ML20206E577

Final Response to FOIA Request for Documents.Forwards Documents Listed in App A.Documents Available in Pdr. Documents Listed in App B Totally Withheld (Ref FOIA Exemption 5)

ML20206E577
Person / Time
Issue date:	11/04/1988
From:	Grimsley D NRC
To:	Chapman N BECHTEL POWER CORP.
Shared Package
ML20205C260	List: IA-88-464, Requests That Generic Issue (GI) 119.1, Pipe Rupture Requirements & Decoupling of Seismic Loads Be Subsumed in Remaining Portion of GI B-6, Loads,Load Combinations & Stress Limits ML20205C255 ML20206E577 ML20206E597 ML20206E604 ML20207P766
References
FOIA-88-464, REF-GTECI-119, REF-GTECI-B-06, REF-GTECI-NI, REF-GTECI-PI NUDOCS 8811180121
Download: ML20206E577 (4)

Text

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Records ovbject to the request that are desenbod in the erselosed Appendices 5 tre W N h tr+# entkety or h part Mr FOtA Esernptona and foe the reanoes set forth below pursuant to 5 U $ C. 562ttd and 10 CFR 9 5 del of NRC Regulators, 1 Tte metAsks eformenon a prc54% caese/ed pwrwart to taccwthe Order 1224 (DIMPTCN 11

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sine eoene ese iron ri U.S. NVCLE AR REGULATORY COMMI55lOP

• • • • FOlA RESPONSE CONTINUATION

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Re: F01A-88-464  :

APPENDIX A RECORDS MAINTAINED IN TFE PDR UNDTR THE ABOVE REQUEST NUMBER

1. 7/13/83 Memorandum from H. Denton and R. Minogue to W. Dircks, subject:  !

Proposal for Reviewing NRC Requirement for Nuclear Power '

Plant Piping, w/ enclosure. (40pages)

2. 9/10/85 Memorandum from H. Denton to R. Minogue, subject: Schedule for .

. Resolving and Completing Generic Issue No. 119 - Piping Review Committee Recomendations, (2 pages), w/ enclosure: Prioritization d

Evaluation Generic Issue No. 119 "Piping Review Committee Recomenda tions . " (17 pages)

3. 6/9/86 Pemorandum from G. Arlotto to T. Speis, subject: Reconsnended >

Actions Regarding Decoupling of Seismic and Pipe Rupture Loads. ,

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4. 10/2/86 Memorandum to Distribution from G. Arlotto, subject:

1 Termination of Proposed Revision to SRP 3.9.3. (2 pages)

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. o Re: F01A-88-464 APPENDIX B RECORDS ~ TOTALLY WITHHELD

1. 4/29/86 Note to Bob Bosnak from Bill Shields, subject: Revision of SRP 3.9.3. (1 page) Exemption 5.

2. 5/15/86 Note to Guy Arlotto from Joe Scinto, re: SRP 3.9.3. (2 pages)

Exemption 5

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.UI.13 T3 MEMORANDUM FOR: William J. Dircks E.necutive Director for operations FRCM: Harold Dentsn Director Office of Nuclear Reactor Regulation Robert 8. Minogue, Directar j

3 office of Nuclear Regulatory Research SUBJEC*: PROPOSAL FOR REVIEWING NRC REQUIRENENTS FOR NUCLEAR MWER PLANT PIPING The proposal you requested in your May 31, 1943 memoraneJa is enclosed. We consulted with V. Stallo, T. Murley and W. Kane in the proposal aevelopment, but staff from RES and NRR were mainly responsible for its preparation. Our consultant, S. H. Bush, provided significant inputs. m of the key element.s of the proposal are as follows:

1. We propose the establistment of an NRC Piping Review Committee made up predominently of MC personnel free the various officas. With assistanca from expert consultants, the cammittee will pull toget.%r all available information inside and outside the MC and review all piping related requirinents. Four task grow s, one each dealing with pipe cracking. -

seismic design of piping, pipe breaks and other dynamic loads / load comeine-tions for piping are proposed to be established unear the NRC Piping Review Committee. Se also propose ta solicit views from irmstry and other intarosted parties in the pursuit of our objectives. .

2. We suggest one cochairman from NRR and one eschairman free RES, R. H.

Vollmer and L. C. Shan respectively, ta acaninister the NRC Piping Review Consittee. Approximataly 12 individuals from NRC will staff the MRC Piping Review Cammittee, with recrosentation from RES, NRR, IE, OEU and the Regions. Most of these will be assigned to one or more of the task groups.

Our consultant,'S. H. Bush, is our nomination for Vica Chairman. A total of 17 individuals from the NRC will be involved in the entire effort.

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3. We estiasta a period of aporoximately 12 months to complete the review and fomulata recausendattons. A four to five man year effort free approxi-antaly 17 NRC staff is planned. Consultant costs are estimated at 4Dout

$300,000.

CMTACT: L C. Shan 443-5904 1

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' W. J. Di rcks 2 g gg g 4 Our priwry deliverables will be recommendations, where appropriate, for revising the present requirements on nuclear power plant piping, and direction as to what work should be done to respond to iss,ues not currently amenable to resolution. Our proposal does not effer quantita-tiye and detailed value iasact analysas for any recommend tion, although qualitative statements regarding the cost and safety benefits that may acc ve will be included. Value impact analysis, in our view, is warranted after our recommendations are acted on, and implementation of the recom-eendations are being undertaken.

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We envision that regulatory actions regarding nuclear power plant piping currently under way will continue to be sede on a timely basis. It is strongly urged in our proposal that no such 4ctions be delayed or deferred pending the recoseencations to be dativered as a procuct of our proposal.

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Harold Centan, Director Office of Nuclear Reactor Regulation

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'd Roeert 5. Minogue, Director

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Enclosure:

Proposal O

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. FOR REVIEW 0F NRC RQUIRDOTS FOR NUC'J.AR PIPING July 7, 1983

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1 P90POSAL TABLE OF CONTENTS

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I Introduction 1 A. General lackground 1

! 5. Regulatory Issues 3 Cbjective 4 II III $ cope 4 IV Accroaca A. Organization 5

3. Plan of Atta:k 4 C. Content 9 V Car ent Status 11 A. Pipe Crack 11

5. Seismic Design 16 I

C. Pipe treak 22

0. Other Dynamic Loads and toad Comeinations 31 i

36 VI' Schedule and Deliverable VII Manpower and lodget 37 l

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I. !NTRODUCTICN A. General Backereund Several of the current twgulatory positions relating to piping design were developed without significant data. As relevant data became available, the consonatisa of some positions became apparent. Moreover, the low probability of postulated events, such u the full flow area break, 'when examined using j crocabilistic or detersinistic fracture mechanics and probabilistic risk analy-sis, was made evident. These conservative positions have resulted in a large nuncer of aessive pipe whip restraints, component supports and snumbers wnien stiffen piping systems. Since stiff piping systans generata nich thersal stresses and no::le loads and can be nors adversely influenced by construction, maintenanca anc inspection errors, many axperts believe they diminish everall safety.

