ML20196B508

Forwards Partial Review of EPRI Rept, SG Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates Database for Alternative Repair Limits, Submitted in

ML20196B508
Person / Time
Issue date:	11/20/1998
From:	Lainas G NRC (Affiliation Not Assigned)
To:	Modeen D NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT &
References
PROJECT-689 GL-95-05, GL-95-5, NUDOCS 9812010117
Download: ML20196B508 (4)

Text

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g UNITED STATES NUCLEAR REGULATORY COMMISSION E

E WASHINGTON, D.C. 20666-0001

.....p November 20, 1998 David J. Modeen, Director Engineering, Nuclear Energy Institute 1

1776 i Street, NW Suite 400 t

Washington, DC 20006-3708

SUBJECT:

EVALUATION OF PROPOSED UPDATE TO SGDSM DATABASE AND l

MODIFICATIONS TO THE METHODOLOGY TO ASSESS STEAM GENERATOR TUBING OUTSIDE DIAMETER STRESS CORROSION CRACKING

Dear Mr. Modeen:

t By letter dated June 5,1998, the Nuclear Energy Institute (NEI) submitted Addendum 2 to the Steam Generator Degradation Specific Management (SGDSM) Database, " Steam Generator i

Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates Database for Alternate Repair Limits." The letter also transmitted responses to two NRC requests for additionalinformation. At this time, the NRC is reviewing a number of updates and proposed l

changes to the methodology for implementing voltage-based steam generator tube repair criteria of Generic Letter 95-05, " Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking." NEl has requested

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that the NRC review the following proposals: (1) updates to the SGDSM Database, (2) a revision i

to the calculated leak rate for Plant S Tube R28C41, (3) data exclusion Criterion 3a, (4) the i

implementation of a voltage-based probability of detection (i.e., Probability of Prior Cycle Detection - POPCD), (5) a correlation relating axial tearing failure loads to bobbin coil voltage for I

outside diameter stress corrosion cracking (ODSCC), (6) eddy current system voltage calibration procedures, and (7) the use of alternate growth rates to address accelerated ODSCC voltage growth for higher voltage indications or deplugged tubes returned to service using l

voltage based repair criteria.

The staff has completed its review ofitems 1,2, and 3. Enclosure 1 documents our evaluation.

We are continuing to review the remaining open issues associated with the proposed changes in evaluating voltage-based steam generator tube repair criteria established per the guidance in Generic Letter 95-05 (items 4 through 7). We willinform you of our conclusions upon completion of this review. If you have any questions with regard to our evaluation of the SGDSM Database or the status of our review, please contact Ted Sullivan at (301) 415-2795.

Sincerely, l

h n

Gus C. Lainas, Acting Director 9{ O}

c<^I'd Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission PROJECT NO.: 689

.g 3 '3 cc: C. Callaway, NEl Qg 9012010117 981120 PDR REVQP ERGNOMRC PDR

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PARTIAL REVIEW OF EPRI REPORT " STEAM GENERATOR TUBING OUTSIDE l

t DIAMETER STRESS CORRCSION CRACKING AT TUBE SUPPORT PLATES DATABASE FOR ALTERNATE REPAIR LIMITS" 4

NP 7480-L, ADDENDUM 2,1998 DATABASE UPDATE i

APRIL 1998 The NRC had requested that the industry reevaluate the French data included in the SGDSM Database to ensure quality assurance requirements were satisfied in obtaining these data.

Addendum 2 to the SGDSM Database, addressed this issue and discussed the evaluation of these data per the data exclusion criteria. A significant fraction of these data were excluded from the burst pressure correlation due to the application of appropriate exclusion criteria.

Addendum 2 also updated the database with datr obtained from domestic pulled tubes. The staff has completed its review of the database and associated correlations for %-inch diameter

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tubes and concluded that they are acceptable for use in voltage-based tube repair assessments.

1 i

The updated database and correlations for %-inch diameter tubing was affected by a proposed modification to the data exclusion criteria and a proposed adjustment to the leak rate for one i

tube previously included in the leak rate correlation. The NRC's original evaluation of the leak i

rate for Tube R28C41 concluded that the calculated leak rate of 2496 #hr would be reduced by 50 percent if a flaw length of 0.63 inches was used in the calculation rather than the value of 0.67 inms that included ductile crack extension that is believed to have occurred during leak rate testing. As indicated in Addendum 2, the CRACKFLO Code accovats for crack tip tearing.

