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AOCBSSION,NBR:8303040536 DOC ~ DATE: 83/03/01 NOTARIZED; NO DOCKET.FACIL:50~369 Nil-liam B,, McGuir e Nuclear 'Station~ Unit 1i Duke Powe 05000369

AUTH,NAME 'UTHOR AFFILIATION«TUGKERiH ~ BE Duke Power )Cod

>RGCIP ~ NAME RECIPIENT AFFILIATION,DENTON~ H,'R ~,Of fi ce„of Nuc1 ear Reactor Regul at i one Dir ec tor'ADENSAM~E~ G ~ Licensing Branch '4

SUBJEGT: Forwards addi info «pe mod to main feedwater lines «to steamgenerators D2/D3 ~to alleviate cer,tain forward;flushingitransients<per ARC 830218

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GHARLOTTE, N.G. 28849TELEPIIOWS

(704) 373&531

March 1, 1983

Mr. Harold R. Denton, DirectorOffice of Nuclear Reactor RegulationU. S. Nuclear Regulatory CommissionWashington, D. C. 20555

Attention: Ms. E. G. Adensam, ChiefLicensing Branch No. 4

Re: McGuire Nuclear Station, Unit 1

Docket No. 50-369D2/D3 Steam Generator Design Modification

Dear Mr. Denton:

My letter dated February 3, 1983 provided a report describing the program to beimplemented by Duke Power Company at McGuire Nuclear Station-Unit 1 to address,the recommendations made by the Design Review Panel in its report "UtilityDesign Review Panel, D2/D3 Steam Generator Design Modification", which wassubmitted on January 17, 1983. This program includes a modification to the mainfeedwater lines to the steam generator which is intended to alleviate certainforward flushing transients.

Ms. E. G. Adensam's (NRC/NRR) letter dated February 18, 1983 indicated that inorder for the NRC to complete its review of this modification additional infor-mation would be required, which was requested in the form of 10 items. Pleasefind attached the requested information.

Should there be any further questions in this matter, please advise.

Very truly yours,

Hal B. Tucker

PBN:jfwAttachment

cc: Mr. James P. O'Reilly, Regional AdministratorU. S. Nuclear Regulatory CommissionRegion II101 Marietta Street, Suite 2900Atlanta, Georgia 30303

Dr. M. W. WambsganssArgonne National Laboratories9700 South Cass Avenue, Bldg. 335Argonne, Illinois 60439

Senior Resident InspectorMcGuire Nuclear Station

8303040536 830301PDR ADOCK 05000369 i I

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McGUIRE NUCLEAR STATION, UNIT 1

D2/D3 STEAM GENERATOR DESIGN MODIFICATIONRESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

1. Q.

A.

Figure 2 attached to the utility's submittal dated February 3,1983 does not indicate the normal position of the valves.Provide a figure-which shows the normal position of the valvesduring startup (with the warmup 1 ine in operation) and a f igureshowing the valves in the normal full power configuration.

See Fig. 1 for valve positions during startup with the warmupline in operation.

See Fig. 2 for valve positions during normal full power operation.

2. Q. Provide .the seismic category, quality group and code classifica-tions of all new piping and v'alves.

A. The classification of all new piping and valves in -the Reactor Building(RB) and the doghouses (DH) is seismic Category 1, ASHE Section III,Class 2.

I

All new piping and va ives in the Turbine Building (TB) are non-seismic ANSI B31.1.

See FIg. 1. New piping and valves shown by dotted lines. Buildingboundaries also shown.

3- Q.Part 1

A.

Provide a discussion of the anticipated operating procedureusing the proposed warmup modification. Include the sequence ofvalve operations as part of the discussion.Refer to Fig. 1 for discussion of A steam generator (typical)Main feedwater flow is supplied to the auxiliary feedwater nozzlethrough valves CF121 (feedwater 'check valve), CF160 (feedwaterpreheated bypass check valve) and CF126B (feedwater preheaterbypass valve). Flow is regulated by the control bypass valveCF104AB below 154 power and by the control valve CF32AB otherwise.The feedwater isolation valve CF35AB is closed to prevent flowto the main nozzle. Valves CF134A, CF151B, and CF150 are closedand CF187 open to prevent flow .to the main nozzle through thewarmup line prior to initiating warming'."

"'o

initiate warming, valves CF134A and CF151B are opened topressurize piping. Pressurization rate and purge flow arecontrolled by orifice downstream of CF188. After opening abovevalves, valve CF187 is opened to initiate flow. Thermocouplesdownstream of valve CF35AB and upstream of CF34 are. monitoredto determine that temperatures are satisfactory to initiateforward feedwater flow through the main nozzle.

When temperature limits are met on all steam generators valveCF187 is closed. Valve CF134A is then'losed.

Valve CF150 is then opened to provide tempering flow to theauxiliary feedwater nozzles through valve CF151.

Forward feed flow can be initiated by opening feedwater isolationvalve CF35AB provided all feedwater temperature and reactorpower limits are met.

Valve CF126B is closed to isolate the feedwater preheater bypass.See Fig. 2.

