Plugging Criteria.^or Steam Generator Tubes-Task Group Report" Availability for Public Comment, June 2,1993.

(2)

~ Steam Generator Degradation Specific Management (SGDSM), An Industry initiative." presented by the EPRI Steam Generator Reliability Project at :he USNRC Offices, April 15,1993. (also reference 26 in Draft NUREG : *77).

l The purpose of this le::er is to comment en reference I which was announced in the Federal Reg: ster / Vol. 55. No. '.26/ Friday, July 2,1993 m/ Notices. These comments are 'reing presented by -he EPRl/ Steam Generator Management Program (SGMPit If commems presentec in this :etter deal.cith information associated with :he.

subiect document's referenced proprietary material, then these comments must be considered preliminary and subject to rev:sion if and when the proprietary information becomes available for industry wide review and comment.

he 5GMP has solicited comrnents on the document from its utility members and those non-members who are allowed to participate in the SGMP on subjects dealing with safety related issues. He SGMP has reviewed these comments and consolidated them in this letter.

Since the SGMP is working on a generic approach of steam generator degradation j

management encompassing a;:ernate tube repair criteria (see reference 2), this letter j

Formeric titled the 5.eam Generator Reliability Prc:ect.

,, s D 'lll2%..,. -.,.. er : :e 5., -;r2 93 e : :

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presents comments motivated by (Draft) NUREG 1477's assumptions, analysis, l

requirements, recommendations and conclusions as they compare to this generic effort.

Additionally, the subject of (Draft) NUREG 1477 is interim plug ng criteria for l

steam generator tubes. Utilities have docketed requests for:!::... cplugging criteria and have had considerable communication with the NRC on this issue. The SGMP has not been involved in these discussions. Therefore, the majority of this letter s comments are divorced from these communications and any agreements 3

reached between these utilities and the NRC on the subject of interim 2!! emet:"-

plugging criteria. But to hel reduce the number of submitted utility comment letters, attached are epeci... tility comments m "" d 'c 3 d "t; r u i.L.;

% o e.-

i

[lPC issues. lhese are m a dition to the SGMP comments and intended for the y convenience of the regulatory review process associated with (Draft) NUREG 1477.

l Comments on (Draft) NUREG 1477 are proportioned into four attachinents.

l e

' identifies areas of agreement and attachment 2 lists areas of concern i

which are titled " imperatives Imperatives are areas requiring accommodation by 4

NRC and industry to allow successful regulatory action on the industrv's steam generator degradation specific management program (SGDSM) initiative.

l lists suggested improvements to the draft document and[tachment 4 l

provides individual utility comments mentioned earlier.

l Sincerely, l

Robert E. Smith Chairman, Executive Group Utility Steermg Committee Steam Generator Strategic Management Program i

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4

i 3-j i

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1 Significant Areas of Agreement with (Draft) NUREG 1477.

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0)' As noted on page 3-352, "the regression fit of the burst pressure data as a function of voltage is valid."

i!

' The utility industry believes NRC recognition that a NDE measured parameter l

such as voltage can be related to structural margin through standard regression i

techniques supports the generic approach of SGDSM. SGDSM maintains that an experimental data base is required to correlate the rupture strength of j

degraded tubes with a NDE measured parameter.

)

As noted on page 3-36,'the method for estimating the distribution of voltage -

.[

changes during the next operating cycle is appropriate for purposes of.

accounting for crack growth during the cycle in assessing burst and leakage integrity, assuming that crack growth and voltage growth rates are within the bounds of previous experience." This methodology is described on page 3-12 of the document and involves using plant-specific average voltage change-observed during previous operating cycles.

1 Although the SGMP is in general agreement with this position, it should be -

noted that specific plants may not have sufficient or appropriate historical l

degradation growth data to apply this methodology. An altemate approach -

l should be the use of an appropriate growth rate alue o tained frem data

'* g ' i gathered from other operating plants wherddentical egradation has been l

3 observed and for which acceptable degradation growth rate data can be a

extracted. A recommended growth rate will be provided in the SGDSM -

approach for use where plant specific degradation growth rate inicrmation is unavailable.

Page 3-27 indicated that the " proposed 150 gpd leak rate limit when combined with an effective leak rate monitoring program will provide reasonable j

assurance that should a significant leak be experienced in service, n will be detected and the plant shut down in a timely manner before rupture occurs and l

with no undue risk to public health or safety "

s It is acknowledged that this is an acceptable " defense-in-depth" stra:egy.

/

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(4)

Page 3-37 indicated that " implementation of the Westinghouse (or similar) i guidelines for NDE data acquisition and analysis is essential for ensuring the reliable detection of low voltage signals and for ensuring that var: abilities m t% voltage responses and voltage measurements are minimized.

