ML17254A993

Forwards Request for Exemption from Considering Large RCS Primary Loop Pipe Breaks in Structural Design Basis (GDC-4), in Response to Generic Ltr 84-04.Generic Issue A-2, Asymmetric Blowdown Loads on PWR Primary Sys Resolved

ML17254A993
Person / Time
Site:	Ginna
Issue date:	10/17/1984
From:	Kober R ROCHESTER GAS & ELECTRIC CORP.
To:	Paulson W Office of Nuclear Reactor Regulation
References
REF-GTECI-A-02, REF-GTECI-RV, RTR-NUREG-0821, RTR-NUREG-821, TASK-A-02, TASK-A-2, TASK-OR GL-84-04, GL-84-4, NUDOCS 8410240250
Download: ML17254A993 (16)

Text

REGULATORY FORMATION DISTRIBUTION SY M (RIDS)

ACCESSION NBR; 8410240250 DOC ~ DATE: 84/10/17 NOTARIZED: NO DOCKET FACIL:50 244 Robert Emmet Ginna Nuclear Planti Uni't iE Rochester G 05000244 AUTH'AtlE AUTHOR AFFILIATION KOBERtRDH ~ Rochester Gas.a Electric IP ~ NAME RECIPIENT AFFILIATION Corp'EC PAULSONEN ~ AD Operating Reactors Branch '5

SUBJECT:

Forwards request for exemption from considering large'CS primary loop. pipe breaks in structural design basis (GDC"4)E in response to Generic Ltr 84 04,Generic Issue A-2E "Asymmetric Blowdown Loads on PAR Primary Sys" resolved, DISTRISUTION CODE: A0010 COPIES RECEIVED:LTR 'NCL g SIZE: g" TITLE: OR Submittal: General Distribution NOTES;NAR/DL/SEP 1cy ~ =05000?44 OL:09/19/e9 RECK P IENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL NRR ORB5 BC 01 7 7 INTERNAL; ADM/LFMB 1 0 ELD/HDS4 1 0 NRR/DE/MTEB 1 1 NRR/DL DIR 1 1 NRR/DL/ORAB 1 0 TB 1 1 NRR/DSI/RAB 1 1 04 1 1 RGN1 1 1 EXTERNALS ACRS 09 6 6 LPDR 03 NRC PDR 02 1 NSIC 05 NTI'S 1 1 NOTES!. 1 1 TOTAL NUMBER,OF COPIES REQUIRED+ LTTR 27 ENCL 24

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szizriirz SaalO~Na ANP 5155'zzzzz I rza~ilEII+I TIaWZ j 5TATE ROCHESTER GAS AND ELECTRIC CORPORATION o 69 EAST AVENUE, ROCHESTER, N.K 14649-0001 ROGER W. KOBKR VKE PREQDENT TELEPHONE ELECTRIC Ek STEA5kt PROOUCTTON AREA cooE TIE 546-2700 October 17< 1984 Director of Nuclear Reactor Regulation Attention: Mr. Walter A. Paulsoni Acting Chief Operating Reactors Branch No. 5 U.S. Nuclear Regulatory Commission Washington> D.C. 20555

Subject:

Generic Issue A-2I Elimination of Postulated Pipe Breaks R. E. Ginna Nuclear Power Plant Docket No. 50-244

References:

1; WCAP 9558I Revision 2 (May 1981) "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack".

2. WCAP 9787 (May 1981) "Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation".

3. Letter Report NS-EPR-2519, E.P. Rahe to D.G.

Eisenhut (November 10I 1981) Westinghouse Response to Questions and Comments Raised by Members of ACRS Subcommittee on Metal Components During the Westinghouse Presentation on September 25I 1981.

4. NUREG-0821I Integrated Plant Safety Assessment Systematic Evaluation Programi R. E. Ginna Nuclear Power Plant, December 1982.

5. Generic Letter 84-04I February li 1984I Safety

,Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops.

Dear Mr. Paulson:

USNRC Generic Letter 84-04 provided the staff Safety Evaluation Report for analysis materials submitted for a group of utilities operating PWRs to resolve generic issue A-2. The staff evaluation concluded that< provided two conditions were meti an acceptable technical basis exists so that the asymmetric blowdown loads resulting from large breaks in main coolant loop piping need k

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ROCHESTER GAS AND ELECTRIC CORP. SHEET NO.

DATE October 17 I 1984 Mr. Walter A. Paulson not be considered as a design basis for the sixteen domestic plants for which the analysis applies. The purpose of this letter is to respond to the open items identified in generic letter 84-04 to obtain final resolution to generic issue A-2.

The two conditions specified in generic letter 84-04 that must be met for staff approval concern verification of bending moment loads at two plants and verification of leak detection capability. Ginna is not one of the two plants for which confirmation of maximum bending moments was requiredI so this condition of approval is not applicable to Rochester Gas and Electric. The second conditionr that leakage detection systems exist to dete'ct postulated flaws utilizing guidance from Regulatory Guide 1.45I with the exception of seismic equipment qualificationr is applicable.

