ML20058N494
| ML20058N494 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 12/15/1993 |
| From: | George Wunder Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9312220073 | |
| Download: ML20058N494 (121) | |
Text
{{#Wiki_filter:_ o-December 15, 1993 Docket No. 50-395 LICENSEE: South Carolina Electric & Gas Company j FACILITY: Virgil C. Summer Nuclear Station, Unit No.1
SUBJECT:
MEETING
SUMMARY
- STEAM GENERATOR REPLACEMENT
-1 On November 30, 1993, members of the s1.aff met with representatives of South ,
Carolina Electric & Gas Company (SCE&C or the licensee) in Rockville, Maryland to discuss the licensee's upcoming sceam generator replacement. The replacement is planned-for refuel:ng outage 8, scheduled to begin in September 1994. The steam generator replacement will result in a larger primary system; ' therefore, certain parameters associated with the primary and reactor protection systems will have to be changed, and certain Technical Specifications (TS) modified. The licensee discussed the TS changes and ' presented safety analyses supporting thase changes. With the exception of the TS changes, the licensee intends to make the replacement under the provisions , of 10 CFR 50.59. l The licensee concluded the presentation by describing their startup testing program. SCE&G also outlined plans for a future power uprate. A list of those in attendance is provided as Enclosure 1, and a copy of the licensee's handout is provided as Enclosure 2. Original Signed by: George F. Wunder, Project Manager l Project Directorate 11-1 : Division of Reactor Projects I/II " Office of Nuclear Reactor Regulation l
- 1. Attendance list
- 2. Meeting Handout -
OFC LA:PD2fdDRhEk) PM:PM14DRPE D:PD21:URP[ > NAME PAndeYs'on G d er:jrm SBa[wak DATE 12/ F /93 l 12//# /93 12/hh93 / /93 0FFICIAL RECORD COPY DOCUMENT NAME: G:\SUMNER\88172.MTS Mk P
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SUMMARY
OF 11/30/93: w/ Enclosures 1 & 2 ! Docket File ! NRC & Local PDRs j PD#11-1 Reading ; L. Plisco, ED0 E. Merschoff, Region II , G. Wunder , I w/ Enclosure 1 T. Murley/F Miraglia L. J. Callan, Acting S. Varga G. Lainas S. Bajwa P. Anderson 0GC E. Jordan R. Jones J. Blake F. Cantrell R. Lobel G. Hornseth N. Wagner R. Goel C. Gratton ACRS (10) cc: Licensee & Service List 1 I 4 l 200000
e December 15, 1993 Docket No. 50-395 LICENSEE: South Carolina Electric & Gas Company FACILITY: Virgil C. Summer Nuclear Station, Unit No. 1
SUBJECT:
MEETING
SUMMARY
- STEAM GENERATOR REPLACEMENT On November 30, 1993, members of the staff met with representatives of South Carolina Electric & Gas Company (SCE&G or the licensee) in Rockville, Maryland to discuss the licensee's upcoming steam generator replacement. The replacement is planned for refueling outage 8, scheduled to begin in September 1994.
The steam generator replacement will result in a larger primary system; therefore, certain parameters associated with the primary and reactor protection systems will have to be changed, and certain Technical Specifications (TS) modified. The licensee discussed the TS changes and presented safety analyses supporting these changes. With the exception of the TS changes, the licensee intends to make the replacement under the provisions < of 10 CFR 50.59. The licensee concluded the presentation by describing their startup testing program. SCE&G also outlined plans for a future power uprate. A list of those in attendance is provided as Enclosure 1, and a copy of the , licensee's handout is provided as Enclosure 2. Original Signed by: George F. Wunder, Project Manager Project Directorate 11-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosures:
- 1. Attendance list
- 2. Meeting Handout OFC LA:PdkDRPd PM:PM)d)RPE D:PD21:UR,P[
NAME PAndeYso b Gker:jrm SBahak DATE 12/ If /93 12//# /93 12/h[93 / /93 0FFICIAL RECORD COPY DOCUMENT NAME: G:\ SUMMER \88172.MTS
1 pn nec u ~ + h UNITED STATES YWg- j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 0001 f December 15, 1993 Docket No. 50-395 LICENSEE: South Carolina Electric & Gas Company FACILITY: Virgil C. Sumer Nuclear Station, Unit No.1
SUBJECT:
MEETING
SUMMARY
- STEAM GENERATOR REPLACEMENT On November 30, 1993, members of the staff met with representatives of South Carolina Electric & Gas Company (SCE&G or the licensee) in Rockville, Maryland to discuss the licensee's upcoming steam generator replacement. The replacement is planned for refueling outage 8, scheduled to begin in Septembe:-
1994. The steam generator replacement will result in a larger primary system; therefore, certain parameters associated with the primary and reactor protection systems will have to be changed, and certain Technical Specifications (TS) modified. The licensee discussed the TS changes and presented safety analyses supporting these changes. With the exception of the , TS changes, the licensee intends to make the replacement under the provisions of 10 CFR 50.59. The licensee concluded the presentation by describing their startup testing program. SCE&G also outlined plans for a future power uprate. A list of those in attendance is provided as Enclosure 1, and a copy of the ' licensee's handout is provided as Enclosure 2. h W / George F. Wunder, Project Manager Project Directorate 11-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosures:
- 1. Attendance list 2.' Meeting Handout I
Virgil C. Summer Nuclear Station CC* Mr. R. J. White Nuclear Coordinator S.C. Public Service Authority c/o Virgil C. Summer Nuclear Station Post Office Box 88, Mail Code 802 Jenkinsville, South Carolina 29065 J. B. Knotts, Jr., Esquire Winston & Strawn Law Firm 1400 L Street, N.W. Washington, D. C. 20005-3502 Resident Inspector / Summer NPS c/o U.S. Nuclear Regulatory Commission Route 1, Box 64 Jenkinsville, South Carolina 29065 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta St., N.W., Ste. 2900 Atlanta, Georgia 30323 Chairman, Fairfield County Council Drawer 60 Winnsboro, South Carolina 29180 Mr. Heyward G. Shealy, Chief Bureau of Radiological Health South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 ! Mr. R. M. Fowlkes, Manager Nuclear Licensing & Operating Experience South Carolina Electric & Gas Company ' Virgil C. Summer Nuclear Station Post Office Box 88 , Jenkinsville, South Carolina 29065 Mr., John L. Skolds, Vice President Nuclear Operations , South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station - Post Office Box 88 Jenkinsville, South Carolina 29065
1
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ENCLOSURE 1 NOVEMBER 30. 1993 j STEAM GENERATOR REPLACEMENT I MEETING ATTENDANCE STAFF SCE&G G. Lainas R. Clary R. Jones L. Cartin S. Bajwa J. LaBorde , J. Blake M. Miltner F. Cantrell A. Rice R. Lobel G. Hornseth N. Wagner R. Goel C. Gratton G. Wunder OTHER B. Carrick RG&E B. Flynn RG&E , J. Smith RG&E ' C. Robinson Duke Power M. Hazeltine Duke Power J. Torre Duke Power D. Ethington Duke Power R. Borsum BWNT R. Beck Bechtel L. Zerr STS
Enclosure 2 ; RF-8 l
- 1994 .
