ML20115G077

Forwards Proposed Methodology for Analysis of Primary Coolant Loops at McGuire Nuclear Station Units 1 & 2 & Catawba Unit 1 for Steam Generator Replacement

ML20115G077
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	10/15/1992
From:	Tucker H DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9210260015
Download: ML20115G077 (7)

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5 Duke ikuer Company list B Tlwn PO But IIKi6 Schio! Iit!l'inidrul Charlatte. NC N01 Itg4 kclear Generation (70ImMDI DUKE POWER October 15,1992 U. S. Nuclear Regulatory Commission ATTN: Document Centrol Desk Washington, D. C. 20555

Subject:

McGuire Nuclear Station Units 1 and 2 Docket Nos. 50-369, 370 Catawba Nuclear Station Unit 1 Docket Nos. 50-413 Methodology for Analysis of the stimary Coolant Loops for Steam Generator Replacement Gentlemen:

Pursuant to 10CFR 50.4, attached is the proposed mtthodology for the analysis d the primary coolant loops at McGuire Nuclear Station Units 1 & 2 and Catawba Unit I for steam generator replacement. Since the B & W replacement steam generators have a slightly higher center of gravity and a greater mass than the original Westinghouse steam generators, the NSSS primary coclant loops must be reanalyzed.

The reanalysis effort consists of a parametric study which will show that the original design basis analysis of the reactor coolant system is conservative and will therefore remain valid after the replacement steam generators are installed. Since the premsed methodology has not been previously used by Duke Power Company, it is requested that the NRC review and indicate acceptance of the proposal inclu&d in Atiachruent 1, in suppret of the current steam genuator replacement schedule by December 1,1992.

Shotild there be any questions concerning this proposal or if additional information is required, please contact Davia V. Ethington at (704) 382-6633.-

Very truly yours,

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11. B. Tucker, Senior Vice President Nuclear Generation 9210260015 921015

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- Attachments xc (w/att):

S. D. Ebneter Regional Administrator, Region II T. A. Reed, ONRR R. E. Martin, ONRR P. K. VanDoorn Senior Resident inspector (MNS)

W. T. Orders Senior Resident inspector (CNS)

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i Doku Powr Company Methodology for Analysis of the Friinary Coolant Loops at McGuire Nuclear Station Units 1&2 and Catawba Unit 1 for Steam Generator Replacement jntLoducilon The existing Westinghouse model D2/D3 steam generators installed in McGuire Nuclear Station Units 1 & 2 and Catawba Unit 1 are sufforirig from tubo degradation probloms. Duko Power has decided that the most cost offective solution to thoso problems is to replace the steam generators in each unit. Accordingly, an order has boon placed with B&W International for twelvo replacement steam generators. The B&W replacement steam generators have a r lightly highor contor of gravity and a greator mass than the original Westinghouse steam generatoro. Those changes require that the NSSS primary coolant loops be roanalyzed. The roanalysis effort consists of a parametric study of tho roactor coolant system responso before and attor steam generator replacement. The parametric study will be performed by Babcock &

Wilcox Nuclear Services Inc. The intent of this paramotric study is to show that the original design t' asis analysis of the reactor coolant loop is conservativo and therefore is still valid. A new design basis will not be created for the reactor coolant loop.

ficacMLC001a0LLQ0panalysis The loading analysis of the primary coolant loop will be divided into two phases with each using a different loop me 'ol. The first phase involves constructing a loop model with the original steam generator similar to that used by Westinghouse. Gravity and seismic loads will be applied to this model consistent with those used by Westinghouse.

The stiffnesses of the component suppods provided to Westinghouse for the original loop analysis will be used in this model. Sdsmic excitation will be provided by the floor responso spectra at the various elevations at v;hich the NSSS component suppods are attached to the Reactor Building Interior Structura. Structural damping in the NSSS math model will be the same as that used by Westinghouse in the original analysis.

The analytical results from this model will be used to benchmark" the analysis assumptions and input data by comparison of the output results with the original Westinghouse results. The intent of this analysis is to show that tho structural proporties of the reactor coolant loop model (mass, stiffness, and boundary conditions) and the analytical techniques applied to it will glvo results comparabl3 to those provided by Westinghouse before the next phase of the NSSS analysis. Exact duplication should not be expected due to differences in modeling techniques, arialytical methods, computer codes, etc. The loop model will be checked for only one of the throo units which will undergo stonm generater replacement. This will be sufficient to validate the analysis approach as the three reactor coolant systems are almost (but not exactly)

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l The socend ph ;se of the analysis involves modification of the math model constructed

'in phase one to incorporato the replacement steam generators in this model, more advanced analytical techniques will be used. The NSSS primary coolant loop will be linked to tho structural model of the Reactor Building Interior Structure using aprings which represent ino individual component s ipports. Solsmic excitation will be provided by the top of basemat responso spectra rather than from the floor rosponso spoetra at tho various olovations at which the component supports are attached to the Interior Stnicture. This will reduce the solsmic input through the elimination of the effects of spoetra broadoning in floor responso spectra. ASME Code Case N-411 1, "Alternativo Damping Values for Responso Spectra Analyals of Class 1,2 and 3 Piping" will be used as allowed by Rogu!atory Guido 1.84 to reduce the solsmic input into the primary coolant loop even fudhor. This code caso was previously ovaluated for McGuire and Catawba and is included in the plant FSARs (Section 3.7.1.3) as acceptablo alternativo damping values. Leady stato thormal loads and gravity loads reflecting the new steam gonorator will also be considered.