Otaer area's covered by regulatory positions, such as seismic critaria pertaining to piping, load concination critaMa, constfuction of floor respense spectra and damming values, are in need of a reusessment whica could lead to revisions that world reflect a more realistic response of piping systars to faulted or axt see ac:1 cent conditions.

l White axnerimental and analytical evidence confirss the conservatise of some positions, service experience may raisa questions concerning our knowleege of the actual response of piping to failure mecaanisms suca u intargranular st sss enrrosion cracking (ICSCC) in largs gWR piping. This lattar condition is exacercated by the uncertainty associated with ersck detection and si:ing of l IGSCC in austanitic stainless steel when using conventional ultrusnic tasting precedures.

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2 An example of changing positions in the United States made on the basis of the availantlity of significant uperimental data and analytical studies is the proposec elfeination of the full flow area break as a design critarion in car-tain PwR primary systans. The greatly axpanded data base pertaining to nuclear ,j piping and its response to operating and accident conditions should permit an oejective review of existing regulatory cr taria.

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B. Reculatorv Issues ,

Four gnups of regulatory issues have been identified as indicated below:

(1) Pipe crackino due to intergranular stress corrosion has occurred rore extansively than previously forecasted in larger-diameter piping. Issues relata to inservice inspections (MOE), evaluating repair techniques (including replacement metaMals), and allowing continued operation. Pipe cracking due to j

vibrations in small-ciameter pipes and thersal fatigue are also issues, as are actions NRC may require to reduce the potential for pipe cracking.

(2) Seismic desien issues relate to pipe dascing and the fact that ,

the CBE, although not directly safety related, usually controls design because of Icwer allomaale sins: 1evels anc lower damping values. Also cJnsidered are critaMa for piping with witiple incepenoent supports, peak becacening require-sents for floor response spectra, and incustry design practices. .

(3) Pice Snak issues relata to requiring full flow area pipe breaks,

catemining break locations, and replacement critaMa.

I (4) Cartain 1 cad coseinstions, particularly the t.CCA plus SSE lead i

c:acination, recrosent a severt design mquirement leading to massive swports on piping. No studies support a causative relation betwen pipe break and earthquake, and for the particular case of the pMaary loop of a PVR, l it has teen canonstrated that earthquakes are extremely unlikely to induce a l 1

full flow ans pipe bnat. l i

Other eimasic leads treated uncer this issue include hycrocynamic j J

loads such as watar hasser and loads resulting from SRV discharge, and vibra-tional loads.

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t. - 4 lI. CBJECT1vE The ocjective of this proposed review is to evaluate cur ent piping regulatory requirements for lignt water reactor nuclear power plant design using available domestic and foreign infomation in order to provide recommendations on wtiere and how we should modify our current requirements, and to identify areas for fur *Jter action. The review will not impact on axisting ongoing regulatory actions prior to acceptanca of the final report, nor impede the resolution of any specific piping problem. However, some preliminary recommencation* approved my NRC sanagement say be utilized on a case-by-Case basis.

!!:. SCOPE

'he scoce of this review covers all safety-related piping systans and those higti energy systass wnich are important to safety in new and operating nuclear pl ants. The review will be perfomed on a system integrated basis can- .

sidering all ongoing prograss. Safety-related piping systans are defined as*

. those piping systans neeced to assure the integrity of the RCL pressure bounc-ary, ts snut oawn the reactar and maintain it in the shutdown esncition and to sitigata the consequences of accidents. (See USNRC meno datec Movencer 20, 1931, l to Distribution fras H. R. Dentan, entitled "Standard Definitions for commonly Usec 3afety Classification Terss.")

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' 5 IV. apot0ACH Nc*.:s will be given to identifiable problems arising froa present requirements and from plant eperating ax:erience. The suggested approach is to establish an NRC Piping Review Committee, gather all available information, and with the help of consultants and industry, evaluata this information for use in the ngulatory process. By conducting a review of existing NRC piping requirements, a res, ort will ,4 he precared which will present recommendations and conclusions relating to NRC nuclear power plant piping critaria.

~he organi:ational structure, the plan of attack, and the recort contents associated with this approach are described as follows:

A. Orqani:stion Our organitation attempts to einfaire tapact on axist{ng NRC programs car *ied out by NRC personnel, to maximi:e compliance with the intent of the assesament anc to yield maximum credibility to the recor*.. The following are essantial k

to tne preposed organization: r

1. the suestantial use of csnsultants to prepart position pacers, review or assess submit *.ed data, and prepart summaries for further reivew;

2. the estan11shment of an MRC Piping Review Camaitsee scorting directly to the Directar, NRA anda up predominantly of NRC personnel from NRR, RES IAE, AE00 EU, ACAS and the Regional Offices. The ED recrosentative vill sene as a consultant, and the AC15 representative will be an obsener. This l

Committee is expected to involve neout 12 individuals. There will be four i,

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tecnnical tast groups, each group responsible for a specific technical issue (see the organizational chart on page 7). About 17 individuals from NRC will be involved in the task groups and the NRC Piping Review Com-eittee. Expert consultants in the fields of seismic design, piping design, systans, fracture mechanics and metallurgy will be used to critically .

evaluate and review the proposed recossencations and conclusions; i

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3. the esordination with industry and other interested parties in order to octain inforsation on operating experience, design, construction, maintenance and inspection problems and research results. Comments on t.% proposed re-commendations and conclusions will also be sougnt from insustry. There have seen considerable efforts rotated to proposed incrovement of piping design criteria and perforsance on the part of AIF, PVRC and E7RI. .

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OstGANIZAllIWAL CilANT --

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IIRC PirlfsG REVIEW Cot 91lTTEE .

Cechairman (2)

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Vice Chairman (1) ~

l Mee6ers: NRC Staff (9) -

ne- Censultants

, . IIIOU$TRY COORDINATION

. Chairma.(1) IIRC Staff Secretary (1)

TA5K GROUP 018 FIPE CaACKS TASK GA0tr 011 SEI5MIC DESIGli TA5K GALUP 01t PIPE BREAK TA5K GROUP OII Sint OYllANIC AII0 LOAS Chairman (1) 00MBINAllell5 Members: IIRC 5Laff Same as Seea a*s -

Pipe Cracks Task Group Pipe Cracks. Task Group  % ,,

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'Spe Cracks Task Gr Censultants e

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. B. Plan of Attack .