i l V C Summer SGR Project l SCE&G i l Presentation to NRC November 30,1993 Ron Clary
~
l SCE&G Attendees: j Jamie LaBorde
. April Rice
- Mark Miltner i
Lou Cartin i 1
. j V. C. Summer Nuclear Station Steam Generator Replacement Project
[$2P!"?"b?tf%%1l@!!!Iffj$"1$sf @$Ei a I 9 Introduction Ron Clary l Basic Facts ! Organization / Management j History of the Project l Overview of Project Schedule ! SIG Fabrication Status i Implementation Status Engineering Status ! 9 Engineering Activities Jamie LaBorde : Summary of Modification Packages
# Submittals to NRC April Rice /
Replacement Philosophy Lou Cartin : Previous Submittals . October 1993 Submittal
- Future Submittals NRC Approval Schedule for Technical Specification Changes ;
9 Startup Testing Program Mark Miltner Testing Philosophy _ Description of Tests we Plan to Run j 9 Other Issues April Rice : NRC Feedback on 03/12/93 Piping Analysis Le t ter i S/G Recycle Facility Plans .! 9 Plans for Power Uprate Ron Clary Scope of Project & Schedule
.-,1
V. C. Summer Nuclear Station Steam Generator Replacement Project 4qqogisWONM;ssI@.$,Vl.MF_si!@-. m que #m w, . 5 eMMDM ya!%w$w@ay .c4mee,m@@y. a.ai2
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# 3 Loop Westinghouse PWR S Original Startup - 1982
# 2775 MWt Core Power
# Westinghouse Model D-3 Steam Generators 8 Fall 1994 - RF 8 S/G Replacement Outage 9 Spring 1996 - RF 9 Outage to Implement Power Uprate to 2900 MWt Core Power G Terms Unique to V C Summer Station
' MRF - Design Change Package
' Emergency Feed Water - Auxiliary Feed Water )
V. C. Summer Nuclear Station Steam Generator. Replacement -Project j [ t $ p'E T g # @ $ i @ lg B Rf M i$ f 4 1% ssit q
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9 S/G Project _ Group Manages Project 1 S Bechtel Contracted to Perform Major S/G . Replacement Engineering and f Installation Activities ; i O Bechtel Programs / Procedures Integrated into l V C Summer Programs ; F S V C Summer Replacement Activities Based on ,! Site Visits / Lessons Learned from Previous i Replacements . . . ,
# D C Cook !
' Palisades
# Turkey Point
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# Millstone i
/ North Anna !
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8 V. C. Summer Nuclear Station Steam Generator Replacement Project IsifdEinf@5f5IIEMEeAE@EANTIdi!5$Iff505f5IS 8 Presentation to NRC on 01/09/92 ... Key Points . . .
' Reviewed the Basic Replacement Approach Supporting Analysis l Engineering Scope j
=
Power Uprate Licensing Schedule 4 l l 1
/ Question Posed - Can NRC Support Fall 1994 Replacement Schedule (Uprate Remaining in Spring of 1996)
~
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Received Positive NRC Feedback
1 V. C. Summer Nuclear Station i Steam Generator Replacement Project ! If58AKid1H3EifsTEt#il%illT&MenhwsE _;__ _ i o I e Follow-up Presentation on 08/12/92 ... l Key Points . . . j Discussed Analysis Strategy Including Small Break LOCA i 1
/ Reviewed Engineers Scope i
Presented Licensing Submittal Schedule Discussed Old Steam Generator Disposal Options ;
Reviewed Implementation Schedule '
l i. 4
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1994 Steam Generator Replacement Schedule 1990 1991 1992 1993 1994 1995 1996 ' 1 Planning & Decision Analysis E@.@hNM t Primary Systems Secondery/ BOP Systeme , Containment Walkdow n Select S/G Fabrication Contractor S81E f l l Purchase / Fabricate S/G's [yy n t21 @ Ri W ipth l j l l l Detailed Engineering Analyses [!?uW2(22WeiiMM1 Primary Systeme j Secondary & T/G BOP & Contelament Licensing Submittal Development S8W 8'.8M d NRC Approval [12 Mohs l ! Engineering l;26BiGB!HF1 Engineer Facilities Rigging Te mp. HVAC, Plat f or m e. E TC. Procedures, Construotton Packages Plant Preparation b"**'b'l Build S/G Storage Facility Wasehouse, ETC. S/G Inspectione , Work Packages Relacement Outage Activities MSR '2 FW, S/G Replacement b
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V. C. Summer Nuclear Station Steam Generator Replacement Project 3M?M*$8E@I5f@(91E*3 ail $ 9 Delta 75 Steam Generators being Fabricated by Westinghouse in Pensacola, Florida O Component Status: S/G 'A'
- Tubing Installation Complete
- Installing Anti-Vibration Bars
- Assembling Upper Shell S/G 'B'
- Tubing Installation Complete
- Installing Anti-Vibration Bars
- Assembling Upper Shell S/G 'C' _
- Installing Tubes
- Assembling Upper Shell O Shipment by Barge / Rail - August 19,94
w Steam Generctor Fabrication
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~ i V. C. Summer Nuclear Station Steam Generator Replacement Project ; i Mgaj{$jjj@QLyjy({ , i Implementation Process is Proceeding on Schedu.'e ; I 9 Core Project Management / Engineering Personnel - are Currently Assigned ' 9 Some Support Work Completed in RF-7 9 Two Support Facilities Complete . . . Remainder to be Complete in 1994 9 Beginning to Develop Detailed Implementation Packages _ _ 9 Integrated Outage Schedule is in Initial Development 9 Startup Program is in Initial Development i
I i V. C. Summer Nuclear Station ' Steam. Generator Replacement Project Fi$5slaggg[S$5(Qgfeylll$fggnmegjgfff$$4 ; Design Process Is On Schedule S Engineering Modification Packages are in Final Design Review Stage
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9 Licensing Package Submittals are on Schedule
' Major Submittal Made 10/93 l
i l 9 Decision Made to Store Retired Steam Generators l on Site in Lieu of Burial at Barnwell l l
1 Engineering Activities UQmle _asorc e i i 9 9 e I 4
L V.C. SUMMER STEAM GENERATOR REPLACEMENT DESIGN CHANGE
SUMMARY
REPLACE N STEAM GENERATORS /y s f 'g l i I ; REFLACE IO S FIEAM GlhTJLATORS y REVISE MRF 90001 g+ INSTRUMENTATION i RIGGING & , STRUCTURAL ! I MRF 90002 I I IAC f, 1 MRF 9oses l l INSULA 110N I ( l I&C SETTOINTS MRF 90005 *,
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\ j REROUTE ANALYSIS j MRF 90010 i !
