ML20106H982

Exemption from Portion of 10CFR50,App A,Gdc 4 Re Need to Analyze Large Primary Loop Pipe Ruptures as Structure Design Basis

ML20106H982
Person / Time
Site:	Vogtle
Issue date:	02/05/1985
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:	GEORGIA POWER CO.
Shared Package
ML20106H987	List: ML20106H982 ML20106H989
References
NUDOCS 8502150460
Download: ML20106H982 (15)

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UNITED STATES NUCLEAR REGULATORY COMMISSION In.the Matter of

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GEORGTA POWER COMPANY, ET. AL i

Docket Nos. 50-424

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and 50-425 (Vogtle Electric Generating Plant,

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Units 1 and 2)

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EXEMPTION I.

On August 1,1972, the Georgia Power Company, the Municipal Electric Authority of Georgia, thE Og'lethorpe Power Corporation, and the City of Dalton, Georgia (the licensees) tendered an application for licenses to construct Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle or the facility) with the Atomic Energy Commission (currently the Nuclear Regulatory Commission or the Commission). Following a public hearing before the Atomic Safety and

, Licensing Board, the Consnission issued Construction Permit Nos. CPPR-108 and CPPR-109 permitting the construction of Units 1 and 2, respectively, on June 28, 1974 Each unit of the facility is a pressurized water reactor, containino a Westinghouse' Electric Company nuclear steam supply system, located at the

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licensee's site in Burke County, Georgia.

On June 30, 1983, the licensees tendered an application for Operating Licenses for each unit of the facility, currently in the licensing review process.

II.

The Construction Permits issued for constructing the facility orovide, in certinent part, that the facility units are sub.iect to all rules, regulatiors 8502150460 850205 PDR ADOCK 05000424 A

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-2, and Orders of the Comission. This includes General Design Criterion (GDC) 4 of Appendix A to 10 CFR 50. GDC. 4 requires that structures, systems and com-ponents important to,afety shall be designed to accomodatekhe effects of, and to be compatible with, the environmental conditions associated with the normal operation, maintenance, testing and postulated accidents, including loss-of-coolant accidents. These structures, systems and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whi.pping, discharging fluids that may result from equipment failures, and from events an,d conditions outside the nuclear power unit.

In a submittal dated October 25, 1983, the apolicants enclosed Westinghouse Reoort MT-SME-3082 (Reference 1) containing the technical basis for their request to: (1) eliminate the need to postulate circumferential and longitudinal pipe breaks in the Reactor Coolant Sys'cem (?',5) primary loop (hot leg, cold leg and cross-over leg piping); (2) eliminate the need to install pipe whip restraints and jet impingement shields associated with previously postulated breaks in the RCS primary loops and; (3) eliminate the need to corsider dynamic effects. and loading conditions specifically associated with previous 1v oostu-lated pipe breaks in the RCS primary loop, including.iet impingement loads, cavity pressure loads, blowdown loads in the RCS and attached piping, and sub-i compartment pressure loads. By a subsequent submittal date,d April 2, 1984, the applicants requested an exemption from a portion of the reouirements of GDC 4 related to the above, in support of the prior request. The applicants also stated in their submittals that the exemption request does not apply to the design bases for the containment' including the desion basis for structural loading of subcompartment walls and floors, the emergency core cooling system, l

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or environmental qualification. The applicants also stated that the design of i

their reactor coolant system supports would remain unchanged.

Based onJjts. review of the applicants' October 25th submittal, the NRC staff requested additional information and provided comments on the reports i

l (References 1 and 9) which were transmitted to the applicant in the fonn of questions by NRC letter dated March 19, 1984 (Reference 2).

i By a submittal dated May 17, 1984, the applicants responded to the staff's questions, providing a new report identified as Westinghouse Report WCAP-10551 (Reference 3).

In a separat,e submittal, dated April 2, 1984, the applicants provided a value-impact analysis which, together with the technical information contained in the Reference 3 report, provided a comprehensive justification for i

requesting a partial exemption from the requirements of GDC 4.

By letter dated December 21, 1984, the applicants described their present

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installation status of the pipe whip restraints and,iet impingement shields for both Unit 1 and Unit 2.

For Unit 1, of the twenty-four (24) pipe whip restraints per unit, only support structures for sixteen (16) are installed. No bearing f,

bars or attachments have been installed and no shimming has begun. For Unit 2 eight (8) are similarly partial?y installed.

Installation has not begun for the j

remainder of the Unit 1 and Unit 2 restraints. Ndditionally, none of the,iet impingement shields for either unit have been installed.

