ML20138M389
| ML20138M389 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Midland, Bellefonte, 05000000, Washington Public Power Supply System |
| Issue date: | 12/12/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20137Y048 | List: |
| References | |
| TASK-A-02, TASK-OR GL-84-04, NUDOCS 8512200449 | |
| Download: ML20138M389 (10) | |
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ATTACHMENT THE B&W OWNERS GROUP 287, 302 312, 313, 329, DOCKETN05.50-269,270,439,&466 346, 438, SAFETY EVALUATIDN REPORT ON THE ELIMINATION OF LARGE PRIMARY LOOP RUPIURE5 A5 A DESIGN BASIS Section A Engineering Branch Division of PWR Licensing-B INTRODUCTION By letter dated September 7, 1984, the B&W Owners Graup (B&WOG), on behalf of participating utilities with B&W designed facili*.jes, submitted a generic report (Reference 1) on the technical bases for eliminating large primary loop piping ruptures as a design basis.
Reference 1 presented the results of a bounding evaluation for the following B&WOG members:
Licensee or Applicant Facility Arkansas Power & Light Co.
ANO-1 Consumers Power Co.
Midland-2 Duke Power Co.
Oconee.'., 2, 3 Florida Power Corp.
Crystal River 3 Sacramento Municipal Utility District Rancho Seco Supply System WNP-1 Tennessee Valley Authority Bellefonte 1, 2 I
Toledo Edison Co.
Davis-Besse 1
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2-The Reference 1 submittal was made to provide technical justificati.on for the preceding licensees and applicants of the B&WOG in regard to 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.
After meeting with the B&WOG, the staff formally responded by letter (Reference 2) dated March 12, 1985, to transmit the staff's comments and questions on the submittal. The response to the staff's concerns resulted in a revision to the submittal, Reference 3,andanadditionalreport(Reference 4)onmaterialproperties data, both of which were transmitted to the NRC on October 22, 1985.
By means of deterministic fracture mechanics analyses, the B&WOG contends that postulated double-ended guillotine breaks (DEG8s) of the primary loop reactor coolant piping will not occur in the facilities addressed in References 3 ad 4 and therefore need not be considered as a design basis for installing protective devices such as pipe whip restraints to guard against the dynamic offacts associated with such postulated breaks.
No other changes in design requirements are addressed within the scope of the referenced reports; e.g., no changes to the definition of a LOCA nor its relationship to the regulations addressing design requirements for ECCS (10 CFR 50.46), containment (GDC 16, 50), other engineered safety features and the conditions for environmental qualification of equipment (10CFR50.49).
The Commission's regulations require provision of 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 asymmetric loads on pressurized water reactor (PWR) vessels and their internals which could result from these large postulated breaks at discrete I
locations.in the main primary coolant loop piping.
This led to the establish-mentofUnresolvedSafetyIssue(USI)A-2,"AsymmetricBlowdownLoa'dsonPWR Primary Systems."
. 'B The NRC staff, after several review meetings with the Advisory Committee on ReactorSafeguards(ACRS)andameetingwiththeNRCCommitteetoReview Generic Requirements (CRGR), concluded that an exemption from the reguia-tions would be acceptable as an alternative for resolution of USI A-2 i
for 16 facilities owned by 11 licensees in the Westinghouse Owner's Group (oneofthesefacilities,FortCalhounhasaCombustionEngineeringnuclear steamsupplysystem). This NRC staff position was stated in Generic Letter 84-04, published on February 1, 1984 (Reference 5). The generic letter states thattheaffectedlicenseesmustjustifyanexemptiontoGDC4onaplant-specific basis. Other PWR applicants or licensees may request similar exesiptions 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 analyses that the detec-tion of small flaws by either inserv' e 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 DEG8 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 of appropriate remedial measures, if indicated, can reduce the probability of catastrophic failure to insigniff-cant values.
Advanced fracture mechanics technology was applied in topical
- reports (References 6,7,and8)submittedtothestaffbyWestinghouseonbehalfof 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
4 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 5.
Probabilistic fracture mechanics studies conducted by the Lawrence Livermore National Laboratories (LLNL) on both Westinghouse and Combustion Engineering nuclear steam supply system main loop piping (Reference 9) confirm that both the probability of leakage (e.g., undetected f1pw growth through the pipe wall by fatigue) and the probability of a DEG8 are very low. The t
results.given in Reference 9 are that the best-estimate leak probabilities for Westinghouse nuclear steam supply system main loop piping range from
-8
~7 1.2 x 10 to 1.5 x 10 per plant year and the best-estimate DEGB proba-
-12
-12 bilities range from 1 x 10 to 7 x 10 per plant year.