Consequently, the appropriate crack length input into the code for computation of the leak rate is the lower value of 0.63 inches. On this basis, the NRC concludes that the reduction in leak rate i

for Tube R28C41 (Plant S) from 2496 Uhr to 1250 Uhr is appropriate and acceptable for inclusion in the %-inch tube database.

Data exclusion Criterion 3a was proposed by the industry to eliminate the introduction of non-relevant data into the leak rate database due to anomalies encountered during leak rate a

testing. Addendum 2 identified two leak rate data points presently included in the %-inch tube diameter leak rate data base that may be inappropriately included in the leak rate correlation.

These data were noted to have low measured leak rates relative to the mean regression line for the bobbin coil voltage versus leak rate correlation. Although this correlation is empirically based and scatter in the data may be a reflection of the empirical nature of the model, the industry noted that the measured leak rates for these tubes were much lower than expected based on mechanistic considerations.

Unlike the voltage-based leakage correlation, a model can be constructed using fundamental principals to demonstrate a general relationship between through-wall crack length and leak rate. The industry has applied one such modelto leak rate data for steam generator tubing and demonstrated consistency between measured 1eak rates and mechanistic models for predicting leakage. The leak rates for the two tubes in question appear to be outliers with respect to mechanistic predictions of leak rate based on the observed crack lengths of the subject specimens ~. Therefore, the industry has proposed additional data exclusion criteria to address the apparent problem with the leak rate data obtained from the two specimens. Data exclusion Criterion 3a permits data to be removed from the leak rate data base provided it lies below the ENCLOSURE

David J. Modosn, Director Engine ring, Nucl:ar En rgy Instituta 17761 Street, NW Suite 400 i

Washington, DC 20006-3708

SUBJECT:

EVALUATION OF PROPOSED UPDATE TO SGDSM DATABASE AND l

MODIFICATIONS TO THE METHODOLOGY TO ASSESS STEAM GENERATOR TUBING OUTSIDE DIAMETER STRESS CORROSION CRACKING

Dear Mr. Modeen:

By letter dated June 5,1998, the Nuclear Energy Institute (NEI) submitted Addendum 2 to the Steam Generator Degradation Specific Management (SGDSM) Database, " Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates Database for Alternate Repair Limits." The letter also transmitted responses to two NRC requests for additionalinformation. At this time, the NRC is reviewing a number of updates and proposed changes to the methodology for implementing voltage-based steam generator tube repair criteria of Generic Letter 95-05, " Voltage Based Repair Criteria for Westinghouse Steam i

Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking." NEl has requested that the NRC review the following proposals: (1) updates to the SGDSM Database, (2) a ievision to the calculated leak rate for Plant S Tube R28C41, -(3) data exclusion Criterion 38, (4) the implementation of a voltage-based probability of detection (i.e., Probability of Prior Cycle Detection - POPCD), (5) a correlation relating axial tearing failure loads to bobbin coil voltage for outside diameter stress corrosion cracking (ODSCC), (6) eddy current system voltage calibration procedures, and (7) the use of alternate growth rates to address accelerated ODSCC voltage growth for higher voltage indications or deplugged tubes returned to service using voltage based repair criteria.

l The staff has completed its review of items 1,2, and 3. Enclosure 1 documents our evaluation.

We are continuing to review the remaining open issues associated with the proposed changes in evaluating voltage-based steam generator tube repair criteria established per the guidance in Generic Letter 95-05 (items 4 through 7). We will inform you of our conclusions upon completion of this review. If you have any questions with regard to our evaluation of the l

SGDSM Database or the status of our review, please contact Ted Sullivan at (301) 415-2795.

