ML20235S810

Safety Evaluation Supporting Proposed Redesign of Steam Generator Upper Support of Reactor Coolant Loop Sys for Unit 2 & Proposed Revised FSAR

ML20235S810
Person / Time
Site:	Vogtle
Issue date:	10/07/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20235S801	List: ML20235S797 ML20235S810
References
NUDOCS 8710090145
Download: ML20235S810 (3)

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UNITED STATES g .g NUCLEAR REGULATORY COMMISSION in f j WASHINGTON, D, C. 20555

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ENCLOSURE SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION STEAM GENERATOR SUPPORTS REDESIGN & FSAR REVISION V0GTLE ELECTRIC GENERATING PLANT UNIT ?

By letters dated July 1, and August 6, 1987, Georgia Power Company (the applicant) submitted a discussion of a proposed redesign of the steam gene, stor upper support of the reactor coolant loop system for the Vogtle Electric Generating Plant, Unit 2. The applicant enclosed an evaluation to verify the adequacy of the redesign and a proposed revision of the Final Safety Analysis Report (FSAR).

The support redesign and FSAR revision will permit Vogtle Unit 2 to reduce the number of large hydraulic snubbers in the steam generator support' systems from five to two for all four steam generators. The technical basis for the FSAR revision is the use of. " leak-before-break" principle which would eliminate the dynamic (ffects of postulated pipe ruptures from the primary piping design basis. The cvaluation showed that the configuration of the redesign will be able to withstand all remaining loadings, including those caused by the safe shutdown earthquake and the limiting high energy line breaks at branch nozzles.

Specifically, the evaluation indicated that stresses in the reactor coolant ,

loop piping are within the FSAR allowables with adequate margins of safety. '

The Vogtle Unit 2 reactor coolant system consists- of four reactor coolant -l loops (RCLs) with one steam generator per loop. An identical ~ design was used j for the supporting system of all four steam generators. The upper support is '

! an octagonal ring girder with curved bearing surfaces placed around the steam I generator shell and hung from the steam generator trunnions by four tie rods.

These tie rods support the dead weight of the ring assembly and position the ,

ring girder vertically. Three rigid struts carrying loads from the steam generator through the ring girder are anchored to the secondary shield wall to restrict the movement of the steam generator. One strut is located on ,

each of the three sides of the steam generator not facing the reactor. Five hydraulic snubbers located parallel to the hot leg are connected to the ring girder on the reactor side of the steam generator. The' snubbers will pemit the unrestrained thermal loop movement to the final hot operating position.

The proposed redesign effort is limited to the upper support and will reduce the number of enubbers from five to two in each support.

In Volume 3, " Guillotine Break Indirectly Induced by Earthquakes," of NUREG/CR-3660,

" Probability of Pipe Failure in Reactor Coolant Loops of Westinghouse PWR

. [ Pressurized Water Reactorl Plants," the Lawrence Livermore National- Laboratory performed an independent review on Westinghouse topical. reports WCAP-10551 and WCAP-1055?. The results indicated that the design and construction practices used in Westinghouse PWR plants will not significantly affect the probability of a double-ended guillotine break; thus the elimination of large primary loop pipe rupture as a structural design for Westinghouse PWRs can be pemitted.

This change results'in reduced loading level, which in turn, requires less upper support rigidity for dynamic events considered in the plant design.

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Loadings considered in the redesign are those caused by deadweight, internal pressure, thermal movement, seismic events (operating basis earthquake and safe j shutdown earthquake), and postulated pipe ruptures at nozzles. l 1

The staff's evaluation of the RCL leak-before-break analysis for the original l support configuration is attached to the staff's exemption to General Design Criterion 4 for Vogtle dated February 5,1985. A comparison has been made of the original loads with the revised loads from the steam generator upper lateral support redesign. Based upon this comparison it was verified that the staff's original conclusions regarding the RCL leak-before-break analysis remain valid.