l' EFW PIPE REROUTE rown MRF 900U l i REACIDR BLDG FIFING
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ELIMINATE l COPPER I l-DELETE :
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INCREASE ' CHARGING FLCW l l TOLERANCE 3 i ./ \ ,
/ h CIIARGING MRF 90011
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90001 THE GENERATOR REPLACEMENTITSELF THIS WILL INCLUDE REMOVAL OF THE : OLD S/Gs, INSTALLATION OF THE NEW i S/Gs, FIT-UP AND WELDING OF THE RCS PIPING, REWORKING THE MAIN STEAM PIPING AND INSTRUMENT TAPS, REMOVAL OF ANY UNNEEDED RC JET SHIELDS OR WHIP RESTRAINTS AND ! REWORK OF SEVERAL INTERFERENCES. 1
MRF 90001 REPLACEMEhT OF TIIE STEAM GENERATOlG S RER OF MAIN STEAM PIPE WHIP RESTRAINTS e RER OF SG PRIMARY SUPPORT, TEMPORARY SG SUPPORTS ABANDONMENT OF RCS WHIP RESTRAINTS ~ SG SNUBBER REDUCTION 4 PRIMARY REACTOR COOLANT LOOP TEMPORARY PIPE RESTRAINTS 8 MAIN STEAM TEMPORARY PIPE RESTRAINTS ' S FEEDWATER, EMERGENCY FEEDWATER, BLOWDOWN, & SHELL DRAIN TEMPORARY PIPE RESTRAINTS e REACTOR COOLANT PIPE CUTTING & WELDING e MAIN STEAM PIPE CUTTING & WELDING
# FEEDWATER PIPE CUTTING & WELDING G EMERGENCY FEEDWATER PIPE CUTTING & WELEING SECONDARY SIDE BD AND SHELL DRAIN PIPE CUTTING & WELDING
# NUCLEAR SAMPLING PIPE CUTTING & WELDING e INSTRUMENT LINE TUBE CUTTING & WELDING e PERFORMANCE OF LASER TEMPLATING e PIPE END DECONTAMINATION O
TEMPORARY REMOVAL OF INTERFERENCES - RW SURFACE VENTILATION
# TEMPORARY REMOVAL OF INTERFERENCES - RV HEAD VENT e PREPARATION OF NEW STEAM GENERATORS
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90002 STRUCTURAL EVALUATIONS FORINSTALLATION ~ THIS WILL PROVIDE THE EVALUATIONS FOR FLOOR LOADING, PULLING POINTS, ! JACKING TOWERS, POLAR CRANE ' UPRATE,ETC. ANY FIELD WORK, ! EITHER PERMANENT OR. TEMPORARY, l REQUIRED BASED ON THE ABOVE' EVALUATIONS .WILL BE CONTROLLED BY ! THIS MRF. THE LIFTING OF THE S/G's IS ' CONTROLLED BY THIS MRF. O
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.l' THE NEW S/G'S WILL HAVE DIFFERENT !
MAIN FEEDWATER AND EMERGENCY I FEEDWATER NOZZLE LOCATIONS. THE t 3 ANALYSIS, ROUTING AND INSTALLATION l
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l 90004 : FORWARD REVERSE FLUSH ; DELETION , THIS MRF WILL REMOVE THE LINES, ! VALVES, INSTRUMENTS AND ASSOCIATED SUPPORTS FOR THE FLUSH PIPING. IT [ WILL ALSO REMOVE OR SPARE' THE CIRCUITS ASSOCIATED WITH THE DELETED VALVES AND INSTRUMENTS, , REMOVE THE UNNECESSARY FWIV INTERLOCKS AND REWORK THE MCB FOR THE DELETED SWITCHES. e
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I i 90005 INSULATION ii NEW INSULATION WILL BE REQUIRED DUE TO NEW NOZZLE LOCATIONS PIPE : ROUTING ETC. THE INSTALLATION OF ' THE REFLECTIVE METAL INSULATION , I INSIDE THE REACTOR BUILDING WILL BE COVERED BY THIS MRF. I l i i i
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i 90006 j IMPULSE LINE AND TRANSMITTER CHANGES ~ t THE NEW S/Gs AND THEIR INSTRUMENT TAP LOCATIONS WILL REQUIRE RE- . ROUTING OF IMPULSE LINES AND POTENTIALLY REPLACEMENT OF SOME TRANSMITTERS DUE TO -SPAN CHANGES. THIS M.RF IS ONLY FOR IMPULSE LINES, AND TRANSMITTER. ' I 1
3 J 90007 ' INSTRUMENT SETPOINTS AND : i CONTROLS l - I VARIOUS SETPOINTS WILL BE CHANGED FORTHE NEW S/Gs SUCH AS S/G LEVEL j i TRIPS, OVER POWER AND OVER ! TEMPERATURE DELTA T TRIPS AND : RUNBACKS AND T AVG AND DELTA T ! PROGRAMS. THE S/G LEVEL CONTROL l PROGRAM WILL ALSO BE CHANGED AND THESE CHANGES WILL BE CONTROLLED BY THIS MRF. ALLPROCESS CABINET WORK WILL BE CONTROLLED BY THIS MRF. ; l
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-1 l MRF 90007 INSTRUMENT SETPOINTS AND CONTROLS ,
FW Loop Temperatures: Delete the bistable for FWIV closure and the ! associated wiring. The MCB indication remains. j Forward Flush Flow Loops: Delete the entire
- loop ~ with its associated wiring and process cabinet cards. l l
FW Loop Flows: Delete the low flow FWIV closure signal and the associated status lights and wiring. Steamline Pressures: Delete the low _S/G pressure IVIV . closure signal and the associated status lights and wiring. ! S/G Narrow Range Level: Delete the lo-lo-lo S/G level FWIV closure .a signal and the associated status lights and wiring. Delete the ' program input for S/G level control and for the S/G lo and-lo-lo level reac'mr trips as well as change the setpoints. Change the i high level feedwater isolation setpoint. Provide new scaling for the transmitters to account.for new tap locations and the analysis assumptions. Power Range NI's: Delete the input for programming S/G level I control and protection. Delete the negative rate trip. AMSAC: Revise the logic and setpoints. Steam-Dump System: Add a new card to improve the system response i and stability as well as revise time constants. > Pressurizer Pressures: Delete the rate portion of the circuit for the pressurizer low pressure reactor trip. i FW Pump Speed Control: Provide new setpoints for the lower pressure
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differential required by the feedring style S/G. Delta T: Provide new setpoints for Over-power and Over-temperature ; trips and runbacks.
