ML20154M305

Forwards Safety Evaluation Approving Rev 1 to BAW-1847 & BAW-1889P Re Elimination of Postulated Pipe Breaks in PWR Primary Main Loops.Protection Against Dynamic Loads Requires Request for Exemption from GDC 4 Per Generic Ltr 84-04

ML20154M305
Person / Time
Site:	Washington Public Power Supply System
Issue date:	02/18/1986
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Mazur D WASHINGTON PUBLIC POWER SUPPLY SYSTEM
References
GL-84-04, GL-84-4, NUDOCS 8603140033
Download: ML20154M305 (3)

Text

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February 18, 1986 b6Clk I

L Docket No. 50-460 t

Mr. D. W. Mazur Managing Director Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99352

SUBJECT:

SAFETY EVALUATION OF B&W OWNERS GROUP REPORTS DFALING WITH ELIMINATION OF POSTULATED PIPE SREAKS IN PWR PRIMARY MAIN LOOPS Re: WPPSS Nuclear Project No.1 (WNP-1)

The NRC staff has reviewed the B&W Owners Group reports BAW-1847, Rev. 1, and ,

BAW-1889P which apply " leak-before-break" technology as an alternative 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 technical basis has been provided to eliminate, as a design basis, the dynamic effects of large ruptures in the main loop pipirg of those B&W Owners Group facilities Ifsted in the enclosure. 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 (GOC 4).

Such exemptions must 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, " Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," February 1,1984. 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.

Generic letter 84-04 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 probability of rupturing piping is extremely low under design basis conditions. On July 1,1985, the Comission 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 5

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Mr. D. W. Mazur 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,

% IGIXALsI m g John .5 ,'0'i rector PWR Project Directorate #6 Division of PWR Licensing-B

Enclosure:

As Stated cc w/ enclosure:

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I Mr. D. W. Mazur WPPSS Nuclear Project No. 1 l Washington Public Power Supply System (WNP-1)

! CC:

Mr. V. Mani Nicholas D. Lewis, Chairman United Engineers & Constructors, Inc. State of Washington 30 South 17th Street Energy Facility Site Evaluation Philadelphia, Pennsylvania 19101 Council Mail Stop PY-11 Nicholas S. Reynolds, Esq. Olympia, Washington 98504 Bishop Liberman, Cook, Purcell and Reynolds 1200 Seventeenth Street, N.W., Mr. Eugene Rosolie Suite 700 Coalition for Safe Power l Washington, D. C. 20036 Suite 527 408 South West Second Street Mr. E. G. Ward Portland, Oregon 97?04 Senior Project Manager Babcock & Wilcox Company Nina Bell P.O. Box 1260 Nuclear Information and Resource Lynchburg, Virginia 23505 Service l

1346 Connecticut Avenue, N.W.

i Resident Inspector /WPPSS NPS Washington, D. C. 20036 c/o U.S. Nuclear Regulatory Comission l

P.O. Box 69 l

Richland, Washington 99352 l Mr. R. B. Borsum l

Nuclear Power Generation Division i Babcock & Wilcox 7910 Woodmont Avenue, Suite 220 l Bethesda, Maryland 20814 G. E. Craig Doupe, Esq.

, Washington Public Power Supply System l

P.O. Box 968 Richland, Washington 99352 Regional Administrator, Region V U.S. Nuclear Regulatory Comission, 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 I

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1. 'o UNITED STATES l' ^ ,t NUCLEAR REGULATORY COMMISSION B ,I WASMNGTON D. C. 205$5 December 12, 1985

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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 p a tulated breaks of primary main loop piping will require an exemption from Gerneral 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. Crystal River 3 8. Davis-Besse 1 0, , . { (D _

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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 probability 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.

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

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

Sincerely, Wd DennisM.Cru{.tchfi'eTd , <4 Director Y

51stant for Technical Support Division of PWR Licensing-B Office of Nuclear Reactor Regulation

Enclosure:

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

o ATTACHMENT THE B&W OWNERS GROUP .

DOCKET NOS. 302 50-269,270,287,&466 346, 438, 439, 312, 313, 329, .

SAFETY EVALUATION REPORT ON THE ELIMINATION OF LARGE PRIMARY LOOP RUPIURES A5 A DESIGN BASIS -

Section A  !

