ML20154M753

Advises That Proposed Mods to GDC 4 in Process of Final Rulemaking.Promulgation of Final Rule Will Eliminate Need for Exemption Requests & Performance of Safety Balances. Requested Info on Leakage Detection Sys Should Be Submitted

ML20154M753
Person / Time
Site:	Arkansas Nuclear
Issue date:	02/18/1986
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Jeffery Griffin ARKANSAS POWER & LIGHT CO.
References
GL-84-04, GL-84-4, NUDOCS 8603140459
Download: ML20154M753 (3)

Text

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• February 18, 1986 h b Ck k 7

i Docket No. 50-313 Mr. John M. Griffin, Senior Vice President of Energy _ Supply

. Arkansas Power and Light Company P. O. Box 551 Little Rock, Arkansas 72203

SUBJECT:

SAFETY EVALUATION OF B&W OWNERS GROUP REPORTS DEALING VITH ELIMINATION OF POSTULATED PIPE BREAKS IN PWR PRIMARY PAIN LPDPS Re: Arkansas Nuclear One, Unit No.1 The NRC staff has reviewed the B&W Owners Group reports BAW '.847 Rev.1, and BAW-1889P which apply " leak-before-break" technology as an riternative to designing against dynamic loads associated with postulated ruptures of primary coolant loop piping. As discussed in the enclosed letter to the B&W Owners Group, we have concluded that an acceptable technt;al basis has been provided to eliminate, as a design basis, the dynamic ef ects of large ruptures in the main loop piping of those RAW Owners Grr,up facilities listed in the enclosure. Authorintion by the NRC to nut provide protection against the dynamic loads resulting from postulated breaks of primary main loop l

- piping will require an exemption from General Design Criterion 4 (GDC4).

Such exemptions must be justified on a facility specific basis. Each request for an exemption should fr.clude a safety balance in accordance with the guidance provided in NRC Generic Letter 84-04, " Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main loops," February 1,19'34 In addition, information for each facility should be submitte>J to demonstrate that leakage j -detection systems installed at the facility comr,1y with Regulatory Guide 1.45. .

Generic Letter 84-04 informed all operating PW licensees, construction l permit holders and applicants for construction permits of the staff's intent I

to proceed with rulemaking changes to GDC 4 to permit the use of analyses

- that demonstrate the probability of rupturing piping is extremely low under design basis conditions. On July 1, 1985, the Commission published a proposed modification to GDC-4 which would permit the use of such analyses for PWR primary coolant loop piping. The NRC staff is currently in the 3 g g 1

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Mr. Griffin process of final rulemaking. Promulgation of the final rule will eliminate the need for exemption requests and performance of safety balances; however,

< the- requested information on leakage detection systems should be submitted..

Sincerely,

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John F. Sto z, Director PWR Project Directorate #6 Division of PWR Licensing-R l

Enclosure:

As Stated cc w/ enclosure:

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Mr. J. M. Griffin Arkansas Power & Light Company Arkansas Nuclear One, Unit I cc:

Mr. J. Ted Enos, Manager Licensing Arkansas Power & Light Company P. O. Box 551 Little Rock, Arkansas 72203 Mr. James M.. Levine, General Manager ,

Arkansas Nuclear One P. O. Box 608 Russellville, Arkansas 72801 Mr. Nicholas S. Reynolds Rishop. Liberwen, Cook, Purcell A Reynolds 1200 Seventeenth Street, NW Washington, D.C. 20036 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 Resident Inspector U.S. Nuclear Regulatory Commission

. P. O. Box 2090 Russellville, Arkansas 72801 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive Suite 1000

• Arlington, Texas 76011 Mr. Fr:nk Wilson, Director Division of Environmental Health Protection Department of Health Arkansas Department of Health 4815 West Markham Street Little Pock Arkansas /2201 Honorable Ermil Grant Actino County Judge of Pope County Pope &ountyCourthouse Russellville, Arkansas 72801 l l