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Tave: Provide new setpoints for various reduced temperature modes of operation. - l RCS Flow: Provide new scaling for new design flow. i S/G Wide Range Level: Provide new scaling information as required ) by the new upper tap location. l Reactor Vessel Level Indication System: Provide new narrow range scaling to compensate for the new vessel flow. I
L- . 90008 FACILITIES OFFICE SPACE, NEW AND OLD SG ' STORAGE BUI-LDINGS, WORKSHOPS, ACCESS CONTROL POINTS, AND OTHER TEMPORARY (OR PERMANENT) FACILITIES REQUIRED TO SUPPORT THE PROJECT WILL BE CONTROLLED BY TH.E FACILITIES MRF. THOSE ITEMS REQUIRED TO PHYSICALLY INSTALL THE S/Gs ARE NOT INCLUDED IN THIS SCOPE OFWORK. . 1 W
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l l 90009 l MOISTURE . SEPARATOR l 1 REHEATER UPGRADE ! l THE ORIGINAL MSR's CONTAINED l COPPER NICKEL TUBE BUNDLES. COPPER l r IS A CONTRIBUTOR TO SG TUBE ' PROBLEMS AND MUST BE MINIMIZED. THE COPPER CONTAININ~G -TUBES WERE : REPLACED WITH NON-COPPER ! CONTAINING TUBES DURING RF-7 THE Li TUBE CONFIGURATION WAS ALSO - CHANGED TO MAKE PERFORMANCE AND RELIABILITY GAINS. W e m =
q F 90010 . ANALYSIS : i THE ANALYSIS FOR THE REPLACEMENT ! 1 IS BEING CONSOLIDATED INTO ONE' j PACKAGE TO HELP. STREAMLINE THE , LICENSING PROCESS. THIS PACKAGE: WILL CONTAIN THE ANALYSIS RESULTS, FSAR MARKUPS AND TECH. SPEC. i I l CHANGES FOR SG REPLACEMENT. THIS l t PACKAGE WOULD NOT CONTROL ! SPECIFIC SUB-TASK FSAR CHANGES i . I REQUIRED BY THE SUB-TASKDETAILS: EVEN IF THE JUSTIFICATION 'OF THE SUB-TASK IS INCLUDED.
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L :- . i 90011 i CHARGING MINIFLOW TEST CONNECTIONS l
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THE SMALL BREAK LOCA ANALYSIS ! REQUIRES MORE STRINGENT i SURVEILLANCE TEST DATA TO MAINTAIN AN ACCEPTANCE CRITERIA : WITH AN ADEQUATE OPERATIONAL BAND. MINIFLOW FLOW . TEST CONNECTIONS WILL BE INSTALLED TO q GATHER DATA TO SUBSTANTIATE THE ELIMINATION OF UNCERTAINTIES ACCOUNTED FOR IN PREVIOUS ANALYSES.
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l l 1 90012 ; ELECTRICAL SUPPORT FOR SG , REPLACEMENT . 4 POWER REQUIREMENT FOR STEAM GENER.ATOR REPLACEMENT REQUIRES A SIGNIFICANT POWER SOURCE, SUCH AS ; THE "B" REACTOR COOLANT PUMP- - MOTOR FEEDER. 1 SEVERAL ELECTRICAL i INTERFERENCES ALSO EXIST. THE j i TEMPORARY POWER FOR REPLACEMENT- l ACTIVITIES AND THE ELECTRICAL ! INTERFERENCE WORK IS CONTROLLED BY THIS MRF. i 6 1 I i
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b h i h f 8 o e 1 4 eme Submittals : to NRC e Apri Reice ; i
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1 2 l 1 l 1 I I i l
8 V. C. Summer Nuclear Station . Steam Generator Replacement Project j . 355@$kN$NNN NNNND O W5ks5AAN!asFTi$ 9 S Replacement Philosophy i
/ Utilizing 50.59
/ Some Technical Specification-Changes are !
Required due to New Type of S/G (Delta 75) and Revised Operating Conditions ' L e
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O h V. C. Summer Nuclear Station Steam Generator Replacement Project bygggg@$sgsybg((($1(#ppfpgggggggggjggggg p 9 Submittals to Support Technical Specification Changes
/ Delta 75 Design Report Submitted 09/03/92 (WCAP-13480) f
/ Leak before Break - Approved by NRC SER i dated 01/11/93
/ Initial Analysis Submittal made 04/30/93
/ Final Analysis Submittal and SGR Technical Specification Changes Made 10/29/93
/ SBLOCA Analysis and Related Technical ,
Specification Changes to be Submitted by 12/31/93 5
' P V. C. Summer Nuclear Station i Steam Generator Replacement-Project ,
N d M ?is E 3 f s $ @ @ E @ N I N A I M M E l W S B N @ T @ M ! 9 Other Submittals i j
/ March 12,1993 Description of Piping -
Analysis Performed to Support SGR i i
/ Future Submittal . . .