3 i

2 Page numbers refer to (Draft) NUREG 1477 (see reference 1).

y)Qh

4 1

It should be recognized that the generic SGDSM initiative will establish an industry standard of data acquisition and analysis for each degradation that has i

associated with it an alternate repair criteria which satisfies the protocol established by SGDSM.

5) Page 4-6 indicates that changes to the reactor coolant specific activity technical specifications would permit a licensee to operate with predictions of primary to l

secondary leakage during accident conditions higher than that which would be allowed using Standard Technical Specifications. This would still ensure the same level of protection to the public.

This operational requirement is intended in part to minimize both simple and compound iodine spiking. Although the SGMP is in general agreement with this position, it should be recognized, as indicated in NUREG-0933, "A Prioritization of Generic Safety Issues," July 1991, item B-65, that iodine spiking is a significant effect in only non-core melt accident consequences, which are not major contributors to nuclear plant risk." It is suggested that NUREG 1477 reaffirm this position of spiking not being major contributors to t

nuclear plant risk for plants operating under steam generator alternate tube repair criteria.

6) Page 4-39 indicates that the NRC concluded that because calculations using leak rates which are several orders of magnitude above the expected leak rates f: om a plant using IPC, and because the fraction of the initial RWST inventorv calculated to remain was so large, no plant specific calculations are required to demonstrate adequate RWST inventorv.

i The SGMP believes that SGDSM will allow higher than IPC allowable leak rates, but lower than assumed values used in these NRC calculations for the reason of adequate RWST inventory. Analysis by the SGMP was previously presented to the NRC (see reference 2) substantiating this conclusion.

Therefore, the NRC conclusion of no need for plant specific calculations should remain valid for alternate repair criteria under SGDSM and should be so indicated in the NUREG.

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f "2

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Imperatives l

1) Pages 3-14 and 3-17 identify a requirement to pull tubes to meet specified.

objectives. These objectives are:

1. Enhance and validate the empirical burst and leakage correlations, f

u*

2. Confirm that axial ODSCC continues to be the dominant degradation

-[

mechanism at the TSP intersections, and j

3. Provide additional data for assessing the reliability of the inspection methods (e.g., provide verification of "non detectable indication" as indicated by NDE interrogation of the degradation).

The obiectives are intended to be met by pulling a minimum of nine tube TSP

nteractions exhibiting voltages ranging from less than 1 volt to the maximum

>bserved per plant outage.

ATP4 Within the framework of the SCDS proacghe above noted objectives and p

method of attaining these objectives are not completely valid or warranted.

i

~d Limited tube pulls across the industry (not on individual plant-by-plant per y

outage basis) should be performed to verify degradation's burst and leak.

4 )

correlations maintained in the industry's SGDSM program. The number of

(

p

ube pulls over time should not be open ended. Engineering criteria will be y

[4

< stablished identifying tube pull needs. Obtaining useful tube pull data.

~

xhibiting a high benefit / cost ratio should be the objective. Practical tube pull

.equirements are required. For exampiglesired tube intersections for ODECC

>:xamination should be taken from the Tower bundle elevation to maximize a

>uccessful tube pull effort. Also verification of correlations should be done for the range of independent variables which have the largest safety significance and/or exhibits unacceptable dependent variab.e uncertainty. Areas of safety

)

significance is located at the safety criteria's limit value; not at values far i

removed from the limit.

Plants that haven't pulled tubes establishing degradation morphology may

)

need to do so to a limited but practical extent. But sufficient NDE data may be available from the plant to establish the degradation mechanism by comparing to fully characterized NDE signatures availabgr example, in an industn 5GDSM data base or acceptable alternative. To date approximately 200 tube intersec: ions from field pulled tubes exhibit censistent eddy current signa:ures for ODSCC at tube supports.

~

6 Cost / benefit associated with pulling tubes to assess the adequacy of ISI techniques used in SGDSM is unacceptably high and may not enhance safety.

ISI procedures should be adequately evaluated and qualified by reliability assessment programs before plant application. Such a program is exhibited by SGDSM with its requirement to implement the EPRI PWR Steam Generator Examination Guidelines document which contain protocol requirements for data analyst and ISI equipment qualification. It is emphasized that this reliability assessment program is in addition to and does not substitute for a plant's own site specific ISI qualification program which can be satisfactory.

An example of an unacceptable cost / benefit ratio for a tube pull requirement is that associated with verifying "no detectable defect (NDD)." If an acceptable ISI reliability assessment program is mandated under SGDSM, justification requiring verification of in-field NDD calls can not be substantiated.