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1. #g UNITE 3) STATES 8 o NUCLEAR REGULATORY COMMISSION y I WASHINGTON, D. C. 20555 g [ December 12, 1985 Mr. L. C. Oakes, Chairman B&W Owners Group Leak-Before-Break Task Force Washington Public Power Authority P.O. Box 460 3000 George Washington Way Richland, Washington 99352

SUBJECT:

SAFETY EVALUATION OF B&W OWNERS GROUP REPORTS DEALING WITH ELIMINATION OF POSTULATED PIPE BREAKS IN PWR PRIMARY MAIN LOOPS

Reference:

1. B&W Owners Group report BAW-1847, Rev.1, " Leak-Before-Break Evaluation of Margins Against Full Break for RCS

, Primary Piping of B&W Designed NSS," September 1985.

2. B&W Owners Group Report BAW-1889P, " Piping" Material Properties for Leak-Before-Break Analysis, A. L. Lowe, Jr. , K. K. Yoon, and R. H. Emanuelson, October 1985, proprietary.

3. NRC Generic Letter 84-04, " Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Breaks in PWR Primary Main LOOPS,"

February 1, 1984.

The NRC staff has completed its review of the referenced B&W Owners Group reports which apply " leak-before-break" technology as an alternative to designing against dynamic loads associated with postulated ruptures of primary coolant loop piping.

The staff evaluation concludes that an acceptable technical basis has been provided to eliminate, as a design basis, the dynamic effects of large ruptures in the main loop primary piping of the B&W Owners Group facilities.*

Authorization by the NRC to not provide protection against the dynamic loads resulting from postulated breaks of primary main loop piping will require an exemption from General Design Criterion 4 (GDC4). Such exemptions must

• 1. ANO-1 5. Rancho Seco-

2. Midland-2 6. WNP-1

3. Oconee 1,2,3 7. Bellefonte 1,2

4. Crysta] River 3 8. Davis-Besse 1 I$

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c-be justified on a facility specific basis. Each request for an exemption should include a safety balance in accordance with the guidance provided in NRC Generic Letter 84-04 (Reference 3). In addition, information for each facility should be submitted to demonstrate that leakage detection systems installed at the facility comply with Regulatory Guide 1.45.

Reference 3 informed all operating PWR licensees, construction permit holders and applicants for construction permits of the staff's intent to proceed with rulemaking changes to GDC-4 to permit the use of analyses that demonstrate the prot, ability of rupturing piping is extremely low under design basis conditions. On July 1, 1985, the Commission published a pro-posed modification to GDC-4 which would permit the use of such analyses for PWR primary coolant loop piping. The NRC staff is currently in the process of final rulemaking. Promulgation of the final rule will eliminate the need for exemption requests and performance of safety balances; how-ever, the requested information on leakage detection systems should be submitted.

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By copy of this letter with enclosed safety evaluation report, Mr. J. F. Walters of Babcock & Wilecx is being informed of this action.

This information is also being transmitted to participating licensees and applicants of the B&W Owners Group.

Sincerely,

}&y ~7)

DennisM.Cru(,chied, t

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sistant Director for Technical Support Division of PWR Licensing-B Office of Nuclear Reactor Regulation

Enclosure:

As stated cc: J. F. Walters, B&W

2- ATTACHNENT

. THE B&W OWNERS GROUP DOCKET N05. 50-269, 270, 287, 302, 312, 313, 329, s

346, 438, 439, & 460 .