The WESTDYN computer code was used as the analytic tool. This code was discussed in Westinghouse topical report WCAP-8252, Revision 1, " Documentation of Selected Westinghouse Structural Analysis Computer Codes." The staff approved this report i by letter dated April 7,1981, from R. L. Tedesco to T. M. Anderson of Westinghouse. '

The mathematical models used consisted of mass and stiffness representations for all four RCLs and the reactor vessel. Three directional seismic analyses were perfonned using envelope response spectra with damping values of two and four i percent for operating basis earthquake and safe shutdown earthquake, respective ~1y. l The three directional seismic responses were combined by the square-root-sum-of-squares (SRSS) method. Combination of closely spaced modes was conducted i according to the approved FSAR method. Time history forcing functions for the postulated breaks at five nozzles (the pressure surge, the accumulator, the residual heat removal, the main steam, and the feedwater line nozzles) were used to determine the most severe pipe rupture loading, and the SRSS method was '

used to combine seismic and pipe rupture loads.

The adequacy of the redesign was verified by comparing the analytic results with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III (Code) requirements. Vogtle Unit 2 Fas adopted the requirements of the 1977 Code Edition up to and including the Sunmer 1979 addendum (Subsection NB) and the Summer 1977 addendum (Subsection NF).

The applicant's submittal confinned that piping stresses did not exceed Code stress limits and loadings on nozzles did not exceed the allowables loading provided by the manufacturer in the equipment specifications.

Paul Munroe hydraulic snubbers were used for the upper supports. The service record of these snubbers showed that no service oriented failures have been reported. Because Vogtle Unit 2 will follow all maintenance requirements recommended by the manufacturer, high reliability of snubber performance is expected.

The applicant redesigned the steam generator upper supports for Vogtle Unit 2 by applying approved piping analysis methods and using qualified equipment.

Based on its review, the staff concludes that the results are acceptable.

Therefore, the staff approves the redesign and the proposed revised FSAR.

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z Mr. J. P. O'Reilly Georgia Power Company Vogtle Electric Generating Plant cc:

Mr. L. T. Gucwa Resident Inspector Chief Nuclear Engineer Nuclear Regulatory Comission Georgia Power Company P. O. Box.572 P.O. Box 4545 Waynesboro, Georgia 30830 Atlanta, Georgia 30302 Mr. Ruble A. Thomas Deppish Kirkland, III, Counsel Vice President - Licensing Office of the_ Consumers' Utility j Vogtle Project Council Georgia Power Company / Suite 225 Southern Company Services, Inc. 32 Peachtree Street, N.W.

P.O. Box 2625 Atlanta, Georgia 30303 Birmingham, Alabama 35202 James E. Joiner I Mr. Paul D. Rice Troutman, Sanders, Lockerman, Vice President & Project General Manager & Ashmore Georgia Power Company Candler Building j Post Office Box 299A, Route 2 127 Peachtree Street, N.E. 1 Waynesboro, Georgia 30830 Atlanta, Georgia 30303 -

Danny Feig Mr. J. A. Bailey 1130 Alta Avenue ',

Project Licensing Manager Atlanta, Georgia 30307 Southern Company Services, Inc. )

P.O. Box 2625 Carol Stangler Birmingham, Alabami 35202 Georgians Against Nuclear Energy  ;

425 Euclid Terrace i Ernest L. B?ake, Jr. Atlanta, Georgia 30307 Bruce W. Churchill, Esq.

Shaw, Pittman, Potts and Trowbridge 2300 N Street, N. W.

Washington, D. C. 20037 Mr. G. Bockhold, Jr.

Vogtle Plant Manager Georgia Power Company Route 2, Box 299-A Waynesboro, Georgia 30830 1

Regional Administrator, Region II i U.S. Nuclear Regulatory Comission )

101 Marietta Street, N.W. , Suite 2900 i Atlanta, Georgia 30323  ;

Mr. R. E. Conway Senior Vice President and Project Director '

Georgia Power Company Rt. 2, P. O. Box 299A '

Waynesboro, Georgia 30830

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