Similarly, the 1
best-estimate leak probabilities for Combustion Engineering nuclear steam
~0 supply system main loop piping range from 1 x 10 per plant year to
-8 3 x 10 per plant year, and the best-estimate DEGB probabilities range
~13 from 5 x 10~14 to 5 x 10 per plant year.
In addition, LLNL recently conducted an evaluation of BW nuclear steam supply main loop piping with the result that the best-estimate leak and DEG8 probabilities are nominally identical to those calculated for the Westinghouse and Com-
~~
i bustion Engineering studies. 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 1
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 5.
Even for new plant's, these devices tend to restrict access for future inservice inspection of piping; or if they are remove,d and i
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, hign occupational radiatica exposure (ORE) would be incurred while public risk reduction would be very low.
Removal and reinsta11ation for inservica inspection also entail significant ORE over the life of a plant.
i PARAMETERS EVALUATED BY THE STAFF The B&WOG facilities evaluated in Reference 3 include both 177-FA and 205-FA plants and configurations of the lowered-and-raised-loop designs.
The pri-mary coolant loop piping of these facilities are comprised of straight sections and elbows in each of four pipe sizes - 28, 32, 36 and 38 inch diameters. The piping materials in the primary main loops are low alloy ferritic~ steels (SA-106 GrC, SA-508 C1 1, and SA-516 Gr 70) and wrought stainless steel safe ends (SA-376 TP 316).
In its review of References 3 and 4, the staff evaluated the B&WOG analyses and materials data with regard to:
the location of maximum stresses in the piping, associated with the combined loads from normal operation and the Safe Shutdown Earth-quake (SSE);
potential cracking mechanisms; size of postulated 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-low alloy, ferritic steel piping, wrought stainless steel safe ends and associated weld material.
i
t -'
6-STAFF CRITERIA USED IN THE EVALUATION 1
i j
The NRC staff's criteria for evaluation of the above parameters are l
delineated in the Report of the U.S. Nuclear Regulatory Commission Piping
}
i Review Committee, NUREG-1061,-Volume 3 " Evaluation of Potential for i
Pipe Breaks." These criteria are enumerated in Chapter 5.0 of Volume 3 l
l of the NUREG and are as follows:
f-(1) The loading conditions should include the static forces and moments l
(pressure, deadweight and thersal expansion) due to poreal operation, l
- and the forces and moments associated with the safe shutdown earth-i quake (SSE). These forces and soments should be located where the i
i highest stresses, coincident with the poorest material properties,
)
are induced for base materials, weldsents and safe-ends.
4' (2).Forthepipingrun/systemsunderevaluation,allpertinentinformation j
which demonstrates that degradation or failure of the piping resulting from stress corrosion cracking, fatigue or water hammer is not likely, I
should be provided.
Relevant operating history should be cited,'which j
includes system operational procedures; system or component modifica-l l
tion; water chemistry parameters, limits and controls; resistance of
[
I materialtovarjousformsofstresscorrosion,andperformanceunder cyclic loadings.
l (3) A through-wall crack should be postulated at the highest stressed l
locationsdeterminedfrom(1)above. The size of the crack should j
j be large enough so that the leakage is assured of detection with at least a factor of ten using the minimum installed leak detection l
capabilitywhenthepipeissubjectedtonormaloperationalloads.
(4) It should be demonstrated that the postulated leakage crack;is stable i
unde'r normal plus SSE loads for long periods of time; that is,. crack I
growth, if any, is minimal during an earthquake. The margin, in terms l
]
I l
i I
- of applied loads, should be at least the [ T and 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 designloads)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 should be determined by comparing leakage-size crack to critical-size cracks. Under normal plus SSE loads, it should be demonstrated that there is a margin of at least 2 between the leakage-
. size crack and the critical-size crack to account foF 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, weldsents and safe-ends, the materials properties including the J-R curve used in the analyses, and long-tern effects such as thermal aging and other limitations to valid data (e.g., J maximum, maximum crack growth).
