ML20148K019

Summary of Subcommittee on Sequoyah Nuclear Plant 800709 Meeting W/Westinghouse Re Status of Open Review Items, Hydrogen Control & Implementation of near-term OL Requirements

ML20148K019
Person / Time
Site:	Sequoyah
Issue date:	10/10/1980
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-1768, NUDOCS 8012020033
Download: ML20148K019 (24)

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' . 4 MINUTES OF THE SEQUOYAH SUBCOMMITTEE '

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l MEETING g@ JULY 9. 1980 //d Q / g

]* 4) Y g WASHINGTON,DC gB@p The ACRS Sub^ committee on the Sequoyah Nuclear Plant, Units 1 and 2 met on July 9, 1980, to discuss the Tennessee Valley Authority (TVA) application for a license to operate the Sequoyah Nuclear Plant, Units 1 and 2 at full power. Presentations were given by the NRC Staff, TVA, and the Westinghouse Electric Corporation. The principal topics which were discussed were the status of the NRC Staff's open review items, the implementation of the NT0L requirements, hydrogen control, and the Applicant's and NRC Staff's work on risk assessment for ice condensers and the applicability of filtered vented containment. The Subcommittee also discussed the differing professional opinion of the adequacy of the review of weld repairs made to the pressurizer relief pipe on. Unit 1. Notice of this meeting was published in the Federal Register on June 24, 1980. A copy of this notice is included as Attachment A and a list of attendees is included as Attachment C. Portions of the materia'l provided to the Subcommittee at this meeting are included as Attachment D. The complete set of mate. rial provided 'to the Subcommittee is in the ACRS files.

No oral statements were given by members of the public nor were there any requests for time to make oral statements. No written statements were submitted. ,

A transcript was kept of the Subcorriittee meeting proceedings.

The ACRS members in attendance were Dr. J. Carson Mark, Subcommittee Chairman, and lir. W. Mathis. The ACRS consultants present were Dr. I. Catton, Dr. W.

Lipinski, and Dr. Z. Zudann and Dr. R. Savio of the ACRS Staff was also present. Dr. Savio was the Designated Federal Employee for this meeting. The entire meeting was held in open session. _ _ _ , _ _ ,__

TUh M W U MS STATUS OF THE NRC REVIEW - C. Stahle, NRC Staff p ,

Mr.Stahlesummarizedthestatusoftheopenreviewihsues-currentlyundercon-sideration. The Staff has listed 13 full-power non-TMI issues and 40 full-power TM1 related issues for Sequoyah. Actions on 8 of the non-TMI issues and on 15 of the TMI related items are essentially cmplete. Thirteen of the TMI related items are to be resolved by agreed upon dates and one of these items (degraded core) will be resolved in rulemaking. Actions on all remaining incomplete items are expected to be completed within the next two to three weeks. Listings of these items are on pages 1-3 of Attachment D.

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Sequoyah July 9, 1980 <

Mr. Stable noted that there was-the difference of professional opinion on the part Mr. John Halapatz of the NRC Staff regarding the adequacy of the Sequoyah Unit 1 weld repair of the pressurized relief pipe. The pressurizer relief pipe.had been bent during the hot functional test as a rescit of a failure of the relief pipe to slide through a pipe hanger support. The pipe r

in question is fabricated from six-inch 31655 pipe. The pipe was repaired by ,

grinding two 270 degree, two-thirus thickness depth grooves in the pipe '

opposite to and straddling the bent section. The grooves were filled with weld material, reground to remove the weld material and then filled a second time with weld material. Weld metal shrinkage provided the forces to straighten the affected section of the pipe. The pipe was not subjected to a hydrostatic test after the weld repair. ASME codes would have required that the test be per-formed had the pipe section been penetrated. Mr. Halapatz's concerns were related to the" lack of hydrostatic testing and the possibility that the pipe material in the area of the weld may have been sensitized. The matter was discussed at some length. It was generally agreed that, in view of the degree of inspection the weld area had received, a hydrostatic test of the system would not increase the assurance that the system would perform as designed. The issue of pipe metal sensitization was not, however, resolved. Construction of and examination of a prototypical mock-up and third party inspection were sug- -

gested as ways of resolving the. dispute as to the adequacy of the pipe system after the repair. Mr. Halapatz has summarized his concerns in the doc'ument included as Attachment F.

FLOOD PROTECT 10tl - M. Burzynski, TVA Mr. Burzynski indicated that the probable maximum flood for the Sequoyah plant was-based on a three day storm, occurring over the 21,400 sq. mile Of water-shed, with a total rain f all of 16.8 inches and preceded by a thrae day storm occurring three days earlier with a total rainfall of 6.7 inches. Th'e seismic design base flood was based on the failure of four upstream dams coincident with a flood crest equal to that of one-half the probable maximum flood. The flood protection plan calls for the immediate controlled shutdown and cooldown, the call-up of additional personnel, and the activation of the diesel generator system. These steps would be taken when the flood warning is received. Follow- j ing this the flood protection plan calls for using the high pressure fire protection system as a replac'ement for the auxiliary feedwater system, replacing I

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. Sequoyah July 9, 1980 I

component cooling water with the essential raw coolant water, filling radwaste ,

tanks to prevent flotation, sealing drains to the diesel generator and emer-gency raw coolant water buildings, and disconnecting all batteries situ'ated below the design base flood level. The flood warning system provides a mini-  ;

mumof24h$ursnoticebeforetheflocJcrestreachestheplant. This would allow implementation of all parts of the flood protection plan.