SCE&G Position on 10 CFR 50 -! Appendix J (ILRT). Credit will be - Taken for the ASME Section-XI Hydro Test. -1 h p
1 V. C. Summer Nuclear Station . Steam Generator Replacement Project j BI$MF1T(135M5$E!!@N$W!$KWYS$itdNGEERP3f$l i . f G NRC Approval of Technical Specification Changes ; 1 1 l
/ Approval Requested by 08/31/94 :
b
/ Latest Date would be Mode 5 Entry (Late October 1994) e b
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. 1 SCE&G's LICENSING SUBMITTAL l AN OVERVIEW l
l l F L. R. CARTIN SENIOR ENGINEER . SCE&G - t NOVEMBER 30,1993 ,
TOPICS ! INTENT l SCOPE i MAJOR RESULTS TECH SPEC CHANGES l 1 l
% )
LICENSING SUBMITTAL ANALYSES TO SUPPORT LICENSING OF THE REPLACEMENT SGs. REFLECTS THE NEW SGs AND ,WHERE POSSIBLE, THE IMPACT OF A CORE POWER LEVEL AS HIGH AS 2900 MWT - THE PLANT'S ORIGINAL ENGINEERED SAFEGUARDS RATING. OUTLINES CHANGES TO THE PLANT'S DESIGN BASIS REQUIRED TO PRESERVE THE BASIS OF THE NEW ANALYSES. PROVIDES TECH SPEC CHANGES AND THE SUPPORTING SIGNIFICANT HAZARDS EVALUATION. e
NSSS PERFORMANCE 1 PARAMETERS : Tavg: 572 to 587.4 F i 6 SG Tube Plugging: 0 to 10% : Flow: 92,600 to 102,600 gpm per 1 loop I l J
- - - - -1
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- SUBMTITAL SCHFDULE AND FORMAT TO SUPPORT STEA.V 'ERATORS REPLACEMENT TECHNICAL SPECIFICATION CHANGES FOR THE VIRGIL C. SUhBfER NUCLEAR STATION Submittal Title 1 2 3 4
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i List of Tables X X .X List of Teres X- X X > List Acronyms and Abbreviations X X X Executi7e S-===ry X 'X X .f 1.0 Introduction-Description of Licmse Amendment Request X X , 1.1 Purpose for Change I 1.2 Cunent license Basis and Function ofIdentified Technica] ! Specifications ! 1.3 - Description of Proposed Change . 2.0 Basis for Evaluations / Analyses Performed X X l 2.1 Design Power Capability Parameters X 2.1.1 Discussion of Pammeters I 2.1.2 References l 2.2 NSSS Design Transients 2.3 Control System Serpomts -I 2.4 Reactor Protection System / Engineered Safety Feannes Actuation ! System Setpoints 3.0 Safety Evaluations / Analyses j 3.1 less of Coolant Accident Analyses ! 3.1.1 Large Break LOCA X ! 3.1.2 Small Break LOCA X ; 3.1.3 Post-IDCA long Term Core Cooling Suberiticality X ., 3.1.4 Hot Leg Switchover to Prevent Potential Boron Precipitation X -i t 3.1.5 References X 3.2 LOCA Hydraulie Forces X-3.2.1 Introduction i 3.2.2 Method of Analysis ;
'3.2.3 Results 3.2.4 , References i Te ret Submmal Dates 1: August 31.1992 "
2; Aptd 30.1993 '
- 3. October 29,1993 4 December 31.1993 R$G-TOC;.COE: 908'93 i I
1
9 Submittal Title 1 2 3 4
)
i 3.3 Non-LOCA Analyses X 3.4 High Energy Line Bd Analyses X 3.4.1 LOCA Mass & Energy Releases X X 3.4.1.1 Long Term LOCA Mass and Energy Releases 3.4.1.2 Short Tenn LOCA Mass and Energy Releases 3.4.2 Short Term Contamment Analysis -IDCA Reactor Building X X Sub~Wwt Analysis 3.4.3 Main Ste= mime Break Mass / Energy Releases X 3.4.3.1 Inside Contamment 3.4.3.2 Outside Contamment 3.4.4 Img Term Contamment Analysis X 3.4.4.1 Main Str= mime Break Contamment Integrity Analysis 3.4.4.2 LOCA Reactor Building Integrity Analysis X 3.4.5 Environmental Conditio cs - Steam Ilne Break (SLB) Outside X Contamment I 3.4.6 Equipmmt Qualification X 3.4.7 References X X 3.5 Steam Gmerator Tube Rupture Accident Analysis X X 3.6 Reactor Cavity Pressure Evaluation X ] 3.6.1 Introduction I 3.6.2 Evaluation Results 3.6.3 References 3.7 Radiological Analysis X X 3.7.1 Introduction 3.7.2 Source Terms 3.7.3 Radiological Conm-3.7.3.1 Loss of Offsite Power 3.7.3.2 Waste Gas Decay Tank Rupture 3.7.3.3 Break in a CVCS Line . 3.7.3.4 Large Break LOCA 3.7.3.5 Main Steam Ilne Break l 3.7.3.6 Steam Generator Tube Rupture i 3.7.3.7 Locked Rotor Teer* Submrcal Detes 1; Augur. 31,197.