Seve-al steps are considered necessary to develop a cefinitive nport. These are: ,

1. A clear-cut definition of the problem in each probles area should be formulated. This proposal accomplishes this in preliminary fashion.

Additional effort will be required.

2. It is recognized that current NRC decisions being made on specific plants will affect some of the issues ta he mytowed. Obvious asamoles include Generic Issues and IE tulletins. The status of the GeneHc Issues and IE Bulletins should be reviewed anc reported. NRC-funced researca impacting on regulatsr/ accisions in thesa proelen areas will aise be reviewed and reoorted. It is asstmed that consultants will be prepaMng the reviews.

wrxing with approcriata personnel within NRC. This output ultimately will become a part of an appendix adenssing the specific probles area.

3. Efforts by industry and other interested parties in spec 1fic probles areas g should be assembled and reviewed.. Since organiutions such as DRI. INPQ.

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PVRC, AIF, A/E's, MESS vendors and utilities have different interests, axper-t tise and.opirrions, it will be destraale for the NRC Piping Review Cammittee to deal with one er tw focal organizations dich reflect the vaMous wiews of the insustry and the professional societies. It is anticipataa that the NRC Pising Reviow Committee will siScuss vith ineustry grows how this ocjectiwe can be achieved. Pertinent foreign activittas in research and regulatary cevelop unt appropMate ta the proeles areas should be reviewed and assemeled.

The caseinstion of (1) regulatary efforts and (2) industry and foreign prograss i

will lead ta appendicas coverir=9 each pmbles area. Responsibility for diges-tion of the inforestion and preparation of the position pacers will be given ta selected consultants.

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A. The NRC Piping Review Committee, as an entity, would have the responsi-cility for the preparation of the final report for subsittal to the Director. NRR. They would rely on direct assistance from their consultants to facilitata report preparation.

C. Content ,

• be following presents an cuttine of the contents of the final report delineat-ing specifically what will and will not be included.

1. There will be several appendices, each addressing a given proelen. These acconcicas would be developed from a 11tarature review, operating axseHence, and ongoing programs as well as from submittals made try interested organi-tations.

2. . A hist.orical section would cover the initial recognition of the problem and

. a enronolog of regulatory 'and industry actions including a detailed

• listing of such actions appearing u tranch Tecnnical Positions, Aegulatory Guides, Regulations, "Eneric Issues, or in National Standarts and Codes such as ASMC Sections III or XI. ,

3. A section containing an assessment of the adecuacy of availacle evidence for developing accropriata conclusions and recosusencations. Current research try MAC, industry and foreign sourcas will be reviewed and future researen deemed necessary to finally resolve the proeles will be cited.

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This section and the proceding section would deal with the tecnnical issues. -

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4. The final section would consist of conclusions and recoemendations with acclication to specific plants, including legal input where required, for each problem and subprotles wnere the NRC Piping Review Committee feels it nas sufficient confidence to offer guidance. helicit stataments will be .

written for those situations where the NRC Piping Review committee feels it is unable to drew a conclusion, and the reasons why no conclusions can be sace will be identified. The felication of any recessendations made and an indication of how they affect the reliability and overall safety of piping syntans will be given. Conclusions and reconsentations should be directly related to relevant tranen Technical Pcsitions, #egulatory Guides, Regulations, anc Generic Issues,

5. De report will g contain the following:

I (a) a quantitative at4 detailed value-imact analysis for any of the MAC Piping Review Casmittee's rv ndations or conclusions; nonetheless, qualitative stataments relating to the safety and cast benefits wnich may accrue will be attainoted.

(b) suggested specific wording to revised tranen Tecnnical Positions, J

3egulatory Guides, Regulations, or Generic Issues.

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v. CURRENT STATJS The cresent status of the four defined problem areas is given in tarss of the tecnnical issues, the regulatory efforts, and the research ocjectives and sc pe.

This status information is given entirely feca the NRC perspective; in the report ceveloped free this proposal, however, industry and foreign views will also te given.

A. ,ios creek ...

1. gwR Dice crackinc As a result of stress correston cracking fouac in the smaller bypass lines in thR rectreulating systans (WREG-75/047 and the implementation document WREG-0313), and the concem over the susceptibility of larger-diameter piping .

in the same,'systas, USI A-042 was established. A technical resolution was acnieved as documentea in W REG-0531. The results presented in that report

'smed the ' bases for the implement,ation document, WREG-0313. Revision 1 (for l c:meent) whica was later formalind to incorporate public comments. A multi-plant action was estas11shed (MPA-605) ta inclement the staff positions con-tained in WREG-0313. Revision 1.

Curing the process of implementation, and as a result of cracking found in larger-tiameter piping at Mine Mile Point in 1M2, NRC issued two bu11stins (ID 82-03 and ID 13-02) requesting that furt$er inspections be concucted coin-cicent with plant specific refuating outages. Also, by memoraneum cated Oecencer 12, IN2, the EDO assigned NRA the lead responsibility to review and evalusta the plant specific inspection results. Subsequently, the NRA staff proposed an action plan (which we.s cassented on by RES, CST /NRR and OL/NRR) whien called for a second revision to WREG G313, Levision 1, to incon: orate i

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'h s*,ons for continued operation tse latest covelopments on the technic'J ,.

" A draft of of af f ected plants and the long-tam B;u. u. sr r . the situation.

this revision is scheculed to be availacle by July 11, 1983.

Althougn somewhat suryrised by the extensiveness of cracking, the staff asin-tains the position that the basic phenomenon of stress corresion cracking in .4SS-2C4 used in WR piping systans is the same and has not be A fully unopected fo This later ennetusion was l

larger-etaaeter piping, given longer operating time.

learly statac in NURG-C211 Revision 1, which also cited the cracking exper ence of larger-diameter piping at a German !WR in 1978 (FRM-7390-41-M)..

I

2. Put Pioe Crackinq f

Severst instances of cracting in M feecwater pining were recorted in 1979.

loth industry and the letC staff have indeoendently studied the cracking that oc:ur-ed in M feeewater piping and have enncluded that the cracking secnanism is cua to thersal f atigue. Aupented inservice inspections of feedwatar afsing

• in the vicinity of staas generator noz::les appear to be prucent in erter to cetect any cracts that signt develop before more serious prootees oc:ur.

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l The M Pipe Cract Stuey Group's recommencations to reduce the cracting of l

f feecwatar lines caused by thermal fatigue are cantained in Table I of WREG-l 0691.

The short tors solution requirsa that plants witch have axserieced If long-tars remedial ac*,f ons are cracting undertata aupented inspections.

taken and prove to be effective, the 'nspection requirements say be recuced to ASME Sect'on XI requirements.

