ML17261A610

Notifies That Util Intends to Pursue Steam Generator Snubber Replacement Program.Objective of Program to Replace Six of Eight Hydraulic Snubbers During Feb 1988 Refueling Outage. Brief Technical Description of Analysis Encl

ML17261A610
Person / Time
Site:	Ginna
Issue date:	09/22/1987
From:	Kober R ROCHESTER GAS & ELECTRIC CORP.
To:	Stahle C NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NUDOCS 8709290085
Download: ML17261A610 (20)

Text

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ACCESSION NBR: 8709290085 DOC. DATE: 87/09/22 NOTARIZED: NO DOCKET 0 FACIL: 5C-244 Robert Emmet Qinna Nuclear Planti Unit ii Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION KOBERe R. W. Rochester Gas Rc Electric Corp.

RECIP. NAME RECIPIENT AFFILIATION STAHLEa C. Document Control Branch (Document Control Desk)

SUBJECT:

Notifies that util intends to pursue steam generator snubber replacement program. Obgective of. program to replace six of-eight hydraulic snubbers during Feb 1'F88 refueling outage.

Brief- technical description of. analysis encl.

DISTRIBUTION CODE: AOOID COPIEB RECEIVED: LTR J. ENCL J SIZE: l I TITLE: OR Submittal: General Distribution NOTES: License Exp date in accordance with 10CFR2i 2. 109(9/19/72). 05000244 REC IP IENT COPIES RECIP IENT COP IES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 LA 1 0 PD1-3 PD 5 5 STAHLEi C 1 1 INTERNAL'RM/DAF/LFMB 1 0 NRR/DEBT/ADS 1 NRR/DEST/CEB 1 NRR/DEST/MTB 1 1 NRR/DEST/RSB 1 1 NRR/DOEA/TSB 1 I

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'iutggi fizzzez i ~IS <<iuVVZZ ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.y. 1464g-pppI ROGER W, K08ER TEI.CPHONC VICe PNeaDeNT ELECTRIC PRODUCTION AIICA CODE 'llO 546-2700 September 22 I -1987 U.S. Nuclear Regulatory Commission Document Control Desk O. C/l Attn: I1r. Carl Stahle PWR Project Directorate No. 1 Washington> D.C. '0555 QP n

CA

Subject:

Steam Generator Snubber Replacement Program R. E. Ginna Nuclear Power Plant CA Docket No. 50-244

Dear Nr. Stahle:

Rochester Gas and Electric intends to pursue a steam generator snubber replacement program. The objective of this program is to replace six of the eight hydraulic snubbers per steam generator with rigid structural members during the refueling outage beginning in February 1988. This replacement will reduce required maintenance activities and aid in keeping radiation exposures as low as reasonably achievable (ALARA).

Enclosed is a brief technical description of the analysis that will be performed to provide the justification for the snubber replacement. This analysis will involve the application of both updated pipe break design criteria to reduce maximum loads on the reactor coolant loop support systemI and the change to NEB BTP 3-1 (Generic Letter 87-11) to eliminate arbitrary intermediate breaks in the main steam lines.

Consistent with your discussions with members of the RGGE staff I it is our intention to perform this replacement as a plant modification subject to review per 10CFR50.59. There are no changes to the Technical Specifications required beyond those previously submitted on August li 1983 (last updated July 24I 1987) which retains operability and surveillance requirements for snubbers but which removes the lists of snubbers from the Technical Specifications. We anticipate that approval of that proposed change to the Technical Specifications will occur by the 1988 outage.

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We are planning to meet with the NRC Staff on September 28 f 1987 to discuss this matter further. We request that a written indication of your concurrence with this application of 10CFR50.59 be received no later than October 23<,1987 as changes or additional reviews beyond this date could prevent',the replacement from occur-ring during the 1988 outage. If no written indication is received by this date> we will assume your concurrence and proceed with the program as planned.

Very truly yours>

Roger W. Kober Enclosure

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ROCHESTER GAS & ELECTRIC CORPORATION STEAM GENERATOR HYDRAULIC SNUBBER REPLACEMENT PROGRAM TECHNICAL

SUMMARY

1.0 Introduction The upper portion of each of the two steam generators at Ginna Station are currently restrained against lateral seismic and pipe break loads by eightI large (532<000 lb.

capacity) hydraulic shock arrestors (snubbers) as shown in Figure l. An independent set of supports provide lateral and vertical restraint to the lower portion of each SG.