- Apry 30,1993 3: October 29.1993 4: Decemoet 31,1993 R5G-TOC 2.CGE: 908% ii
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l l l l l Submittal Title 1 2 3 4 3.7.3.8 Fuel Handling Accident 3.7.3.9 RCCA Ejection l 3.7.4 References l l 3.8 Prunary Components Evaluations
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3.8.1 Reactor Vessel X , l 3.8.1.1 Reactor Vessel Structural Evaluation l 3.8.1.2 Reactor Vessel Briule Fracture Integrity 3.8.2 Reactor Internals X l 3.8.2.1 ~aermal-Hydraulic Performance ' 3.8.2.2 Bypass Flow Analysis 3.8.2.3 Hydrsulic Ilft Force Analysis 3.8.2.4 RCCA Scrum Performance Evaluation 3.8.3 Steam Generators X 3.8.3.1 'hermal-Hydraulic Performance Evaluation
- 3.8.3.2 Structural Evaluation
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3.8.4 Teb X ; 3.8.5 keactor Coolant Pumps (RCPs) and RCP Motors X 3.8.6 Control Rod Drive Machmium X ; 3.8.7 Reactor Coolant Piping and Supports X 3.8.8 Application of leak-Before. Break Methodology X X l 3.8.9 Conclusions X l 3.8.10 References X 3.9 Fluid and Auxiliary Systems Evaluations X l 3.9.1 Introduction 3.9.2 Discussion of Evaluations Performed 3.9.2.1 Fluid Systems Evaluation 3.9.2.2 Annhery Equipment Evaluation i 3.9.2.3 NSSS/ Balance of Plant Interface I 3.9.3 Conclusions i 3.10 Fuel Structural Evaluation X 3.10.1 General Considerations 3.10.2 Fuel Assembly Structural Evalution 3.11 Summary of Technical Specification Changes X l Terree Sub-m! Detes 1: Augua 31. !F. L April 30.1993 - 3: Octar 29.1993 l 4: December 31.1973 RSG TOC.CGE: 9'28!93 iii l i
I Submittal Title t- 3 .4 4.0 Conclusions X
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Appendix 1 10 CIR 50.59 (Assessmmt of UsurJ;;d Safety Questions) .X X Appendix 2.10 CFR 50.92 (No Significant Hazards Determin=* ion) X~ X ,
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Appedix 3 Proposed Technimi Specification Changes X X- : Appendix 4 WCAP 13480 " Westinghouse Delta 75 Steam Gmerator Design X X and Fabncation Infonnation for the VCSNS" Appendix 5 WCAP-13605 "Prunary Loop [mak-Before-Break Reconciliation X X to Acxount for the Effects of SGR/Uprating" Appedix 6 VCSNS 75AR Chapter 15 Write Ups - X X_ l 4
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T.mei suteni D.g 1: Aups 31,1992
- Apnl 30.1993
- 3. October 29,1993 4 December 31,1993 RsG TOC 2 CGE: 10!:193 iv
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i -Intended tor new plant and replacement i Highest thermal capacity of any steam
; steam generator apphcanons generator in this shell size 1 Idenocal shell size to the 51 Senes. ~* Upratmg compadbihty up to 1050 MWt Model D, and Model F steam generators per steam generator i - Exceeds the func6onal performance - Lowest T-hot capability provides capabihties of cusung steam generators excellent tuel etF.eiency and lower l corrosion rates
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M85 - Addabonal core DSB marpn Based on demonstrated strengths of the - Wide water level control span , Model F steam generator . . . l j Feednng design perrnits simplest tube
# ~ 7 -~ "7 bundle, maumizmg tube rehabihty whde .
-Twelve access ports, including two !
providing addiuonal operaung marpn pnmary manways and ten secondary side ) l ; 84 steam generators with similar teatures access ports i operaung for an aserace of 6 yean -Enhanced penpheral tube access trom pnmary and secondarv sides j 2 -- > Studs, rather than bolts, on pnmarv and i
-Uses thermalh treated Alloy 690 tubmg secondarv manwavs for faster mstallauon
- the state of-the art tube matenal and removal
- based on its corrosion resistance a Forged shell boundanes eliminate Tn2ncular p
- tch tube arrav maumires lonptudmal weids and help to reduce I tace bundle surface area inservice inspeccon ome 3 Full depth hydraube espansion tutse-to- Channel head claddmg machmed to a tubesneet io,nts with minima] residual smooth surface tinish to reduce depost-l Stress tion, thereby helpmg to reduce man-tem 5tuniess steel tube supports enhance ciposure corrosion res; stance and reduce wear #1ntegral sludge collector to promote
- Broached tube supports enhance riuid sludge deposidon away from tne tube g tlushing at the tube mtertace, to limit bundle chemicai concentracon around the tube > Capable of 3 percent condnuous 4
- Tube suppons and tion distnbunon blowdown flow rates badie hase tlar surfaces wnere the tubes > Moisture carrvoser no more than 0.10 4 touch the supports to reduce drvout percent at tull power cperation l - Three sets of U-bend supports with > Wide water level control span helps to mtmmum par construccon enhance reduce the potential for tnps dunng tuoc ubration and a car marpns transient operauon ! \
1 l i t i b J e w -- . _ _ . - . - _ . . . - . _ . - - , - . . . . - - - - _ _ ._ _- _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . . _ _ . . . ,4-. - . . ..---,c.. + - , ,
6 , WESTINGHOUSE PROPRETARY CLASS 2 l TABLE 1-3 MODEL D3 AND DELTA 75 GEOMETRIES Model D3 Delta 75 ! OVERALL DIhENSIONS Overall Length, inches 812.00 812.00 lower Shell ID, inches 129.38 129.38 1 Upper Shell ID, inches 168.50 168.50 TUBE BUNDLE DIMENSIONS Heat Transfer Area, sq ft [48,000 75,1807 , Number of Tubes [4,674 6,307T Tube OD, inches 0.750 0.688 Tube Wall, inches 0.043 0.040 ) Tube Material [ Alloy 600 MA* Alloy 690 Tr"T Tube Pitch, inches 1.0625 0.980 Pitch Arrangement Square Triangular Minimum U-Bend Radius, inches [2.25 3.25T Wrapper ID, inches [124.50 123.007 Wrapper Opening Height, inches [6.00 12.00T Tube Bundle Height, inches [328.4 418.4T Number of Tube Support Pla.es/ 7/1 (Hot Leg) [9/1T