SINGLE UNIT /TWO UNIT DESIGN CHANGES - D. Williams, TVA Mr. Williams summarized the design changes which would be implemented in the Sequoyah plant before two unit operation. TVA was required tp expand their ,

service water heat dissipation facility at Sequoyah as a result of the passage of the Federal Water Pollution Centrol Act of 1972. New natural draft cooling towers were added to allow full power operation under warm weather conditions.

The design of these cooling towers is such that they discharge into the condenser water intake pumping station. Design studies were performed which show that under certain conditions,- the discharge from the natural cooling towers would exceed the design temperature of the ERCW pump. A decision was made to build a new ERCW pumping station. The present pumping station is adequate to accom-modate the operation of Unit 1. The new pumping station will be for the operation of Unit 2. In addition, an interim auxiliary building secondary containment enclosure has been added to provide a more effective barrier to airborne con-taminants. This barrier will no longer be needed after the completion of the construction associated with Unit 2.

STATUS OF THE LOW POWER TEST PROGRAM - C. Stahle, NRC Mr. Stable indicated that the review of the low power test program was essen-tially completed and that approval of the program was expected by July 11, 1980. l A similar program has been approved for North Anna Unit 2 and is currently underway.

REACTOR VESSEL N0ZZLE UNDER CLAD CRACKING - E.. Toddy, Westinghouse Mr. Toddy indicated that the concerns relative to the existence of cracking in the Sequoyah reactor vessel nozzles were related to the discovery by the Westinghouse-French licensee of the cracking in pressure vessels manufactured in Europe. The cracking is believed to be hydrogen-induced and as a result of the welding process / heat treatment used in applying the cladding. The l

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Sequoyah July 9, 1980 Sequoyah Unit 1 nozzle was manufactured in Rotterdam and was accordingly inspected. Underclad cracking was found in the vessel nozzles. All cracks were below the designated ASME critical flaw size but were quite numerous in some sections of the nozzles. The Prairie Island Unit 1 vessel nozzles were clad using similar techniques. This vessel will be inspected in about six months at the refueling outage. The vessel has been in operation for about nine years. Infor- ,

mation on the existence of underclad cracking and propagation should be obtained from this inspection.

STATUS REPORT ON ICE CONDENSER RISK ASSESSMENT STUDIES - R. Christe, TVA Mr. Christe indicated that there were four programs currently underway. The ,

Systems Interaction Methodology Applications Program being conducted at Sandia under NRC sponsorship is using the Watts Bar plant for the l study and is similar to the Sequoyah plant. It was concluded that the facility f was well protected against interactions which were considered within the scope  !

of the study.

l The Reactor Safety Study Methodology Applications Program is being conducted by RES-PAS. The objective is to determine dominant accident sequences using the methodology developed in WASH-1400. The plants being studied are Sequoyah, Grand Gulf, Calvert Cliffs, and Oconee. The Sequoyah ice condenser study is not yet completed. The retults at this point indicate that ice condenser ptants have at different domina'nt accident sequences but the risk associated with the plant is similar to-what would be expected in larger dry containment plants.

TVA has contracted with Kaman Sciences, Inc. for the performance of the reliability evaluation of the Sequoyah Unit 1 auxiliary feedwater system. The G0 code was employed in this analysis. The GO code had been used extensively for defense applications studies and is' oriented to success tree rath'er than fault tree sequences. The results indicated that the probability of success-fully starting an auxiliary feedwater system upon demand and providing adequate water fluid pressure to at least two steam generators was 0.99999. In a loss-of-offsite power with diesel generators and battery backup available, the auxiliary feedwater system supply success probability was calculated at 0.99997.

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July 9, 1980 Sequoyah In addition, TVA intends to use the GO methodology to assess plant availa-bility and plant safety. Phase 1 of this work is to be conducted with simplified plant models. it was initiated on July 1, 1980 and is expected to f be completed'by December 31, 1980. Phase 2 which involved expansions of the simplified model and the application of improved reliability data will be initiated on January 1, 1981 and will be completed within a year.

STATUS OF PAS WORK ON ICE CONDENSER RISK ASSESSMENT - M. Taylor, NRC/ PAS Mr. Taylor described the ice condenser reliability study being conducted within the Reactor Safety Study Methodology Applications Program. This program l involved a reassessment of the WASH-1400 work to obtain an improved baseline for  ;

the WASH-1400 model plants and studies more current LWR designs, one of which was the Sequoyah ice condenser. The estimated probability of severe core damage resulting in these studies is given on Table 1 in Attachment D. It is noted that the risk associated with the Sequoyah plant is comparable to that asse-ciated with the Surry and Peach Bottom WASH-1400 plants. Site characteristics are incorporated in.these-estimates. The Indian Point studies involved risk estimates for the Indian Point site with various reactor types. A sumary of these results is given on pages 4-5 of Attachment D. Mr. Taylor noted that the ice condenser failure most affecting risk is the overpressure failure and _

that, while the ice condenser dominant risk sequences are somewhat different from the WASH-1400 PWR and PWR designs, the overall risk seems to be comparable within the uncertainties of the analysis.