Additionally, an infinite amount of data is required to prove a " negative", that is, to verify NDD, resulting in an infinite cost / benefit ratio.

[(2)

Page 3-20 indicated a lack of a proven correlation between leak rate and voltage and pas;e 3-21 recommends a leak rate relation which is not a function of voltage.

40 Standard statistical tests should be used establishing the very low probability of y y

,g a zero slope linear relationship (i.e., no functional relationship between leak I

rate and voltage).

Althou;;h one could arrive at a relatively low correlation coefficient for the subject data, this does not necessarily mean a correlation does not exist. Data scatter :ausing a relatively low correlation coefficient can be directiy related to some physical phenomenon. In the case of ODSCC at tube support intersecnonghe phenomenon is morphology variations ti.e., number of ligameras) associated with the cracking. But uncertainties associated with the regress:en model and supporting data can be applied to account for this data scatter.

Cogent statistical arguments concerning these issues were presented at a previous utility industrv/NRC meeting (see reference 2).

Similar statistical techniques and not arbitrary qualitative judgments should be used establishing correlations for other degradation mechanisms. Correlations under 5GDSM should not be precluded by NUREG 1477 assumptions, such as the distnbution of leak rate not depending on vo'ltage, unless adequately supported by standard statistical arguments.

)

Page M3 provides a probability of detection (POD) value oi O.6 for ODSCC at t

tube su: port plates. As noted this value was the average POD for 20-percent f(f.y I

7

~

through-wall to 100-percent through wall cracks. Page 3-29 recommends a

consideration of multiple ISI techniques as a means for enhancing POD.

It is SGMP's understanding this average POD value was based on the resuits of the Surry project as summarized in NUREG/CR-5117, PNL-6446,5:eam Generator Tube Integrity Program / Steam Generator Group Project. Final Project Summary Report. May 1990. The SGMP believes this is an inappropriate use of the data. The Surry data was taken without following o[f Nc,

p. ( specific ISI examinatinn cuidelines. This is not the case under the industrv's

-proposed SGDSM progranf Specific inspection guidelines will be followed which have supporting POD information considerably higher than the 0.6 value recommended in the draft NUREG document. Justification for a higher value is provided in the SGDSM program. Specifically, for ODSCC at tube support plate intersections, the applicable value vs. voltage is greater than 95%

at a 90% confidence for the 1.0 to 3.0 voltage range (the range of most interest).

A lower probability value will be allowed for lower voltages that are not structurally significant. Additionally, structural margin is not compromised even when growth rate is evaluated with these initiallower voltage values for the next operational cycle of the plant.

Although the industry provided POD is acceptable, the actual value is higher because of independent dual analyst interrogation of the inspection data as required under SGDSM. For example, if the probability of detecting degradation of a certain extent is 0.9, then the probability of a single miss (POM) is 0.1, the probability of two misses is 0.1 x 0.1 = 0.01, and so the POD for two inspections is 1.0 - 0.01 = 0.99, assuming independence between inspections.

Credit for dual, independent analyst inspection of the data should be allowed.

Other ISI techniques to develop an enhanced effective POD appears unnecessary. Since the POD, for example that associated with eddy currem bobbin coil probes used to interrogate ODSCC at tube support plates. is acceptable, use of complementary devices is superfluous and only results m additional occupational radiation exposure and O & M expense.

i

4) Page 3-17 indicates that all input parameters for setting the voltage limit shouid be evaluated at consistent 95-percent confidence values. Additionally,it is stated, "the task group concluded that the deterministic methods used to O,

establish the IPC, generally, satisfy the requirements of Regulatorv Guide 'd21.

Ju even when consistent 95-percent confidence values of the input parameters y

(i e., voltage variability, voltage growth, and burst pressure) are used.

i it is inappropriate to require high probability values of for example. 95% 95%

on all input values to ensure that certain limits specified in Regulatorv Guide l.121. such as 3 times normal operating pressure for tube burst, no: be exceeded.

This suggested methodology produces an(;Innecessarih3high probability that a I

safety limit will not be exceeded.

(

5fb U/)

i s

g' It should be noted that the margin to burst factor of 3 specified in Regulatorv Guide 1.121 is an engineering safety factor intended to bound uncertainties which have not been quantified. This conservative, deterministic l

methodology of setting limits is analogous to the ASME code requirement of fatigue where fatigue test data is arbitrarily modified by factors of 20 on stress or 2 on cycles, whichever is limiting, to produce a conservative design fatigue l

curve.

l l

The 95%/95% or acceptable probability value on for example, tube burst, derived with uncertainties accounted using appropriate values, should be specified at actual normal operating pressure or main steam-line break conditions.