Sincerely, 1

Gus C. Lainas, Acting Director l

Division of Engineering Office of Nuclear Reactor Regulation l

U.S. Nuclear Regulatory Commission l

PROJECT NO.: 689 cc: C. Callaway, NEl Distribution:

See attached list DOCUMENT NAME: G:\\ RUSH \\LETR-NEl.WPD INDICATE IN BOX:"C"eCOPY W/O ATTACHMENT / ENCLOSURE, "E"= COPY W/ATTIENCL. "N"=NO COPY i '"i (AMDE l

OFFICE EMCB:DE EMCB:DE EMCB DE NAME PJRush*

ELMurphy*

EJSullivan*

G inas DATE 10/28 /98 10/30/98 11/09/98 11 /

/98 OFFICIAL RECORD COPY

I

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3 in order to minimize the potential of eliminating data that exhibit scatter consistent with the

,o uncertainties irrplicit in mechanistic models, the staff recommended by letter dated February 26,1998, that the industry consider utilizing a more stringent confidence interval than typically accepted by the NRC (99% versus 95%). The use of a larger confidence interval should minimize the exclusion of data due to expected scatter within each leak rate model as additional data are incorporated into the leak rate database. The NRC has concluded that data exclusion Criterion 3a is acceptable as proposed in Addendum 2. In addition, leak rate data from specimens MB598-3 and MB604-2 should be appropriately deleted in accordance with Criterion 3a from the database for %-inch tubing.

Addendum 2 included modifications to the %-inch diameter tubing database based on the presumption of NRC acceptance of a proposed reduction in leak rate for Tube R28C41 (Plant S) and data exclusion Criterion 3a. As indicated previously, these changes are acceptable. In addition, riew pulled steam generator tube data was added to the %-inch tube database as documented in fuoendum 2. The staff has reviewed these data and concluded that the modifications to tae database and the correlations with these data are appropriate. Therefore, the NRC concludes that the database for %-inch diameter steam generator tubing and the associated correlations for burst, probability of leak, and leak rate included in Addendum 2 are acceptable. Future changes to the SGDSM Database involving the inclusion or exclusion of pulled tube data (% and Winch diameter tubing) should be made in accordance with the database protocol.

The NRC is actively reviewing the remaining open issues associated with industry's proposed changes in evaluating voltage-based steam generator tube repair criteria established per the guidance in Generic Letter 95-05 (items 4 through 7 listed previously). The NRC will inform you of our conclusions upon completion of this review. If you have any questions with regard to our evaluation of the SGDSM Database or the status of our review, please contact me at (301)415-2795.

Sincerely, Gus C. Lainas, Acting Director l

Division of Engineering Office of Nuclear Reactor Regulatian l

U.S. Nuclear Regulatory Commission PROJECT NO.: 689 i

cc: C. Callaway, NEl Distribution:

l File Center PUBLIC EMCB RF BSheron GLainas JStrosnider SMagruder CBeardslee SCoffin l

AKeim JTsao l

DOCUMENT NAME: G:\\ RUSH \\LETR-NEl.WPD MI // /

[y/f6 INDICATE IN BOX: "C"= COPY W/O ATTACHMENT! ENCLOSURE,"E"= COPY WIATT/ ENCL,"N"=NO COPY OFFICE EMCB:DE EMCB:DE EMCB:DE (A)D:DE NAME PJRush' ELMurphy*

EJSullivan*

GClainas DATE 10/28/98 10/30/98 11/09/98 11 /

/98 OFFICIAL RECORD COPY 1

l 3

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In order to minimize the potential of eliminating data that exhibit scatter consistent with the uncertainties implicit in mechanistic models, the staff recommended by letter dated l

February 26,1908, that the industry consider utilizing a more stringent confidence interval than typically accepted by the NRC (99% versus 95%). The use of a larger confidence interval should minimize the exclusion of data due to expected scatter within each leak rate model as l

additional data are incorporated into the leak rate database. The NRC has concluded that data exclusion Criterion 3a is acceptable as proposed in Addendum 2. In addition, leak rate data from specimens MB598-3 and MB604-2 should be appropriately deleted in accordance with Criterion 3a from the database for %-inch tubing.

