ML20127H080
| ML20127H080 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 01/11/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20127H084 | List: |
| References | |
| NUDOCS 9301220184 | |
| Download: ML20127H080 (3) | |
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SAFETY LyALMARQN BY THE OFFICE OF NUCLEAR Rf ACTOR REGULATION EELATED TO THE.USE OF LEAK-BEFORE-BREAK TECHNOLOGY SOUTH CAROLINA ELECTRIC & GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATIQN. UNIT NO. I D_0CKET NO, 50-395
1.0 INTRODUCTION
On June 22, 1992, South Carolina Electric & Gas Company (the licensee) requested that the dynamic effects of postulated pipe ruptures in the reactor coolant loop piping be eliminated from the design basis for the Virgil C.
Summer Nuclear Station, Unit No. 1 (Summer).
The request was based on a leak-before-break (LBB) analysis performed by Westinghouse (Reference 1) as permitted by General Design Criterion 4 (GDC-4) of Appendix A to 10 CFR Part 50.
The design basis for the Class 1 piping requires that the dynamic effects of pipe breaks be evaluated and that pipe whip restraints and jet impingement barriers be installed to protect safety systems.
Since the mid-1980's, the NRC has determined that such breaks have been shown to be unlikely and may be deleted from the design basis if the piping system can be shown to qualify for leak-before-break.
GDC-4 allows the use of the plant-specific LBB analysis to eliminate the dynamic effects of postulated pipe ruptures in high energy piping from the design basis.
A licensee, with an NRC approved LBB analysis, may remove pipe whip restraints and jet impingement barriers.
The acceptance criteria for the LBB analysis are defined in NUREG-1061 and draft Standard Review Plan (SRP) 3.6.3.
They are summarized as follows:
The LBB analysis should provide data on materials specifications and limitations, and age-related degradations such as thermal aging.
The piping materials must be free from brittle cleavage-type failure over the full range of the system operating temperature.
The analysis should consider the forces and moments due to pressure, deadweight,'(SSE). thermal expansion, operating basis earthquake, and safe shutdo earthquake The analysis should identify location (s) at which the -
highest stresses are coincident with the poorest material properties for base metals, weldments, and safe ends.
The analysis should postulate a through-wall flaw at the highest stressed locations.
The flaw size should be large encugh so that any leakage is assured of being detected with at least a margin of 10 on leakage using the minimum installed leak detection capability when the pipe is subjected to normal operational loads.
9301220184 930111 PDR ADOCK 05000395 P
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. The analysis should show that the postulated leakage flaw is stable under faulted conditions (normal plus SSE loads).
The leakage flaw should also be stable under larger loads at least 1.4 times the normal plus SSE loads.
However, the margin of 1.4 may be reduced to 1.0 if the individual normal and SSE loads are summed absolutely.
Under normal plus SSE loads, the safety margin should be at least a factor of two between the leakage-size flaw and the critical-size flaw to account for the uncertainties inherent in the analyses and leakage detection capability, The analysis should provide operating experience to show that the pipe will not experience stress corrosion cracking, fatigue, or water hammer.
The operating history should include system operational procedures; system or component modification; water chemistry parameters, limits, and controls; resistance of piping material to various forms of stress corrosion; and performance of the pipe under cyclic loadings.
For Class 1 piping, a fatigue crack growth analysis should be performed to show that the postJlated flaw (s) at highest stress location (s) Will not grow significantly during 40 years of service.
2.0 LVALU_AT ION The reactor coolant system (RCS) piping at Summer has an as-built outside diameter of 33.90 inches with a minimum wall thickness of 2.205 inches.
The piping material is austenitic wrought stainless steel SA376 TP304N, and the elbow fittings are cast stainless steel SA351 CF8A.
For cast stainless steel, thermal aging must be considered because it decreases the fracture toughness of the material.
The licensee showed that the fracture toughness of the elbow material is as good as or better than cast stainless material that was thermally aged and tested in a Westinghouse study.
In the stability analysis, the licensee used the conservative material
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properties from the Westinghouse study for the elbow.
The applied J value and applied tearing modules were shown to be less than the maximum J value and tearing modules of the elbow material.
The staff finds that the licensee has addressed the thermal aging issue satisfactorily.
The licensee used loads from effects of pressure, deadweight, thermal
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expansion, and SSE in the track stability analysis.
The critical stress locations and analysis methods were selected based on the fact that cast v
stainless steel (elbow) has a lower fracture toughness than the austenitic y
stainless steel (pipe).
The critical location, based on the applied load, is
,y located at the junction between the reactor vessel outlet nozzle and hot leg.
Based on fracture toughness, the critical locations are located at the elbow p
between each hot leg and each steam generator and at the elbow between the
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cold leg and the reactor vessel.
To determine crack stability at critical locations, the licensee applied the modified limit load method as specified in i
draft SRP 3.6.3 to austenitic stainless steel and the J-integral method to cast stainless steel.
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. The licensee showed that the postu14ted leakage flaw is stable under normal plus SSE loads.
The applied loads were combined absolutely and the safety margin on loads was shown to comply with the recommended value of 1.0 in NUREG-1061.
The licensee showed that the margin between the leakage-size flaw and the critical-size flaw satisfies the recommended value of 2.0 for'all the load cases.
e The licensee stated that the leak detection system for the reactor coolant pressure boundary meets the intent of Regulatory Guide 1.45 which recommends that a leakage of one gallon per minute in I hour be detected.
The licensee used a margin of 10 on leakage in calculating the leakage crack size.
This is consistent with the LBB criteria in NUREG-106).
For the fatigue crack analysis, the licensee calculated the crack growth in 40 years using crack growth equations in Appendix A to Section XI of the ASME Code.
Thermal transients, including number of cycles and temperature 1
differentials, were used.
The maximum crack size at end of life, propagated from a postulated 0.425 inch deep crack, wa. calculated to be 0.474 inch.
The staff finds both the fatigue crack growth evaluation and results are acceptable.
The licensee showed that for Westinghouse plants there is no history of stress corrosion cracking in the RCS piping because of controls in the water chemistry, and there is a low probability for water hammer because the RCS is designed and operated to preclude the voiding condition necessary to generate severe water hammer transients.
The staff finds that the licensee has addressed stress corrosion cracking and water hammer satisfactorily.
3.0 CONCLUSION
The NRC staff has performed independent flaw stability calculations to evaluate the licensee's LBB analysis of the reactor coolant piping at Summer.
The staff concludes that the licensee's LBB analysis is consistent with the criteria in NUREG-1061, Volume 3, and draft SRP 3.6.3.; therefore, the analysis complies with GDC-4.
Thus, the probability of large pipe breaks -
occurring in the RCS line is sufficiently low that-dynamic effects associated with postulated pipe breaks need not be a design basis.
4.0 REFERENCES
WCAP-13206, " Technical Justification for Eliminating Large-Primary Loop Pipe Rupture as the Structural Design Basis for The Virgil C. Summer Nuclear Power Plant," Westinghouse Electric Corporation, April 1992 (proprietary).
Principal Contributor:
J. Tsao Date:
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