SAFETY EVALUATION REPORT ON THE ELININATION OF LARGE PRIMARY LOOP RUPIURt5 A5 A DESIGN BASIS Section A Engineering Branch Division of PWR Licensing-B

- INTRODUCTION By letter dated September 7, 1984, the B&W Owners Group (B&WOG), on behalf of participating utilities with B&W designed facilities, submitted a generic

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report (Reference 1) on the technical bases for eliminating large primary loop piping ruptures as a design basis. Reference 1 presented the results of a bounding evaluation for the following B&WDG members:

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Licensee or Applicant Facility Arkansas Power & Light Co. ANO-1 Consumers Power Co. Midland-2 Duke Power Co. Oconee 1, 2, 3 -

Florida Power Corp. Crystal River 3 i Sacramento Municipal Utility District Rancho Seco Supply System WNP-1 Tennessee Valley Authority Bellefonte 1, 2 , ,

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Toledo E.dison Co. Davis-Besse 1 C k<

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The Reference 1 submittal was made to provide technical justification for

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l- the preceding licensees and applicants of the B&WOG in regard to a request for an exemption to General Design Criterion (GDC) 4 of Appendix A to 10 CFR Part 50 in regard to the need for protection against dynamic effects from postulated pipe breaks. After meeting with the B&WOG, the

! staff formally responded by letter (Reference 2) dated March 12, 1985, i

to transmit the staff's comments and questions on the submittal. The  !

! response to the staff's concerns resulted in a revision to the submittal,

l. Reference 3, and an additional report (Reference 4) on material properties data, both of which were transmitted to the NRC on October 22, 1985.

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l By means of deterministic fracture mechanics analyses, the B&WOG contends that postulated double-ended guillotine breaks (DEGBs) of the primary loop reactor coolant piping will not occur in the facilities addressed in i References 3 and 4 and therefore need not be considered as a design basis for installing protective devices such as pipe whip restraints to guard against the dynamic effects associated with such postulated breaks. No

- other changes in design requirements are addressed within the scope of the referenced reports; e.g., no changes to the definition of a LOCA nor its relationship to the regulations addressing design requirements for ECCS(10CFR50.46), containment (GDC16,50),otherengineeredsafety

,. features and the conditions for environmental qualification of equipment (10CFR50.49). .

The Commission's regulations require provision of protective measures against the dynamic effects of postulated pipe breaks in high energy fluid system piping. Protective measures include physical isolation from postulated pipe l

rupture locations if feasible or the installation of pipe whip restraints, jetimpingementshieldsorcompartments. In 1975, concerns arose as to the i

asymmetric loads on pressurized water reactor (PWR) vessels and their l internals which could result from these large postulated breaks it discrete locations in the main primary coolant loop piping. This led to the establish-mentofUnresolvedSafetyIssue(USI)A-2,"AsymmetricBlowdownLoa'dsonPWR l Primary Systems."

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The NRC staff, after several review meetings with the Advisory Committee on Reactor Safeguards (ACRS) and a meeting with the NRC Committee to Review Generic Requirements (CRGR), concluded that an exemption from the regula-tions would be acceptable as an alternative for resolution of USI A-2 for 16 facilities owned by 11 licensees in the Westinghouse Owner's Group (one of these facilities, Fort Calhoun has a Combustion Engineering nuclear steamsupplysystem). This NRC staff position was stated in Generic Letter 84-04, published on February 1, 1984 (Reference 5). The generic letter states thattheaffectedlicenseesmustjustifyanexemptiontoGDC4onaplant-

,l specific basis. Other PWR applicants or licensees may request similar exeniptions from the requirements of GDC 4 provided that they submit an acceptable technical basis for eliminating the need to postulate pipe breaks.

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The acceptance of an exemption was made passible by the development of ,

advanced fracture mechanics technology. These advanced fracture mechanics j techniques deal with relatively small flaws in piping components (either '

postulatedorreal)andexaminetheirbehaviorundervariouspipeloads.

Theobjectiveistodemonstratebydeterministicanalysesthatthedetec-tion of small flaws by either inservice inspection or leakage monitoring f systems is assured long before the flaws can grow to critical or unstable

.. sizes which could lead to large break areas such as the DEG8 or its ,

equivalent. The concept underlying such analyses is referred to as

" leak-before-break" (L88). There is no implication that piping failures cannot occur, but rather that improved knowledge of the failure modes of piping systems and the application of appropriate remedial measures, if indicated, can reduce the probability of catastrophic failure to insignifi-cant values.