Addendum 2 included modifications to the %-inch diameter tubing database based on the presumption of NRC acceptance of a proposed reduction in leak rate for Tube R28C41 (Plant S) and data exclusion Criterion 3a. As indicated previously, these changes are acceptable. In addition, new pulled steam generator tube data was added to the %-inch tube database as documented in Addendum 2. The staff has reviewed these data and concluded that the modifications to the database and the correlations with these data are appropriate. Therefore, the NRC concludes that the database for %-inch diameter steam generator tubing and the associated correlations for burst, probability of leak, and leak rate included in Addendum 2 are acceptable. Future changes to the SGDSM Catabase involving the inclusion or exclusion of pulled tube data (% and %-inch diameter tubing) should be made in accordance with the database protocol.

i The NRC is actively reviewing the remaining open issues associated with industry's proposed

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changes in evaluating voltage based steam generator tube repair criteria establit hed per the guidance in Generic Letter 95-05 (items 4 through 7 listed previously). The NRC will inform you of our conclusions upon completion of this review. If you have any questions with regard to our evaluation of the SGDSM Database or the status of our review, please contact me at (301)415-2795.

Sincere!y, Edmund J. Sullivan, Acting Chief Materials and Chemical Engineering Branch l

Division of Engineering l

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission PROJECT NO.: 689 cc: C. Callaway, NEl Distribution:

File Center PUBLIC EMCB RF BSheron GLainas JStrosnider SMagruder CBeardslee SCoffin AKeim JTsao DOCUMENT NAME: G:\\ RUSH \\LETR-NEl.WPD

• See Previous Concurrence To receive a copy of this document, indicate in the box:

"C" = Copy without attachment /encloture

'E" = Copy with ettachment/ enclosure "N* = No copy OFFICE DE:EMCB lC DE:EMCB l

DE:EMCB C

4 NAME PJRush' ELMurphy" EJSullivan GA6 U

DATE 10/28/98 10/30/98 11 A /98 OFFICIAL RECORD COPY

3 in order to minimize the potential of eliminating data that exhibit scatter consistent with the uncertainties implicit in mechanistic models, the staff recommended by letter dated February 26,1998, that the industry consider utilizing a more stringent confidence interval than typically accepted by the NRC (99% versus 95%). The use of a larger confidence interval should minimize the exclusion of data due to expected scatter within each leak rate model as additional data are incorporated into the leak rate database. The NRC has concluded that data exclusion Criterion 3a is acceptable as proposed in Addendum 2. In addition, leak rate data from specimens MB598-3 and MB604-2 should be appropriately deleted in accordance with Criterion 3a from the database for %-inch tubing.

4 Addendum 2 included modifications to the %-inch diameter tubing database based on the presumption of NRC acceptance of a proposed reduction in leak rate for Tube R28C41 (Plant S) and data exclusion Criterion 3a. As indicated previously, these changes are acceptable. In addition, new pulled steam generator tube data was added to the %-inch tube database as documented in Addendum 2. The staff has reviewed these data and concluded that the modifications to the database and the correlations with these data are appropriate. Therefore, l

the NRC concludes that the database for %-inch diameter steam generator tubing and the associated correlations for burst, probability of leak, and leak rate included in Addendum 2 are acceptable. Future changes to the SGDSM Database involving the inclusion or exclusion of i

pulled tube data (% and %-inch diameter tubing) should be made in accordance with the j

database protocol.

At this time the NRC is continuing its review of the remaining open issues assoc ated with industry's proposed changes in evaluating voltage-based steam generator tube repair criteria established per the guidance in Generic Letter 95-05 (items 4 through 7 listed previously). The j

NRC willinform you of our conclusions upon completion of this review. If you have any questions with regard to our evaluation of the SGDSM Database or the status of our review, i

please contact me at (301) 415-2795.

Sincerely, i

Edmund J. Sullivan, Acting Chief Materials and Chemical Engineering Branch

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Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission PROJECT NO.: 689 cc: C. Callaway, NEl Distribution:

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PUBLIC EMCB RF BSheron GLainas JStrosnider SMagruder CBeardstee SCoffin AKeim JTsao DOCUMENT NAME: G:\\ RUSH \\LETR-NEl.WPD Ta receive a copy of this document, indicate in the bor

• C' = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No co7y OFFICE DE:EMCB lC D#$4/JCE DE:EMCB l

l NAME PJRush/[//A EbAu/pdy EJSullivan DATE 10/Dr/98" 10M/98 10/ /98 OFFICIAL RECORD COPY

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'j Distribution: Letter to David Modeen. Dated:

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