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.- s 13 While the staff encourages consideration of remedial measures, it could be some ifee before their effectiveness is demonstrated. Thus, it is believed that

sntinued augmented insanica inspections of the feeewater lines in the vicinity of staea generator noules for Westinghouse and Combustion Engineering facilities should be tuplemented and continued until it is demonstrated by axperience that they are no longer needed. (labcock and Wilcox facilities have a different l feeewater systas design and have not aspeHenced feedwater line cracking.)

l he NRR staff is precaring guidance for supented inservice inspections of M a

feeewater lines in the vicinity of staan genera *.cr no ules. This guidance will i ne essentially the same as that developed my ILE in 19h and as recausended by tne Pipe Craet Study Group.

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Coac4 nations of faarication, stress, and envirormental esnditions have resultad in isolated in' stances of stress corrosion crecting of low pressure schedule 10 tyte 304 stain 1ess steel piping systans in Ms. Resolution of this issue was l

puolisned in NUREG-Cell, lased on operating alperience, it was esncluded that i

curnnt ISI requirwinta for thin-walled piping in Ms are adequata Oracting as found in the normal makeup /hign pressure injection nocles of four

! R&W f act11 ties. In esca case, the associated thermal sleeve has been loose or was sissing. The esoair consisted of hard rolling in new thersal sleeves, whether or not any further remedial asuures art necessary will cepend on future 1

inspection results.

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3. General 04seussion

r. sunnary. there are several pipe cracking mechanisas that uist in nucle'ar reactor f acilities--vibration and/or thersal f atigue, stana erosion, and stns ccerosion cracting. Some, such as thersal or vibration fatigue, an directly related to plant operating conditions, and once recognized, usually can be Others, such as corrected by a modification la plant operating proceduns.

intergranular stns corrusion cracting (IG5C,), C represent a complu interaction bet.een satarial properties, f abrication procedures and reactor environment so that uore is no assurance that a change in one variaale will necessarily I

resolve tne issue.

The sucject of IG5C: in gWts will be further reviewed because of its casoluity, ,

potential saf ety implications, and its severe fecact on ce plant avallamility.

Sota the NRC and the industry have devoted a great amount of resources to study the phenomenon, and to solore various remedial measurts both Icng and sher 1!

A great oeal of thforsation has been generatad. This information needs t o -i .

to be reviewed, independent of day-to-cay 1(cansing activities.I f n ordar to comonstrate the efficxy of asasures taken by incustry. Some of.the issues taat need to be pasined are:

• The adequacy of ultrasonic uamination methods for cract detution and sizing l

The development anc inclucing qualification of penannel and procedures.

l application of imenved state-of-the-art inspection techniques.

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• The potential for the cract to pngtsss casoletely around the pipe and fer l

the crack fnnt to progrees uniforsly, raising the possibility of a OEG3

• Wether there art bending stresses that say promote preferential cract gMwth in one pipe qudrant leading to leak-tefort-breat.

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' The possibility that low tancerature sensiti:ation can occur from residual This elements such as phosphorus, or sulfur rather than free carton alone.

raises the issue of possible IG5CC in extra low carton grades of' piping.

The userimental evidence is not considered sufficiently definitive to .=

nsolve this issue.

The NRC nsenren progras on tne subject was initiated in 1981 and is axcected to covelop information needed to permit an iM,4cendant cacasility for prediction, The progras inclucas:

cetection and control of pipe c-seting in LWR systaas.

1. Develoosent of the means to o4jectively and cuantitatively evaluata leak 4etection systaes and cracts througn NOE;

2. Definition of the role of stress, estallurgical variables, and environment on pice cracting suscoctibility, . including the influence of plant opera-tiens on these wardanles, f

1. Evaluation of incustry proposed fixis and repair procecures.

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6. Seitete Qesica Sef:re seismic 1 cads on piping can be toecuted, it is usually necessary to estimate seisaic ground notions, perform a soil-structure interaction analysis anc calculata building rssponse, sines piping is supported, in generet, try the The discussion builcing, at!.

the building in turf. is supported try the soil.

h t are handled Dy

ore and the proposed resiew ars limited only to areas t a "

Due to large uncertainties in estinating seismic secnanical and piping egineers.

input soil structure intaraction effects and building response and the f act of l

nuc'ser structure f atture information under seismic environmenta, ue natura This practice and easign tancency in seismic design is towarts esnservatism.

philosophy has been isolemanted at each stage of the seisaic assign cMin for piping systas design. The accumulative effect of these conservatisaa has made '

A more rigia piping nuclear piping systems more rigid than nonnuclear picing.

systas normally tv not beneficial for routinely occur-ing thersal trans'f ants This uncalancoc assign between thersal and seismic events introcuces a question as is bother the rigid nuclear piping is sors or less rsliable when the ritu

, In eftent years, increasing cancerv sssociatec with ,all eventa are consioered.

f has led to a joint MAC vith the relation between piping reliacility and stif ass:

and incustry affort to as eas the penales. Sooral is use have been icantified and damoing, cultiply supportd piping, response are unter assessment. They are:

sDeCtr*Je peak hP3adening, the operating basis earthquake (ClO cefinitien, and incustry casign prac*: ice.

1. Camoina Camoing r,quirements *er piping system seismic cesign are given in Regulatory Guice 1.61.

Camping f actors are used in the saisaic dynamic saalysis ta repre*

Due ta tae lack of can-sent energy cissipation capability of piping systass.

f tcence in our uncerstancing of the parameters *nich influence camoteg, lower

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This resulted in ..n excessive number of pipe

• tound camoing values nre used.

sedcor'.sbeingusedindesign. Two types of su::perts are normally used for selsaic design: rigid struts and snubbers. When rigid struts ars used, the piping syrtes is stiffened. A stiff piping system say be less reliable if When snubbers are used in seismic thorsal events are taken into account.

j design, the failure of snubbers in the lock-up syde can introduce highsr thsneal *1 stress on t.ie sys*as. The high failure rata of snubbers has causec concern The best resolution to Mgarding piping reliability even under seismic events.

un procles is to establish realistic dascing values to be used in design.

C.ar ently, the Damoing Task Group of the PVRC Technical Cassittee on Piping Systems is conducting an in-depth assessment to generate realistic dampir g values from vari;us piping tast da**. It is anticipated that within one year, s. .

RES has tw researca roccamencation from the PVRC will be available for review.

programs underway to experimentally develos and intarpret damping behavior in

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piping. Accitionally, inta'enational cooperative efforts are being pursued.