Tho analysis of the phase two model will be performed using all applicablo loadings and load combinations. New displacements forces and moments, piping stresses, and responsa spectra will be calculated. The static and solsmic analysis results will be compared to those from the phase one model with the original Westinghouse steam generators (referrod to heroin as the "baselino analysis"), in order to show that the pipe stresses, component loads and component support loads are still valid. If the results of the loop piping containing the replacement steam generators are less than that of the baseline analysis modo! then the stress repcrts need only be updated to reforonce the Bt NS calculations. If the results of the loop piping containing the replacement steam generators are groater than those in the baseline analysis then the original Westinghouso design reports will be checked and upgradod as nocessary. -

A flow chart which shows the basic stops that will be followed in the reactor coolant syster.1 analysis is included as Attachment 2.

Pipo rupturo loads due to a double ended guillotine break in the primary loop are to be oliminated due to the uso of leak-bofore break criteria which was previously approved by the NRC for the reactor coolant systems at both McGuire and Catawba. Pipo rupture loads due to breaks on the residual heat removat accumulator, and pressurizer surgo ilnos will still be considered on the model, as ;.d as loads due to a break in the main feedwater, main steam, and any other applicable secondary side systems.

The pipo rupture analysis results for the component supports will b3 compared to the original component support loads. If the now loads are obviously bounded by the original results then the original stress reports will be considered valid. It is considered likely that this will be the caso due to the margins introduced by the use of leak-before-break critoria. If the component support loads calculated by BWNS exceed the original support loads, the component sirosses will be shown to satlafy the requireme'its of the Design Specification, ASME Code, and FSAR and olthor now or revised stress reports will be issued. A final check will be mado to verify that the analyses supporting the use of leak before break criteria are still valid for the re.lsed loop analysis results.

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e McGolre and Catawba are similar plants but they are not identical. Thorofore the

'modeling and analysis approaches described above must be performed for occh plant The basic techniques to be used in the analysis are consistont with thoso afroady i

describod in the FSAR for each plant. Some changes may be required due to the use of different computer programs, etc. The FSARs will be updated to reflect these and any other changos.

BeactoLCoojanLLoop30mponents The replacement steam gonorators will be analyzod and designed by B&W to moot the requirements of the Design Specification, the ASME Codo, and the l'SAR.

The reactor vessel and reactor coolant purnps will be reviewed in a manner similar to the reactor coolant loop piping. The basolino (phase one piping math model described abovo) analysis nozzlo loads on each ploco of equipment will be compared with the nozzio. ads from the original Westinghouse primary _ loop analysis to ensure that the baseline analysis is valid. The nozzle loads from the analysis containing the rop'acoment steam generators are then compared to the nozzio loads from the basolino analysis. If the nozzle loads from the analysis containing the replacement steam generators are loss than the baseline analysis no:zlo loads then the oxisting stress reports will be updated to reference the BWNS calculations and no further work will be required, if the nozzlo loads from the analycis containing the replacement steam generators aro greator than the basolino analysis nozzio loads then the original Westinghouso stress reports will be checked and updated as necessary.

All existing reactor coolant loop components will be reviewed to determine the impact of the now thormal transients associated with the replacement steam generators. A fatigue ovaluation will bo included in this revlow if necessary, based on a comparison bo+ ween the now and original design transients, ficantoLCoolanLLoopA9mponenLSuppolts The loads from the BWNS analysis of _the reactor coolant loop on the component supports will be compared to the Duke Power design loads for each support. if the now-loads are less than the design loads the calculations'will ot ly bo updated to reference the BWNS calculations. The supports will be analyzed to evaluato any loads which increase beyond the design loads. Any necessary support modifications will be made.-

BuactotBuilding_Sttuntute If the loads from the BWNS analysis of the reactor coolant loop on any of the component suppsrts increase beyond the design loads for that support then the support will be roanalyzed as stated above. The embedmont loads and building structure loads

-obtained from the support reanalysis will be compared to *.he design loads for the ombodmont and structure. If the now loads are less than the design loads the calculations will only be updated to reference the BWNS calculations. Any of the new loads on the NSSS support embedmonts and building structures which increase Pago 3 t"

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beyorfd the design loads will require that the embodment and building structure be ovaluated. No work will be necessary otherwise.

. i The increased welght of the replacement steam generators will be incorporated into the salsmic analysis of the interior structure for each plant, Now Reactor Building floor responso spectra will be generated uslag methods described in section 3.7 of the plant FSARs. These spectra, along with the associated frequencies and moda, shapos, will then be compared to tho design frequenclos, modo shapes and spectra to ensure no significant changos have occurred in the building responso, i

BIanchPJplugilnes The displacements, rotations and responso spectra at branch lino nozzle locations on the reactor coolant loop and components from the BWNS analysos will be compared to those movements and spectra used for the des,a of the branch piping systems. If the movements or spectra increase then the branch lines will be evaluated and reanalyzed as riocessary; otherwise, no work will be required. Any branch lines off of the reactor coolant loop or secondary sido lines which are rorouted to accommodate the steam gonorator replacoment will be reanalyzod.

CORolusion A paramotric analysis of the reactor coolant piping at McGuire Nuclear Stat' ion Units 1 &

2 and Catawba Nuclecr Station Unit 1 provides adequato assurance that the design of the primary coolant loop, major components, component supports, branch piping F/stGms Lnd the building structuros remain bounded by the origir.al plant design bases after the steam generator replacement.

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NSSS Primary Loop Analysis Methodology Stan V

" Baseline" Modoi Analyzo Loop Model m

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Compare to@Rosults Not OK OK V

Analyze Loop Model w/ Replacement SG V

Compara to " Baseline" Model Results Not OK V

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Stress Reports Update Stress Reports 4

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