STATUS ON HYDROGEN CONTROL STUDIES - G. Oilworth, TVA and W. Butler, NRC Mr. Dilworth sumarized the work that TVA had done on hydrogen control in ice condensers. Mr. Dilworth noted that the studies had been conducted ever the past nine month period and had concluded that the current Sequoyah design can withstand substantial amounts of hydrogen above the design basis. The maximum containment strengths of 33 psig and 42.5 psig, based respectively on the yield and ultimate strengths, have been caluclated. TVA has concluded that the present Sequoyah design can withstand the hydrogen generated at 25% metal-water reaction of the ultimate strength is used as the f ailure criteria. The l assumptions are that the hydrogen is generated and released into the contain-ment and that a fast burn (5 to 30 seconds) occurs and that no shock wave is associated with the hydrogen burn. This would be similar to the postulated i

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Sequoyah July 9, 1980 TMI-2 scenario. A number of concepts for providing additional protection against hydrogen burns were studied. . Vented containments using filtered release, additional containment shells, and coupled containment structures (Unit 1 and Unit 2) were studied. Controlled combustion (ignition sources) and means for preventing combustion (nitrogen inerting and halon suppression) were also studied.

Filtered vented containments were judged not to be effective for rapid pressure transients and to have associated with them a risk of unnecessarily releasjng contamination. The coupled containment concept studied was that of connecting the containments for Unit 1 and Unit 2 through a vaived tunnel. These were judged not to be effective for rapid pressure transients and to have the potenfialfordegradingthesafetyofthesecondunitintheeventofanucident.

Controlledcombustionsourceswerejudgedtohavethehighestp6IEntialfor reducing the risk from hydrogen during most accidents leading to clad oxida-tion. The system also has the advantage of having a moderate initial cost and low operating and maintenance costs. Nitrogen inerting was judged to be impractical for an ice condenser containment. The containments are small and require fre-

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quent entry for inspection and maintenance duties. Backfits which would elimi-nate the need for frequent entry would be difficult and expensive. Halon sup-pression was judged to be potentially effective in preventing hydrogen combustion.

The decomposition products, however, may degrade the long term operability of the plant equipment under accident conditions. The most promising concepts for enhancing the design capability for hydrogen control were ignition sources and halon suppression.

TVA proposed that a distributed ignition source system be installed as an l interim hydrogen control measure. Programs would be continued which would be directed at improving the distributed ignition source system and to further the applicability of halon suppression. Mr. Dilworth indicated that the dis- l tributed ignition source system had the potential for controlling hydrogen up to 70% metal-water reaction. It is expected that the core could not be l contained within the pressure vessel beyond this poitit.

P Mr. Butler indicated that the Staff would review the proposed ignition system.

TVA would be permitted to proceed with the installation of the system but not L6 operate it until NRC concurrence was obtained. It is expected that the l

Sequoyah July 9, 1980 t

installation of this' system and the NRC review would take about three months, in addition, Mr. Butler indicated that the NRR had initiated a user's request for a safety research program directed to evaluating systems for mitigation of degraded core / core melt accidents. This program would be directed toward the development of information on mitigation systems for all LWR containments for use in the upcoming rulemaking proceeding. The short term program (6-12 months) would be directed toward evaluating systems for the mitigation of degraded core accidents in ice condensers and Mark III containments. Hydrogen generation rates for degraded core accidents would be evaluated and containment response during a hydrogen burn would be determined. The effectiveness of various hydrogen control systems under these conditions would be studied.

The long term program (2 years) would be directed towards the study of venting systems and advanced hydrogen r]ntrol systems for all LWR containments. Dis-tributed ignition sources, large thermal recombiners, halon systems, inerting systems, water fog systems, large catalytic recombiners, and oxygen .cavenging systems would be studied.

STATUS REPCRT ON FILTERED / VENTED CONTAINMENT SYSTEMS - J. Meyer, NRC/NL Mr. Meyer described the Staff's current studies on filtered vented containemnt systems. These studies were directed toward the Zion and Indian Point plants. .-

Conceptual designs for filtered vented containments have been studied. These involve venting to water suppression pools with the possible addition of follow- ,

on sand and gravel and charcoal /HEPA filters. Systems would not be able to cope with all ranges of hydrogen release unless very large vent lines were used.

Vent lines of about a three foot diameter would have the capability of dealing with most accident sequences.

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l NOTE: For additional details, a complete transcript of the meeting is '

available in the NRC Public Document Room, 1717 H Street, N.W., -

Washington, D.C. 20555 or from the Alderson Reporting Co., Inc.,

300 7th Street, 5.w., Reporters Building, Washington, D.C. 20024, (202/554-2345).

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24, 1980 / Notices Federal Register / Vol 45. No. u3 / Tuesday, June -

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C432 =- Buildine. Commonwealth and Walnut Di During the initial portion of the Streets. Harrisburg. Pennsylvan:s 17128 ed during the balance of the meeting. the Subcommittee, along withand the York Co!!ege of Pennsylvania, C4

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8 any of its consultants who may be Country Club Road. York. Pennsylvania g7 '

l e 'lir, subcommittee win then hear prernt, will exchar preliminary 17405. A copy will also be filed with the presIntations by and hold discussionsviews regarding matters to be Secretary for the Commission's review CL with representatives Whe NRC Staff. considered durmg the b6!ance of the in accordance with 10 CFR 2.206(c). AsSr Re their consultants and other interested sneetint prodded in to CE R 2.206(c) this decision I persons, The Subcomfnittee will then hear wiu become the final action of the ac Further information regarding topics Commission twenty days representattves after issuance to be discussed. whether the meeting p enentations by and hold discussionswith of De the NR has bien cancelled or rescheduled, the the Tennessee Valley Authority (TVA). unless the Commissic- Cects to review ac Chairm:n's ruling on requests for the their consultants, and other tmerestsd the decision on its owt - ation within su bt tb** pn oppo-tunity to present oral statemema Persons. Dated at Bethesda.Marytand, this 13th day an.