Ginna has several leak detection systems with at least one that is capable of detecting a one gallon per minute leak in four hours. Conservative calculations of leakage from flaws shown to be stable in WCAP 9558 and WCAP 9787I indicate that leak flow rates one to two orders of magnitude greater than one gallon per minute can be expected if these flaws exist in reactor coolant piping (see reference 3). The equipment provided for leak detection> the means of quantifying reactor coolant system leakage and leak detection operability requirements are delinated in Section 3.1.5 of the Ginna Technical Specifications. The Ginna leak detection capability has been evaluated against current regulatory criteria during Systematic Evaluation Program review of topic V-51 Reactor Coolant Pressure Boundary (RCPB) Leakage Detection. Acceptance of our leak detection capability is documented in reference 4. Because Ginna Technical Specifications require the operability of leak detection systems and because these systemsr with margini are capable of detecting leakage from postulated circumferential throughwall flaws> adequate leak detection capability exists to satisfy the staff condition of approval.

Even though the staff concluded that an acceptable technical basis had been provided> Mr. Eisenhut's February li 1984 letter also stated that authorization to remove or not to install protection against asymmetric dynamic loads in the primary coolant loop will require an exemption from General Design Criterion 4 (GDC-4). Rochester Gas and Electric does not believe that an exemption is required becausei among other things> footnote 1 to Appendix A to 10 CFR Part 50 anticipated that further details relating to the type> size, and orientation of postulated breaks in specific components of the reactor coolant pressure boundary would be developed to define Loss of Coolant Accidents (LOCAs) postulated in plant design bases. Thus> the existing design

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ROCHESTER GAS AND ELECTRIC CORP. SHEET NO. 3 oATE October 17 1984 Nr. Walter A. Paulson criteria anticipated that< when developed< justification such as advanced fracture mechanics analyses could be used to define postulated LOCA pipe break sizes less than the double-ended rupture of the largest pipe. In addition< Appendix A sets forth requirements for design criteria that must be included in an application for a construction permit for a proposed facility pursuant to the provisions of 10 CPR 50.34. It has not been established that the application requirements apply to facilities which were already operating prior to the issuance of Appendix A.

Nevertheless> since sufficient justification for an exemption has been presented and< for all intents and purposes> an exemption will have no effect on plant operation> Rochester Gas and Electric Corporation requests that an exemption from GDC-4 be issued as set forth in the enclosed application.

er truly yours<

er W. Kober

t EXEMPTION APPLICATION ROCHESTER GAS AND ELECTRIC CORPORATION R.E. GINNA NUCLEAR POWER PLANT In response to generic letter 84-04< Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks In PWR Primary Nain Loops> Rochester Gas and Electric Corporation requests the elimination of large reactor coolant system primary loop pipe breaks from consideration in the structural design basis of the Ginna Nuclear Power Plant. This request is based upon the use of advanced fracture mechanics technology as applied to primary system piping in Westinghouse Electric Corporation topical reports WCAP 9558, Revision 2 (proprietary) and WCAP 9787 (proprietary) and is the resolution of generic issue A-2, "Asymmetric Blowdown Loads on PWR Primary Systems".

The bases for the request are as follows:

l. Extensive opera'ti'ng experience has demonstra,ted the integrity of the PWR reactor" coolant system primary loop including the fact that there has,never been a leakage .crack.

2. Pre-service> and in-service inspections performed on piping for the Ginna plant minimize the possibility of flaws existing in such piping. The application of advanced fracture mechanics has demonstrated that if such flaws exist they will not grow to a leakage crack when subjected to the worst case loading condition over the life of the plant.

3. If a large through-wall flaw is postulated> large margins against unstable crack extension exist for the Ginna stainless steel primary coolant piping even if subjected to the safe shutdown earthquake in combination with the loads associated with normal operation.

The application of advanced fracture mechanics technology has demonstrated that small flaws or leakage cracks (postulated or real) will remain stable and will be detected either by in-service inspection or by leakage monitoring systems long before such flaws can grow to critical sizes which otherwise could lead to large break areas such as the double-ended rupture of the largest pipe of the reactor coolant system. To date> use of this advanced fracture mechanics technology has been limited because of the definition of a LOCA in Appendix A to 10 CFR Part 50 so as to include postulated double-ended ruptures of piping regardless of the associated probability and regardless of the fact that there is no mechanistic scenario under which this event will occur.

Application of the LOCA definition< without regard to this advanced fracture mechanics technology> to large diameter thick-walled piping such as the primary coolant pipes of a PWR

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imposes a severe penalty in terms of backfit cost and occupational radiation exposure. Massive pipe whip restraints which would be required without the fracture mechanics technology must be installed and then removed for in-service inspections. As Issue documented in the NRC's Value-Impact Statement for Generic A-2i this penalty is unreasonable because these pipes do not have a history of failing or cracking and are conservatively designed.

