ML20137T347
| ML20137T347 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 11/22/1985 |
| From: | Thompson H Office of Nuclear Reactor Regulation |
| To: | PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| Shared Package | |
| ML20137T351 | List: |
| References | |
| NUDOCS 8512060441 | |
| Download: ML20137T347 (15) | |
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'/590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION L
In the Matter of
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PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE, )
Docket Nos. 50-443 and ET AL.*
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50-444 (Seabrook Station, Units 1 and 2)
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EXEMPTION I.
i On March 30, 1973, Public Service Company of New Hampshire, et al.
r (applicants) tendered an application for licenses to construct Seabrook l
Station, Units 1and2(Se$brookorthefacility)withtheAtomicEnergy Comission (currently the Nuclear Regulatory Comission or the Comission).
i 1
Following a public hearing before the Atomic Safety and Licensino Board, the Comission issued Construction Permit Nos. CPPR-135 and CPPR-136 permitting the construction of the Units 1 and 2, respectively, on July 7,1976.
Each I
unit of the facility is a pressurized water reactor, containing a Westinghouse Electric Company nuclear steam supply system, located at the applicants' site in Seabrook, New Hampshire.
On June 29, 1981, the applicants tendered an application for Operating f
Licenses for the facilities, currently in the licensing review process.
8512060441 851122 I
DR ADOCK 05000443 l
- The current construction permit holders for Seabrook Station are: Bangor Hydor-Electric Company, Canal Electric Company, Central Maine Power Company, Central Vermont Public Service Corporation, Connecticut Light & Power Company, Fitchburg Gas & Eiectric Light Company, Hudson Light & Power Department, Maine Public Service Company, Massachusetts Municipal Wholesale Electric Company, t
Montaup Electric Company, New England Power Company, New Hampshire Electric Cooperative, Inc., Public Service Company of f!ew Hampshire, Taunton Municipal Lighting Plant, United Illuminating Company, Vermont Electric Generation and Transmission Cooperative, Inc., and Washington Electric Cooperative, Inc.
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II.
The Construction Permits issued for constructing Units 1 and 2 of the Seabrook Station provide, in pertinent part, that the facility is subject to all rules, regulations and orders of the Commission. This includes General Design Criterion (GDC) 4 of Appendix A to 10 CFR 50. GDC 4 requires that structures, systems and components importa.it to safety shall be designed to a;;commodate the effects of, and to be compatible with, the environmental cenditions associated with the normal operation, maintenance, testing and postulated accidents, including loss-of-conlant accidents. There structures, systems and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, discharging fluids that may result from equipent failures, and from events and conditions outside the nuclear power unit.
In a submittal dated August 9, 1984, the applicants enclosed Westinghouse Report WCAP-10567 (Reference 1) containing the technical basis for their request to eliminate the postulated breaks in the reactor coolant loop (RCL) piping of Seabrook Station, Units 1 and 2.
The applicants further stated that eliminating these postulated breaks would result in eliminating their associated dynamic effects which are specifically defined as the effects of missiles, pipe whipping, and fluid jets.
In the February 1, 1985 submittal, the applicants stated that granting of their request (1) eliminates the need to postulate longitudinal and circumferential pipe breaks in the RCL primary piping (hot leg, cold leg, and crossover leg piping); (2) eliminates the need to install associated pipe whip restraints in the RCL primary piping; and (3) eliminates the requirement to
. analyze and design for the dynamic effects of these breaks including jet impingement, reactor cavity pressurization and load combination assumptions.
The exemption request will not apply to the containment design bases, the emergency core cooling system, or environmental qualification, engineered safety features systems response, or the design of the RCS heavy component supports.
The applicants' submittal of February 1, 1985, contains the results of an analysis of the occupational radiation dose reduction which provides the value-impact analysis for Seabrook Station, Units 1 and 2.
The technical information contained in Reference (1) together with the value-impact analysis, provided a comprehensive justification for requesting limited exemptions from the requirements of GDC 4.
From the deterministic fracture mechanics analysis contained in the technical information furnished, the applicants concluded that postulated breaks up to and including the double-ended guillotine breaks (DEGB) of the primary loop coolant piping in Seabrook 1 and 2 need not be considered as a design basis for installing protective structures, such as pipe whip restraints and jet impingement shields, to guard against the dynamic effects associated with such postuidted breaks. However, the applicants will continue to postulate the equivalent area of a DEGB as the-design basis for.the containment, the ECCS, the engineered safety systems response, environmental qualification and the design of the RCS heavy component supports.
III.