Further information regarding topics cnd the time allotted therefor canbe obtained by a prepaid telephone call to to be discussed, whether t$e meetmg oGune. teen sus the cogninnt Designated Federal has been cancelled or rescheduled. Forthethe Nuclear Regulatory Commission- Boi Employee. Mr. Peter Tam (telephone Chairman's ruling on requests for the gd ae c, caea.

2C2/634-14131 between 8d5 a.m. yortunity and to resent oral strementa g ;; ~ r*~~ " - a' t e

$40 p.m.. E!JT.

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or#:#y'lp,"*2C"c'*u T io ' ten ma Iy;dyY the cognirant Designated Federal W **" c7 Employee, Dr. Richard Savio (telephone ause ccos mww c Ant Advisory Committee. Monogement Oficer.

202/634-3267) between 8:15 a.m. and am.

tra om mam ms.am a er etat ammo caos mwM -

5:00 p.m EST or dEDT.Ihave l datermined.in accordance with DEP ma<ARTM S

./ ubsection 10(d) of the Fe eraAdvisory FederalRanrood Committee Admirustretion Act, that it mayRe: [px be AcMsory Commf ttee on Reactor necessary to close some portions of this Safegus+da, Subcommittee on the Oct meeting to prokct proprietary Public MeeUng To Out!!ne and Discuse Sequoyah Nuclear Plant; Westing informadon. The authonty for such Proposed Guidelines and Procedures a

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%e ACRS Subcotranittee on the closure is Exemption (4) to the Sunshfne Regerding Rock taland Railroad and ,

Sequoyah Nuclear Plant wiU hold a A ct. 5 U.S.C. 552b(cj(4). Employat Asalstance Act and gj) rnating on July 9.1980, in Room 104a. Milwaukee Railroad Restructuring Act

' H St., NW. Washington, DC to Ded kne ta tono. > ras bha C. Hope, On Wednesday. June 25.19e0, at 10:00 w the Tennessee Valley Authority gvi

( a e tir.g wa Adm ry Comautsee.Managesnest ofton pommamsmsmaww am the Faderal Railroad Adminutration (FRA) will hold a

' DS t51 a t. No ce of meeting in Room 8334 of the Nassif pubbshed June 20,1980. ammoe a mm*w  ;

8">P1 in recordance with the procedures Building. 400 7th Street. Sauthwest. Bac)

['.ccW. No. 50421 Washington. D.C. to oudne and discuac  ;

outlined in the Federal Register on g October 1.1E'9. (44 FR 56408), oral or Wetropontan Edtson Co. (Three Mthe the propred guidelines and procedures ,,,

wnden statements may be presented bylaiand Nuclear Statlon. Unit 2); . to be lasued by the Capartmer.1 of ' enge members of the public. recordings will lasuance of Director's Decision Under Transportation (DOT). under which the of ,

be permitted only during thsse portions 10 CFR 2.206 Public may submit applicaffons for bt of the meeting when a transcript is being On September 14.1979, a notice was directed service under section 104 of satis the kept. and questions may be asked onlypublished in the Federal Register that aRock taland Railroad Tr6nsition and! bent 1 by members of the Subconunittee. Its petitan by the Anti.Noclear Geovp Employee Assistance Act(Pub.1.96-consultants. and Staff. Persons desiring } Mets Representing York (ANCRY) was being 254) and section 18 of the Milwaukee into to make oral statements should notdy Railroad Restructuring Act (45 U.S.C.

the Designated Federal Employee as farconsidered under to CFR :.20s.

916) and the proposed criteria which tha ! duth l wan ANCRTs petition requested that ma in advance as practicable so that Commission prepare an environmental Department will use to evaluate those "

refin l appropriate arrangements can 1,e made impact statement concerning the venting applications.  ! bear.

to allow the necessary time durms the of radioactive gases from the reactae ne meeting is open to the pubtle, mieting for such statementa.

l ortgu buildng of the Three Mile island Nuclear including intemted states and  ! as at The entire meeting wdl be cpen to Station. Unit 2. Because this action wdl organizations who are considering l wher public attendance except for those not cause any significant environmental applying for directed service mder Pub.; suthi snsions during whach the Subcommittee impact. it has been determined nut to ( ofthe finds it necenary to discuss proprietary prepare an environmental Wpact W 254. I lasund in Washington. D.C on Jane 24 l Is au-information. One or more closed statement. Accordingly. ANCRTs l the T l sessions may be necessary to discuss petition is denied. 1M " cond suchinformation.(Sunahme Act A copy of the formal decision denying !&haelT.Haley form Exemption 4.)To the extent practicable.the petition is available for inspection in Acting Chef C.aun.et , andr these closed sessions will be held so theasCommission's Public Document rra a= m.mn m.m m e.g l presc to minimize inconvenience to members Room at 1717 H Street NW a us,a co e . a f Decis of the public in attendance. Washington. D.C. 20555 and in the local 972, De agenda for subject meeting shall public document rooms at the State i Thi e as follows: Library of Pennsylvania (Covernment provs Wednesday. July 9.1960. 8:30 a.m. Publications Section), Education until the conclusion ofbusiness. i L