1 From the deterministic fracture mechanics analysis contained in the tech-nical information furnished, the applicants concluded that the 60stulated double-ended guillotine breaks (DEGB) of the primary loop conlant pipinc j

in Vogtle, Units 1 and 2, need not be considered as a design basis for instal-ling protective structures, such as pipe whip restraints and jet impingement

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-4, shields, to guard against the dynamic effects associated with such postulated breaks. However, the applicant proposes to continue to postulate the DEGB as

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the design bas'is for the containment subcompartments, for the ECCS and for environmental qualification.

III.

The Commission's regulations require that applicants provide protective measures against the dynamic effects of postulated pipe breaks in high energy fluid system pipin~g. Protective measures include physical isolation from postulated pipe rupture loca'tions if feasible or the installation of pipe whip restraints, iet impingement shields or compartments.

In 1975, concerns arose as to the asymmetric loads on pressurized water reactor (PWR) vessels and their, internals which could result from these large postulated breaks at discrete locations in the main primary coolant loop piping. This led to the establishment of Unresolved Safety Issue (USI) A-2, " Asymmetric Blowdown Loads on PWR Primary Syrtems."

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 regulations would be acceptable as an alternative for resolution of USI A-2 for sixteen facilities owned by eleven licensees in the Westinghouse Ow'ner's Group (one of these facilities, Fort Calhoun has a Combustinn Engineering nuclear ste'am i

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supply system). This NRC staff position was stated'in Generic Letter 84-04, i

published on February 1,198a (P.eference 4). The generic letter states that the affected licensees must justify an exemption tn GDC 4 on a plant-specific basis. Other PWR applicants or licensees may request similar exemptions from

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5-the requirements of GDC 4 provided that they submit an acceptable technical basis for eliminating the need to postulate pipe breaks.

The acceptance of an exemption was made possible by the development of advanced fracture mechanics technology. These advanced fracture mechanics technioues deal with relatively small flaws in pkping components (either postulated or real) and examine their behavior under various pipe loads. The objective is to demonstrate by deterministic analyses that the detection of small flaws by either inservice inspection or leakage monitoring systems is assured lorg before the flaws can grow to critical or unstable sizes which could lead to large break areas such as the DEGB or its equivalent. The concept underlying such analyses is referred to as " leak-before-break" (LBR).

There is no implication that piping failures cannot occur, but rather that l

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 insignificant values.

Advanced fracture mechanics technology was applied in topical reports (References 5, 6 and 7) submitted to the staff by Westinghouse on behalf of 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 advanced in these topical reports demonstrated that the probability of breaks f

occurring in the primary coolant system main loop piping is suf iciently low -

such that these breaks need not be considered as a design basis for requiring installation of pipe whip restraints or jet impingement shields. Tha staff's l

Topical Report Evaluation is included as part of Reference 4 1

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Probabilistic fracture mechanics studies conducted by the Lawrence Liver-moreNationalLaboratoriesl(LLNL)on.bothWestinghouseandCombustionEngineer-ingnuclearsteamsupplysystemmainlooppiping(Reference 8Iconfirmthat both the probability of leakage (e.g., undetected flaw growth through the pipe wall by fatigue) and the probability of a DEGB are very low. The results given in Reference 8 are that the best-estimate leak probabilities for Westinghouse nuclear steam suoply system main loop piping range from 1.2 x 10-0 to 1.5 x 10~7 per plant year and the best-estimate DEGB probabilities range from 1 x 10-12 to 7.x 10-12 per plant year. S,imilarly,' the best-estimate leak probabilities for Combustion Engineering nuclear steam supply system main loop piping range from 1 x 10-0 per plant year to 3 x 10-8 per plant year, and the best-estimate DEGB probabilities range from 5 x 10-14'to 5 x 10-13 per plant year. 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 for installation of large, massive pipe whip ' restraints and iet impingement shields is not necessarily the mo:t cost effective way to achieve the desired level of safety, as indicated in Enclosure 2, Regulatory Analysis, to Reference 4.

Even for new plants, these devices tend to restrict access for future inservice inspection of piping; or if they are removed and reinstalled for inspection, there is a potential risk of damaging the piping and other,sa'ety-related.

components in this process.

If installed in operating plants, high occupational radiation exposure (ORE) would be incurred while public risk reduction would l

be very low. Removal and reinstallation for inservice inspection also entail sionificant ORE over the life of 'a plant.

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7 IV.

The primary coolant system of Yogtle, Units I and 2, described in Reference 3, has four main loops each comprising a 33.9 inch diameter hot leg, a 36.2 inch diameter crossover leg and 32.14 inch diameter cold leg piping.

The material in the primary loop piping is cast stainless steel (SA 351 CF8A).

In its review of Reference 3, the staff evaluated the Westinghouse analyses with regard to:

the location of maximum stresses in the piping, associated with the combined loads from nonnal operation and the SSE; potential cracking' mechanisms;-

size of 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 the fracture toughness properties of thermally-aged cast stainless steel piping and weld material.