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i Advanced fracture mechanics technology was applied in topical re' ports l (References 6,7,and8)submittedtothestaffbyWestinghousehnbehalfof l the licensees belonging to the USI A-2 Owners Group. Although the. topical

! reports were intended to resolve the issue of asymmetric blowdown loads that

resulted from a limited number of discrete break locations, the technology i

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advanced in these topical reports demonstrated that the probability of breaks

, occurring in the primary coolant system main loop piping is sufficiently low I

such that these breaks need not be considered as a design basis for requirin2 installation of pipe whip restraints or jet impingement shields. The staff's l Topical Report Evaluation is attached as Enclosure 1 to Reference 5.

Probabilistic fracture mechanics studies conducted by the Lawrence Livermore i National Laboratories (LLNL) on both Westinghouse and Combustion Engineering l, nuclear steam supply system main loop piping (Reference 9) confirm that both the probability of leakage (e.g., undetected flaw growth through the

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pipe wall by fatigue) and the probability of a DEGB are very low. The ,

results given in Reference 9 are that the best-estimate leak probabilities

for Westinghouse nuclear steam supply system main loop piping range from 1.2 x 10 -8 to 1.5 r. 10 ~7 per plant year and the best-estimate DEG8 proba-bilities range from 1 x 10-12 to 7 x 10 -12 per plant year. Similarly, the -

best-estimate leak probabilities for Combustion Engineering nuclear steam

! supply system main loop piping range from 1 x 10 -8 per plant year to 3 x 10-8 per plant year, and the best-estimate DEG8 probabilities range

,' from 5 x 10~14 to 5 x 10-13 per plant year. In addition, LLNL recently

! conducted an evaluation of 8W nuclear steam supply main loop piping

! with the result that the best-estimate leak and DEG8 probabilities are ,

nominally identical to those calculated for the Westinghouse and Con-

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bustion Engineering studies. These results do not affact core melt -

probabilities in any significant way.

During the past few years it has also become apparent that the requirement >

forinstallationoflarge,massivepipewhiprestraintsandjetimpingement shields is not necessarily the most cost effective way to achieve the '

l i desired level of safety, as indicated in Enclosure 2, Regulatory Analysis, i to Reference 5. Even for new plants, these devices tend to restrict access ,

f for future inservice inspection of piping; or if they are removed and l reinstalled for inspection, there is a potential risk of damaging the

! piping and other safety-related components in this process. If installed i

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, in operating plants, high occupational radiation exposure (ORE) would be incurred while public risk reduction would be very low. Removal and reinstallation for inservice inspection also entail significant ORE over the life of a plant.

PARAMETERS EVALUATED BY THE STAFF The B&WOG facilities evaluated in Reference 3 include both 177-FA and 205-FA plants and configurations of the lowered-and-raised-loop designs. The pri-mary coolant loop piping of these facilities are comprised of straight sect, ions and elbows in each of four pipe sizes - 28, 32, 36 and 38 inch ,

diameters. The piping materials in the primary main loops are low alloy ferritic' steels (SA-106 GrC, SA-508 C1 1, and SA-516 Gr 70) and wrought stainless steel safe ends (SA-376 TP 316). In its review of References 3 ,

and 4, the staff evaluated the B&WOG analyses and materials data with regard to:

- the location of maximum stresses in the piping, associated with the

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combined Icads from normal operation and the Safe Shutdown Earth-quake (SSE);

potential cracking mechanisms; i - size of postulated 'through-wall cracks that would leak a detectable amount under normal loads and pressure; stability of a " leakage-size crack" under normal plus SSE loads and the expected margin in terms of load; a

margin based on crack size; and - . .

i the fracture toughness properties of low alloy, ferritic steel piping, wrought stainless steel safe ends and associated weld material.