If an engineering safety factor of 3 is mandated, per Regulatory Guide 1.121.

then credit should be allowed for example, for the constraint provided by the j

tube support plate preventing tube burst under normal operating pressure using an appropriate definition of " burst" j

If the above described probabilistic methodology, demonstrating a "high level

)

of probability" exists that safety is maintained, is not allowed, then the ASME

]

code requirement of 1.4 times main stearn line break pressure conditions as i

dictated by Regulatory Guide 1.121 should be retained. It should be met I

without additional conservative bounding confidence limits on the test data.

1 5)

Page 5-1 indicates that "the task group concluded that evaluation of offsite and control room doses can be used in conjunction with the recommended me: hod i

for predicting primarv-to-secondary leakage rates to provide adequate assurance that offsite doses would not exceed establishe

'mits (i.e.,10CFR Part 100 and GDC 19) under postulated accident condition

4 i

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The SGMP be eves that an acceptable probability value be specified for the l

6 safety criteri of interest only, which in the present case is the radiological dose l

\\g limit at th

'A as dictated by 10CFR100. An appropriate probability value should be established for meeting less that 10CFR100 limits.

It is unreasonable to dictate a high probability value (e.g.,95%/95%) on all variables to meet a small fraction of or 10% of 10CFR100 limits. Quantification of uncertainties should remove the need for applying engineering safety factors such as one tenth of 10CFR100 limits. Such safety factors were originally incorporated to compensate for nonquantifiable uncertainties. Additionally, not all input variable uncertainties need be specified at high probability values in setting an acceptable high level of probability that 10CFR100 limits are not exceeded. A modern approach to such analysis is the so-called best-estima:e BE) methodology, in which one predicts realistic results with explicit consideration of and accounting for uncertainties that are introduced at everv l

l

.o 9

stage of the calculation. One would then need to demonstrate to a high level of probability that safety limits of 10CFR100 are met. For example, such an approach in the acceptance criteria for emergency core cooling systems (ECCs) in light water reactors is now allowed per a NRC rule change in 1988.

Unfortunately, in this case a high but unspecified probability (although the Standard Review Plan repeatedly calls for a 95% probability at a 95% confidence level that certain limits should not be exceeded) is dictated for limits (e.g., peak clad temperature and minimum DNBR) which themselves have incorporated j

conservative, possibly 2 sigma tolerance bounds on safety limit distributions.

in any event, the SGMP suggests that similar analysis, with appropriate modification defining acceptable safety margin, be allowed to establish the allowable steam generator primary to secondarv leakage under faulted load conditions.

The SGMP favorably recognizes that somewhat similar analysis is suggested for

)

investigation on page 5-2 of (Draft) NUREG 1477. But the SGMP recommends j

analysis more closely aligned with the 1988 ECCS methodology with appropriate improvements in actually establishing quantifiable, J

probabilistically defined safety margin under SGDSM.

i (6)

Page 4-2 indicates that in order to place the Standard Review Plan (SRP) l radiological dose analysis in perspective, a realistic analysis was performed.

i Specifically, it is stated that,"this analysis calculated offsite doses using assumptions and data closer to the expected conditions, that is, without as

[

much margin as is included in the licensing (SRP) calculations /

b

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The above statement does not appear to be consistent with the so-called realistic analysis presented in the draft NUREG. Specifically, the realistic analysis uses a compound iodine spike for the faulted load event of a main steam line break which is 500 times the release rate from fuel which gives a coolant activity of 11pCi/gm.

The SGMP agrees that 0.1pCi/gm is a more realistic value of the equilibrium 1131 coolant activity. But the compound spiking value of 500 is not. NRC contractor data indicates that the compound iodine spike is one to perhaps two orders of magnitude less than 500. This information was previously presented by the SGMP to the NRC on April 15,1993 (see reference 2).

It should be noted that NUREG 0844,"NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Genera:or Tube Integrity, September 1985, " specifies a best estimate value of 10-20% of the 500 value.

Finally. Supplement 13 to NUREG-0933, "A Prioritization of Generic Safety Issues. December 1991, Item 67.5.1 refers to a reassessment of radiological i

i

3o consequences associated with SGTR accidents and reevaluation of the applicability of the assumptions in the Standard Review Plan, Section 15.6.3.

