ML20235T663
| ML20235T663 | |
| Person / Time | |
|---|---|
| Site: | Seabrook, 05000000 |
| Issue date: | 01/18/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20235T530 | List:
|
| References | |
| FOIA-87-51 GL-84-04, GL-84-4, NUDOCS 8707220223 | |
| Download: ML20235T663 (11) | |
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i ATTACHMENT SEABROOK STATION, UNITS 1 AND 2 DOCKET N05. 50-443/444 SAFETY EVALUATION REPORT ON THE ELIMINATION OF LARGE PRIMARY LOOP RUPTURES AS A DESIGN BASIS Component Integrity Section Materials Engineering Branch Division of Engineering INTRODUCTION
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By letter dated August 9,1984, the applicant for Seabrook, Units 1 and 2, submitted a report (Reference 1) on the technical bases for eliminating large primary loop piping ruptures as a structural design basis. This submittal was made in support of a request for an exemption to General Design Criterion (GDC) 4 of Appendix A to 10 CFR Part 50 in regard to the need for protection against dynamic effects from postulated pipe breaks.
By means of deterministic fracture mechanics analyses, the applicant contends that postulated double-ended guillotine breaks (DEGBs) of the primary loop reactor coolant piping will not occur in the Seabrock units and therefore need not be considered as a design basis for installing protective devices such as pipe whip restraints and jet impingement shields to guard against the dynamic effects associated with such 8707220223 870717 PDR FOIA GOREN87-51 PDR
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postulated breaks.
No other changes in design requirements are addressed within the scope of the referenced report; e.g., no changes to the definition of a LOCA nor its relationship to the regulations addressing design require-ments for ECCS (10 CFR 50.46), containment (GDC 16, 50), other engineered safety features and the conditions for environmental qualification of equipment (10 CFR 50.49).
The Commission's regulations require that applicants provide protective measures against the dynamic effects of postulated pipe breaks in high
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energy fluid system piping.
Protective measures include physical isola-tion 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 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 (U51) A-2,
" Asymmetric Blowdown toads 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 facilities owned by 11 licensees in the Westinghouse
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. 0.ner'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, published on February 1,1984 (Reference 2).
The generic letter states that the affected lic. isees mast justify an exemption to GDC 4 on a plant-specific basis.
Other PW'R applicants or licensees may request similar exemptions from the requirem.ents of GDC 4 provided that they submit an acceptable technical basis for eliminating the need to postulate pipe breaks.
Tne acceptance of an exemption was made possible by the development of advanced fracture mechani/s technology.
These advanced fracture rechanics 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 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 concept underlying such analyses is referred to as " leak-before-break" (LBB).
There is no implication that piping i
failures cannot occur, but rather that improved knowledge of the
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1 failure modes of piping systems and the application of appropriate remedial measures, if indicated, can reduce the probability of cat.astrophic failure to insignificant values.
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i 4-Advanced fracture mechanics technology was applied in topcal reports (References 3, 4, and 5) submitted to the staff by Westinghouse on behalf of the licensees belonging to the USI A-2 Owners G-oup.
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 attached as Enclosure 1 to Reference 2.
Probabilistic fracture mechanics studies conducted by the Lawrence Livermore National Laboratories (LLNL) on both Westinghouse and Combustion Engineering nuclear steam supply system main icop 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
W system main loop piping range from 1 x 10 8 per plant yea
- to 3 x 10 8
- +r plant year, and the bestestimate DEGB probabilities range from 5 x 10-14 to 5 x 10 1 per plant year.
These results do not affect
- ore melt probabilities in any significant way.
Curing the past few years it has also become apparent that the require-i sent for installation of large, massive pipe whip restraints and jet
'mpingement 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 cevices tend to restrict access for future inservice inspection of piping; or if they are removed and reinstalled for inspection, there
's a potential risk of damaging the piping and other safety-related components in this process.
If installed in operating plants, high c cupational radiation exposure (ORE) would be incurred w1i1e public risk reduction would be very low.
Removal and reinsta11ation for inservice inspection also entail significant ORE over the life of a plant.
I PARAMETERS EVALUATED BY THE STAFF i
The primary coolant system of Seabrook, Units 1 and 2, described in Reference 1, have four (4) main loops each comprising a 33.9 inch diameter hot leg, a 37.5 inch diameter crossover leg and 32.4 inch
. ciameter cold leg piping.
The material in the primary loop piping is wrought stainless steel with cast stainless steel fittings and associated welds.
In its review of Reference 1, the staff evaluated tne 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 fittings and weld material.
STAFF CRITERIA USED IN THE EVALUATION 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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The loading conditions should include the static forces and moments J
j (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 :noments should be located where the 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 po.stulated 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-l bility when the pipe is subjected to normal operational loads.
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STAFF EVALUATION AND CONCLUSIONS Eased on its evaluation of the analysis contained in Westinghouse Report WCAP-10567 (Reference 1), the staff finds that the applicant nas 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, 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 Ref erence 3, and accepted by the staff in Enclosure 1 to Reference 2.
I Specifically:
(1) The loads associated with the highest stressed location in the main l
i 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 stresses used by Westinghouse in Reference 3.
l (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 I
which demonstrates its inherent stability.
This includes a l
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. low susceptibility to cracking failure from the effects ef corrosion (e.g., intergranular stress corrosion cracking), water hammer, or fatigue (low and high cycle).
This operating history totals over l
400 reactor years, including five (5) plants each having 15 years of operation and 15 other pl&11ts with over 10 years of operation.
(3) The leak rate calculations performed for the Seabrook units, using an initial through wall crack of 7.5 inches are identical to those of Enclosure I 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 through the postulated flaw results in a factor of at least 10 relative to the sensitivity of the Seabrook leak detection systems.
(4) The margin in terms of load of the Seabrook units based on fracture mechanics 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 exists and is within the bounds of Section 4.2.3 of Enclosure 1 to Reference 2.
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In addition to the wrought stainless steel pipes, the Seabrook I
units have cast stainless steel fittings and associated welds in i
the primary coolant system.
As an integral part of its review, the j
staff's evaluation of the material properties data of Reference 7 is enclosed as Appendix I to this Safety Evaluation Report.
The applied J for Seabrook in Reference 1 for cast stainless steel 2
fittings and associated welds 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 Seabrook, Units 1 and 2 is sufficiently low such that protective
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devices associated with postulated pipe breaks in the primary coolant systems of these facilities need not be installed.
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(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 Reponse 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 proprietary.
(6) Lawrence Livermore National Laboratory Report, UCRL-86249, " Failure Probability of PWR Reactor Coolant Loop Piping," by T. Lo, H. H. Woo, G. 5. 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.
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