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STAFF CRITERIA USED IN THE EVALUATION The NRC staff's criteria for evaluation of the above parameters are delineated in the Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, NUREG-1061, Volume 3, " Evaluation of Potential for Pipe Breaks." These criteria are enumerated in Chapter 5.0 of Volume 3 of the NUREG and are as follows:

(1) The loading conditions should include the static forces and moments (pressure, deadweight and thermal expansion) due to normal operation,

• and the forces and moments associated with the safe shutdown earth-quake (SSE). Those forces and moments should be located where the highest stresses, coincident with the poorest material properties, are induced for base materials, weldsents and safe-ends.

(2) For the piping run/ systems under evaluation, all pertinent information which demonstrates that degradation or failure of the piping resulting from stress corrosion cracking, fatigue or water hammer is not likely, should be provided. Relevant operating history should be cited, which includes system operational procedures; system or component modifica-tion; water chemistry parameters, limits and controls; resistance of material to various foms of stress corrosion, and performance under ,

cyclic loadings.

(3) A through-wall crack should be postulated at the highest stressed locations determined from (1) above. The size of the crack should be large enough so that the leakage is assured of detection with at least a factor of ten using the minimum installed leak detection capabilitywhenthepipeissubjectedtonormaloperationalloads.

(4) Itshouldbedemonstratedthatthepostulatedleakagecrackfisstable under normal plus SSE loads for long periods of time; that is,. crack growth, if any, is minimal during an earthquake. The margin, in terms

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of applied loads, should be at least the /T and should be determined by a crack stability analysis, i.e., that the leakage-size crack will not experience unstable crack growth even if larger loads (larger than designloads)areapplied. This analysis should demonstrate that l crack growth is stable and the final crack size is limited, such that a double-ended pipe' break will not occur.

(5) The crack size should be determined by comparing leakage-size crack to critical-size cracks. Under normal plus SSE loads, it should be l

demonstrated that there is a margin of at least 2 between the leakage-

. size crack and the critical-size crack to account fo'r the uncertainties inherent in the analyses, and leakage detection capability. A limit-load analysis may suffice for this purpose; however, an elastic plastic

! fracture mechanics (tearing instability) analysis is preferable.  ;

(6) The materials data provided should include types of materials and materials specifio tions used for base metal, wel bents and safe-ends, the materials properties including the J-R curve used in the analyses,  !

l, and long-term effects such as thermal aging and other limitations to

. valid data (e.g., J maximum, maximum crack growth).  ;

i j STAFF EVALUATION AND CONCLUSIONS Based on its evaluation of the analysis contained in BAW-1847, Rev. 1 (Reference 3)andthematerialsdatapresentedinBAW-1889P(Reference 4),

thestafffindsthattheB&WOGhaspresentedacceptabletechnicaljustifi-

cation, addressing the preceding criteria, to eliminate, as a design basis, ,

the dynamic effects of large ruptures in the main loop primary coolant piping of the B&WOG facilities evaluated. Specifically:

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(1) The loads associated with the highest stressed location in the main loop primary system piping are 1,685.7 kips (axial), 37,171 in-kips (bending moment) and result in maximum stresses of about 51% of Service Level D limits specified in Section III of the ASME Code.

(2) For the B&WOG facilities, there is no history of cracking failure in reactor primary coolant system main loop piping. The reactor coolant system primary loop has an operating history which demon-strates its inherent stability. This includes a low susceptibility  ;

tocrackingfailurefromtheeffectsofcorrosion(e,g., inter-

.granularstresscorrosioncracking),waterhammer,orfatigue(Iow and,highcycle). This operating history totals over 53 reactor-years spanning 13 years of operation.

