ML20096B450
| ML20096B450 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/28/1984 |
| From: | Miraglia F Office of Nuclear Reactor Regulation |
| To: | TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| Shared Package | |
| ML20096B451 | List: |
| References | |
| CON-#185-543 OL, NUDOCS 8409040147 | |
| Download: ML20096B450 (16) | |
Text
7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION
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In the Matter of TEXAS UTILITIES GENERATING COMPANY Docket Nos. 50-445 and 50-446 (Comanche Peak Steam Electric Station, Units 1 and 2)
EXEMPTION l
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- 6 On July 20, 1973, the Texas Utilities Generating Company (the applicant)
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tendered an application for licenses to construct Comanche Peak Steam Electric 5
Station, Units 1 and 2 (Comanche Peak or the faciHty) with the Atomic Energy Comission (currently the Nuclear Regulatory Comission or the Comission).
Following a public hearing before the Atomic Safety and Licensing Board, the Comission issued Construction Permit Nos. CPPR-126 and CPPR-127 permitting the construction of Units 1 and 2, respectively, on December 19, 1974.
Each Unit.of the facility is a pressurized water reactor, combining a Westinghouse Electric Company nuclear steam supply system, located at the applicant's site in Somervell/ Hood Counties, Texas, approximately 40 miles southwest of Fort Worth, Texas.
On February 27, 1978, the applicant tendered an application for Operating Licenses for each Unit of the facility, currently in the licensing review process',
with Unit I licensing to occur in the nea.r I
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7590-01 r
II.
The Construction Permits issued for constructing the facility provide, in pertinent part, that the facility Units are subject to all rules, regulations 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 components important to safety shal be designed to accomodate the effects of and to be compatible with the environment ~al conditions associated with the normal operation, maintenance, testing and postulated accidents, including l
loss-of-coolant accidents. These structures, systems and components shall be t
appropriately protected against dynamic effects, including the effects of
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missiles, pipe whipping, discharging fluids that may result from equipment failures, and from events and conditions outside the nuclear power unit.
By a submittal dated October 31, 1983, the applicant requested an exemp-tion from a portion of the requirements of GDC 4 to: (1) eliminate the need to postulate circumferential and longitudinal p'ipe breaks in the Reactor Coolant System (RCS) 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) to elimina'te 'the need to consider dynamic effects and loading condi-tions associated with previo~usly postulated pipe breaks in the RCS primary 1 cop, including jet impingement loads, cavity pressure loads, blowdown loads in the RCS and attached piping, and subcompartment pressure loads.
In support of this exemption' request, the applicant's submittal enclosed Westinghouse Re-3 port MT-SME-3135 (Reference 1) containing tiie technical basis for their request.
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l 7590-01 Based on its review of_the applicant's submittal, the NRC staff requested additional information and provided comments on the reports (References 1 and 9)
-which were transmitted to the applicant in the form.of questions by NRC letter
. 1,
'l dated March 2, 1984, (Reference 2).
By a submittal dated April 23, 1984, the applicant responded to the staff':
questions (Reference 2) and provided a revision to the Reference 1 report iden-tified as Westinghouse Report WCAP-10527 (Reference 3).
In a' separate submittal, also dated April 23, 1984, the applicant provided a value-impact analysis which, together with the technical *information contained in the Reference 3 report.
- provided a comprehensive justification for requesting a partial exemption from the requirements of GDC 4.
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From the deterministic fracture mechanics analysis contained in the technical information furnished, the applicant stated that the postulated double-ended guillotine breaks (DEGB) of the primary loop coolant piping will not occur in 9
Comanche Peak Units 1 and 2 and, therefore, need not be considered as a design basis for installing protective struct0res, such as pipe whip restraints and jet impingement shields to guard against the dynamic effects associated with such postulated breaks.