The conclusion listed in this item indicates in part that the, " DST agreed that a "best estimate" analysis modeled after plant experiences, like Ginna. could be beneficial in more realistically determining the risk and conservatisms inherent in the current SRP requirements." It further goes on to say, "if this limited scope comparison of the SRP model with best estimate analysis is followed, this issue could be considered as an improvement to current licensing positions (a licensing issue)." Since this SRP reevaluation would obviously include a reassessment of the compound 1131 spike of 500, the SGMP recommends that theNRC initiate reassessment of the SRP, Section 15.6.3, as

, -rtmommended b,QAEin item 67.5.1 of NUREG-0933 (Supplement 13). Such a

,,, y(

reassessment shou 1 Fuse NRC contractor data gathered to support Generic Issue 67.5.1. Such a reassessment and its conclusions would also apply as it relates to I 131 spiking models used in analysis of the main steam line break event with concurrent steam generator leakage.

(7) Page 3-11 provides discussion on a NRC model which assesses crack growth, fracture behavior, leakage rates, and is a traditional mechanics based approach.

It concludes that sensitivity studies using the model indicated the difficulty in reliability modeling complex phenomena like IGA /ODSCC. It goes on to (j

further state, "the model can use, as input data, the flaw lengths and depths derived from inservice inspection results of the flaws in tubes that are left in service," and " improvements in the ECT technology should enable the use of length - and depth - based criteria for plugging and repairing, thereby providing

(

the recuired link to inservice inspection data for the NRC modelf Additionally page 3-39 suggests that,"over the long term, the use of length -

and/or depth-based criteria for plugging and repair would be a preferable approach, and the indusry should pursue improved nondestructive testing technology to support such criteria."

The SGMP takes issue with the suggestion that the above described approach can become a viable option to interrogate the structural integrity of steam generator tubing experiencing ODSCC at tube support intersections. The SGMP believes that pursing this approach for this type of degradation would be a wasteful utilization of finite resources with a low probability of success in the foreseeable future. But of course advances in NDE technology is desirable and being pursued by the utility industry. The questions which arise on the presem subject are issues of practicality and timeliness.

As noted on page 3-29 of the (Draft) NUREG, "the most desirable situation is to define a plugging limit that (1) is reliability enforceable and (2) provides resonable assurance that tubes that satisfy the limit will continue to have structural and leak-tight integrity." The SGMP believes that a voltage based repair limit for ODSCC n tube supports satisfies this desired objective of the NRC.

.11 i

4 NUREG 1477 clearly indicates, as noted above, "the difficulty in reliability ' _

modeling complex phenomena like IGA /ODSCC." In addition to this problem.

state-of-the-art NDE technology available today and in the foreseeable future should not and can not provide the necessary, highly accurate inputs to compensate for the inadequacies of a recognized, unreliable mechanistic model y

of IGA /ODSCC.

The SGMP believes that such an approach will ultimately result in more uncertainty in its application than a voltage based criteria.

1 (8) Page 3-33 indicates that IPC is not applicable at any TSP intersections where the

~

intersection cannot be inspected with the full-sized bobbin probe specified in the Westinghouse guidelines. Additionally, page 3-31 indicates that the task l

group recommends that bobbin coil probe variability be limited to 5 percent.

d it must be recognized that smaller type probes are needed to interregate

,g previously sleeved tubes. Such probes should be allowed as long as they meet y

,f acceptable performance requirements (i.e., adequately qualified) wnh required.

calibration. The major concern with the use of smaller probes is ::MedW probe centering within the tube. This has not been a problem to date.

I Industry lacks capability to meet a probe variability specification of 5 percent. In any event, this specification is unnecessary. Realistic probe to probe variability is already taken into account in the correlations database because zarious probes where used in its generation. Additionally, allowance for a maxirnum of 15% probe wear induced voltage change encompasses this voltage variability l

i concern.

Finally, under the SGDSM initiative equipment qualification req =remen:s

'l will establish acceptable performance standards on such probes resulting in weil defined voltage variability uncertainties which can be appropriately handled in best estimate analysis establishing well defined safety nargins.

h e,IJ f Y N N W -

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t Suggested Improvements (1)

Page 2-1 states that the NRC evaluates each applicant's SG design, water chemistry, and inspection program in accordance with the regulatory guidance in the Standard Review Plan (SRP) NUREG-0800 before issuing a license.

t It is SGMP's understanding that the SRP was first issued in 1975 (NUREG-75/087). A number of PWR plants were issued their construction permit after 1975, but were not licensed to the SRP. It is suggested that the above noted i

paragraph be clarified to recognize this fact. Additionally, appropriate discussion should be added explaining why plants not licensed to the SRP must j

now be evaluated with respect to it for application of alternate steam generator tube repair criteria that satisfy safety margins dictated by Regulatorv Guide 1.121 and ASME code requirements.

(2)

Page D-1 recommends that a probe wear measurement be made automat:cally on an in-line standard each time a tube is inspected.