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

report (Reference 1)onthetechnicalbasesforeliminatinglargeprimary loop piping ruptures as a design basis. Reference 1 presented the results of a bounding evaluation for the following B&WDG members:

Licensee or Applicant Facility l

Arkansas Power A Light Co. AND-1 Consumers Power Co. Midland-2 Duke Power Co. Gconee 1, 2, 3 Florida Power Corp. Crystal River 3 Sacramento Municipal Utility District Rancho Seco Supply System WNP-1

, Tennessee Valley Authority Bellefonte 1, 2 ,,

.- i Toledo Edison Co. Davis Besse 1 ',

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The Reference I submittal was made to provide technical justification for the preceding licensees and applicants of the B&WOG in regard to a request for an exemption to General Design Criterion (GOC) 4 of Appendix A to 10 CFR Part 50 in regard to the need for protectica 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,

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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, Reference 3,andanadditionalreport(Reference 4)onmaterialproperties  !

data, both of which were transmitted to the NRC on October 22, 1985.

By means of deterministic fracture mechanics analyses, the B&WOG contends thatpostulateddouble-endedguillotinebreaks(DEG8s)oftheprimaryloop l reactor coolant piping will not occur in the facilities addressed in l 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 l against the %namic effectif associated with such postulated breaks. No  !

other changes in design requi?oments art addressed within the scope of the referenced reports; e.g., to changes to the definition of a LOCA nor its relationship to the regulations addressing design requirements for ,

ECCS(10CFR50.46),containernt(GDC16,50),otherengineeredsafety features and the conditions for environmental qualification of equipment l (10CFR50.49). .

i The Commission's regulations require provision of protective measures against  ;

the dynamic offacts 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, l jetimpingementshieldsorcompartments. In 1975, cerns arose as to the asymmetricloadsonpressurizedwaterreactor(PWR)v.sselsandtheir internals which could result from these large postulated breaks et discrete locations in the main primary coolant loop piping. This led to the establish-mentofUnresolvedSafetyIssue(USI)A-2,"AsymmetricBlowdownLoa'dsonPWR Pr* _rj ,t .

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i 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-l ,

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 (oneofthesefacilities,FortCalhounhasacombustionEngineeringnuclear ,

l i I steamsupplysystem). This NRC staff position was stated in Generic Letter - 84-04,publishedonFebruary1,1984(Reference 5). The generic letter states l

J that the affected licensees must justify an exemption to GDC 4 on a plant-specific basis. Other PWR applicants or licer. sees may request similar l

' exesiptions from the requirements of GOC 4 provided that they submit an

' acceptable technical basis for eliminating the need to postulate pipe i breaks.

i The acceptance of an exemption was made possible by the development of advanced fracture mechanics, technology. These advanced fracture mechanics techniques deal with relatively small flaws in piping components (either l l postulated or real) and examine their behavior under various pipe loads. l l-l The objective is to demonstrate by deterministic analyses that the detec-tion of small flaws by either inservice inspection or leakage monitoring systems is assured long before the flaws can grow to critical or unstable f sizes which could lead to large break areas such as the DEGB or its l.* equivalent. The concept underlying such analyses is referred to as l " leak-before-break" (LB8). There is no implication that piping failures  ;

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cannot occur, but rather that improved knowledge of the failure modes of l

piping systems and the application of appropriate remedial seasures, if

! indicated, can reduce the probability of catastrophic failure to insignifi-7 f

cant values.

Advanced fracture mechanics technology was applied in topical ~re' ports f

! (References 6,7,and8)submittedtothestaffbyWestinghouseonbehalfof the licensees belonging to the USI A-2 Owners Group. Althoughthe, topical l

• • corts w m intended to resolve the issue of asymmetric blowdown loads that l

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

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

Probabilistic fracture mechanics studies conducted by the Lawrence Livermore National Laboratories (LLNL) on both Westinghouse and Combustion Engineering nuclear steam supply system main loop piping (Reference 9) confirm that

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I both the probability of leakage (e.g., undetected flaw growth through the pipe wall by fatigue) and the probability of a DEG8 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 x 10~7 per plant year and the best-estimate DEG8 proba-l 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 l, 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 B&W 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 Com- j bustion Engineering studies. These results do not affect core melt -

probabilities in any significant way.