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7590-01 By letter dated June 7,1984 (Reference 10), the applicant clarified the scope of its request for exemption from GDC 4 requirements. Since the Westinghouse Report WCAP-10527 provided analyses encompassing other structures in both
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Comanche Peak Units 1 and 2, and seemed to be in conflict with the scope of the exemption requested in an earlier letter dated February 17,1984(Refer-ence 11), the applicant stated in the Reference 10 letter that, although the analyses contained in the Report WCAP-10527 encompassed relief from the need to install pipe break protective devices 'in both Units 1 and 2, the exemption being requested pertained solely to the installation of jet impingement shields associated with such breaks in eight (8) ' locations per loop in Comanche Peak Unit 1, as~speci.fied in Section 4.0 of the value-impact analysis submitted by the applicant's letter dated April 23, 1984.
III.
The Commission's regulations require that applic' ants provide protective measures against the dynamic effects of postulated pipe breaks in high energy fluid system piping. Protective meas 0~res include physical isolation from L
postulated pipe rupture locations if feasible or the installation of pipe whip restraints, jet impingement shields or compartments.-
In 1975, concerns arose as to the asymmetric l'oads on pressurized water reactor (PWR) Nessels and 1
I their internals which could result from these large postulated breaks at l
j discrete locations in the main primary coolant loop piping. This led to the 7
j establishment of Unresolved Safety Issue (USI) A-2, " Asymmetric Blowdown Loads
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on PWR 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 Re-view Generic Requirements (CRGR), concluded that for certain facilities an
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exemption from the regulations _ would be acceptable as an alternative for reso-i lution of USI A-2 for sixteen facilities owned by eleven licensees in the West-inghouse Owner's Group (one of these facilities, Fort Calhoun has a Combustion Engineering nuclear steain supply system). This NRC staff position was stated inGenericLetter84-04,publishedonFebruaryi,1984(Reference 4).
The generic letter states that the affected licensees must justify an exemption to GDC 4 on a plant-specific oasis. Other PWR applicants or licensees may request
' similar exemptions from 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 techniques deal with relatively small flaws in piping components (either postulated or real) and examine their behavior under various pipe loads. The objective is to demonstrate by detenMnistic analyses that the detection of small flaws by either inservice inspection or leakage monitoring systems is assured long before the flaws can grow to critical dr 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" (LBB).
There'is no implication that piping failures cannot occur, but rather that improved knowledge of the failure modes of piping systems and the application
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of appropriate remedial measures, if indicated, can reduce the probability of catastrophic failure to insignificant values.
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I 7590-01 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 discrets break locations, the technology 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 cohsidered 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
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4.
Probabilistic, fracture mechanics studies conducted by the Lawrence Liver-l more National Laboratories (LLNL) ori both Westinghouse and Combustion Engineer-ing nuclear steam supply system main loop piping (Reference 8) confirm that both the probability of leakage (e.g., undetected flaw growth through the pipe wall by fatigue) and the probability of a DE'B are very low. The results given G
in Reference 8 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 DEGB probabilities range from 1 x 10-12 to 7'x 10-12 per plant' ye'ar.
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 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.
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7590-01 During the past few years it has also become apparent that the requirement for installation of large, massive pipe whip restraints and jet impingement shields is not necessarily the most cost effective way to achieve the desired "5
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 safety-related components in this process.
If installed in operating plants, high occupational radistion 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.
IV.
The primary coolant system of Comanche Peak Units 1 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 normal operation and the SSE; potential cracking mechanisms; size of through-wall cracks that would leak a detectable amount under normal loads and pressure; l
stability of a " leakage-size crack".under normal plus SSE loads and the expected margin in terms of loada
7590-01
-8 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 delin-eated in its Topical Report Evaluation, Enclosure 1 to Reference 4, Section 4.1, "NRC Evaluation Criteria", and are as follows:
(1) The loading conditions should includ.e the 4tatic forces and moments (pres-sure, deadweight and thennal expansion) due to normal operation, and the forces and moments associated with the safe shutdown earthquake (SSE).
These forces and moments should be located where the highest stresses,
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coincident with the poorest material properties, are induced for base mateFials, weldments 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 modification; water chemistry parameters, limits and controls; resistance of material to various forms of stress corrosion, 'and performance under cyclic loadings, n
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musm (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 adequate margin using the
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minimum installed leak detection capability when the pipe is subjected to normal operational loads.