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2. Multioly Suecorted pioing

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The Standard Review Plan (SRP 3.9.3) requires a very conservative procedure in order to computa the seismic effects of piping supported between floors or betwen buildings. Specifically, the inertial or dynamic casconent of the loacing is calculated using an envelop spectra of all the suoports, the psuedo-static or seismic anchor movement component of loading is calculated using tae sost unfavorable coscination of support sotions, r3suming that peak displace-aants occur simultaneously, and the e ,amic and psuedostatic components are concined by the absoluta sus rule (in contrast to the SRP requirement, the staff has begun to ac:ept the SRSS rule). This position was developed at a time when 1

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3 the urgency to develop a position did not allow for a full and complete under-stancing nf the behavior af multiply supported piping subject to independent saismic inputs. Since that time, RES-funded studies have been casoleted, or are nearing completion, which indicata that the SRP position is unreasonable and that more realistic critaria can be estabitshed. This particular technical issue .i likely be resolved within a year.

3. Resoonse Soectrus #eak troadening To account for uncertainties in the natural frequencies of supporting structures, Regulatory Guice 1.122 requires taat the peaks of floor response spectra be broadaned, typically 215L It has been pointed out that for piping this could have the effect of placing a few modes at the peak floor responsa spectrum value, since piping natural frequencies taad to be relatively cle,sely spaced.

Because peaks in the floor response specta usually are shar;, and pointad (with

~ the consequence that no more than one mode could have the peak or near peak ' l floor resuonse spectrum value), the regulatiry guide position imodses an 6nneces- t sary conservatisa on piping design. It has been recommended that a acre fational way of dealing with uneartainties in the natun1 frequencies of supporting structures muld be peak shifting rather than peak broadening. With peak shift-ing, the shape of the floor response spectrum is not changed, but instand, the response spectrus is shifted bachards and forwards along the frequency scale within given limits, and the most unfavorable locacion selected for design. '

This procedure has the advantage of dealing with uncertainties in si.acorting structure frequencies without requiring that acre than ore We be at the peak value of the floor response spectns. As the procedur' is al eady available, and as its adequacy has already been demonstrated 'adt. research performed by RES contractors, immediata isolamentation of this relaxation is possible.

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a. Ooeratino Bass. Earthouake (CBE) Definition Accorcing to 10 CFR 100, Appencix A, when an earthquake exceeding the CBE Licensees sust then occurs, shutdown of the nuclear power plant is required.

cemonstrate that no functional das:ge has occurred to those features necessary Secause for continued operation without undue risk to public health and safety, lower allowable stressas and damping values are permitted with the CSE, many d

piping cesigns at nuclear power plants are controlled by the 08E rather than the more severs Safe Shutdown Earthquaka (15E). Loads for both the CBE and the SSE are estimated using elastic dynamic analysis proceducts; however, the allowabia for the SSE implies local nonlinear behavior.

It has been argued that it is unreasonanle to allow an earthquake which sust be resisted without sustaining casage to control the sir.ing and proportioning of pipe supports and other elements.

While the two-earthquaka approach to design is commonly adopted in many countries for nuclear reactor design, there is a growing sentiment that the CBE eust be su to a lower value or that allowable i stresses must be increased such that piping design is not controlled by an event not directly related,ts ,

safety. This specific issue, once anopted for nqsar reactor piping, is likely to have incertant isolications for other stmetura3 faatures at nuclear power plants.

5. Industrv practices Thors are an infinita variety of difderent pipe su;rport system ::astqns and mceling pra'cticas which can be isolemented to satisfy bom ASME esce and NRC reouirements. These different designs and modeling practicas, however, are not equally reliable or equally cost effective according to results obtained frea The assumptions made in modeling piping sys*as parametars, RES-funded studies.

such u those regar: ting support stiffness, gaps, overlap techniques and so on, eu affect the final result, but say be uniscortant when otner design conservatises 4.

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are present. It is expected that the NRC Piping Review Coemittee will explore ,

this matter. It is common industry practica to isolement the first iteration in  !

tne design process which satisfits the requirements ratner than optimize the pipe support systes in terms of reliability or costs. Nonetheless, procedures are becos.Ing available which can achieve these goal:. It is unclear at this time how the NRC should deal with these goals and three basic issues to be ,f decided are: (1) does the NRC have confidence in the technical fowWation in the optimization algoHthes? (2) should the NRC serely encourage or Mther enforce utilization of the algoH thes? (3) is it necessary to revise industry oractice in modeling if the conservatisas in design are removed? Recommenda-tions from the NRC Piping Review Committee are sought on both these questions.

Staff regulatory efforts and their respective status c:nsist of the following:

1. GeneH e Tssues A-40 and A-41 were created. The short-tsrs effort (A-40) is to revise SRP' 3. 7.1 -3'. 7. 3. The long-tern effort (A-41) is' the Seismic Safety Margins Reseafh Program (55MRP). The A-40 is ongoing and will be coraleted in FY 1984.i The SSMRP has developed a cosorehensive probanilis-tic methodology to assass plant seismic Hsk and will also develop simpli-fied methods for calculating seismic Hsk.

2. Generic Issue A-13 was created for snubber oparamility. Solutions were proposed in NUREG-0371; which consists of evaluation of industry practices asscciatad with snubber design and oualification tasting, and development of Technical Specifications and regulation changes. SRP 3.9.3, STS 3/4.7.9 for PWRs and STS 3/4.7.5 for SWRs were revised. A draft Regulatory Guide to assure a hign level of snubber operability is casoleted.

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3. Generic Issue B-25 deals with piping benchmarks. The program is providing the staff with a capability to conduct independent verification of computer coces used by licensees for his dynamic analysis of ASME Class 1, 2, and 3 nuclear piping. Generic Issue 8-25 is an ongoing program which has the capability to verify time domain and frequency dwain analysis, elastic and inelastic analysis, and can use either sultisupport spectra input or j single-enveloped spectrum input. Test verification of the codes will be completed in FY 1984 4 Generic Issue B-51 was created for assessment of inelastic analysis tec.e nicues used in piping and support analysis. Since ce ASME Code pemits Level 0 stress limits for low probability events, anc large (nelastic strains may occur under such stress limits, it is important that properly ,

cualified analysis techniques are used. Generic Issue B-51 is an ongoing ,

issue for continuous monitoring of the ASME Code rule changes.

5. : Current regulations require that components and stractures shall be designed Due to

f. for two levels of seismic loads, namely, the CSE and the SSE.

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. different requirements in load definitions, load coecinations, and stress allowanies, the CBE generally controls the design. The staff has investi-gated the implications of having beo levels of seismic design and has proposed a revision to seismic requirements in a report to the Comissioners cated Acril 27, 1979.

