ML20126E523
| ML20126E523 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/05/1985 |
| From: | Thompson H Office of Nuclear Reactor Regulation |
| To: | NORTHEAST NUCLEAR ENERGY CO. |
| Shared Package | |
| ML18023A046 | List: |
| References | |
| NUDOCS 8506170112 | |
| Download: ML20126E523 (15) | |
Text
l UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of NORTHEAST NUCLEAR ENERGY COMPANY Docket No. 50-423 (Millstone Nuclear Power Station, Unit 3)
EXEMPTION
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I.
On February 10, 1973, the application for a license to construct Millstone Nuclear Power Station, Unit 3 (Millstone Unit 3 or the facility) tendered by Millstone Point Company and joint applicants was docketed by the Atomic Energy Commission (currently the Nuclear Regulatory Comission or the Comission). Following a public hearing before the Atomic Safety.
and Licensing Board, the Commission issued Construction Permit No. CPPR-13 permitting the construction of Unit 3 on August 9, 1974. The facility is a pressurized water reactor, containing a Westinghouse Electric Company nuclear steam supply system, located at the licensee's site in the town of Waterford, New London County, Connecticut on the north shore of Long Island Sound.
On October 29, 1982, the licensee tendered an application for an Oper-ating License for the facility, currenity in the licensing review process.
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II.
The Construction Permit issued for constructing the facility provides, in pertinent part, that the facility is subject to all rules, regulations and Orders of the Comission. ThisincludesGeneralDesignCriterion(GbC) 4 of Appendix A to 10 CFR 50. GDC 4 requires that structures, systems and components important to safety shall be designed to accomodate the effects of and to be compatible with the environmental conditions associated with the nonnal 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 whipping, discharging fluids that may result from equipment failures, and from events and conditions outside the nuclear power unit.
By a submittal dated September 12, 1984, the applicant enclosed Westing-house Report WCAP-10586 (Wesinghouse Non-Proprietary) and WCAP-10587 (Westing-house Proprietary) (Reference 1) containing the technical basis for their request to: (1) eliminate the need to design for pipe whip, jet impingement l
l and the asymmetric effects of cavity pressurization due to primary loop pipe breaks; (2) eliminate the need for pipe whip restraints and jet impingement shields on primary loop piping; (3) eliminate primary loop LOCA load evaluation on primary loop piping, branch line piping and branch line supports (branch line LOCA loads would be retained in the design basis); (4) eliminate the need to include primary loop LOCA loads in the design of the reactor coolant pump P1 snubber in loops I and 2 (2 out of 44 snubbers).
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3-By a subsequent submittal dated October 18, 1984, the applicant submitted a modification to request for exemption from General Design Criterion 4 to dis-regard all reference to considerations for including the P-1 snubbers in the exemption request.
The applicant also stated in their submittal that the exemption request does not affect the emergency core cooling system design bases, containment and subcompartment design bases, equipment qualification bases and engineered safety features systems response. The applicant also stated that the design of the reactor coolant system heavy component supports will continue to assume a double ended primary loop pipe break with a break area equal to that which would occur if pipe whip restraints were installed.
The applicant provided a value-impact analysis in its September 12th sub-mittal which, together with the technical information contained in the Reference 1 report, provided a comprehensive justification for requesting a partial exemption from the requirements of GDC 4.
Finally, by letter from J. F. Opeka to B. J. Youngblood dated May 7,1985, the applicant requested that a partial exemption to GDC 4 be granted for the first two cycles of operation.
From the deterministic fracture mechanics analysis contained in the tech-nical information furnished, the applicant concluded that the dynamic loading effects associated with postulated double-ended guillotine breaks (DEGB) and longitudinal breaks in the primary loop coolant piping in Millstone Unit 3, need not be considered as a design basis. These dynamic loading effects include pipe whip, jet impingement, asymmetric pressurization transients and break associated transients.
. Therefore, structures such as pipe whip restraints and jet impingement shields, to guard against the dynamic effects associated with such postulated breaks may be eliminated.