(4) 'It should be demonstrated that the postulated leakage crack is stable 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-siz.e crack will not experience unstable crack growth even if larger loads (larger than design loads) are. applied. This analysis should demon-strate that crack growth is stable and the final crack size is limited, such that a double-ended pipe break will not occur.
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(5) The crack size should be determined by comparing leakage-size crack to critical-size cracks.
Under normal plus SSE loads, it should be demon-
.strated that there is adequate mirgin between the leakage-size crack 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 fra'cture nechanics (tearing instability) analysis is preferrable.
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' 7590-01 (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
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effects such as thermal aging and other limitations to valid data (e.g.
J maximum, maximum crack growth).
V- :.
Based on its evaluation of the analysis contained in Westinghouse Report WCAP-10527 (Reference 3), the staff finds'that the applicant has presented an acceptable ~ technical justification, addressing-the above criteria, for not
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installing _ protective devices to deal with the dynamic effects of large pipe ruptures in the main loop primary coolant system hiping of Comanche Peak, Units 1 and 2.
This finding is predicated on the fact that each of the para-meters evaluated for Comanche Peak is. enveloped by the generic analysis per-formed by Westinghouse in Reference (5), and accepted by the staff in Enclo-sure 1 to Reference 4.
Specifically:
(1) The loads associated with the highest stressed location in the main loop primary system piping are considerably lower than the bounding loads used by Westinghous't in Reference 5, or those established by the staff as limits (e.g. a moment of 42,000 in-kips in Enclosure 1 to Reference 4).
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' 1 (2) For Westinghouse plants, there is no history of cracking failure in reac-l tor primary coolant system loop piping. The Westinghouse reactor coolant system primary lcep has an operating history which demonstrates its inherent '
i stability. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g. intergrannular stress corrosion cracking), water hammer, or fatigue (low and high cycle). This operating history totals over
' 400 reactor-years, including five plants each having 15 years of operation and 15 other plants with over 10 years of operation.
(3) The results of the leak rate calcula' ions performed for Comanche Peak.
t using an initial through-wall crack are identical to those of Enclosure 1 to Reference'(4). The Comanche Peak plant has an RCS pressure boundary leakdetectionsystemwhichisconsistentwibtheguidelinesofRegula-tory Guide 1.45, and it can detect leakage of one (1) gpm in one hour.
The calculated leak rate through-the postulated flaw is large relative to the sensitivity of the Comanche Peak plant leak detection system.
(4).The expected margin in terms ofToad for the leakage-size crack under normal plus SSE loads is within the bounds calculated by the staff in Section4.2.3ofEnclosure(1)toReference4.-Inaddition,thestaff found a significant margin in terms of. loads larger than normal plus SSE loads.
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I 12-(5) The margin between the leakage-size crack and the critical-size crack was calculated. Again, the results demonstrated that a significant margin exists and is within the bounds of Section 4.2,3 ~of Enclosure 1 to Refer-ence 4.
(6) As an integral 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) p1' ants, including the Comanche Peak Units, are. presented, and lower bound or " worst case" materials properties were identified and used in the analysis performed in the Reference 3 report 2
by Westinghouse. The staff's upper bound of 3000 in-lb/in on the applied J (. refer to Appendix 1, page 6) was not exceeded; the applied J for Comanche 2
Peak in Reference 3 was substantially less than 3000 in-lb/in,
In view of the analytical resultspresented in the Westinghouse Report for Comanche Peak (Reference 3) and the staff's evaluation findings related above, a
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the staff concludes that the probability or likelihood of large pipe breaks
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occuring in the primary coolant systeh loop of Comanche Peak Units 1 and 2 is sufficiently low such that such pipe breaks need not be considered as a design l
basis for requiring protective devices. However, the pipe whip restraints have already been inst'alled in Unit 1, and the applicant has 1imited the scope of its exemption request to the installation of jet impingement shields in Unit 1
1 only. The requested exemption from GDC 4 is limited to exemption from the need to instal 1 jet impingement' shields at specified locations in Unit 1.
