ML20024D797

Forwards Response to 830721 Request for Info Re BWR Stainless Steel Piping Insp.No Previously Inspected Welds Appear to Meet Sensitivity for Detection Criteria Per IE Bulletins 83-02 or 82-03

ML20024D797
Person / Time
Site:	Browns Ferry, University of Virginia
Issue date:	08/04/1983
From:	Mills L TENNESSEE VALLEY AUTHORITY
To:	Harold Denton Office of Nuclear Reactor Regulation
References
IEB-82-03, IEB-82-3, IEB-83-02, IEB-83-2, NUDOCS 8308080228
Download: ML20024D797 (10)

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TENNESSEE VALLEY AUTHORITY CHATTANCOGA, TENNESSEE 37401 400 Chestnut Street Tower II August 4, 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Denton:

In the Matter of the ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 By letter from D. G. Eisenhut to H. G. Parris dated July 21, 1983, we received a request for information regarding inspections of BWR stainless steel piping for the Browns Ferry Nuclear Plant unit 3 That request for information was made pursuant to 10 CFR 50.54(f).

Enclosed is our response to that request.

The enclosed response is based on NRC staff concurrence with TVA's ongoing repair program for Browns Ferry unit 1. That repair program-was submitted by my letter to J. P. O'Reilly dated July 18, 1983 As you know, if any problems arise regarding the acceptability of our repair program, the enclosed response may be significantly impacted. At the request of your staff, further details on planned weld overlays will be submitted in the immediate future.

-Very truly yours, TENNESSEE VALLEY AUTHORITY L. M. Mills,!4anager Nuclear Licensing

-Subscrib fore me is gdswornto d y of 11/W2 1983 Jh . W&

Notary Public My Commission Expires / G' "

' Enclosure cc: .See page 2 8308080228 830804 \

PDR ADOCK 05000259 G PDR \k

, 1983-TVA 5OTH ANNIVERSARY An Equal Opportunity Employer

2-Mr. Harold R. Denton August 4, 1983 oo (Enclosure):

U.S. Nuclear Regulatory Commission Region II

-ATTN: James P. O'Reilly, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 Mr. R. J. Clark Browns Ferry Project Manager U.S. Nuclear Regulatory Commission

, 7920 Norfolk Avenue Bethesda, Maryland 20814 i

Nuclear Regulatory Comaission Document Control Desk Washington,'D.C. -20555 l.

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i ENEOSURE

. RESPONSE '10 D. G. EISENBUT'S LEITER 10 H. G. PARRIS DATED JULY 21, 1983 REQUEST FOR INFORNATION REGARDING BWR STAINLESS STEEL PIPING INSPECTIONS BROWNS FERRY NUEEAR PLANT UNIT 3 Item 1 - -

Justification for continued operation of your f acility prior to completing the inspections described by Attachment A in view of the increased evidence of cracking' since the issuance of IE Bulletin 83-02.

_ Response A. Leak-before-break criteria is still valid.

Based on the f act that stainless steel is a very tough and ductile material, the complete f racture of piping with IGSCC-type cracking when ,

subj ected to design, seismic, or accident loads is not possible.

Another argument is based on the f act that IGSCC crack growth depends on the asymmetry of the weld sensitization and bending loads. It is an accepted f act that stainless piping can have a complete circumferential crack equal to 63 percent of the wall thickness and still meet the saf ety margins a s required by the original construction codes. Before the crack depth reaches this size, the circunferential variations of welding residual stress and material susceptibility in the heat-af f ected zone, combined with the asymmetric bending loads, will lead to asymmetric crack growth and the formation of a short throughwall crack with limited leakage.

This hypothesis is verified by extensive field experience. As an added precaution, plant instrumentation which is capable of reliably detecting such limited leakage is utilized during operation.

B. Cracks found to date f all within the crack growth models.

It is a known f act that IGSCC-type cracking takes a long time to --

initiate to a threshold level that can be detected-by ultrasonic examination techniques. Cracks in small-diameter piping (i.e., riser lines) propagate rapidly to throughwall localized leaks once crack initiation occurs. Cracks in large-diameter piping grow fairly rapidly for approximately 20-percent throughwall and then level off. Crack growth beyond 30-percent. throughwall in large-diameter piping is very j

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sl ow . All cracking' observed to date f alls within the model described I above. JTVA rej ects the idea that the cracks detected in Browns Ferry I unit 1 initiated and grew within the last couple of years. The reason l that so much cracking has been detected since IEBs 82-03 and 83-02  !

inspections began is because the number of welds inspected has been f ar j greater than that require'd' by ASNE Section II and the sensitivity of '

IEBs 82-03 and 83-02 techniques with regard to detecting IGSCC indications is much greater than conventional ASME Section 11 exuaination techniques. In addition, IGSCC indications of significant depth may have been detected using earlier ASME Section II techniques, but the maplitude f rom the indications was such that recording or evaluating further was not required by ASME Section II. Given the above factors, it is not at all surprising that the amount and extent of detected cracking have increased significe.ntly since IEBs 82-03 and 83-02 examinations have begun.