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DRATT AGENDA ARSR PRESENTATIC?4 TO ACRS WG-6 ON JULY 9, 1980 l

ff:30 - 8:40 EXECUTIVE SESSION 8:40 - 9:00 INTRODUCTION - C. KELBER, NRC 9:00 - 9:30 REACTOR SAFETY MODELING AND ASSESSMENT - H!JMMEL, ANL 9:30 - 10:00 3-D CODE DEVELOPMENT - SHA, ANL 10:00 - 10:45 SSC CODE DEVELOPMENT AND IESTING - GUPPY, BNL BREAK (10 MINUTES) 10:55 - 11:15 THERMALHYDRAULIC LMFBR SAFETY EXPERIMENTS - GINSBURG, BNL 11:15 - 11:40 AEnoSOL MEASUREFE::TS AND MODELING FOR FAST REACTOR SAFETY - GIESEKE, BCL 11:40 - 12:30 AEROSOL RELEASE AND IRANSPORT FROM LMFBR FUEL - KRESS, ORNL P

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r-INCOTLETE MON-TMI ISSUES ON SE000YN1 UNIT NO.1

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l. SEISMIC AUDIT pen ACRS LETTER 8. ATWS - REVIEW AND APPROVE PERATING PROCEDURES

2. POSITION REQUIRED REGARDING l

FOUDATION MONITORING ON SETTLEMENT 9. COMPLIANCE OF IE BULLETIN

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3. POSITION REQUIRED ON CONTAIMMENT INSTRUMENTATION 1 CONTROL ROon gg gg, SYSTEM 90 RING 9PERATION 14 . ECCS EVALUATION MoDEL CONCERNING FUEL CLAD SWELLING COMPLIANCE WITH R.G. 1.193

5. POSITION REQUIRED REGARDING PROCESS AND NUREG/CR-0560

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CcNTROL PROGRAM TOPICAL REPORTS WCAP-9225, 11.

S. Eau!P. QUALIFICATIONS COMPLY MITH 9230 AND 9236 RELATED TO THE GUIDELINES OF 'IUREG-0538 %in STEAM & FEEDLINE BREAK ACCIDENTS

7. PAD 3-3 PERFORMANCE CODE - COMPLETE EVALUATION REGARDING RESTRICTION IN \ 12. Q-LIST COMPLETE REVIEW OF THE USE OF THIS CODE 'Q-l!ST" REQUIREMENTS

13. COMPLIANCE OF OIE BULLETIN 80-05 RELATED TO 3Y-Pass.

OVERRIDE, RESET CIRCUlT3

_ _ _ _ _ _ _ _ _ _ _ _ ______ ______ _ ____________-______m-- _ _ _ _ _ -_ _ _ m __________________________ _ _ _

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~ INCOMPLETE (FULL-POWER) TMI ISSUES ON SEQUDYAH UNIT NO. 1 i

1. SHIFT TECH ADVISOR -1/81 -lfi.~~COMTAINMENT DEDICATED PENETRATION - 1/81 .

2. IMMED. UPGRADE OF SRO & R0 90AL.-8/SO 17. CONTAINMENT ISOLATION DEPENDABILITY

18. ADD. ACC, MONITORING INSTRUMENTATION -

3. ADMIN. OF IRAINING PROGRAM FOR

! LICENSING EXAMS - 8/80 1/81

/ 4. REV. SCOPE & CRITIERA FOR UORMAL 19. INADEQUATE CORE COOLING INSTRUMENTS- 1/81 LICENSING EXAMS - 8/80 OF B&O TASK FORCE

20. FINAL RECOM

5. REV. SCOPE & CRITERIA FOR SIMUL.

21. UPGRADE EMERGENCY PREPARDNESS EXAMS

22. UPGRADE EMERGENCY SUPPORT FACILITIES- 1/81

6. - PROC FOR VEIRiFICATION OF

24. COMMUNICATIONS T.

CORRECT PERF. OF OP. ACTIVITIES

7. CONTROL ROOM DESIGN REVIEW 25. IMPL. OF MRC AND FENA RESPON.

8. REACTOR COOLANT SYSTEMS VENTS- 1/81 26. OFFSITE DOSE MEASUREMENTS

9. Post-ACCIDENT SAMPLING - 1/81 27. IN-PLANT RADIATION MONITORING- 1/31

10. TRAINING FOR MITIGATING CORE DAMAGE 28. CONTROL ROOM HABITABILITY c

11. ANALYSIS OF NYDROGEN CONTROL 29. POWER - ASCENSION IEST

12. DEGRADED CORE - RULEMAAING .

13.. RELIEF AND SAFETY VALVE IEST REO.- 6/81

14. AFW RELIABILITY EVALUATION

15. AFW INITIATION AND INDICATION -

1/81

_ _ . _ _ _.. . , _ - .. _ _ . _ ~ _... _ . _ _ _ _ _ . _.. . _ _ - . . . _ . .