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7590-01' The' staff also reviewed the value-impact analysis provided by the appli-l'
- cant'for:not providing_ protective structures aga' inst postulated' reactor coolant
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system loop pipe breaks to assure as low as. reasonably achievable (ALARA) expo-sure to plant personnel. Consideration was given to design features for reduc-ing doses to personnel who must operate, service and maintain the Comanche Peak instrumentation,: controls, equipment, etc. Nonnally, facilities and equipment are designed to save person-rems; however, the Comanche Peak value-impact anal-ysis shows that the addition of protective devices for_RCS pipe breaks will cost about 2 person-rems annually due to the -slowing-down of normally anticipated work, and increasing the scope of routine maintenance in radiation areas that would be involve.d. The analysis provides a reasonable estimate for this addi-tional radiologica'l-cost.
In view of the very low probability of pipe b'reaks at the specified locations covered by this exemptionh the reduction of occupational exposure resulting from this exemption outweighs the potential accident exposure i
reduction that might result from installation of the jet impingement barriers.
VI.
In view of the staff's evaluation findings, conclusions, and recommenda-tions above, the Commission has detennined that, pursuant to 10 CFR 50.12(a),
this Exemption is authorized by law and will not endanger life' or prop-erty or the common defense and security, and is otherwise in the public inter-est. The Commission hereby approves the requested limited exemption from GDC 4 of Appendix'A to 10 CFR Part 50, to permit the licensee not to install i
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l 7590-01 jet impingement shields associated with postulated pipe breaks of the eight (8) locations per loop in the Comanche Peak Unit 1 primary coolant system, as specified in Section 4.0 of the value-impact: analysis submit by the applicant's letter dated April 23, 1984. This Exemption does not pertain to the installation of pipe whip restraints, already installed in Unit 1, or to the installation of pipe whip restraints and jet impingement shields in Comanche Peak Unit 2.
The portion of the request concerning Unit 2 will be dealt with in a separate NRC action.
The Commission has determined that the issuance of the exemption will have no significant e.nvironmental impact on the environment (49 FR 33945
).
FOR THE NUCLEAR REGULATORY COMMISSION 50A'b b
FrankMiraglial.DeMyDirector Division of Licensing Office of Nuclear Reactor Regulation Dated atptthesda, Maryland this Jayofg 1984 9
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7590-01
. t REFERENCES (1) Westinghouse Report MT-SME-3135, " Technical Bases for Eliminating Large Primary Loop Pipe Ruptures as the Structural Design Basis for Comanche Peak Units 1 and 2," October 1983, Westinghouse Class 2 proprietary.
(2) Letter to R. J. Gary of Texas Utilities Generating Company " Request for Additional Information Concerning Leak-Before-Break Analysis for Comanche Peak Steam Electric Station (Units 1.and 2)," dated March 2, 1984.
(3). Westinghouse Report WCAP-10527, " Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Comanche Peak Units 1 and 2," April 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 Main Loops," February 1, 1984.
(5) Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack, WCAP-9558, Rev. 2, May 1981, Westinghouse Class 2 proprietary.
.. (6) Tensile an( Toughness 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 Coninents 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, West 4nghouse Class 2 proprietary.
(8) Lawrence Livermore National Laboratory Rpeort, 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) Texas Utilities Generating Company letter TXX-4197, "Requ'est for Partial Exemption" (H. C. Schmidt to B. J. Youngblood) dated June 7, 1984.
(11) Texas Utilities Generating Company letter TXX-4118, " Request for Partial Exemption," (R. J. Gary to 8. J. Youngblood) dated February 17, 1984.
j Notes: See next page
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7596-01
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REFERENCES NOTE:
Non
-in '.-proprietary versions of References 1, 3, 5, 6, 7 and 9 are available b
the NRC Public Document Room as follows:
~ MT-SME-3136 WCAP 10528
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.WCAP.9570'
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(6 WCAP 9788 (7) Non-proprietary version attached to the Letter Report
-(9) WCAP 10457 4
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