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 barriers. 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 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 16 facil-ities owned by 11 licensees in the Westinghouse Owner's 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, pub-lished on February 1, 1984 (Reference 2). The generic letter states that
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-5 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 detenninistic 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 or unstable sizes which could lead to large break areas such as the DEGB or its equivalent. The con-cept 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 of appropriate remedial measures, if indicated, can reduce the probability of catastrophic failure to very small values.
Advanced fracture mechanics technology was applied in topical reports (Refer-ences 3, 4, and 5) submitted to the staff by Westinghouse on behalf of the licensees belonging to the USI A-2 Owners Groep. Although the topical reports were intended to resolve the issue of asymetric 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 instal-lation of pipe whip restraints or jet impingement shields. The staff's Topical Report Evaluation is attached as Enclosure 1 to Reference 2.
Probabilistic fracture mechanics studies conducted by the Lawrence Liver-more National Laboratories (LLNL) on both Westinghouse and Combustion Engineering nuclear steam 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 leak probabilities for Combustion Engineering nuclear steam supply system main loop piping range from I 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.
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 Refer-ence 2.
Even for new plants, these devices tend to restrict access for future inservice inspection of piping; or if they are removed and reinstalled for
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inspection, there is a potentia 1 risk of damaging the piping and other safety-related components in this process. If installed in operating plants, high occupational radiation exposure (ORE) would be incurred while public risk reduction would be very low. Removal and reinstallation for inservice inspec-tion also entail significant ORE over the life of a plant.
IV.
The primary coolant system of Millstone 3 described in Reference 1, has four (4) main loops each comprising a 33.9 inch diameter hot leg, a 36.2 inch diameter crossover leg and 32.2 inch diameter cold leg piping. The material in the primary loop piping is cast stainless steel (SA 351 CF8A). In its review of Reference 1, 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;
- 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 2, Section 4.1, "NRC Evaluation Criteria," and are as follows:
(1) The loading conditions should include the static forces and moments (pressure, deadweight and thermal expansion) due to normal operation, and the forces and moments associated with the safe shutdown earth-quake (SSE). These forces and moments should be located where the
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highest stresses, coincident with the poorest material properties, are induced 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; 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 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 capa-bility when the pipe is subjected to normal operational loads.
. (4) It should be demonstrated that the postulated leakage crack is stable under nomal plus SSE loads for long periods of time; that is, crack growth, if any, is minimal during an earthquake.
The margin, in tems of applied loads, should be detemined 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 comparing 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 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).
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Based on its evaluation of the analysis contained in Westinghouse Report WCAP-10587 (Reference 1), the staff finds that the applicant has presented an acceptable technical justification for eliminating the dynamic loading effects associated with the postulated full flow area circumferential and longitudinal pipe ruptures in the main loop primary coolant system of Millstone 3.
These dynamic loading effects include pipe whip, jet impingement, asymmet'ric pressurization transients and break associated dynamic transients in unbroken portions of the main loop and connected branch lines (branch line LOCA loads would be retained in the design basis). This finding is predicated on the fact that each of the parameters evaluated for Millstone 3 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 2032 kips (axial), 28,789 in-kips (bending moment) and result in maximum stresses of about 78% of the bounding stresses used by Westinghouse in Reference 3.
(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 of corrosion
. (e.g., intergranular stress corrosion cracking), water 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 Millstone 3, using an initial through-wall crack of 7.5 inches are identical to those of Enclosure 1 to Reference 2.
The Millstone plant has 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 through the postulated flaw results in a factor of at least 10 relative to the sensitivity of the Millstone 3 detection systems.
(4) The margin in terms of load of the Millstone unit based on fracture mechanics analyses for the leakage-size crack under nonnal plus SSE loads is within the bounds calculated by the staff in Section 4.2.3 l
of Enclosure 1 to Reference 2.
Based on a limit-load analysis, the l
load margin is about 2.8 and based on the J limit discussed in (6) t below, the margin is at least 1.5.