COMPLETE (FULL POWER) Till ISSUES Otl SEQUOYAH UNIT N0.1 ,

1. REACTOR INSPECTOR AT OPERATIllG REACTORS

2. SHORT TERM ACC. ANALYSlS AND PROC. REVISI0fl

3. NSSS VEllDDR REVIEW 0F PROC.

4. PILOT MONITCRING 0F SELECTED EMERG. PROC. FOR NTOL APP.

5. LOW POWER TESTING TPAINIt!G

6. PLANT SHIELDING

7. EMERG. POWER FOR PRESSURIZER HEATEP.S

8. PRIMARY COOLANT SOURCES OllTSIDE C0tiTAIMMENT i

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- June IL 1990 f I

Docket Nos. 50a327/328  ;

I MEMOFANDUM FOR: S. S. Pawlicki Chief Materials Engineering Branch Division of Engineering FROM: J. H sl a pat z Pateria'Is Engincuring 7. ranch  !

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SUBJECT:

EXPRES$10N OF DI'rVERIN~3 PROFESSIONAL OPINIQi IN TKE MATTER i OF THE A0EQUACY OF SE200 YAH UNIT ONE WELD CRAWBEAD REPAIR OF PRESSURIZER RELIEF PI TE The author of this memorandum, hereinaf ter referred to as the minority, herewith expresses his minority opinion in the matter of the adequacy of the  ;

weld drawbead repair of the Segacyah Unat One pressurizer relief pipe. The ,

minority expresses its dif fering professional opinion in accordance with i Stction II. A. 3.J of the mcmorandum, Strael J. Chilk to William J. Di r c ks , i dated May 1, 1980, subject, "FY 1932-66 Policy Planning and Frugram Guidance ,

l f, P P PG ) . "

Non Conformance Report NdR 54P-79-S-8 disclosed, that during the tot l functional testing of Soquoyah Unit one, 1-H0H-93 pipo support for the ,

pressurizer relief piping failed to flide in the vertical direction as the l pressurizer expanded during heatup cf the reactor coolant system. As a re r.u l t l the 6-inch, schedule 160 (nos. .718 c.all), Type 316 stainless steel '

/

pressurizer relief pipe was bent. The related safety implicatilon was that f ailure of this pi. ping could lead te, an uncontrolled blowdown of the reactor coolant system.  ;

1 As corrective actio,ns, TVA had two options. The first option vis to cut out ,

the damaged pipe and replace it. 'this option, however, .would require a sy st erm pressure test in accordance with ('77) ASME Code Section XI IWA-4400(a),

which requires that after repairs by welding on the pressure retaining j boundary that a system pressure te st be performed. The 5?eend option va s to ]

straighten the pipe by a repair procedure which would be exe .pted f rom nystem ]

hydrostatic testing. TVA, to avcid cutting out the damaged pipe, nought this j exemption through IWA-4400(b)(3), which exempts f rom hydrostatic testing repairs by welding on the pressure r<taining boundary provided that the repairs did not penetrate through the pressure boundary. ,

j Ihe corrective action used by TVA to straighten the pipe was the weld drawbead I

technique. Two 270* grooves were ground in the pipe opposite to and straddling the kink. The grooves were filled with weld metal, reground to ,

remove that weld metal, then filled a second time with weld metal. Weld metal shrinkage provided th> stressing to plastically straighten the pipe.

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J S. Su Pawl ic ki ' )

The repair was 4: cept,4 by the Meterialt Engineering 3 ranch via the remoranda1, Tawlitki to R denstern, dat d De :e-bor 4, 1979, 5abj e:t , j

"' Tennessee Valley Authority, ferf aoyah H :loar Unit No. 1.* TY A j u st i f ! >.1 t he ,

exemption from hydrostatic testi.ng of the syst.em af t.er t.be rep 4 4r vr. tN ! eti s

'of T4A-4400lb) {1), claiming th4t the process of welding to rr slign (be pips did not , result in teletration raf the relector ecolant loundary. The minority challengt d acceptance e,f t.he vepair on the basis that more inf ormalim was needed.

1 The menorar. dun, Gusta f son to rawlicki, dated January 25, 1960, vubject, " Tr i p i Re port cf Vir,it to Tennessen valley Authority Saquo/sh Nuclear Plant, Unit-1,"

which reported.on a visit to the Sequoyah site, f ound the 'r epair acceptabic.

The ninarity, alter review of this memorandu.n'and do:umentation related ,

t her eto, re:omr?cnded in the memorandum, Hilapatz to Pawlichi, dated Febr uary 77, 1980, tubj ec t , "Segunyah Unit One Weld Drawbead Realignment of 6" )

Pre s s u ri zett Rrtlief Pipe , that the Materials Engineering Branch def er l acceptance of the repair! pending the develop ent and review of adiitiesal l information. The minority was then aivisad by his assistant dir ector that he 1 was to personally examlne the sold mo:Lup used to qualify the r<rpsir ehich had  ;

been mide. The inemorundum, Pawlicki to Rabenstein, dated rebruiry 25, 1930, subject, " Tenner.see Valey Aathority, Sequoyah 1:u:lcar Plant, Uait No. 1, '

Fe a l i gnmen t, of Pressurizer Relief Pipe," then reiterated accept.ance of the repair ar.d recomen+sd that the minority meet with TVA per sonnel an! exa. tine metallographic 52 p7es. On turch 5 and 6, 19B0, the minority visit ed TVA at .