3. Q.Part 2

Explain how the valve bypassing the containment isolation checkvalve on the main feedwater line will be controlled during thewarmup procedure and power operation.

A. See Fig. 1 for discussion of A steam generator.

CF34 is not the containment isolation check. Containment isolationis provided by check valve CF121. CF34 originally served as thestress analysis boundary between piping analyzed for steam gen-erator design temperature and piping analyzed for feedwater systemdesign temperature. Piping is currently analyzed for steam generatordesign temperature back to isolation valves CF35AB, CF175 and CF187.Therefore an open bypass around CF34 does not adversely affect anycurrent safety analysis.

Identify which valves are operable from the contro'I room, fromthe remote shutdown., panel, and from both the control room andthe remote shutdown panel. For those valves which are operablefrom,.both locations, verify that transfer switches will beinstalled to isolate the valves'from the control room in theevent of a fire in the control room. Also discuss the positionindication of the above valves.

All motor operated valves in the warmup, flow path are controlledfrom. and have position indication in the control. room only. Thewarmup line is a potential leak path from spurious valve opera-tion due to a control room or cable spreading room fire; however,flow is limited by orifices to approximately 40 gpm per steamgenerator which when combined with previously analyzed leakagedue to fires does not hinder the ability of the auxiliary feed-water pump to supply adequate flow for heat removal.

5. Q. Chapter 15 of the FSAR discusses the design basis accidentsand how they would affect the McGuire Nuclear Station Unit 1.For each design basis accident for which the additional waterloss due to the proposed containment isolation check valvebypass valve being open could have an effect, provide theresults of revised analysis to verify that the FSAR analysisbounds the case in.which this valve. is open. The valve couldbe open either intentionally or assumed to fail in the open

. position.

A. Refer to Fig. 1 for discussion of steam generator A. Seequestion 3 for discussion of check valve. CF34 bypass. Valves CF134Aand CF151B receive auxiliary feedwater pump automatic startsignals from the corresponding trains to isolate, thus preventingwater loss beyond design basis accident analyses. Therefore,check valve CF34 bypass has no adverse affect on any safetyanalysis.

6. Q. Provide a discussion of the. procedures to test. and inspect thenew valves and piping in order to meet the requirements of GeneralDesign Criteria 45 and 46.

A. No additional active valves were added. as a result of thismodification. Existing inserv.ice inspection procedures willtherefore be used to verify opera'bility of these valves andsafety related piping added. .Hanuai valves added will beverified in the valve check list in the operating procedures.

Eight (two per steam generator) active valves in the warmup flowpath receive additional engineered safety feature signals toclose on automatic start of the corresponding train motor drivenauxiliary feedwater pumps. These valves will require additionaltesting per Technical Specifications for response times for thesignal added.

7. Q. identify which piping and valves will be provided with tornadomissile and tornado protection. for those components where thisprotection is not provided; verify that the failure of thesecomponents concurrent with the loss of offsite power and themost limiting single failure will not result in unacceptableconsequences.

A. All piping and valves inside the Doghouses and Reactor Buildingare protected against tornado missiles and tornadoes. Pipingand valves outside doghouses and Reactor Building are not pro-tected. See Fig. 1 for building boundaries. Loss of theseunprotected components concurrent with the loss of offsite powerand the most limiting single failure will not result in un-acceptable consequences.

8., Q. Since the piping valves will be subjected to a temperature higher than 200 FPart 1 or a pressure greater than 275 psig, provide the results of a high energy pipe

break analysis in accordance with the Standard Review Plan Section 3.6.1.Provide drawings which show the piping configurations, postulated breaklocations, the jet cones which result fran the escaping fluid, and the'ocationof the pipe whip restraints. For those portions of the proposed system whichdo not have a temperature higher than 200 F and do not have a pressure greaterthan 275 psig, provide the results of the moderate energy pipe crack analysisin accordance with the acceptance criteria of Standard Review Plan Section 3.6.1.

A. The piping and valves associated with the Steam C~er8tor modification shall besubjected to approximate operating temperatures of 441 - 556 F, an approximateoperating pressures of 1105 — 1052 psia. As such this piping is high energy andis analyzed in accordance with the requirenetns of the Standard Review PlanSection 3.6.1. The main and auxiliary feechmter systen design parametersassociated with the Steam (aerator changes are delineated in the McGuireNuclear Station modification (NSM) No. 1100.

The modifications to the feedwater system piping have resulted in seven (7) newint~iate break locations on Duke Class B f~ter piping in the outbiddoghouse. There are no changes- in intermediate break locations on feedwaterpiping in the irdxxu".d doghouse. An initiallydered intermediate breaklocation on &abater piping in the inboard doghouse was not changed, since atthe new second highest stress location none of the following conditions exist:

1. Maximum stress range exceeds the threshold level of 0.8 (1.2 S + S )H A

2. A change is reauired in pipe parameters, such as major differencesin pipe size or wall thickness.

3. The ncw highest stress location is 0.4'1.2 S + S ) < S <0.6 ( 1.2 SH + S ), and the stress at the new stress locationis ) 20% higher @an the original break location stress, andresul.ts in unacceptable consequences to safety related systems.