The NRC staff's criteria for evaluation of the above parameters are delineated in its Topical Report Evaluation, Enclosure 1 to Reference 4, Section

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4.1, "NRC Evaluation Criteria", and are as follows:

(1) The loading. conditions should include t' e static forces and moments h

(pressure, deadweight and thermal expansion) due to normal 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 and the lowest material toughness are coincident for base materials, weldments and safe-ends.

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-8, (2) For the oiping run/ systems under evaluation, all pertinent information which demonstrates that degradation or failure of the piping resulting l

from stress corrosion cracking, fatigue or water hammer is not likely, J

should be provided. Relevant operating history shnuld be cited, which includes system operational procedures; system'or component modifica-tion; water chemistry parameters, limits and controls; resistance of material to various forms of stress corrosion, and performance under cyclic loadings.

(3) A through-wall cra,ck should be postulated at the highest stressed locationsdeterminedfrom(1)above. The size of the crack she::1d be large enough so that the leakage is assured of detection with adequate margin 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 crack is stable t

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 of applied loads, 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 design loads) are applied. This analysis should demonstrate that 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 comparino the leakage-size crack to critical-size cracks'. Under normal plus SSE loads, it should be demon-strated that there is adequate margin between the leakace-size crack

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and the critical-size crack to account for 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 specifications used for base metal, weldments and safe-ends, the materials properties including the J-R curve used in the analyses, and long-term effects such as thermal aging and other limitations to valid data (e.g. J. maximum, maximum crack growth).

_V :.

Based on its ev.aluation of the analysis contained in Westinghouse Report WCAP-1055T (Reference 3), the staff finds that the applicants have presented an acceptable technical justification, addressing the above criteria, for not installing protective devices to deal with the dynamic effects of large pipe ruptures in the main loop primary coolant system piping of Vogtle, Units 1 and 2.

This finding is predicated on the fact that each of the parameters evaluated for Vogtle is enveloped by the generic analysis performed by Westinghouse in

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Reference 5, and accepted by the staff in Enclosure I to Reference 4.

Specifically:

(1) The loads associated with the highest stressed location in the main loop primary system piping are 1,962 kips (axial), 28,810 in-kips (bending moment) and result in maximum. stresses of about 75% of the bounding stress used by Westinghouse in Reference 5.

Further, these loads are approximately 70% of those established by the staff as limits (e.g. a moment of 42,000 in-kips in Enclosure I to Referertce 4).

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(2) For Westinghouse plants, there is no history of cracking failure in 4

reactor primary coolant system loop piping. The Westinghouse reactnr j

coolarit system primary loop has an operating history which demonstrates its inherent stability. This includes a low susceptibility to cracking failure from the effects of corrosion'le.g. intergranular stress corrosion cracking), water hammer, or fatigue (low and high cycle). This operating history totals over 400 reactor-years, includin.g five (5) plants each having 15 years of operation and 15 i

other plants with over 10 years of operation.

(3) The results of the leak rate calculations performed for Vogtle, using i

an initial through-wall crack of 7.5 inches, are identical to those of Enclosure 1 to Reference 4.

The Vogtle plant has an RCS pressure bo'undary leak detection system which is consistent with the guidelines of Regulatory Guide 1.45, and it can detect leakage of one (1) gpm in one hour. The calculated leak rate through the postulated flaw results I

in a factor of at least 10 relative to the :,ensitivity of the Vogtle plant leak detection system.

(4) The margin in terms of load based on fracture mechanics analyset for the leakage-size crack under c'rmal plus SSE loads is within the bounds calculated by the staff in Section 4.2.3 o.f Enclosure 1 to Ref' erence 4.

Based on a limit-load analysis, the load margin is

- about P.9 and based on the J limit discussed in (6) below, the margin l

is at least 1.5.

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(5) The margin between the' leakage-size crack and the critical-size crack j

was calculated by a-limit load analysis. Again, the results demon-e

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the bounds of Section 4.2.3 of Enclosure I to Reference 4.

(6) As an in'tegral part of its review, the staff's evaluation of the material. properties data of Reference 9 is enclosed as Appendix 1 to this exemption.

In Reference 9, data for ten (10) plants, including the Vogtle units, are presented, and lower bound or " worst case"'

materials properties were identified and used in the analysis per-fonned in the Reference 3 report by Westinghouse. The applied J for Vogtle in Referenc,e 3 was substantially less than 3000 in-lb/in,

e Hence, the staff's upper bound of 3000 in-lb/in2 on the applied J (refer to Appendix 1, page 6) was not exceeded.