STAFF EVALUATION AND CONCLUSIONS Based on its evaluation of the analysis contained in BAW-1847, Rev. 1 (Reference 3) and the materials data presented in BAW-1889P (Reference 4),
thestafffindsthattheB&WOGhaspresentedacceptabletechnicaljustifi-cation, addressing the preceding criteria, to eliminate, as a design basis, the dynamic effects of large ruptures in the main loop primary coolant piping of the B&WOG facilities evaluated.
Specifically:
S
8-(1) The loads associated with the highest stressed location in the main loop primary system piping are 1,685.7 kips (axial), 37,171 in-kips (bendingmoment)andresultinmaximumstressescfabout51%of Service Level D limits specified in Section III of the ASME Code.
(2) For the B&WOG facilities, there is no history of cracking failure in reactor primary coolant system main loop piping. The reactor coolant system primary loop has an operating history which demon-strates its inherent stability. This includes a low susceptibility to cracking failure from the effects of corrosion (e,g., inter-
.granularstresscorrosioncracking),waterhammer,orfatigue(Iow andhighcycle).
This operating history totals over 53 reactor-years spanning 13 years of operation.
(3) The leak rate calculations perfomed for the B&WOG facilities used initial postulated throughwall flaws larger in size thai those of to Reference 5.
B&WOG facilities have an RCS pressure boundary leak detection system which is consistent with the guide-linesofRegulatoryGuide1.45suchthatleakageofone(1)gpain one hour can be detected. The calculated leak rate through the postulated flaw is large relative to the staff's required sensitivity of plant leak detection systems; the margin is at least a factor of ten (10)onleakage.
(4) The margin in terms of load based on fracture mechanics analyses for the leakage-size crack under normal plus SSE loads (Service Level D loads)meetsNUREG-1061, Volume 3,guidanceonmargins. Based on a limit-load analysis, the load margin is at least fT.
Similarly, based on the J limit, the margin is at least U.
~
- e (5) The margin between the leakage-size crack and the critical-size crack w'as calculated by a limit load analysis.
Again, the results demonstrated that a margin of at least 2.0 exists and is within i
the guidelines of NUREG-1061, Volume 3.
(6) In their review of the reactor coolant piping, the B&WOG first listed all the base metals and weld metals represented.
From a review of published test data -- J-R curves and tensile properties --
the materials from the list that were most likely to be limiting were identified. A test program was then conducted to obtain the toughness
' and tensile data required.
From these data, a limit'ing J-R curve and the associated tensile stress-strain curve was selected for the fracture analyses of the base metal and weld metal in the straight sectiont tid elbows of the piping identified for evaluation. The staff concludes that the choice of limiting materials is satisfactory.
In view of the analytical results presented in Reference 3, the materials data contained in Reference 4, 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 the 8&WOG facilities is sufficiently low such that dynamic effects associated with postulated pipe breaks in these facilities need not be a design l-basis.
e O
L
10 -
1.
M W Owners Group Report BAW-1847 " Leak-Before-Break Evaluation of Marg"insAgainstFullBreakforRUSPrimaryPipingofB&WDesigned NSS, September 1984.
2.
Letter to L. C. Oakes of the B&W Owners Group," B&WOG Leak-Before-Break Report, BAW-1847," dated March 12, 1985.
3.
B&W Owners Group Resort BAW-1847, Rev.1 " Leak-Before-Break Evaluation of Margins Against tull Break for RCS Primary Piping of B&W Designed NSS," September 1985.
4.
B&W Owners Group Report BAW-1889P, " Piping Material Properties for Leak-Before-Break Analysis," A. L. Lowe, Jr., K. K. Yoon und R. H.
Emanuelson, October 1985, proprietary.
5.
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.
6.
Westinghouse Report WCAP-9558 Rev. 2 " Mechanistic Fracture Evaluation ofReactorCoolantPipeContaIningaEcstulatedCircumferentialThrough-wall Crack," May 1981, Class 2 proprietary.
7.
Westinghouse Report WCAP-9687, " Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation,"
May 1981, Class 2 proprietary.
8.
Westinghouse Response to Questions and Comments Raised by Members of ACRS Subcosmittee on Metal Components During the Westinghouse Presenta-tion on September 25, 1981 Letter Report MS-EPR-2519, E. P. Rahe to Darrell G. Eisenhut, November 10, 1981, Westinghouse Class 2 proprietary.
9.
T. Lo, H. H. Woo, G. S. Holman and C. K. Chou, " Failure Probability of PWR Reactor Coolant Loop Piping," presented at the ASME PVP Conference and Exhibition, June 17-21, 1984, San Antonio, Texas.