The Commission's regulations require that applicants provide
. protective measures against the dynamic effects of postulated pipe breaks in high energy fluid system piping. Protective measures include physical isolation from 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 asymetric 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, "Asymetric Blowdown Loads on PWR Primary Systems."
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 Owners' Group (one of these facilities, Fort Calhoun, has a Combustion Engineering nuclear steam supply system).
This NRC staff position was stated in Generic Letter 84-04, published on February 1, 1984 (Reference 2). The generic letter states that the affected licensees must justify an exemption to GDC 4 on a plant-specific basis. 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 deterministic analyse ~ that the detection of small flaws by either f
inservice inspection or leakage monitoring systems is assured long before the flaws can graw 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-beford-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 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 3, 4, and 5) 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 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 included as a part of Reference 2.
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. Probabilistic fracture mechanics studies conducted by the Lawrence Livermore National Laboratories (LLNL) on both Westinghouse and Combustion Engineering nuclear steem supply system main loop piping (Reference 6) confirm that 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 6 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 year.
Similarly, the best-estimate
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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. The 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 jet impingement shields is not necessarily the most cost effective way to achieve the desired level of safety, as indicated in Enclosure 2, Regulatory Analysis, to Reference 2.
Even for new plants, these devices l
tend to restrict access for future inservice inspection of piping; or l
if they are removed and reinstalled for inspection, there is a potential risk of damaging the piping and other safety-related components in this l
process.
If installed in operating plants, high occupational radiation l
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.
4 IV.
The primary coolant system of Seabrook, Units 1 and 2 described in Reference 1, has four main loops each comprising a 33.9 inch diameter (outside) hot leg, a 37.5 inch diameter crossover leg and 32.4 inch diameter cold leg piping. The materials in the primary loop piping are wrought stain-less steel pipe with cast stainless steel fittings and associated welds.
In its review of Reference 1, the staff evaluated the Westinghouse analyses with i
regard to:
the location of maximum stresses in the piping, associated with combined loads from normal operation and the SSE; potential cracking mechanisms; size of through-wall cracks that wou a 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 wrought and 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 2, Section 4.1, "NRC Evaluation Criteria", and are as follows:
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(1) The loading conditions should include static forces and moments (pressure, deadweight and thermal 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 and the lowest material toughness are coincident for base materials, 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; systen 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 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 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;
_g-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 (largerthandesignloads)areapplied. 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 margin should be determined by comparing the leakage-size crack to critical-size cracks. Under normal plus SSE loads, it should be demonstrated that there is adequate margin 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 the purpose; however, an elastic-plastic fracturemechanics(tearinginstability)analysisispreferable.
(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-tem effects such as thermal aging and other limitations to valid data 'e.g. J maximum, maximuin crack growth).
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Based on its evaluation of the analysis contained in Westinghouse Report WCAP-10567 (Reference 1), 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 Seabrook. Station, Units 1 and 2.
This finding is predicated on the fact that each of the parameters evaluated for Seabrook is enveloped by the generic analysis performed by Westinghouse in Reference 3, and accepted by the staff in Enclosure 1 to Reference 2.
Specifically:
(1) The loads associated with the highest stressed location in the main loop primary system piping are 2332 kips (axial),
37,045 in-kips (bending moment) and result in maximum stresses of about 97% of the bounding stress used by Westinghouse in Reference 3.
Further, these loads are approximately 88% of those established by the staff as limits.
(2)
For Westinghouse plants, there is no history of cracking failure in reactor primary coolant system loop piping. The Westinghouse reactor coolant system primary loop has an operating history which demonstrates its inherent stability. This includes a low susceptibility to cracking failure from the effects ofco-rosion(e.g.intergranularstresscorrosioncracking), water
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hammer, or fatigue (low and high cycle). This operating history totals over 400 reactor-years, including five (5) plants each having 15 years of operation and 15 other plants with over 10 years of operation.
(3) The leak rate calculations performed for the Seabrook plants using an initial through-wall crack of 7.5 inches are identical to those of Enclosure 1 to Reference 2.
The Seabrook plants have an RCS pressure boundary 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 i
through the postulated flaw results in a factor of at least 10 relative to the sensitivity of the Seabrook plants' leak detection system.
(4) The margin in terms of load based on fracture mechanics i
analyses for the leakage-size crack under normal plus SSE loads is within the bounds calculated by the staff in Section 4.2.3 of Enclosure 1 to Reference 2.
Based on a limit-load analysis, the load margin is about 2.0 and based on the J-limit, the margin is at least 1.1.
(5) The margin between the leakage-size crack and the critical-size crack was calculated by a limit-load analysis. Again, the results demonstrated that a margin of at least 3 on crack size exists and is within the bounds of Section 4.2.3 of to Reference 2.