In v.iew of the analytical results presented in the Westinghouse Report for Vogtle (Reference 3) and the staff's evaluation findings related above, the staff concludes that the probability or likelihood of large pipe breaks occur-

' ring at the eight (8) locations in each primary coolant system loop of Vogtle, Units I and 2 is sufficiently low such that such pipe breaks and their associated dynamic loads as indicated in the applicants' October 25 letter need l

not be considered as a design basis for requiring pipe whip restraints and iet impingement shields. Eliminating the need to consider these dynamic loads for this particular application does not in any way affect the design bases for the containment, the emergency core cooling system, or the environmental qualifi-cation for Vogtle.

The staff also reviewed the value-inpact analysis provided by the appli-cant in their April 2,1984, submittal for not providing protective structures I

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12 against postulated reactor coolant system loop pipe breaks to assure as low as reasonably achievable (ALARA) exposu.re to plant personnel. Consideration was giventodesign'fe'aturesforreducingdosestopersonnelwho.bustoperate,'

service and maintain the Vogtle instrumentation, controls, equipment, etc. The Vogtle value-impact analysis shows that the elimination of protective devices for RCS pipe breaks will save an occupational dose for plant personnel of approximately 700 person-rem for both units over their operating lifetime. The staff review of the analysis shows it to be a reasonable estimate of dose savings. Therefore, with re,spect to occupational exposure, the staff finds that there is a radiological benefit to be gained by eliminating the need for the protective structures.

VI.

In view of the staff's evaluation findings, conclusions, and recommenda-tions above, the Comission has determined that, pursuant to 10 CFR 50.1P.(a),

this exemption is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest. The Commiss' ion hereby approves the limited exemption from GDC 4 of Appendix A to 10 CFR Part 50, to permit the applicants not to further install pipe whip restraints and.fet impingement shields and not to consider dynamic effects l

and loading conditions a's* detailed in Part II of this exemp" tion associated with postulated pipe breaks of the eight (8) locations per loop in ihc Vogtle, l

Units 1 and 2 primary coolant system, as specified in Enclosure D of the l

l applicants' letter dated October.25,1983.

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. Pursuant to 10 CFR 51.32, the Commission has determined that the issuance of_the exemption will have no significant impact on the environment (50 FR 4605

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The exemption will become effective upon date of issuance.

FOR THE NilCLEAR REGULATORY COMMISSION

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isen u, virectbr Division of Licensing Office of Nuclear Reactor Regulation n ted at Bethesda, Maryland a

this 5th day of February 1985 9

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REFERENCES

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(1) Vestinghouse Report MT-SME-3082, " Technical Aases' for Elfminating large

. Primary Loop Pipe Ruptures as the Structural Design Basis-for.Vogtle,

. Units l'and f,"' September 1983, Westinghouse Class 2 proprietary.

(2) Letter to D. O. Foster of Georgia Power Company, " Request for Additional

-Information Concerning Leak-Before-Break Analysis for Vogtle Electric Generating Plants, (Units 1 and 2)," dated March 19, 1984.

(3) ' Westinghouse Report WCAP-10551, " Technical Bases for Eliminating large Primary Loop Pipe Rupture as the Structural Design Basis for Vogtle Units 1 and ?," May 1984, Westinghouse Class 2 proprietary.

(4) NRC Generic letter 84-04, " Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Breaks in PWR Primary j

Main loops,"-February 1,, 1984.

(5) Mechanistic. Fracture Evaluation of R'eactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack, WC4d-9558, Rev. 2 May 1981, Westinghouse Class 2 proprietary.

(6) Tensjle and Trughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation, WCAP-9787, May 1981, Westinghouse Class 2 proprietary.

(7) Westinghouse Response to Questions and Comments Raised by Members of ACRS '

Subcommittee on Metal Components During the Westinghouse Presentation on September 25, 1981, Letter Report NS-EPR-2519, E. P. Rahe to Darrell G.

Eisenhut, November 10, 1981, Westinghouse Class 2 proprietarv.

(8) Lawrence Livermore National Laboratory Report, UCRL-86249, " Failure Prob-ability of PWR Reactor Coolant Loop Piping," by T. Lo, H. H. Woo, G. S.

Holman and C. K. Chou, February 1984 (Preprint of a paper intended for publication).

(9) Westinghouse Report WCAP-10456, "The Effects'of Thermal Aging on the Structural Integrity of Cast Stainless Steel Piping for Westinghouse Nuclear Steam Supply Systems," November 1983, Westinghouse. Class 2 proprietary.

(10) Georgia Power Company letter, " Alternative Pipe Break Design Consider-ations" (D. O. Foster to H. R. Denton) dated April 2,1984 I

Notes: See next page i

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REFERENCES

- NOTE: Non-proprietary versions of References 1, 3, 5, 6, 7 and 9 are available in the NRC Public-Document Room as follows:

r (1) MT-SME'-3082, non-proprietary (3

WCAP 10552 (5

WCAP 9570 (6

WCAP 9788 (7) Non-proprietary version attached to the Letter Report (9) WCAP 10457 O

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