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1 C. Pice Break Current requirements specify that piping and supports of the reactor coolant pressure boundary be designed to e-data the dynamic effects of postulated pipe break events, including a LOCA induced by a doubicended rupture of the .

largest primary pipe. Subsequent licansing positions set by Regulatory Guida 1.46 and SRP 3.6.2 also indicated that such large breaks should be universally postulated for all high energy lines inside and outside the containment, and at several locations per pipe run. Implementing these positions has resulted in installation of numerous pipe whip restraints, snubcers, and jet shielding structures, that may not provide additional safety, and in fact, may detrac ,

from safety in some ways. For instance, piping so designed is stiffer and exoeriencas greater strsss when subject to thersal expansion, and has become sore inaccassible for fr. service inspection. Thus, a revision of current pipe breeK postulation critaria is necessary in order to acnieve better balance in casign for both norsai and abnormal plant situations. Staff efforts and their respective status consist of t;a following:

1. Generic Issue A-18 was created to upgrada pipe break critaria. The task has completed the following short-tars objectives:

(a) SRP 3.6.2 was revised which provides consistant pipe break critaria for high energy lines both inside and outside the containment.

(b) Critaria for the pipe break exclusion region at the enntainment pene l trations were defined, and pipe rupture loads to the guard pipe were investigated.

(c) An investigation of the deleterious effects of pipe whip restraints and snucbers to piping normal operation were conducted and resorted.

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., 23 Generic Issue A-18 became inactive shortly after the TMI accident due to prioritization considerations. Its long-tars objectives were carried on by several research programs including investigation of pipe whip, jet ispingement load:, pipe-to pipe impact tests, piping failure mechanisas, and pipe break probability.

Generic Issue 40 was created to review the safety concernt associated with j

2. 1 pipe breaks in SWR scram systems. Options acceptable to the staff consist of establishing either (a) a low probability for the event, (b) accaptable consequences for the event, or (c) availability of altarnata cooling sys-taas and initiation equipment for the event. NUREG-0803 was issued wnidi g1ves guidance for resclution on a plant specific basis. GE has recent!y provided adcitional inforntion to strengthen the generic resolution using ,

the argument of low event prcoability. The new submittal is currently und,ar staff review. .

3. Generic Issue 61 was created to review the concarn of postulated break at .

the safety / relief valve line inside the airspact of a BVR Mark I and II containment, sinca the staan bypassed suppression pool will rapidly j

l pressurize the containment. NUREG-0487 was issued w11cn required Mark II plants under construction to perform a piping fatigue analysis. Operating The Mart I plants were required to conduct periodic visual inspections.

staff considers that the most effe:tive resolution is utilizing the con-tairment spray. New requirements regarding operational procedures and/or design modifications of containment spray systans remain to be developed.

1

24

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The staff is planning to revise its break postulation critaria in SRP 3.6.1 and SRP 3.6.2.

Piping f ailures generally occur at nigh stMss and fatigue locations, such as " at r* tarsinal ends or at locations where corrosive environment and fatigue exist. The NRC, in the development of positions for postulating pipe ructure, selected stress level and fatigue usage factor as the criteria for For instance, for a high energy, Class 1 piping, pipe oecicing pipe rupture.

ruptures are mquired to be postulated at the tarsinal ends and at fatarinediata locations wnere the calculated maximum stress range exceeds 2.4 5, or the .

1 cumulative usage factor is larger than 0.1. At least two intamediata locations If two intemediata locations cannot be datamined by the must be selected.

acove str'ess or usage factor criteria, the two highest stnss , locations are then selected. Sinca these locations are not f.ixed early in the construction process, s .

A/Es fino it very difficult to isolement such criteria.

As a result of the above requirements, 300 to 400 pipe weip restraints will have Total costs for design, procurement to be installed for a typical iNR plant.

and construction of those restraints are est mated in the range of 20 to 40 i

For a plant under construction, these requirements millien dollars per unit.

For an existing plant, back-ine ease construction cast and design cosciexity.

Also, there are fitting would be required in order to meet current standards.

For example:

several negative aspects to having pipe whip restraints.

1.

Access for raintenance and inservice inspection is isoeded, therefore incnasing radiation exposure to intpectors. ,

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2. Higner thermal stress caused by the pipe coming in contact with the restraints is introduced.

In 1979, NRC initiated studics in both NRR and RES to assess the likelihood of 4 naving a double-ended guillotine pipe break (DEGB) in the PWR primary loop.

i NRC programs in the area of piping fractu.M sechanics will concentrate on vali-

.ating tuctile fracture analysis techniques, developing a coscrenensive piping materials data tue, and conducting pipe tests to detarsine the failure modes of piping for use in developing improved pipe break critaria. .

To data, the NRC has sponsored the development of expeMaantal techniques for enaractarizing ductile fracture preperties of piping satarials along with intar-meciate size pipe tasts to validata the use of taaring instanility fracture menanics. This work has beu performed at USNRCC in Annapolis. Development of experimental techniques for sataMais charactaMration included investigation of the effects of specimen geometry and development of single-specimen casouter controlled unloading comp 11anca J-resistance cune tasting. Fut:are wort in this area includes datarsination of environmental and loading histoMes on ductite i

fracture properties. The pipe tests conducted at USNRDC wre performed using 8-inen-diameter, circunferentially-flawed, A106 grade B piping loaded in four point bending. The loading train was made compliant through the use of springs so that conditions for both stable and unstable crack extansion could be e eated.

Using J-resistance curves generated from the pipe tasts, taaring instability Future pipe analysis accurataly predicted stable and unstable crack extension.

tests to be conducted at USHRCC will be perforsed on 8-inc. t diasatar A106 cirung with circurferentially-flawed welds.

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During the next 3 years, NRC's sajor effort in the area of piping fracture The objective of this program is mechanics will be the Degraded Piping Program.

to further validata and improve the ability of ductile fracture mechanics analyses to accurately predict the loading capacity and failure mode of cracked '

piping and to provide expeMaental data which can be used to develop 11 prove Phase I of this program began in September 1981 at Eatta11e pipe creak cMtaMa.

The Phase I effort included the development of faproved Columous Lacoratories.

cuctile fracture mechanics analyses, review of previous and ongoing pipe fracture programs, and development and costing of a comprehensive pipe test program.

Phase II of the program, to begin in the fall of 1983, will include low energy .

The and nign energy (typical LWR pressure and tamparature conditions) tests.

nigh energy tasts, in particular, will provide necessary data for developing improvec pipe break critaria to replace the currently postulated double-ended guillotine break.

Finally, beginning in fall 1983, NRC will sponsor a data base development pro-i

.- N wnica will include generation of necessary fracture mechanics propert es I

for piping satorials and development of a cescutaH :ed data systas.

In NRR's ' effort, a datantinistic fracture mechanics assessment bued on met toughness asasurements was made to evaluate PVR pMaary loop piping integ loop piping The results of the study suggested that the rupture of PWR primar/

l is unlikely.

