ML20136C984
| ML20136C984 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Vogtle |
| Issue date: | 08/20/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML082840446 | List:
|
| References | |
| FOIA-84-663 NUDOCS 8601040068 | |
| Download: ML20136C984 (12) | |
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glph B f ATTACHMENT GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT UNITS 1 & 2 DOCKET NOS. 50-424 AND 50-425 SAFETY EVALUATION REPORT ON THE ELIMINATION OF LARGE PRIMARY LOOP RUPTURES AS A DESIGN BASIS 4
Component Integrity Section Materials Engineering Branch Division of Engineering
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. INTRODUCTION, By letter dated October 25, 1983, Georgia Power Company, the applicant for Vogtle Units 1 & 2, submitted a report (Reference 1) on the technical I
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 2
Part 50 in regard to the need for protection against dynamic effects from postulated pipe breaks. After meeting with the applicant and I
Westinghouse, the staff Tomally responded by letter (Reference 2) dated March 19, 1984 to transmit the staff's comments and questions on the submittal. The response to the staff's concerns' resulted in l
j a revision to the report, Reference (3), which was submitted to the NRC on May 17, 1984. By means of deterministic fracture mechanics analyses, the applicant. contends that postulated double-ended guillo-tine breaks (DEGBs) of the primary loop reactor coolant piping will not occur in Vogtle Units 1 & 2 and therefore need not be considered
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0601040060 851127' i
BELL 84-663 4
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effects associated with such postulated breaks. No ~ other changes in design requirements are addressed within the scope of the referenced i
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reports; e.g., no changes to the definition of a LOCA nor its relation-ship to the regulations addressing design requirements for ECCS (10 CFR i
50.46), containment (GDC 16, 50), other engineered safety features and the conditions for environmental qualification of equipment (10 CFR 50.49).
The Commi.ssion's regulations require that applicants provide protective
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measures against the dynamic effects o'f postulated pipe breaks in high 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.
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l This~1ed to the establishment of Unresolved Safety Issue (USI) A-2,
.!j "Asyneetric Blowdown Loads on PWR Primary Systems."
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q The NRC staff, after several review meetings with the Advisory Consittee on Reactor Safeguards (ACRS) and a meeting with the HRC Committee to Review Generic Requirements (CRGR), concluded that an exemption from -
the ngulations would be acceptable as an alternative for resolution of USI A-2 for 16 facilities owned by 11 licensees in the Westinghouse l
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Owner's Group (one of these facilities, Fort Calhoun has a Combustion
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. Engineering nuclear steam supply system). This NRC staff position was stated in Generic Letter 84-04, published on February 1, ~1984
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i (Reference 4). The generic letter states that the affected licensees I
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 analyses that the detection of.small flaws by either inservice inspection or leakage monitoring systems is assured long before th'e flaws can grow to critical or unstable sizes which could lead to large break areas such as the DEGB or l
its equivalent. The concept underlying such analyses is referred l
l to as " leak-before-break" (LBB). There is no implication that piping failures cannot occur, but rather that improved knowledge of.the a
!j failure modes of piping systems and the application af appropriate i
remedial measures, if indicated, can reduce the probability of catastrophic failure to insignificant values.
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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
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blowdown loads that resulted from a limited number of discrete break l
locations, the technology advanced in these topical reports demonstrated i
that the probability of breaks occurring in the primary coolant system i
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 impingemdt shields.
The staff's Topical Report i
Evaluation is attached as Enclosure 1 to Reference 4.
1-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 l
(Reference 8) conflin 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 8 are that the best-estimate leak probabilities for Westinghouse nuclear
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~7 steam supply system main loop piping range from 1.2 x 10 to 1.5 x 10 per plant year and the best-estimate DEGB probabilities range from
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1 x 10 to 7 x 10 per plant year.
Similarly, the best-estimate i
leak probabilities for Combustion Engineering nuclear steam supply l
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-8 system main loop piping range from 1 x 10 per plant year to 3 x 10 per plant year, and the best-estimate DEG8 probabilities range from
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-13 5 x 10 to 5 x 10 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 require-ment 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 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 radiation 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.
PARAMETERS EVALUATED BY THE STAFF The primary coolant system of Vogtle Units 1 and 2, described in Reference 3, has four (4) main loops each comprising a 33.9 inch diameter hot leg, a 36.2 inch diameter crossover leg and 32.14 inch t(
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. diameter cold leg piping. The material in the primary loop piping is cast stainless steel (SA 351 CFBA).