Knoxvill'a and performed a ' metallurgical examination tof the meekup ascad for the qualification of the weld drawbead realignment of the Segooyah Unit one press,urizer re11cf pipe. Metallographic evidence was docatented which shosed

, that the nockup weld was fully penetrated. Dall penetr ation of t.he m3ckup weld, whieb was 1.upposed to represent t he d e l.d ropa i r o f the da* 0)(*3 pressuriser relief piper, obveusly did not desanstrate compliance with l fiection XI IWA-4400(b)(3). This finding, in itrcit., provided caurie for donial

• of exemption f. rom hydrostatic testing of TVA's weld drawboad rarpair of the -

l pressurizer 're11ef pipe which had been made. Other inconsistencies were noted l between the mtsekup and the a,ctual relief pipe. For ex.s9ple, a diffurent material was used in the mockup. Fur ther, while the mcckup had only one weld groove, the < actual relief pipe repaitt used two weld groove 9. In aidition, mets 11ographic evide net was document ed which r.how ed throug a-wall sensitization to a r.ignificant degrete, indicating that a potential through-wall crack propagation path existed. Since the propagation of cracks through the pipe wall is the essential concern with r espect to the integrity of the reactor j

coolant boundary, it is the minority opinion that intergranular corro sion tests which would expase to the test environment specimer.s which represent the j through-wall microstructure should be perfonned. Howev ett , only tests of ID l specime, surfaces wer e performed.

l l

Given that the mockup wold wa s fully Ier.ctrated, the minority concluded that TVA h/ad not quallfled its exemption to system hydrostatic testing.

Disclo'sure of the above information let to a :treeting of TVA and NRO on !*. arch 13, 1980. It was agree) that TVA would perform Ln sita metallography to eva.luate sensitizat ion in the actual relie f pipe repair and re-radiograph the repair to determine whether or not the pressure boundary had been fully

l.

l S. S. Fawlicki .

penetrated. The examination, reportel in the mamaranium Mills to O'ceilly, dat.ed April 11, 1980, subject, "Feqaoyah Nucl. car Plant Unit t - Pr mn:.ri. car Felief Piping Support - NOR SWP 79-S Supplemental Inforr9 tion" f .S! the ,

we:ld heat af fected zone to be unsensitized and theref ore, trat ensiti:43 Lise  ;

m'etal und'erlying the weld did not encroach on the pipe ID. In allition, on t.he basis of radiographic examination of the repair, it was concluded tha t the veld did not encroach on the pipe ID, i.e., did not f ully renetrate the rear. tor coolant pressure boundary. The se results were concurred in by OIE-RII #

L in the memorandum, Murphy to Thornburg, dated April 22, 1980, subject. "MII Ecport No 50-327/80-12 Concerning In=pection Par formed to Ev al uat e Eopiir cf Sequoyah Unit 1 Pressurizer Kelief Line."

Tne minority considers that scaningful metallurgical conclusions cinnot and should not be made f rom Xerox reproiuctions of the in situ met alloj raphy, which have been made available. Given the carbon content (.052/.059%) of the pressur;ner relief pipe, the minority fin 1s it anomalous thst tb: weld hcit af fected zones did not Show some r.cnsitization, since then it is infarr-3 that the base metal at any distance from the molten weld metal essentially d u 't experience some time in the 800'T to 1500'F sensitization rango daring ,

weld cooling.

The matter of the sensitization of austenitic stainless steels is enveloped in controversy. Arguments are made that the weld drawbed.d repiir w 21 to are no

• dif ferent than adjoining full p2netrated installation welds. In the abe )

of identical metallurgical histories, hoesver, this argament is tenuous. .he minority notes the safety implication involved, viz., that failure of the repaired piping cannot be isolated, which as a consequence, could lead to an uncontrolled blowdown of the reactor coolant system. The minority is of the opinion that thir matter be examined to a much more definitive 4,3 con:!umive end. It should also be kept in mind tha t the environment , which will be -

experienced in service by the repair, will be .a calculated 0.2 ppm maximum oxygen bearing steam rather than reactor coolant water containing a residual oxygen concentration during power operationn of 0.005 ppm. SWR pipe crack experience and the lack of corrosion data on the performance of sensiti zed i austenitic stainless steel weldments in 0.2 ppm oxygen bearing stcam would suggest caution in acceptance of the Sequoyah wcld drawbead repair of the pressurizer relief pipe. The argument that PAR service experience has not identified a problem with pressurizer relief pipes is tenuous, because it is unknown how many, if any, operating plants include pressurizer relief pipes which have been repaired as has Sequoyah's. Given this uncerta inty, which the minority feels is related to the in situ metallo' graphy performed, the more l definitive laboratory examination and corrosion testing of boat samples ;urted l f rom the weld drawbead repaired Sequoyah pressurizer relief pips is proposed for consideration.

With respect to the finding that the weld repair did not full penetrate the  ;

reactor coolant boundary, it is the minority opinion that it has not been demonstrated that the radiographic technique used has the capability to develop this conclusion. While evidence that the 2T hole in an AS M No. 12 penetrameter was visible to TVA EcVel III film interpreters and OIE-RII personnel may demonstrate that det ects are not present, these criteria may not necessarily demonstrate the capability of the technique to discriminate in a

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.c S. S. pawlicki ra!iograph bet-cen sound ald 94tal and souni wrnujht basa matti unlerlying I the wold metal. The technique m ist bc . ble to provide for this distinct ion in order te confirm whether or not the weld has fully penatr 4tn! the re : tar coolant boundary. The capability of the technique could be confirse f or denied by radiographing a known fully penetrated weld and a known p rtially penetrated weld in the Same material and 05t. rving if a di= tinction can be made in film. density dif f erences in the -eld root area bet-ven weld metal and wrought base meta.