Resulta free RES's probabilistic assassment of the probability of l

a DEG3 in the PwR primary loop due to direct and indirect causes confirmed N findings. This stucy employed a probanilistic accel to assess the likalthood of a DEGB caused by innirect causes, such as structural support systas failure

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The probabilities for both direct and indirect leading to the pipe rupture.

This ansessment causes are very low for the Westinghouse PWR plants studied.

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was conducted on austenitic stainless stael piping used for Westinghoese p an na methodology developed in this assessment is equally capable of hardling

• An effort is currently underway in RES to assess J.; 4rritic ferritic steels.

We expect similar conclusions will be reached steel CE and EW primary loops.

for CE anc %W plants. In the lettar from H. R. Centon to M. Ecelsan of the Atomic :ncustM al Forum, dateo May Z,1983, appropriate changes to curre.it cHtaria on pipe rupture requirements for the PWR primary loco are indicated.

West Gem has almacy Malaced the full flow ama break requirement with a*

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Dreak area for pipe whip and dynamic loading of nactor vessel intarnals.

The full flow ana break is still required in Germany for containment sizing, .

EC".S casign, and major component suppor*J. .

De prooabilistic methodology developed by RES is cacaele of handling s An effort is currently corrosion f ailure mechanisms in BVR pMaary systaas.

Historically, stress corrosion uncerway in RES to assess BWR pHaary systems.

cracking has occurnd in SWR Reactor Coolant Loop Piping, and preliminary ations suggest that the SWR full flow area bnak usessment will be less favor-Curnnt planning is to concleta the SWR acle than was the P%R assessment.

Whether there will be any Reactor Coolant Loop assessetnt by the end of 1984 recommendations to relax present requirseents is difficult to predict.

Piping systees othe? than the pMaary loop an more complex dur to large ations in configuration, natarial, loading, environment, function, and suggert (1) those systass for type. Piping systans can be diviced into three groups: ,

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which leak-before-break can be easily establiihed, (2) those systass for which the operating environment and piping mataMa's suggested that leak-before-break will be very hard to be established, and (3) those systans which belong to neither one of the two groups above; that is, systans where leak-before-bruck is ,

marginally established.

4 For the first group, it should not be very hard to develop a technical basis to assess the intarmediata break requirement. The available probabilistic sethod-ology, with support from the planned dog' aced piping program to provide proper f ailure modes for cartain selected piping systaas, snould be able to reach some conclusions in a reasonanle time. In order ta eliminata the intarsediate break requirement based on the leak-before-break hypothesis, it is necessary ,

that the reliability of leak detection capamility be established.

It is anticipated For the other two groups, the decision will be scre difficult.

snat a auch larger effort will be needed to battar understand piping failun sechanises and davelop prover;tive action which can improve piping reliability.

The genesis of the laak-before-break issue goes bacx to a traditional staff interpretation of General Design CMtarion (GDC) 4 in the contaxt of the defi-nition of a LOCA as inclading a bnax equivalent in size to the double-ended ructure of the largest f.ipe in the reactor coolant systas (Appendix A to 10 CFR 50).

This interpretatioa ha9 perseated through vaMous staff guidance documents; i.e. , Regulatory Guide 1.46, SRP 5ections 3.f.1 and 3.6.2, and has enated a situation such that latar vintage plants are nouired to have massive pipe whip restraints and older plants were found not able to take asymmetrical LCCA loads. The latar concern over the asyneetMeal LOCA loads led to the establish-sent of USI A-2.

l 28 l

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As an attaset to resolve the USI A-Z issue. Westinghouse submitted topical recorts, on behalf of 11 licensees of 16 PWR units (one of which is a CE designed f plant), using advanced fracture mechanics to demonstrata that the detection of ,

small flaws (postulated or reai) either by 151 or by leakage monitoring can be assured before the flaws can grow to critical sizes which could lead to a large break area "equivalent" in size to a double-ended guillotine LOCA. Tne staff

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nas evaluated these topical reports, independently verified by research progras ,

anc tecnnical assistance program the methodology contained in these topical moorts, anc has concluded that a valid technical basis exists for exempting An exesotion package tasse plants from the regulation as currently intarpreted.

was preparec anc comments incorporated, and will be reacy for NRA Director's

-eview anc submittal to CRGR before the end of June 1983. .

In carsliel with the above action, the Division of Engineering staff has also craftec a Cosmission paper for NRA management review that proposes granting similar exesotions for other PWRs not covered by Westinghouse topica.ls.

~he third steo to be undertaken is to initiata a rulemaking to seek a long-tars esolution of this situation. There are some ancillary issues, such as intar-mediate breat critaria, which would naturally be resolved as the staff revises and cavelops those staff guidance documents (SRP 3. 6.1, 3. 6.2, and 3.6. x) to succort rulemaking.

The majority of technical wort is cosclete or has reached the stage that is deemed sufficient to support these changes. The NRC Piping Review Committee will be asked to review the results of all work sponsored either by NRC or EED j

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f by the industry in order to confiris the statement above. However, the staff efforts including the init.tation of the rulemaking will not be held in abeyance pencing the outcase of this proposed review.

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.. 31 D. Other Ovnamic t.oads and t.oad Cescinations

- i Sicing and supports are designed to withs*Jnd a spectrus of events, and events  ;

in coesinations. GDC 2 requires that earthquake events should be appropriataly comoined with LOCA events. Licansing position delineated in Aegulatory Guida 1.a8 and SRP 3.9.3 specify that comportents and supports should be daalgned to accommodate incividual and conoined loads due to norsal operating conditions, }

,7 system transients, and postulated low probability events. Loads induced by ,

'i earthquake, LOCA, and systas transients ury with time and their peak responses f:r purpose of design wers assmed to occur simultaneously. This esaservative amoroacn resulted in stif* piping systans which are not only costly, but also say not be very reliacle. Because of this concarn AIF has estan1{shed a sub-

saittee on Load Comoination to address this issue. ,

For developing a mort rational . treatment of loads, load combinations, and stress iisits Generic Issue 8-4 was created. The s'aff efforts to investigata and estan11sh licensing positions on response coacination methodology were casolated anc reported in WAEG4s84, Revision 1. Rules to justify the use of the SRS3 method were delineated based on specific ac:eptable nonexceedance pret. abilities free cumulative distribution functions. These positions were acolied in licens-ing rivi ns of Mart II and Mart III plants.

I SAP 3.9.3 was revised to reflect nw' positions on load comoinations and stress 1

i limits.

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l In addition, stress allowables specified in the A94E Code for piping dasign saka no distinction between'whether the load is st.atic or dynamic. The staff plans to uncertake a revision to SRP 3.9.3 in order to incorporsta secarata stress limits for static and dynamic loads.