Accordingly> for design purposes associated with protection against dynamic effects> we request that postulated pipe breaks in the reactor coolant system primary loop be eliminated from the structural design bases where established by appropriate analysis.

This request does not extend to specifying design bases for containment> the emergency core cooling system< or environmental effects.

The use of advanced fracture mechanics would permit a deterministic evaluation of the stability of postulated flaws or leakage cracks in piping as an alternative to the current mandate of overly conservative postulations of piping ruptures. This request is consistent with the provisions of footnote 1 to 10 CFR Part 50> Appendix A> which contemplated the development of "further details relating to the type> size and orientation of postulated breaks in specific components of the reactor coolant pressure boundary".

As support for this request< in addition to the two Westinghouse topical reports referred to above, we request consideration of the following:

l. Memorandum from Darrell G. Eisenhut (NRC) to All Holders Operating PWR Licensees< Construction Permit and Applicants for Construction Permits dated February 1, 1984

Subject:

Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops (Generic Letter 84-04).

2. CRGR resolution of generic issue A-2< September 28> 1983.

3. ACRS letter dated June 14 1983> re: "Fracture Mechanics Approach to Pipe Failure".

4. Memorandum from William J. Dircks< EDOi to ACRS dated July 29> 1983, re: "Fracture Mechanics Approach to Postulated Pipe Failure".

,These documents and Westinghouse topical reports WCAP 9558 and WCAP 9787 provide a substantial and adequate basis for limiting postulated design basis flaws in stainless steel reactor coolant system piping.

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A detailed value-impact analysis.hhs been performed by Pacific Northwest Laboratory (PNL) to assess the relative costs of using advanced fracture mechanics techniques to justify design bases for several operating PWRs instead of modifying these plants

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to conform to piping restraint designs used in more recent plants.

This analysis clearly establishes that the costs< both in dollars and radiation exposure> are greater for modifying the plants than are the money and radiation exposure costs due to guillotine pipe ruptures considering the low probability of such events.

Rochester Gas and Electric supports the conclusions reached in this analysis.

The PNL value-impact analysis is not specific for each of the evaluated plants< but the analysis inputs are reasonable.

Estimates of occupational radiation exposure rates conservatively correspond with dose rates that are experienced at Ginna in locations where modifications would be required. Portions of the estimates of modification costs and manhours of occupational exposure are based on estimates from utilities with operating PWRs and thus should be realistic. It should be noted> therefore>

however< that the cost estimates are no longer current and are>

probably low. The estimates of guillotine pipe break frequency contained in the analysis are probably too high. The estimates are based on data which is not specific to guillotine breaks of large diameter> stainless steel< nuclear grade piping and, therefore< overestimate the probability of reactor coolant system double-ended pipe ruptures. All of these factors lead to the conclusion that the PNL analysis result is correct but that the analysis understates the relative value of using deterministic techniques to define design bases for the affected plants. The value-impact analysis clearly establishes that advanced fracture meehan'ics analysis is an acceptable alternative to designing and installing plant modifications to mitigate the consequences of unrealistically postulated double-ended guillotine breaks.

i It "is not clea'r 'that 'the,use of advan'ced fracture mechanics

,is not already permitt'ed'by Appendix"A,to',1'0 CFR Part 50 to define LOCA pipe 'br'eak sizes'. Neither is it clear that 10 CFR 50.34 and Appendix A apply to plants already operating at the time these requirements for construction permit applications were issued.

Nevertheless< Rochester Gas and Electric Corporation hereby applies> pursuant to 10 CFR 50.12(a)> for an exemption from the provisions of 10 CFR 50 Appendix A authorizing alternative pipe break analyses to establish the structural design bases resulting from pipe breaks in connection with license DPR-18.

to 10 CFR 50.12(a)> we believe the requested exemption Further'ursuant will not endanger life or property or the common defense and security and is in the public interest.

The NRC staff Safety Evaluation Report (SER) attached to Generic Letter 84-04 assessed the relative costs and public risks of using advanced fracture mechanics techniques to justify revised design bases for several operating PWRs. The results of the assessment indicate that the effects of loss of coolant accidents which occur as a result of the postulated pipe ruptures< and which have been previously analyzed> are little7changed. Only a small reduction in core melt frequency (1 x 10 events/reactor year) and 'only a small reduction in public risk (3 1/2) man-rem total

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for the nominal case for all 16 plants considered) would be achieved if modifications to the plants were required in lieu of using the advanced fracture mechanics techniques. WCAPs 9558 and 9787, which were submitted by the affected utilities> established that there is a large margin against unstable extension of a crack in reactor coolant system piping. Based on this evaluationi this application does not involve a significant increase in the probability or consequences of an accident previously evaluated>

does not create the possibility of a new accident or one different from any previously evaluated> and does not involve a reduction in the margin of safety. Therefore> the proposed amendment does not involve a significant hazards consideration as defined in 10 CPR 50.92.

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