4

Table-1 identifies the seven (7) new break locations/designations and the new

protective device requirements resulting fran the Steam Cinerator modification.

Figures CAP-A-1 thru 4 and CAO-D-1 & 2 provide the jet cones which result franthe escaping fluid issuing frcm the new break locations requiring protective ~

devices.

The location of the new pipe whip restraints which are required by the SteamGenerator modification are shown on isanetric Math Nodels CAP (sh. 2 of 3) andCAO (sh. 3 of 3).

8. q.Part 2

A.

As part of. the pipe failure analysis, include a discussion ofthe means to isolate the pipe break/crack, a discussion of poten-tial flooding including the means to remove the water,. and verifi-cation of the capacity of the means to remove the water, and adiscussion of the affects of wetting nearby equipment due towater impingement, splashing or dripping.

Refer to Fig. 1 for a discussion of Steam Generator A.

All new break locations resulting. from this modification arelocated in the doghouses. The doughouses are provided withsafety related level switches which generate a feedwater isola-tion signal on high level to isolate forward feedwater flow intothe doghouses. Valves CF151B and CF134A receive a feedwaterisolation signal to close as do the feedwater isolation andcontrol and preheater bypass isolation valves. The doghousesand safety related equipment. have been analyzed for the floodlevel and impingement resultin'g from a double ended break in thelargest line with complete loss of steam generator inventory andcontinued auxiliary feedwater addition for 30 minutes. Allinventory from this break is contained in the doghouse below thelevel of safety related equipment.

9. g. Verify that the new valves will not produce internally generatedmissiles, or verify that safety-related equipment is providedprotection from these missiles, in accordance with the StandardReview Plan Section 3.5. 1. 1 and 3.5. 1.2. For each internallygenerated missile, verify that no secondary missiles will begenerated.

A. Valves are not considered credible missile sources, as statedin FSAR Subsection 3.5.2.4.

10. Q. Verify that no sources of internally generated missiles are locatedin the same compartment as. the proposed piping and valves. I fthis cannot be verified then either. provide protection from themissile or verify the failure, of the new component from the missile'illnot have any adverse effect on the plant.

A. We have reviewed the effect of cr'edible missiles on the new feed-water piping and found the piping is shielded from all crediblemissiles listed in section 3.5.2,of the FSAR. The results ofthis review are listed below.

1) Turbine Building

No safety related piping was added in the Turbine Building.

2) Doghouses

No missiles from section 3.5.2 wil3 affect piping in thedoghouse with the exception of valves, subsection 3.5.2.4.For the same reasons stated in this subsection, valve failuresshould not be considered credible missiles. The new piping's shielded from all other credible missiles listed in 3.5.2by the reactor building wall and the doghouse wall.

3) Reactor Building

No credib'le missiles from section 3.5.2 can affect the newpiping in the Reactor Building. The new piping is protectedby the reactor building wall and crane wall from externalmissiles and by the steam generators, feedwater piping andreactor vessel shield wall from internal missiles.

AUX FDW

Fig. 1 Startup Valve Positions (with warmup line in operation)I

AUX FDW DH RF

AB

DH

AB

DH

A CA65

AB

DH

AB

DH

CA61

STM GEN B

TEMPFLOW

STM GEN C

TEMP

FLOW

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DH RF

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CA CF.

STM GEN D

TEMPFLOW

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CHEMICALADDITION

SYSTEM

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Fig. 2 Normal Valve Positions - Full Power

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CHEMICALADDITION

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TABLE-1

HICK ENERGY BREAK IDCATIONS

BREAK NO. SYSTEM DESCRIPTION STRESSRATIO IQCATION PBOZECTION

RHQUIKZKNl'S

CAl-118

CAlM50B CA

CAlM48B CA

CAlM48A CA

2" CF line at the outletof 4" x 2" reducer upstreamof valve 1CF 134A.

2" CF line at inlet oftluxd elbow upstream ofvalve lCF 134A.

2" CF line at inlet offirst elbow upstream of-valve 1CF 134A.

2" CF line at the outlet offirst elbcer upstream ofvalve 1CF 134A.

1.033

1.016

1.095

1.138

CAP-A-1

CAP-A-2

CAP-A-3

CPP-A-4

AB-750-26

AB-750-26

AB-750-26

AB-750-26

None for forwardflow (FF)RR for reverseflow (RF)

None for FFRR for RF

RRs for FFNone for RF

RR for FFRR for RF

CA1-C58A CA 2" CF line at the inlet offirst 90o e33xw upstream ofvalve 1CF 178.

0.860 AB-750-26 RR for FFRR for RF

CAlM58B CA

CAlM48B CA

2" CF line at the outlet offirst 90o elbow upstream ofvalve 1CF 178.

6" CF line at the inlet offirst elbow upstream ofvalve 1CF 163

0.860

0.821

CAO-D-2

N/A

AB-750-26

AB-750-26

RR for FFNone for RF

None for FFNone for RF

Note: RR —Rupture Restraint.

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