ML20237B165

Safety Evaluation Supporting Elimination of Primary Component Support Snubbers from Primary Coolant Loops

ML20237B165
Person / Time
Site:	Beaver Valley
Issue date:	10/21/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20237B138	List: ML20237B165
References
NUDOCS 8712160068
Download: ML20237B165 (6)

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- Enclosure 2 ,

i UNITED STATES

[%p@'%,% g NUCLEAR REGULATORY COMMISSION Es -f WASHINGTON. D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i

ELIMINATION OF PRIMARY COMPONENT SUPPORT SNUBBERS BEAVER VALLEY POWER STATION-UNIT 1 DOCKET NO. 50-334 INTRODUCTION j The application of the leak-before-break (LBB) approach to prevent ruptures of the primary coolant loop piping can eliminate the requirement to design the loop piping for these previously postulated pipe rupture events. If accepted, this approach will permit elimination of certain selected primary component support snubbers which principally carry pipe rupture loads. Because large-bore snubbers are located in the reactor containment cubicles, their deletion will eliminate a significant source of occupational exposure caused by periodic 1

functional testing and maintenance.

Support system reliability will also be improved with the removal of these active component supports. Inadvertent lockup, bleed rate variance, and hydraulic

~ fluid leakage are possible large-bore snubber problems that can be eliminated.

The licensee, in its submittal of June 1, 1987, made a request to a licensing amendment *for Beaver Valley Power Station Unit 1. The amendment would permit the elimination of the following 24 large bore (12 in. I.D.) snubbers:

1. Both snubbers acting parallel to the hot leg in each of the three lower steam generator (SG) supports.

• The licensee's proposed action is a simple removal of pipe whip restraints and jet impingement barriers . Therefore, in accordance with the staff's discussion in 51 FR 12502, no amenchent to the license is needed.

8712160068 871209 PDR ADOCK 05000334 P PDR

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2. Both snubbers acting parallel to the cold leg in each of the three reactor coolantpump(RCP) supports.

3. Four snubbers acting between the RCP support and SG lower support in each cubicle.

Also, during this modification, the two snubbers located at each SG upper support and acting in a direction perpendicular to the hot leg will be replaced by rigid struts. The snubbers to be retained are the two 7,nubbers located at each of the SG upper supports and acting in a direction parallel to the hot leg.

This evaluation addresses the acceptability of the reactor coolant piping, components, and supports with the modified support configuration in withstanding all remaining loads, including those due to the design basis earthquake, with an acceptable margin of safety.

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- EVALUATION Two independent analyses of the primary RCS loop were performed by Stone & Webster Engineering Corporation (SWEC) and Westinghouse Electric Corporation.

The RCS with the revised support configuration was analyzed for the following conditions:

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• Deadweight l

• Internal pressure Thermal expansion Seismicevents(0BEandDBE)

Dynamic effects of postulated pipe ruptures in other systems as specified in the UFSAR (pressurizer surge, accumulator, residual heat removal, main 1

steam, and feedwater lines).

These loading conditions cover all UFSAR-specified loadings other than the dynamic effects of postulated pipe ruptures in the main reactor coolant piping, and are found acceptable to the staff.

For the seismic analysis, peak spread amplified response spectra (ARS) for 2 percent equipment damping (for both OBE and DBE) were used. This damping is within that permitted by ASME Code Case N-411 which was approved for use for Beaver Valley, Unit 1 l l . t

~ in the staff letter of April 8,1987 to the licensee. Three components of earthquake l

input motions, one vertical and two horizontal, were considered in the piping analysis.

The resultant responses were obtained by root-sum-square (SRSS) method. The com-bination of closely spaced modes also conformed to the requirement of Regulatory l Guide 1.92, Rev. 1. For the design of component supports, responses at support

-interfaces obtained from the piping analyses were evaluated in accordance with the UF3AR by combining the vertical response load absolutely with the larger of the two horizontal response loads. The staff finds the above methodology to be acceptable.

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In the analysis of RCS piping stresses, a comparison is made between the maximum stress in the reactor coolant loop piping for the existing and proposed support configurations for controlling load combinations. The results show that the stresses are comparable and are all well within allowables. l In the SWEC's evaluation of support loads on primary RCS components for the proposed support system configuration, it was found that the frequencies of most vibrational modes are virtually unchanged from those corresponding to the original configuration. The postulated terminal end and intermediate breaks in the pressurizer surge, residual heat removal, accumulator, main steam, and feed-water lines were reviewed by SWEC to determine those breaks which would cause the most severe loadings on the revised support configuration with snubbers removed.

These loads were combined with seismic DBE loads by SRSS. The resultants were then combined with deadweight and pressure loads by absolute sum. In all of the evaluated cases, the support loads were found to be within UFSAR and Code allowables.

~ It was also found that the proposed change to eliminate the lower snubbers has a negligible effect on the loads on the remaining upper steam generator snubbers and struts. Besides, the struts carry very small thermal loads because, as stated earlier, they are oriented perpendicular to the hot leg and the steam generator has insignificant thermal movement in this direction.

The staff has found the approach and results of the above evaluation performed by the licensee to be acceptable.

As mentioned previously, two independent analyses were performed by SWEC and' Westinghouse for a single primary loop. The results indicated that the RCS natural frequencies / modes and pipe stresses obtained from both analyses were in good agreement. Close agreement was also found in the support loads obtained.

Note that in both analyses the support configuration is essentially the same as the original configuration, except-for the removal of snubbers. The reactor coolant pump and lower steam generator support stiffness matrices, without snubbers, are essentially the same as originally used for the deadweight and thermal cases. Also, the stiffness matrix for the steam generator upper support reflects substituting two rigid struts for two snubbers.

The staff has found the above independent verification to be valuable in

~ providing added assurance of the adequacy of the proposed piping / support system when subject to the original design basis loadings, with the exception of those associated with large primary loop pipe rupture.

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g. 4 CONCLUSION Based on the above licensee's evaluation of the reactor coolant system with the. j proposed support configuration the staff concurs that the piping, components and supports are stressed within USFAR allowable limits under all the pertinent design basis loading conditions, excluding the loads associated with large pricary loop pipe rupture.

J Principal Contributor Arnold Lee Dated )

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