(3) The leak rate calculations performed for the B&WOG facilities used initial postulated throughwall flaws larger in size than those of Enclosure I to Reference 5. B&WOG facilities have an RCS pressure boundary leak detection system which is consistent with the guide- 7 linesofRegulatoryGuide1.45suchthatleakageofone(1)gpain

! one hour can be detected. The calculated leak rate through the

, postulated flaw is large relative to the staff's required sensitivity t

! ,, of plant leak detection systems; the margin is at least a factor of i ten (10) on leakage. ,

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(4) The margin in terms of load based on fracture mechanics analyses for  !

theleakage-sizecrackundernormalplusSSEloads(ServiceLevelD loads)meetsNUREG-1061, Volume 3,guidanceonmargins. Based on ,

l alimit-loadanalysis,theloadmarginisatleast(T. Similarly,  !

based on the J Ilmit, the margin is at least [T.

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(5) The margin between the leakage-size crack and the critical-size crack w'as calculated by a limit load analysis. Again, the results demonstrated that a margin of at least 2.0 exists and is within the guidelines of NUREG-1061, Volume 3.

(6) In their review of the reactor coolant piping, the B&WOG first listed all the base metals and weld metals represented. From a review of published test data -- J-R curves and tensile properties --

the materials from the list that were most likely to be limiting were identified. A test program was then conducted to obtain the toughness

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and tensile data required. From these data, a limiting J-R curve and the associated tensile stress-strain curve was selected for the fracture an'a lyses'of the base metal and weld metal in the straight sections and

- elbows of the piping identified for evaluation. The staff concludes that the choice of limiting materials is satisfactory.

In view of the analytical results presented in Reference 3, the materials data contained in Reference 4, and the staff's evaluation findings related above, the staff concludes that the probability or likelihood of large pipe breaks occurring in the primary coolant system loops of the B&WOG facilities is sufficiently low such that dynamic effects associated with postulated pipe breaks in these facilities need not be a design ,

basis.

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1. B&W Owners Group Report BAW-1M7 " Leak-Before-Break Evaluation of Marg"insAgainstFullBreakforRd5PrimaryPipingofB&WDesigned NSS, September 1984.

2. Letter to L. C. Oakes of the B&W Owners Group," B&WOG Leak-Before-Break Report, BAW-1M 7," dated March 12, 1985.

3. B&W Owners Group Re mrt BAW-1M 7, Rev. 1 " Leak-Before-Break Evaluation of Mar ins instrllBreakforRCSPrImaryPipingofB&WDesigned u

NSS," t r 1985.

4. B&W Owners Group Report BAW-1889P, " Piping Material Properties for Leak-Before-Break Analysis," A. L. Lewe, Jr., K. K. Yoon and R. H.

Emanuelson, October 1985, proprietary. ,

5. " Safety Evaluation of Westinghouse Topical NRC ReportsGeneric DealingLetter M-M,ination of Postulated Breaks in PWR Primary with Elim Main Loops," February 1,1984.

! 6. Westinghouse Report WCAP-9558 Rev. 2 " Mechanistic Fracture Evaluation

! ofReactorCoolantPipeContaIningahostulatedCircumferentialThrough-j wall Crack," May 1981, Class 2 proprietary.

7. " Tensile and Toughness Properties of i Westinghouse Primary Piping WeldReport WCAP-9687,Use Metal for in Mechanistic Fracture Evaluation,"

l* May 1951, Class 2 proprietary.

! 8. Westinghouse Response to (>stions and Comments Raised by Members of -

l AfRS Subcommittee on Meta' Components During the Westinghouse Presenta-tion on Se j Darrell G.ptember Eisenhut,25, November 1981Westinghouse 10, 1981, Letter ReportClass NS-EPR-2519, 2 proprietary.E. P. Rahe to i.. '

l 9.

T.Lo,H.H.

PWR Reactor Woo,G.S.HolmanandC.K. Chou Coolant Loop Piping," presented {"FailureProbabilityof a the ASME PVI Conference and Exhibition, June 17-21, 1984, San Antonio, Texas.

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