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. (6) In addition to the wrought stainless steel pipes, the Seabrook units have cast stainless steel fittings and associated welds in the primary coolant system. As an integral part of its review, the staff's evaluation of the properties data of Reference 7 is enclosed as Appendix I to this exemption.
In Reference 7, data for ten (10) plants are presented and lower bound or " worst case" materials properties were identified and used in the analysis performed in the Reference 1 report by Westinghouse. The applied J for Seabrook in Reference 1 for cast stainless steel fittings was less than 3000 in-lb/in2 Hence, the staff's upper bound 3000 l
in-lb/in2 on the applied J (refer to Appendix I, page 6) was not
- exceeded, l
In view of the analytical results presented in the Westinghouse Report l
for Seabrook (Reference 1) and the staff's evaluation findings related above, the staff concludes that the probability of large pipe breaks occurring in the primary coolant system loops of Seabrook Station, Units 1 and 2, is suffi-ciently low such that pipe breaks and their associated dynamic loading effects as indicated in the applicants submittals need not be considered as a design basis for requiring pipe whip restraints and jet impingement shields.
These dynamic loading effects include pipe whip, jet impingement, missiles,
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reactor cavity overpressurization, and load combination assumptions.
Eliminating the need to consider these dynamic l'oading effects for this particular application will not in any way affect the design bases for the containment, the emergency core cooling system, the design of RCS heavy component supports, the engineered safety features systems response, or the environmental qualification of equipment for Seabrook.
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l 1 However, in order to provide the Commission with an opportunity to consider the long term aspects of the NRC staff's recent acceptance of the " leak-before-break" approach, these limited exemptions are restricted to a period extending until the completion of the second refueling outage of Seabrook Station, Unit 1, pending the outcome of Commission rulemaking on this issue.
The staff also reviewed the value-impact analysis provided by the applicants in their submittal for not providing protective structures against postulated reactor coolant system loop pipe breaks to assure as low as reason-ably achievable (ALARA) exposure to plant personnel. Consideration was given to design features for reducing doses to personnel who must operate, service and maintain the Seabrook instrumentation, controls, equipment, etc. The Seabrook value-impact analysis shows that the elimination of protective devices for RCS pipe breaks will save an occupational dose for plant personnel of approximately 1400 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 respect to occupa-tional exposure and ALARA considerations, 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 recommendations above, the Commission has determined that, pursuant to 10 CFR 50.12(a), the following exemption is authorized by law and L
- will not endanger life or property or the common defense and security, and are otherwise in the public interest. The Commission hereby approves the requested schedular limited exemption from GDC 4 of Appendix A to 10 CFR Part 50, to permit the applicants to eliminate the protective devices and the dynamic loading effects, as described in Part II of this exemption, associated with the postulated primary loop pipe breaks for Seabrook Station, Units 1 and 2.
The exemption will expire upon completion of the GDC 4 rulemaking changes but no later than the second refueling outage of Seabrook Station, Unit 1.
Pursuant to 10 CFR 51.32, the Connission has determined that the issuance of the exemption will have no significant impact on the environment (50 FR 47468).
The exemption will become effective upon date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION i,
u h L. Thompso Jr,
frector Di sion of Licensi Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland this 22 day of November,1985
REFERENCES (1) Westinghouse report WCAP-10567, " Technical Bases for Eliminating Large Primary Loop Pipe Ruptures as the Structural Design Basis for Seabrook, Units 1 and 2, June 1984. Westinghouse Class 2 proprietary.
(2) 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.
(3) Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack, WCAP-9558, Rev. 2, May 1981, Westinghouse Class 2 proprietary.
(4) Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation, WCAP-9787, May 1981, Westinghouse Class 2 proprietary.
(5) Westinghouse Response to Questions and Coments Raised by Members of ACRS Subcommittee on Metal Components During the Westinghouse Presentation of September 25, 1981, Letter Report NS-EPR-2519 E.P. Rahe to Darrell G. Eisenhut, November 10, 1981, Westinghouse Class 2 proprietary.
(6) Lawrence Livermore National Laboratory Report, UCRL-86249, " Failure Probability of PWR Reactor Coolant Loop Piping," by T. Lo, H. H. Woo, G. S. Holman and C. K. Chou, February 1984 (Reprint of a paper intended for publication).
(7) 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.
NOTE: Non-proprietary versions of References 1, 3, 4, 5 and 7 are available in the NRC Public Document Room as follows:
WCAP 10566 WCAP 9570 WCAP 9788 Non-proprietary version attached to the Letter Report
- 7) WCAP 10457 F
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