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The issue of load concinations became more controversial in recent years. Other dynamic loads such as BWR pressure suppression pool swell loads have further casolicated the issue.

,b These enanges have raised questions concerning implementation of new regula-tiens, increased construction costs, increased radiation e.xposure to maintenanca crws perforsing inspection and maintenance actions, and reduced reliaatlity of stiffer systems under normal operating tre.tsients.

The major NRC concern regarding these issuer can be summarind in four questions:

(1) Which events are required in the piping systas design basis, and fehich events should be considered in costination?

(2) Mcw do we define other piping dynamic loads, for example, their segnitude, duration, and frecuency enaractaristics?

(3) What method should be used in concining piping systas respcases resulting from the above loads?

(4) For each coscination case, what should be the most appreoriata stress allowacle for design to reflect the actual cacacity of the satarial to withstand dynamic loads?

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1. Combination of Lame LOCA and SSE for the PVR P*1marv System - Research Information Lattar No.117 identifies the following results for the NR primary systas:

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(a) Fatigue crack growth due to all transients, including earthquakes, is

. an extremely unlikely mechanisa for inducing a large LOCA. The contHbution of earthquakes to the oc:urrence of this unlikely event is a small percentage of the total probability.

(b) Leak probacility is several orders of segnituce higner than double-ended guillotine break probability. This supports the lead-tefore- ,

breat hypothesis.

As a result of this study, the staff plans to recommend appropMata changes to curunt'cMtaMa on large LOCA and SSE combination for 'NR primary.

systes. In its June 14, 1983 lettar to the EDO, the ACRS stated, 'Ne find .,

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... tne decoucling of the loss of coolant ac:icant and seismic loads to be appropM ata. '

2. Concination of Pipe Break and $$E for Systems Other Than 'he Prisaw Leco in PwR slants - For systans other than the pMaarf loop in NR plants, the contHbution of seismic loads to pipe failure is unknown. From the f ailure sechanism point of view, pipe r.oture due to satssic load is unlikely to occur except den the piping systas is constructed of cor osion susceptive natarfels ,and the systas is operated under fatigue conditions in a corrosive environment. After a careful review of piping fatture data and results free axisting fracture mechanics research, it is possible that an iseroved requirement on caseination of pipe break and SSE can be estaa-lished for syttaes other than WR pMaary loop piping.

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3. Cemeination of Larce LOCA ano SSE for BWR Pioino Systems - A great deal of intergranular stress corrosion cracking (IGSCC) has been reported in BWR piping systans. In recent years, IGSCC on larger dissetar piping has been reported. IGSCC represents a complex interaction phenomena between mate- b 3

rial properties, fabrication procedures, and the operating envirorment. A piping systas under IGSCC attack will have higher rupture probability under a seismic event. The assessment of tae consination requirement of large LOCA and SSE for SWR piping systans will have to take IG5CO effects into consiceration. The probanilistic fracture mechanics methodology developed by RES is capable of evaluating the seismic contribution to pipe rupture

• under IGSCC environments.

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4. Cther Dynamic Loads - Othe'r dynamic loads in piping, such as SRV discharge loads, vibrational loads, and water hammer loads, any require air assessment to datarsine how they should be treated individually Er in conoination. i Particular attention is given to the high frequency contact of these loads which could be damaging to equipsant sounted on piping and also to support systass embedded in concreta. The frequency contant of these loads normally exceeds that associated with earthquakes. Tne mejor NRC concarn with regard to dynamic loads on piping has been directed to seismic loads; however, such leads are better clusified as pseudostatic, and loacs generated by water hammer, water slugging, or valve opening on closure represent loads more severe than'seisart: in tanes of rata, and magnitude.

i In some instancas, major splits or failures of piping have occurred in  !

secondary and tartiary piping systans, and instances have been reportad of pipe suoports being torn from the walls because of water hammer, without,  ;

however, fa11 tug the piping.

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5. Desien Limits - Current piping

• systes design is governed by the ASME Soiler anc Pressure Vessel Code. The design and service allowables stated in the ASME Code were developed for static conditions. When the structural systas is subjected to short duration dynamic loads with high strain ratas, the capability of structures to withstand the loading effect can be auch higher than the static limits. The PVRC Technical Committee on Piping Systaes has j

~1 a task group on dynamic allowables that is currently assessing the avail-able data to battar define the dynamic allowables for piping design. It is recessended that this task group closely c:mmunicate with the PVRC task group in this cavelopment and start to review the PVRC data base. It is anticipatec that a technical resolution say become availabla f-en the PVRC for NRC review. ,

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VI. SCHEDUCE AND DEL:VEUBLES Periodic status reports will be prepared as deliveracles during the course of this effort. Additionally, informal discussions will be used to keep NRC management advised as to progress being ande.

A draft final report will be available in about 9 months. Caspletion of all ,

tasks is estimated to occur in about 12 months from the data of authorization.

The final product will be a report reviewing the status of nuclear piping in the contax of regulatsry and licensing actions. The chief product will be con-clusions and recommendations. They will be directly correlatants to nievant equlatory positions such as Iranch kecanical Positions, Regulatory Guidas. I&E, Sulletins, Regulations, and Generic Issues. The f acact of proposed actions on axisting research will be discussed. Suggestions for future research will be zade wnen it is concluded that no position may be takan until additional infor-sation becomes available.

As indicated previously, the conclusiona and recommendations will be based on the technical judgment of the participants. They will not represent regulatory sositions validated by quantitative value-iscact studies and writtan in language amorocriata to regulatory doc.ments. These actions are assumed to constituta follow-on actions presuming tae conclusions and recossendations are accepted.

However, qualitative statanerrts regarcing the cost and safety benefits tnat say acen e will be included in the report.

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. . . . 37 VII. MANPOWER AND 8UDGET ,

The NRC manpower estimate is 4 to 5 nan years, involving about seventeen indi-viouals. E.xpertise will be drawn primarily from NRR and RES, with contribu-tiens from I&E, AE00. ELD, ACR$ and the Regional Offices. Additionally, our consultant, Dr. S. H. Bush will play a predominant role in guiding this effort.

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It is expected that about twenty consultants will participate in the work heretofore described. Total consultant time is forecasted at bse ann years.

Total consultant costs, including travel costs, should be of the on:ar of 1300,000. Consultants will be drawn from the Lawrence Liversore National

• Lamoratory, Argonne National Laboratory, Idaho National Engineering Lanoratory, Brookhaven National Laboratory, Battelle-Pacific Morthwest Laborataries, and ,

other sources, including privata consultants and academia.

Additional NAC manpower will be needed after the

• completion of this effort to revise

• Regulations, Regulatory Gui'das, the SRR, and tranch Technical Positions.

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