M The SGMP believes it is not practical, efficient, or necessary to check the probe

-h wear after each tube has been inspected. He guide tube standard is not capable f

of reproducing the same path for probe travel after each inspection. The guide I

tube standard also creates abnormal probe movement due to missed alig:unent

/gpj [Jwear at the beginning and end of a calibration periodj6 s with the tube end. A more reasonable approach may be to inspect for probe p

j that would reproduce the vertical position of the probe in a smooth}

[* pi 3

f (fb [j standard / conduit environment.fProbe wear uncertainty is taken mto account estaolishing the repair limit. His in combination withye iodic and accurate j

probe wear measu'rementsh mt of p :. i..."ir:E.. should be accep;able.

]

/

If the probe is found to be out-of-spec.. re-examination of previously interrogated tubes with a new probe should be performed.

(3)

Page 3-31 stipulates that noise criteria should be incorporated that would require a certain specified noise level not be exceeded, consistent with the j

objectives of the inspection.

It should be noted that a signal /noi= 'c N) issue is a generic one irrespective u

of alternate tube repair criteria. It exists under a plant's present steam generator repair criteria which is 40% depth based. It is difficult to set objective noise critergspecially if required to proporrion limits per noise inducing mechamsm (e.g., tube noise, electrical noise, etc.). Present plant inspect:on procedures account forjS/h] issues and should be followed during application of altemate tube repair criteria. The SGDSM process will provide guidance in

his area. The draft NUREG should be modified by removal of this noise specification and reflect present plant procedures for dealing with S/N issues.

s

4) Page 3-16 and 3-36 indicate that data outliers should not be removed from the data set from which the alternate tube repair correlations are developed.

O The SGMP agrees that data outliers should not be arbitrarily removed. Data f

3 points should be removed if they are clearly associated with a so called bad test f

or inappropriate test specimen (e.g., damaged by other factors). Additionally, data should be removed if they are associated with degradation morphologies which do not represent the damage mechanism for which the alternate tube repair criteria is developed. This has been done to the SGDSM data set only when data removal has a conservative effect on the correlation line. Outlier removal should be allowed under these circumstances.

5) Page 3-23 discusses the need to test other functional forms of the probability of leakage curve.

The SGMP believes this is a reasonable request for the SGDSM approach. But it

[g should be noted : hat SGDSM incorporates an uncertainty distribution with the v

P.g probability of leakage (POL) curve. This compensates for incomplete

\\

knowledge of the functional POL relationship. Additionally, as in any data-fitting endeavcnthe adecuacy of the model should be examined using standard statistical techdf6ues. It is suggested that the discussion on page 3-23 reflect this approach for future application of SGDSM.

) Page 3-28 recomr ends that an average POD value be used to determine the number of undetected flaws for use in the leakage calculation.

g SGMP believes that the POD in the range of interest is sufficiently high under y

SGDSM and that ncorporating so called undetected indications in the leakage calculation is unwarranted. For the voltage range exhibiting low POD values, g6[

such indications nave been shown to be structurally insignificant.

Additionally, these indications are not expected to grow over the subsequent operating cycle to an extent which causes a leakage or burst concern. It is recommended that the draft NUREG document be modified to support this conclusion.

(7; Page 3-36, fourth bullet appears to imply that probabilistic analyses on a per outage basis should be used to evaluate the conditional probability of tube burst during a postula:ed MSLB.

O The SGMP recommends that any such analysis on an outage basis be avoided.

~

In this regard, the gener:c SGDsM initiative will provide a methodology that avoids this plant rutage requiremen: of a probabilistic calculation ensuring marg:n to burst i:: the faulted load, main steam line break event. It is recommended tha: the craft NUREG be modified to avoid requiring a probabilistic caicciation on a rer ou: age basis if an alternate method is found

14 acceptable as presented in the generic SGDSM initiative. Avoiding involved probabilistic analysis on an outage to outage basis is consistent with communicated NRC's desires in previous SGMP/NRC annual review meetings.

8)

Page 3-19 raises a concern about very large amplitude indications that have i

recently been identified in the field.

()O The SGMP recommends that the NUREG be rewritten to put these incidents in g

the proper perspective. High growth rate degradation is not uniquely associated with alternate tube repair criteria. Irrespective of the repair criteria, g

whether it is 40% depth, crack length, or voltage based, high degradation growth must be adequately handled to maintain safety margin as it has been done in the past. The SGMP believes that this process can be enhanced through the SGDSM initiative with its requirement of more extensive, engineering based inspection criteria allowing early identification of steam generator tube degradation.

[9) Page 3-33 requires that a minimum of 100 TSP intersections below 1 volt be inspected with RPC, regardless of bobbin voltage amplitude.