C. Analysis f or structural integrity of all cracks f ound to date has shown that an adequate margin of saf ety still existed. The reason for this is that, even with a crack 360 degrees around the pipe, the crack depth could be up to 63 percent throughwall before it would infringe on the required saf ety margins. Cracks of this magnitude and circumfer-ence have not been found and substantiate the position stated in' paragraph A.

D. Browns Ferry unit 3 has operated for a shorter period of time than unit 1_or 2; therefore, cracking, if present, should not have progressed to the point of unit 1, which still has structural margin remaining. The piping for all three units was supplied by the same i pipe f abricator; therefore, the main variable affecting crack l initiation and propagation is time of operation. At the time of shutdown for refueling, unit 3 will still have less operating time than unit 1.

E. Since a short operating time remains bef ore shutdown for refueling, the probability is extremely low that any further crack growth, if cracks are present, will progress to the point where saf ety margin is below -

that required by the original construction codes.

F. Browns Ferry units are designed and have been analyzed for large LOCA.

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' Item 2 Identify any weld inspections which appear to satisfy the sensitivity for detection specified by IE Bulletin 83-02. He information provided should include a list of these inspections, the dates of the inspsetions, the extent and results 'of those inspections, and a description of the technique or equilment used. If you have concluded that these previous inspections should influence the scope or schedule of the inspections described in Attacaent A, 'please provide the basis for your conclusion. Fur ther, describe any other unique saf ety related feature, information or action that would justify not accelerating your current test and inspection schedule in accordance with IE Bulletin 83-02.

Resnonse No previously inspected welds appear to meet the sensitivity for detection criteria specified in IEBs 83-02 or 82-03.

Item 3 Describe any special surveillance measures in eff ect or proposed for primary system leakage in addition to the current Technical Specifica tion requirements for your facility.

Resecase A technical specification amendment f or unit 3 was submitted to NRC on March 25,1983, which limits a 2 spa increase in unidentified leakage into the drywell in a 24-hour period. He requirements of this amenhent have been administrative 1y in place since July 1,1983.

The current administrative measures which are in eff ect at Browns Ferry require drywell floor drain sump monitoring to be performed at an increased

- interval of once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with a requirement to be in cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the event that leakage increases by more than 2 gpa in a 24-hour period. This is only in eff ect when the reactor is in run mode and is exempted during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the run mode following a startup.

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Item 4 Direct and indirect costs and impact, including effects on other safety related activities, of conducting the inspections described in Attachment A: (a) at a time which you would commit to conduct'the inspections consistent with Chaimaan Pall'adino's suggestion to the staff and licensees that a realistic schedule for the inspections be developed 'with the idea of accelerating the inspection as much as possible,' and (b) at the time of your aszt , scheduled refueling outage.

Resnonse The earliest date for conducting the IEB inspection on Browns Ferry unit 3 is November 11, 1983, which is based upon TVA's analysis of the best '

available time period that does not impact our major saf ety-related activities and allows f or proper preparation of work by available manpower.

- Our major concern in our analysis is completing Browns Ferry unit 1 anspections and weld repairs while completing the refueling activities and placing unit 1 back online with our present accelerated schedule. This schedule will make available the necessary manpower to be utilized for pre-shutdown planning of unit 3 outage work including preparation of IGSCC inspections.

This November 11, 1983 shutdown date coincides with our present unit 3 cycle 5 scheduled outage and agrees with our justification of continued operation as identified in item 1. As shown in attachment 1, the direct costs for this inspection are expected to be approximately $820,000 and indirect costs are commensurate with a scheduled refuel outage.

Item 3 The direct and indirect costs and impact, including effects on other saf ety related activities, of suspending operation to initiate the inspection described in Attachment A within each of three possible times: (g) 30 days.

(b) 60 days, and (c) 90 days from August 15, 1983.

Resoonse The direct and indirect costs associated with the three possible shutdown times are provided in attachment 1, and the impacts to saf ety-relat.ed activities are shown in attachment 2.

l Item 6 A discussion of the availability of qualified inspection personnel to perform the inspection described in Attachment A at your f acility for the various options in items 4 and 5, above, and the steps you have taken to obtain the services of such personnel.