. Flow Distribution Baffles Support / Baffle Material Carbon Steel 405 SS*"
Number of U-Bend Supports 2 Sets [4 Sets? U-Bend Support Material Alloy 600 [405 SS""r (Chrome Plated) MOISTURE SEPARATOR DESIGN Number of Primary Separators 12 (18T Primary Separator ID, inches [19.50T [19.507 Primary Separator Design Centrifugal Centrifugal / Low Pressure Drop
. Secondary Separator Design Two Tier Single Tier
[(8 Banks)T [(6 Banks)T
- MA - Mill Annealed
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[ " TT - Thermally Treated (small radius U-bends are stress relieved)T
"* SS - Stainless Steel >
WP1486:lD/090192 1 14 ,
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DELTA-75 STEAM GENERATORS FEEDRING SG WITH SAME OUTSIDE DIMENSIONS AS MODEL D3s i i LARGER HEAT TRANSFER AREA (75000 VS 2 48000 FT ) j LARGER NR LEVEL SPAN (240 VS 233 IN) l CONSTANT SG LEVEL CONTROL (60% FROM ' 0 TO 100% RTP) i LESS MOISTURE CARRYOVER (0.1 VS 0.25%) i HIGHER CIRCULATION RATIO (3.4 VS 2.2)
WESTINGHOUSE PROPRIETARY CLASS 2 TABLE 6-1 MODEL D3 AND DELTA 75 CALCULATED FULL LOAD PERFORMANCE Model D3 Delta 75 PLANT RATING = 2787 MWt (929.00 MWt/SG) PRIMARY SIDE - ' Pnmary Flow Rate (1), gpm 103.100 102,600 Pnmary Inlet T+.me(2), *F 617.5 617.5 Primary Average Tw.uie(2), *F 587.4 587.4 Pnmary Outlet Temare(2), *F 557.4 557.3 Pnmary Pressure Drop, psi 36.5 37.4 Primary Side Volume, eu ft 935. 1200. Pnmary Side Mass, lbs 41,300. 53,100. SECONDARY SIDE Feedwater Tw.uie, *F 435.0 435.0 Steam Temperamre(3), *F 544.8 546.2 Steam Pressure (3), psia 1001. 1014. Secondary Pressure Drop, psi 31.6 25.4 Secondary Side Volume, cu ft 5,947. 5,583. ; Secondary Side Mass, Ibs 102,500 118,400 Normal Full lead Water Level, in 492. 532. PLANT RATING = 2912 MWt (970.67 MWt/SG) i l PRIMARY SIDE i Pnmary Flow Rate (1), gpm 103.100 102.600 ! Primary Inlet Temperamre(2), *F 618.7 618.8 I Primary Average Temperamre(2), *F 587.4 587.4 Primary Outlet Temperamre(2), *F 556.1 556.1 Pnmary Pressure Drop, psi 36.5 37.4 Primary Side Mass, lbs 41,300. 53,100. SECONDARY SIDE Feedwater Temperamre. *F 440.0 440.0 Steam Temperature (3), 'F 542.2 544.6
)
i Steam Pressure (3), psia 980. 1000. ; Secondary Pressure Drop, psi 31.6 25.4 i Secondary Side Volume, cu ft 5.947. 5,583. l Secondary Side Mass, !bs 101,200 116,700 Normal Full lead Water Level, in ! 492. 532. [1) These pnrwv flow ratas are bem estunas
- 2) Calcuhned performance bened on operation er T evg = 587.4 *F.
- 3) $ team tempers:ures/ pressures are calculand based on the best esumste flow razes and faulmg factors. }"" '
WP1486:1D/090192 6-4
1 DESIGN TRANSIENTS DESIGN TRANSIENTS UPDATED TO REFLECT NEW SGs AND UPRATE TWO SETS OF CONDITIONS ARE PROVIDED TO COVER THE FULL RANGE OF Tavg (572 - 587.4 F) ARE USED AS THE BASIS FOR THE SG DESIGN AND NSSS COMPONENT EVALUATIONS
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CONTROL SYSTEM CHANGES : SETPOINTS DEFINED TO COVER FULL RANGE OF Tavg, UPRATE AND THE DELTA-75 SG STEAM DUMP - ADDED LEAD LAG COMPENSATION TO THE Tref SIGNAL i SG LEVEL - CONSTANT LEVEL CONTROL i NORMAL LEVEL = 60.4% ; NR SPAN = 240 INCHES ; I Q
b T ABLE 15 4-2
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L ARGE BRE AK-Rest 1TS DECLG DECLG . DECLG DECLG DECLG (C3-0.8) (Ca=0.6) (Cn=04) . i C.,=G.4 ) < C..-0.J)
# MAXS1 IFBA-isec) (sect (sec) isec) isec)
Peak Clad Temperature (*F) 17M 1767. 1924. 1799. 2007. Peak Clad Locadon ift) 6.75 7.00 - 7.00 6.25 6.25 Local Zr/H,0 Reaction (max) Fe 1.7943 2.5064 3.5711 ^1,6980 6.0186 1 Local Zr/H:0 Location eft) 6.25 6.25 6.00 7.00 6.25 Average Zr/H:0 Reaction 9c < l .0 < !.0 < ! .0 < l .0 < l.o s Hot Rod Burst Time (see) 4 7.65 44.91 46.11 43.53 62.09 Hot Rod Burst Locadon <ft) 6.25 6.25 6.00 5.75 6.25 ; i NSSS Power (MWt) 2787. Peak Linear Power (kW/ft) 13.607 Peaking Factor tF.5) 2.45 Intact Loop Accumulator Water Volume trt') 1039 - irmnimum plus line voiume) ; e 10cc Steam Generator Tube Plugging in each steam generator is assumed. .;
- t a
i k 15.4-65 i
. . . , r-. . . -
~
k LARGE BREAK LOCA l APPENDIX K CALCULATIONS ARE DONE ! TO SUPPORT RSG LICENSING AT 2775 ) MWt ; i i l MAX PCT = 2007 F BEST ESTIMATE LOCA ANALYSIS TO BE DONE TO SUPPORT THE POWER UPRATE 1
)
)
i w J
I
- i SBLOCA ANALYSIS
}
NOT ADDRESSED IN THE CURRENT LICENSING SUBMITTAL. l SBLOCA ANALYSES AND SUPPORTING TECHNICAL SPECIFICATIONS TO BE - l SUBMITTED BY THE END OF 1993. l PRELIMINARY STATUS i 10% MORE SI FLOW REQUIRED TO SUPPORT THE POWER UPRATE , MORE SI FLOW TO .BE OBTAINED BY INCREASING THE SI/CHG PUMP R/O LIMIT FROM 680 TO 708 GPM BASED ON RUNOUT TESTS ON ALL THREE CHG/SI PUMPS DURING LAST REFUELING , PRELIMINARY RESULTS ARE _ AVAILABLE. t Y
PRELIMINARY SBLOCA RESULTS Value Parameter 2 Inch 3 Inch 4 Inch 6 Inch Nominal Nominal Reduced Nominal Nominal Twee Two, Twet Two, Twe, Peak Clad Temperature 1634 1852 1799 1477 1690 (*F) Peak Clad Location (ft) 11.75 11.75 11.5 11- 10.75 Max Local Zr-H 2 O 2.58 3.83 3.36 0.43 0.71 Reaction (%) Max Local Zr-H 2 O 11.75 11.75 11.75 11.25 10.75 Location (ft) Total Zr-H;O Reaction < 1 < 1 < 1 < 1 < 1 (%) sr
~
LONG TERM COOLING EVALUATIONS ARE VALID FOR i CORE POWEPsS < 2900 MWT CONFIRMED THAT THE CORE REMAINS SUBCRITICAL ASSUMING ALL RODS OUT. POTENTIAL BORON PRECIPITATION PREVENTED BY , SWITCHOVER TO HL RECIRCULAION WITHIN 8 HOURS, AND l ALTERNATING HL & CL l RECIRCULATION EVERY 18 - HOURS.