The SGMP believes that a voltage based repair criteria for ODSCC at tube support plates, such as being proposed under SGDSM, justifies leaving as a minimum indications less that I volt as indicated by bobbin coil in service.

h g/ -[p The SGMP can not identifv any reasonable justification requiring supplementary RPC inspections of these indications which clearly do not j

'I compromise safety margin. Such interrogation increases O & M cost with little if any incremental increase in safety.

[10) Page 3-33 presents discussion that implies that the industry has not developed an acceptable POD for RPC inspection of ODSCC at tube support plates.

U Although under SGDSM supplementary RPC requirements is not as h,g [v extensively defined as it has become during the IPC initiative, the development b

of an acceptable POD for RPC interrogation of ODSCC at tube support plates will be an objective of the utility industry in support of SGDSM.

11) Page 4-38 indicates that the staff intends to examine in detail the effect of primary-to-secondary steam generator leakage greater than allowable under IPC under severe accident conditions.

The SGMP believes that a severe accident is an extremely low probability event, i, b[

not part of the plant's design basis, and should not be considered in the g

regulatory review process dealing with SGDSM. Severe accidents is a complex issue with system wide implications and therefore SGDSM should not be its focal point of resolution. It should be resolved in severe accident space and not

5 e.-

+

in licensing space associated with SGDSM. In spite of this recommendation, the SCMP did investigate the effect of significant primary to seconciary steam generator leakage on the course of a severe accident. Results were presented to the NRC during an all day meeting at NRCs White Flint facility on December 22,1992. The SGMP conclusion was that SGDSM's anticipated, allowable steam generator leakage would not alter the conclusion reached in NUREG 1150 for the TMLB' severe accident sequence. NUREC 1477 should be rewr:tten to -

reflect these findings.

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t 16 Individual Utility IPC Comments 1.

lacreased Voltage Interim Plugging Criteria (IPC)

NRC Position i

The task group believes that use of a 1-volt limit on an interim basis constitutes an appropriately conservative and cautious approach pending additional data and experience and will ensure that mature stress corrosion cracks will not remain in service. (Page 3-38)

Resoonse (Southern Nuclear)

As indicated in draft NUREG-1477, "a strong relationship exists between voltage and burst pressure." As a result, the voltage required to ensure structural integrity during normal operation or during a steam line break :s well defined. The primary issue with increasing the IPC voltage concerns calculation of the predicted leak rate in the event of a steam line break.

t For any IPC, if the predicted steam line break leak rate is less than the acceptable limit; if the leak rate is calculated using an acceptable leak rate model; and if the IPC voltage is a voltage less than the structural limit; the IPC will be safe and acceptable for use. As a result,if a leak rate model acceptable to the NRC is used to predict leakage, an increased voltage IPC should be justifiable.

A 1 volt IPC using the draft NUREG-1477 leak rate modelis extremely conservative. In fact,in the 66 tubes that have been burst tested, no tube nas been found to leak at steam line break differential pressures less than 2.8 volts.

Therefore, a 1 volt IPC is approximately 1/3 of the lowest voltage indicatica found to date to leak at steam line break differential pressures.

One possible method for implementing the increased voltage IPC would be to increase the limit in 3 volt increments from 1 volt to 2 volts. Although i: is not anticipated that any problems would arise, use of the 3 volt incremen:s would allow operating experience to be gained as the voltage limit is increased.

This approach would allow tubes which are currently in service as a result of the implementation of the 1 volt IPC to remain in service as long as they were not a concern from a structural or leakage standpoint.

2.

Aoproval of IPC for Periods in Excess of One Cvele NRC Position The task group concludes that it is appropriate to limit approval of the interim i

1-volt limit to one operating cycle at a time to ensure that future applications of J

voltage-based pluggmg criteria properly reflect the latest data and experience.

(Page 3-38)

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l Resronse (Southern Nuclear)

'I As discussed previously, IPCs up to 2 volts are conservative. Reviewing I volt 1

IPCs on a cycle by cycle basis requires the extensive use of utility and NRC

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resources which are better spent on the review of the steam generator degradation specific management proposal.- Additionally, the IPC and NUREG-1477, when finalized will have received adequate opportunities for public comment. Based on the conservatism contained in the 1 volt IPC, a cycle by i

cycle approval is not justified.

l 3.