Response j Availability of qualified inspection personnel for: l A. Shutdown by September 15, 1983 On September 15, 1983, WA NDE personnel will be utilized as follows:

1. Sequoyah unit 2 in-service inspection 2 supervisors 2 engineers 18 technicians

2. Browns Ferry unit 1 overlay repair inspections 1 supervisor 3 engineers 3 technicians

3. Watts Bar unit 1 preservice inspection 1 supervisor

. I engineer 4 technicians These figures are f or WA personnel and do not include contractor personnel. 'Ihis encompasses 95 percent of WA's NDE work force; consequently, WA would not be able to perform additional inspection -

work or support a contractor to perform 1EB 83-02 examinations on unit 3 at this time. Due to already accumulated radiation exposure, personnel at Watts Bar are not available to work at an operating plant.

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B. Shutdown by October 15, 1983 Overlay repairs and inspections on Browns Ferry unit 1 and the in-service inspection on Sequoyah unit 2 are scheduled to be comp 1ste by October 15, 1983. If no additional inspections or overlay repairs are required on Browns Ferry unit 1, these qualified inspection personnel would be available to begin the IEB 83-02 inspections on Browns Ferry unit 3. How ev er, as discussed in our response to item 4 above, this work cannot be fully supported without significant impact until the cycle 5_ outage which is scheduled to begin on November 11, 1983.

C. Shutdown by November 15, 1983 By November 15, 1963, adequate qualified inspection personnel would be available to perform the IEB 83-02 inspections on Browns Ferry unit 3.

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Attachasst 1 NONETARY EFFECTS 'OF SHUTDOWN OF BROWNS FERRY NUEEAR PLANT UNIT 3 FOR IGSCC INSPECTIONS Shutdown Date Shutdcwn Date Shutdown Date of Sept. 15. 1983 of Oct. 15. 1983 _of Nov. 11. 1983 Cost in. Inspection Inspection Inspection Millions . Oniv Ontaae Oniv _0utane Oniv Ontaae 3 .,

. Direct '

8 1.58 8 1.58 $ 1.58 8 1.31 81.09 i .82 Indirect 16.51 16.51 7.76 7.76 None None Total $18.09 $18.09 $ 9.34 $ 9.07 $1.09 8 .82 Comments:

1. Direct costs are those costs of actually performing the inspection such as

, manpower, contract, etc.

Indirect costs are those costs associated with an early shutdown of the unit such as fuel penalty costs and replacement power costs.

2. For a shutdown date of September 15, the total costs for only conducting IGSCC inspections versus starting the outage at same time are the same. This is because with overlapping outages (BFN unit 1, BFN unit 3, and SQN unit 2) the impact on manpower will involve contracting and hiring additional support manpow er.

3. For a shutdown date of October 15, the cost that would be realized is less compared to the September 15 costs because the unit 1 outage will be near ocupletion and involve f ever replacement power days.

4. A shutdown of November 11 is within the present scheduled outage window:

therefore, no impact on indirect costs but an impact on direct costs .will be realized from unit 3 manpower needed for refueling outage, f

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Attachasat 2 IMPACT TO SAFh"IT-RELATED ACTIVITIES

. CAUSED BY RECIRCULATION PIPING INSPECTIONS BROWNS FERRY NUCLEAR PLANT UNIT 3 l Sootember 15. 1983 Shutdown

1. - Timef rame not adequte f ae preontage planning. -

2. Coordination of personnel during refuel floor activities of cro units would be difficult.

3. Increa' sed manpower levels could degrade plant and personnel saf ety due to high level of activity in controlled areas.

4. A shortage of engineers and QA personnel would not ensure proper engineering coverage, problem resolution, and workplan completion.

5. Reduce our nuclear capacity to 40 percent (SQN unit 1 and BFN unit 2 operating). ,

6. Probability of meeting September peak demand is 86 percent.

7. September is the end of the f ourth quarter for radiation exposure.

Additional exposure could require TVA to hire and train new personnel.

_0ctober 15, 1983 Shutdown

1. Timeframe not adequate f or preoutage planning.

2. Coordination of perscanel during refuel floor activities of two units vould bi difficult.

3. A shortage of engineers and QA personnel would not ensure proper engineering coverage, problem resolution, and workplan completion.

4. Reduce our nuclear capacity to 60 percent (SQN unit 1, SQN unit 2, BFN unit 2 operating) .

November 11. 1983 Shutdown

1. Reduce our nuclear capacity to 80 percent (SQN unit 1, SQN unit 2, BFN unit 1, and BFN unit 2 operating).

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