During the past few years it has also become apparent that the requirement I for installation of large, massive pipe whip restraints and jet impingement l shields is not necessarily the most cost effective way to achieve the

desired level of safety, as indicated in Enclosure 2, Regulatory Analysis, l to Reference 5. Even for new plants, these devices tend to restrict access I for future inservice inspection of piping; or if they are removed and reinstalled for inspection, there is a potential risk of damaging the i 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 EVAL.UA'TED BY THE STAFF The B&WOG facilities evaluated in Reference 3 include both 177-FA and 205-FA

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plants and configurations of the lowered-and-raised-loop designs. The pri-

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mary coolant loop piping of these facilities are comprised of straight sections 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 combined loads from normal operation and the Safe Shutdown Earth-quake (SSE);

potential cracking mechanisms; 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; margin based on crack size; and . .

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

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 includt the static forces and moments (pressure, deadweight and themal expansion) due to poreal operation,

~and the forces and moments associated with the safe shutdown earth-quake (SSE). These 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/syttens under avaluation, all pertinent infomation 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.

]e (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  ;

capability when the pipe is subjected to nomal operational loads.

(4) It should be demonstrated that the postulated leakage crackfis stable under normal plus SSE loads for long periods of time; that is,. crack a-~th if == ir minimal during an earthquake. The margin, in terms

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3 (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.

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(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 4

to cracking failure from the effects of corrosion (e,g., inter-j . granular stress corrosion cracking), water hammer, or fatigue (Iow and high cycle). This operating history totals over 53 reactor-years spanning 13 years of operation.

l (3) The leak rate calculations perfomed for the B&WOG facilities used

initial postulated throughwall flaws larger in size than those of Enclosure I to Reference 5. S&WDG facilities have an RCS pressure

. boundary leak detection system which is consistent with the guide-lines of Regulatory Guide 1.45 such that leakage of one (1) gpa in

, one hour can be detected. The calculated leak rate through the l postulated flaw is large relative to the staff's required sensitivity j, ' of plant leak detection systems; the margin is at least a factor of

ten (10)onleakage. '

! (4) The margin in terms of load based on fracture mechanics analyses for l the leakage-size crack under normal plus SSE loads (Service Level D l loads)meetsNUREG-1061, Volume 3,guidanceonmargins. Based on i a limit-load analysis, the load margin is at least (T. Similarly, based on the J limit, the margin is at least f T .

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_9 (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&WDG first l 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 limit'ing J-R curve and i

the associated tensile stress-strain curve was selected for the fracture analyses 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 i

j above, the staff concludes that the probability or likelihood of large I pipe breaks occurring in the primary coolant system loops of the B&WDG facilities is sufficiently low such that %namic effects associated i

!~ with postulated pipe breaks in these facilities need not be a design ,

basis.

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1. B&W Owners Group Report BAW-1847 " Leak-Before-Break Evaluation of Marg" ins Against Full Break for RUS Primary Piping of B&W Designed NSS, September 1984.

2. Letter to L. C. Oakes of the B&W Owners Group," B&WOG Leak-Before-Break Report; BAW-1847," dated March 12, 1985.

3. B&W Owners Group Report BAW-1847, Rev. I " Leak-Before-Break Evaluation of Margins Against Full Break for RCS Primary Piping of B&W Designed NSS," Septes wr 1985. I

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

Esanuelson, October 1985, proprietary. ,

5. " Safety Evaluation of Westi NRC Generic Reports DealingLetter 84-04,ination of Postulated Breaksn PWR with Elim house Topical Primary Main Loops," February 1,1984.

6. Westinghouse Report WCAP-9558 Rev. 2 " Mechanistic Fracture Evaluation ofReactorCoolantPipeContaIningafostulatedCircumferentialThrough-wall Crack," May 1981, Class 2 proprietary. .

7. " Tensile and Toughness Properties of Westinghouse Primary Piping Weld Report MetalWCAP-9687,Use for in Mechanistic Fracture Evaluation,"

May 1981, Class 2 proprietary.

8. Westinghouse Response to Questions and Comments Raised by Members of -

ACRS Subcommittee on Metal Components During the Westinghouse Presenta-tion on Se Darrell G.ptember 25,November Eisenhut, 1981 Letter 10,Report NS-EPR-2519,Class 1981, Westinghouse E. P.2Rahe to proprietary.

l 9. T. Lo, H. H. Woo, G. S. Holman and C. K. Chou, " Failure Probability of PWRReactorCoolantLoopPipig",resentedattheASMEPVPConference and Exhibition, June 17 21, 1 n Antonio, Texas.

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