Continued Ure of Rotating Pancake (RPC) Probe NRC Position f

The task group concluded that all bobbin indications for tubes not plugged or j

.epaired should be included in the assumed BOC distribution, regardless of vhether the indications are confirmed by RPC. (Page 3-17)

The task group concluded that additional justification is needed to support

eaving tubes in service that exhibit greater than 1 volt bobbin signals not confirmed by RPC at the TSPs. Even if this can be justified, all bobbin indications should be included in the BOC voltage distribution used for evaluating the potential for tube rupture and leakage, regardless of whether the indication has been confirmed by RPC. (Pages 3-34,3-38)

Resronse (Southern Nuclear)

The issue is not whether the bobbin indication is a false call, but whether a real flaw not detected by RPC would be too small to contribute significantly to either the leakage of burst probability. As noted in Section 7.3 of WCAP-13692,

' Response to NRC Questions on NRC Analytical Model for SLB Leakage of ODSCC at the TSPs? the pulled tube database for which field RPC data is available includes 93 specimens. Of these 93 specimens,6S were detected

.ndications and 25 were NDD. The earliest of the tube pulls used in the RPC databases dates to 1986. Prior to that time, reliable RPC systems that could be used at the tube support plate elevations were not available for general use. All tube specimens in the RPC database are also present in the bobbin database.

Crack morphologies represented are only those that are dominantly axial in orientation. making the RPC database prototypical of axial ODSCC at TSP intersections that is the base for IPC applications.

On the basis of the database, RPC detection probability is greater than 70% at

>40% depth and approaches 100% for depths >70%. RPC detection studies on crack length for PWSCC indicate delectability near 100% for verv deep

.ndications >.2 inch length. Therefore, RPC can be expected to detect flaws of significant length for depths greater than 50% to 70% deep. Potential bobbin

ndications not confirmed by RPC inspection would not be a concern for tube

ntegrity relative to burst or leakage during a subsequent operating cycle.

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18 Furthermore, in Information Notices 90-49,91-87, and 92-80 the NRC Staff indicated that the RPC probe was acceptable for detecting outside diameter -

corrosion at various plants. At Trojan, the use of RPC only for detection of tube support plate degradation was characterized as conservative.

In addition to these cases, several plants have operated allowing these flaws to remain in service without incident. The practice of allowing bobbin indicanons to remain in service if they are not detected by RPC has been explicitly detailed in previous IPC submittals. Both in the description of the IPC and in the revised technical specification pages, the repair criteria are noted to allow flaws to remain in service if an RPC inspection does not detect the flaw. These technical specification amendments were submitted and approved

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in accordance with 10CFR40-92. Both amendments were noticed in the Federal Register and received no comments form the public.

To revoke this important part of the IPC is unjustified and will result in the removal of tubes from service that are not a concern from a structuralintegrity or leakage standpoint.

4.

Use of Smaller Probes Where Necessary j

NRC Position IPC is not applicable at any TSP intersections where the intersection cannot be inspected with the full-sized bobbin probe specified in the Westinghouse guidelines (0.610-inch-diameter probes for 3/4-inch diameter tubing: 0.720-inch-diame:er probes fer 7/8-inch-diameter tubing.) (Pages 3-33,3-38)

Resronse ':,uthern Nuclear

• As documented ir WCAP-13692, which was forwarded to the NRC Staff in Westmghouse letter ET-NRC-93-3863 dated April 14,1993, two recent inspecnons have been made using reduced diameter probes centered for 7/8 inch OD tubing. At the first plant,46 TSP intersections were examined with both 0.720 inch and either 0.560 or 0.580 inch diameter probes. It has not been confirmed that the small diameter probes had a centering device for the nominal 0.775 inch tube ID. The normalized amplitudes obtained were compared to determine whether non-conservative results might be obtained due to reduced sensitivity associated with lower fill factor probes. In all cases, the signal observed with the 0.720 inch probe was detected with the smaller probe. Only 4 of 46 signais had less than 100% of the 0.720 inch probe amplitude: 1( 1.43 volts <99%); 2) 1.24 volts (92%); 3) 1.03 volts (91%); and 4) 1.37 volts G3%).

At a second planr.19 imersections were similari:. compared using 0.7% inch and OMO inch probes. In this trial, three signals observed with the 0443 inch

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o probe exhibited less than 100% of the corresponding 0.720 inch probe amplitude: 1) 0.20 volts GS%); 2) 0.26 volts (29%); and 3) 0.5 volts (86%). All 19 specimens produced measurable signals with both probes, and in none was a signal greater than 1.0 volt with the 0.720 inch probe undersized with the 0.640 inch probe.

At voltages up to 20 volts, the smaller probe diameters calibrated to the same l

ASME standard voltage tend to consistently overestimate the voltage. In other j

words, use of a smaller probe results in more conservative voltage calls and has not resulted in any flaws not being detected.

As a result, smaller bobbin probes with the use of prescribed calibration procedures, should be used to inspect areas of tubes that are not accessible with the standard bobbin probe.

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