The required maintenance, in-service inspection and testing are performed during annual refueling outages. Surveillance activities are performed periodically throughout the year.

2 0 ~Pto sam Descti~tios The intent of the proposed upper lateral support modifi-cation is to replace six of the eight hydraulic snubbers per SG with rigid structural members (bumpers) I thereby minimizing the number of hydraulic snubbers in service for this application. In this wayI annual maintenance activities andI consequently> ALARA radiation exposures to maintenance personnel can also be minimized. The hydraulic snubbers replaced with bumpers will be refurbishedI and stored for use as spares. It is expected that spare parts procurementr as well as utilization of shop facilities and rigging equipment> can be optimized as a result of this replacement. A pair of existing snubbers will remain in place at each SG in the direction of reactor coolant loop (RCL) thermal growth as shown in Figure 3. This arrange-ment provides sufficient design load capacity and represents a reliable support configuration. The design of the existing hydraulic snubbers (manufactured by Anker-Holth) employs a passive-type orifice design. Control valves are not used andI therefore> the failure of such valves (the predominant failure mode of other large-bore snubbers as discussed in IE Bulletin 86-102) is not applicable. The rigid structural members (bumpers) which will replace some of the hydraulic snubbers will be equallyI if not morer reliable.

3 0 The SG hydraulic snubber replacement program will assure that adequate support capacity is maintained with respect to the design basis loads. The current controlling design load for the SG upper lateral support system is an intermediate pipe break in the horizontal main steam line near the top of the SG (See Figure 2). NRC Generic Letter 87-11I "Relaxation

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in Avbitrary Intermediate Pipe Rupture Requirements" I provides guidance for elimination of arbitrary interme-diate breaks and will be applied to this program. Previous piping upgrade program analyses (recently reviewed in NRC Inspection 87-11) I using ANSI B31.1 criteria> have been reviewed and will be revised as necessary to reflect changes resulting f rom this snubber replacement program.

Consistent with Generic Letter 87-11I these analyses have established that no intermediate pipe breaks need to be postulated in the Main Steam (NS) or Feedwater (FW) piping.

Terminal-end breaks at the FW inlet nozzles are now the limiting load and are the new design basis loads.

4 0 Pi in S stems Anal sis The effect of the new design basis loads upon the RCS equipment and piping support system are being analyzed by Westinghouse. An independent review by a consultant with broad experience in RCS support design is also planned.

The use of rigid structural members (bumpers) in the SG upper lateral support system will change the degree of stiffness with which the SGs are restrained against dynamic loads (See Figures 4 and 5). These new stiffnesses have been calculated and results so far indicate that RCS stresses and deflections will not change significantly.

Loads from a pipe break postulated to occur in an auxiliary line (RHRI SI accumulator or surge line) branch connection will also be determined using the new upper lateral support stiffnesses to assess the effect on the reactor coolant loop piping, equipment supports and the new SG upper support seismic response spectra and damping values used in configuration.'.0 The this work will conform to Regulatory Guide 1.60 and 1.61I respectively. Nodal responses and spatial components will be combined in accordance with Regulatory Guide 1.92.

Results of the existing RCS Leak Before Break (LBB) analysis will be reviewed (resolution of the Asymetric Loads issue which employed LBB is documented in an NRC letter dated September 9I 1986).

Ri id Structural Members (Bum ers)

The replacement support hardware consists of individual structural assemblies which will be installed wherever an existing hydraulic snubber is removed. A typical assembly is shown in Figure 6. Each assembly is structurally rigid under compression but will allow freedom of movement in the tensile direction. Each assembly is individually adjust-able in the field to ensure that clearances at each bumper position are adequate for expansion but do not exceed those permitted by the analysis. The assembly and its subparts

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are sized< and analyzed> to withstand the new Design Basis Load. The assemblies will be designed and constructed to the requirements of the American Institute of Steel Con-struction (AISC)> which was the industry code used for the original plant major component supports. Detailed design of the rigid structural members is being performed by RGGE.

Fabrication will be performed by a qualified supplier.

Load combinations will be combined in accordance with the Standard Review Plan Section 3.9.3 Appendix A and NUREG-0484.

6 0 Licensin Activities No revisions to the Ginna Technical Specifications are comtemplated. Previous applications have been made to remove the specific listing of safety-related snubbers from Technical Specifications. Operability and surveillance requirements for remaining snubbers have not been changed.

A safety evaluation of the modification in accordance with 10CFR50.59 will be performed. The Ginna Station Updated Safety Analysis Report (UFSAR) will be revised as appropriate.

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