(5) The margin between the leakage-size crack and the critical-size crack was calculated by a limit load analysis. Again, the results l
demonstrated that a margin of at least 3 exists and is within the bounds of Section 4.2.3 of Enclosure 1 to Reference 2.
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. (6) As an integral part of its review, the staff's evaluation of the material properties data of Reference 7 is enclosed as Appendix I to this Safety Evaluation Report. In Reference 7, data for ten (10) plants, including the Millstone unit, 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 2
applied J for Millstone 3 in Reference 1 was less than 3000 in-lb/in and hence the staff's upper bound on the applied J (refer to Appendix I, page 6) was not exceeded.
In view of the analytical results presented in Reference 1 and the staff's evaluation findings related above, the staff concludes that the probability or likelihood of large pipe breaks occurring in the primary coolant system loops of Millstone 3 is sufficiently low such that protective devices associated with postulated pipe breaks at the eight (8) locations per loop in the Mill-stone 3 primary coolant system need not be installed. However, in order to provide the Commission with an opportunity to consider the long term aspects of the NRC staff's recent acceptance criteria of the " leak-before-break" approach, this exemption is ifmited to a period extending until completion of the second refueling outage of Millstone Unit 3 pending the outcome of Com-mission rulemaking on this issue.
Eliminating the need to consider these dynamic loads for this particular application has not affected the design bases for the emergency core cualing system, the overall containment, the response of engineered safety features systems, or the environmental qualification of equipment for Millstone 3.
Also, it does not propose to alter the design bases of reactor cavity and subcompartment pressurization from that originally approved, which were based on the governing piping ruptures.
The staff also reviewed the value-impact analysis provided by the appli-cant in their September 12, 1984 subm'ittal for not providing protective structures against the dynamic loading effects of postulated reactor coolant system loop pipe breaks to assure as low as reasonably achievable (ALARA) exposure to plant personnel. Consideration was given to design features for reducing doses to personnel who must operate, service and maintain the Millstone 3 instrumentation, controls, equipment, etc. The Millstone Unit 3 value-impact analysis show that the elimination of protective devices for RCS pipe breaks will save an occupational dose for plant personnel of approximately 200 person-rem for the facility over its operating lifetime.
The staff review of the analysis shows it to be a reasonable estimate of dose savings. Therefore, with respect to occupational exposure, the staff finds that there is a radiological benefit to be gained by eliminating the need for the protective structures.
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In view of the staff's evaluation findings, conclusions, and recomen-dations above, the Commission has determined that, pursuant to 10 CFR 50.12(a),
this exemption is authorized by law and will not endanger life or property or l
the comon defense and security and is otherwise in the public interest. The l
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. Consission hereby approves the limited schedular exemption from GDC 4 of Appen-dix A to 10 CFR Part 50, to eliminate the requirement to install protective devices and the requirement to consider dynamic effects and loading conditions, as detailed in Part II of this exemption, associated with postulated pipe breaks of the eight (8) locations per loop in the Millstone Unit 3 primary coolant system. This exemption is effective for a period ending at the completion of the second refueling outage, pending the outcome of rulemaking on this subject.
Pursuant to 10 CFR 51.31, the Commission has determined that the issuance of the exemption will have no significant impact on the environment (50 FR '21954 ).
The exemption will become effective upon date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
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pl Hugh L. Thompso, /Jr., Director Division of Licensing Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland this 5th day of June 1985.
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. (1) Westinghouse Report WCAP-10587, " Technical Bases for Eliminating Large Primary Loop Pipe Ruptures as the Structural Design Basis for Millstone Unit 3, 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 Reponse to Questions and Conrnents 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 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 (Preprint 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:
Ncn-proprietary versions of References 1, 3, 4, 5 and 7 are available in the NRC Public Document Room as follows:
(1) WCAP-10586, non-proprietary (2) WCAP-9570 (3) WCAP-9788 (5) Non-proprietary version attached to the Letter Report (6) WCAP-10457
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