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;
- stability of a " leakage-size crack" under nanmal plus SSE loads, and the expected margin in terms of load; margin based on crack size; and i
the fracture toughness properties of thermally-aged cast stainless j
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steel piping and weld material.
r STAFF CRITERIA USED IN THE EVALUATION 6
T,he NRC staff's criteria for evaluation of the above parameters are delineated in its Topical Report Evaluation, Enclosure 1 to Reference 4, i
Section 4.1, "NRC Evaluation Criteria," and are as follows:
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(1) The loading conditions should include the static forces and moments i
j (pressure, deadweight and thermal expansion) due to normal operation, ij and the forces and moments associated with the safe shutdown earth-
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quake (SSE).
These forces and moments should be located where the i
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, j
should be provided.
Relevant operating history should be cited, which
-l 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 ii 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.
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1 (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-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 i
leakage-size crack and the critical-size crack to account for the j
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
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materials specifications used for base metal, weldments and safe-en'ds, 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, aaximum crack l
growth).
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- 9-i STAFF EVALUATION AND CONCLUSIONS I
l Based on its evaluation of the analysis contained in Westinghouse Report WCAP-10551 (Reference 3), the staff finds that the applicant has 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 Vogtle, Units 1 and 2.
This finding is predicated on the fact that each of the parameters evaluated for Vogtle is enveloped by the generic analysis performed by Westinghouse j
in Reference 5, and accepted by the staff in Enclosure 1 to Reference 4.
Specifically:
I (1) The loads associated with the highest stressed location in the main 4
loop primary system piping are considerably lower than the bounding I
loads used by Westinghouse in Reference 5, or those established by the staff as limits (e.g., a moment of 42,000 in-kips in i to Reference 4).
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(2) For Westinghouse plants, there is no history of cracking failure f
i 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 l
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. t 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 i
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of operation and 15 other plants with over 10 years of operation.
(3) The results of the leak rate calculations performed for Vogtle, using an initial through-wall crack are identical to those of to Reference 4.
The Vogtle plant has an RCS pressure boundary leak detection system which is consistent with the guide-lines 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 is large relative to the sensitivity of the Vogtle plant leak detection system.
t (4) The expected margin in terms of load for the leakage-size crack under normal plus SSE loads is within the bounds calculated by 4
l the staff in Section 4.2.3 of Enclosure 1 to Reference 4. - In addition, the staff found a significant margin in terms of loads larger than normal plus SSE loads.
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(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 Reference 4.
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(6) As an integral part of its review, the staff's evaluation of the material properties data of Reference 9 is enclosed as Appendix I to this Safety Evaluation Report.
In Reference 9, data for ten (10) plants, including the Vogtle Units, are presented, and lower bound or " worst case" materials properties were identified and used in the analysis performed in the Reference 3 report by Westinghouse.
2 The staff's upper bound of 3000 in-lb/in or the applied J (refer to Appendix I, page 6) was not exceeded; the applied J for Vogtle 2
in Reference 3 was substantially less than 3000 in-lb/in,
In view of the analytical results presented in Reference 3 and the staff's evaluation findings related above, the staff concludes that the probability
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or likelihood of large pipe breaks occurring in the primary coolant system e
loop of Vogtle Units 1 and 2 is sufficiently low such that protective devices associated with postulated pipe breaks at the eight (8) loca-tions per loop in Vogtle Units 1 and 2 primary coolant system (as specified in the applicant's letter of April 2,1984 which forwarded their safety balance report need not be installed.
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(1) Westinghouse Report MT-SME-3082, " Technical Bases for Eliminating Large Primary Loop Pipe Ruptures as the Structural Design Basis for Alvin W.'Vogtle Units 1 and 2," September 1983, Westinghouse Class 2 proprietary.
(2) Letter to Donald A Foster of Georgia Power Company, " Request for Additional Information Concerning Leak-Before-Break Analysis for Vogtle Electric Generating Plant (Units.1 and 2)," dated March 19, 1984.
(3) Westinghouse Report WCAP-10551, " Technical Bases for Eliminating
_i Large Primary Loop Pipe Rupture as the Structural Design Basis
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for Alvin W. Vogtle, Units 1 and 2, May 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 and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation, WCAP-9787, May 1981, Westinghouse Class 2 proprietary.
(7) 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.
(8) Lawrence Livermore National Laboratory Rt6ert, UCRL-86249, " Failure Probability of PWR Reactor Coolant Loop Piping," by T. Lo, H. H. Woo, i
G. S. Holman and C. K. Chou, February 1984 (Preprint of a paper intended for publication).
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(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.
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