Given the controversy which sometimes attends the interpretation of examination result.t, inspection by third party is desirable. A*tontion is called to an NRC position sta ted in the nc $oran.ium, Rabenstein to rarris, dated September 12, 1979, subject, "Qaalifiestion of Inspectors. Inspoetion specialists, and Inspection Agencies for Sequoyah." The Rabenstein memorandum states the NRO position that TVA institute third part inspe: tion for the Sequoyah nuclear plant. The Rubenstein nemorandum is previded as an a ttachment to this me narandum. The minority opinion conclu les that third l party inspection is required and should be implemented in the citter of - t he acceptance of the weld drawbead repair of the Sequoyah Unit one pressurizer relief pipe.

. A q. ,

J. Iblarat.

Materials Eng Lnee Tranch Divinion of Engineering office of Nuclear Reactor Fegulation

Enclosure:

As stated .

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cc V. S. Noonan R. L. Tedesco i A. Schwencer C. E. Murphy, 'OIE-RII A. R. He r d t , CIE-RII R. M. Oamble C. Stahle P. K. Van Doorn, 01E-RII MTEB Reading File l

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1 Ol5TRIBUT10N SEP 21 lo79 Docket Files bcc: NSIC  :

NRC POR TIC 1' Local POR G S (16)

TEP.A R. Mattson  !

. L'AR-4 File D. Eisenhut l Docket Nos.: 50-327/328 0. Vassallo J. P. Knight

5. Varga L. Shao F. Williams 5. Pawlicki Mr. H. G. Parris L. Rubenstein V. Noonan i'.ana;cr of Power C. Stahle R. Garble i Tennessee Valley Authority M. Service H. Conrad ELD - C. Woodhead C. Y. Cheg SCOA Chestnut Street To er II Chattanooga, Tennessee 37401 IE (3) 5. J. Bhatt c J.-Halapatz J. M. Grant

Dear Mr. Parris:

F. 3. Litton M. Pum C. D. Sellers M. L. Bevle

SUBJECT:

QUALIFICATION OF INSPECTORS, INSPECTION SPECIALISTS, A:10 INSPECTION AGENCIES FOR SEQUOYAH In Amendment No. 61 to the Sequoyah FSAR, you stated that you will pmv(de your own independent review of the Section XI pmgram of the ASME Boiler '

and Pressure Vessel Code through the T/A central office staff in Chattancoga, Tennessee. It is TIA's policy to provide its on inspection services on the basis that T/A is a Federal agency and it is not subject to State or other non-Federal inspectors.

It is our position that TVA is not except from any of the requir:,- ents of 10 CFR Part 50, Section 50.55s(g)(4). Therefore va recuire that T/A 1 institute the third party inspection system of the Sequoyah nuclear povar plant, j

. A letter o'f compliance is revested. ,

Sincerely, i .

Original aisced byi L. $. Rubenstein, Acting Chief Light Water Reactors 3 ranch No. 4 DiYisioC of Project Managerent i

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%.....*' SEP 211373 ,l Docket Nos.: 50-327/328 Mr. H. G. Pa'rris Manager of Fower -

Tennessee Valley Authority 500A Chestnut Street Tower 11  ;

Chattanooga, Tennessee 37401  !

Dear Mr. Parris:

1

SUBJECT:

QUAllFICAT10ti 0F INSPECTORS, l'4SPECTION SPECIALISTS, AND '

INSPECTION AGENCIES FOR SEQUOYAH i

in Amendment No. 61 to the Sequoyah FSAR, you stated that you will pr vide l your onn independent review of the Section XI program of the ASME 5:iler  ;

and Fressure Vessel Code through the TVA central of fice staff in Chattanooga, j Tennessee. It is TVA's policy to provide its own inspection services on the  :

basis that TVA is a Federal agency and it is not subject to State or other .

l non-Federal inspectors. i It is our position that TVA is not exempt from any of the requirc ents of  !

10 CFR Part 50. Section 50.55a(g')(4) Therefore, we require that TVA l institute the third party inspection system of th,e Sequoyah nuclear p;wer plant.  ;

A letter of compliance is requested.  !

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incerely,  !

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.f5 '.7 mL-Q m.

L. S. fu.enstein, Acting Chief l

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Light k'ater Reactors Sranch No. 4 Division of Project Management j i

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CCs*

Itertert S. Sanger, Jr. Esq.

General Counsel -

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400 Co merce Avenue 7,'- ;

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Knoxville, Tennessee 37902 Mr. E. G. Beasley Tennessee Valley Authority '

400 Commerce Avenue- l W10Cl31 C . . , ,

Knoxville, Tennessee 37902- f, i Ilr. Michael Harding .

f Westinghouse Electric Corporation i P. O. Box 355 Pittsburgh, Pennsylvania 15230 l t

Mr.. David Lambert . . .

Tennessee Valley Authurity  ?,

400 Chestnut Street Tower 11 6  :

Chattanooga, Tennessee 37401  !"' .

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X. EARLY FATALITIES (SUPPORTIVE TREATMENT)

NOTh..TRERE ARE.1ARGE UNCERTAINTIES WITH HE. ABSOLUTE VALUES PRESENTED I .

ASSUMPTIONS: 1) INDIAN POINT SITE METEOROLOGY - 91 WEATHER SEQUENCES l

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