g t r ! LOCA HYDRAULIC FORCING ! FUNCTIONS LOCA FORCES GENERATED FOR THE LOOPS, RV, AND SGs LEAK BEFORE BREAK APPLIED TO l ELIMINATE DE PIPE BREAKS IN THE PRIMARY PIPING l LARGEST BREAKS IN THE CL & HL WERE ANALYZED (ECCS ! ACCUMULATOR LINE & SURGE LINE) CONSERVATIVE BREAK OPENING TIME (0.001 SEC) USED IN GENERAL, THE RESULTING LOADS ARE > 40% LOWER IN MAGNITUDE THAN THE ORIGINAL DESIGN BASIS ANALYSIS. l I
t NON-LOCA ACCIDENT ANALYSIS . NON-LOCA LICENSING BASIS IN CH 15 WAS REANALYZED APPROVED METHODS USED DNBR ANALYSIS TO BE BASED ON THE REVISED THERMAL DESIGN , PROCEDURE , 1 i ANALYSES DEMONSTRATE THAT ALL j PERTINENT LICENSING CRITERIA CONTINUE TO BE MET NE L, SIhiILIAR O HOS f IN THE CURRENT LICENSING BASIS. l J : l l
~ SIGNIFICANT CHANGES NON-LOCA ANALYSES RCCA MISALIGNMENT TOOK NO CREDIT FOR THE NEGATIVE FLUX ! RATE TRIP LOSS OF NORMAL FEEDWATER REQUIRED THE ADDITION OF 400 GPM OF EFW TO PREVENT THE PRESSURIZER FROM FILLING FEEDLINE BREAK INITIALLY BEHAVES ! AS A STEAM LINE BREAK DUE TO THE CHANGE FROM A PRE-HEATER TO l FEEDRING SG. i
~
i Y
. )
RADIOLOGICAL CONSEQUENCES 1 NEW SOURCE TERMS GENERATED TO BOUND : FUTURE FUEL CYCLES FUEL TYPE - VANTAGE + . CORE POWER - 2958 MWt AVG DISCHARGE - BURNUP - 65,370 MWD /MTU CYCLE LENGTH - 480 EFPD ENRICHMENT - 5.0 wt-% DOSE METHODS UPGRADED TO CONFORM TO THE REQUIREMENTS OF THE SRPs. . IN GENERAL, DOSES HAVE INCREASED BUT REMAIN WITHIN APPLICABLE ACCEPTANCE CRITERIA. , 4
~
^ HIGH ENERGY LINE BREAK ANALYSES - i MASS AND ENERGY RELEASES-WERE ! EVA.LUATED AND/OR GENERATED FOR THE FOLLOWING: LOCA FOR SUBCOMPARTMENT ANALYSIS 1 LOCA FOR RB INTEGRITY ANALYSIS SLB FOR RB INTEGRITY ANALYSIS SLB FOR HELB CONSIDERATIONS ' OUTSIDE RB + l PRESSURE AND TEMPERATURES , EFFECTS WERE EVALUATED. e I
~
r i LOCA LONG TERM CONTAINMENT ~ ANALYSIS ! DEPS BREAK WITH MIN / MAX ECCS AND THE DEHL BREAK ANALYZED RB P&T RESPONSE CALCULATED WITH ' CONTEMPT-LT26 INITIAL CONDITIONS CONSISTENT WITH I CURRENT LICENSING ANALYSIS RB COOLING CONSISTENT WITH CURRENT LICENSING ANALYSIS EXCEPT ! RBCU HEAT HEAT REMOVAL CAPACITY j REDUCED BY 60% . i k N
TABLE 3.4.4 3 ' COMPARISON OF REACTOR BUILDING PRESSURIZATION RESULTS - LOCA Description Primary System Postulated Pine Break Break Location DEPS DEPS DEHL Safety injection Min Max NA
~
RB Spray Min Min NA RB Fan Coolers Min Max NA Peak Pressure (psig) RSG Uprate/FSAR 43.7/44.7 43.7/44.3 45.1/43 Time to Peak Pressure (sec) RSG Uprate/FSAR 18/280 18/350 15/12.1 Peak Temperature ('F) RSG Uprate/FSAR 265.4/266.7 265.4/266.1 267.4/264.'4 Key: RB Design Pressure = 57 psig , RB Design Temperature = 283'F DEPS = Double-Ended Pump Suction ; DEHL = Double-Ended Hot Leg l 3.4-49 j i I
. t FIGURE 3.4.4-5 V C SUMMER NUCLEAR STATION ltt!ACI'Olt lillli. DING PittiSSUltli 5(I 4tl
<y.-
L T a w i3 ! y { 3tl { uJ V d - i d 211 et n. i ill , i il 25
- l(I 15 211 11 5 TIMii(SliCONDS)
_.,_ dot lllt .li fiNDI!D llOT I. fig ilitliAK , 1
- _. _ . . . . _...~._.1.._. . . - , - . , ...;..._..__..4_-._.. .. . . , _ _ _ , , ..___.,...m.
O' , I . FIGURE 3.4.4-1 4 V C SUMMER NUCLEAR STATION
. III3 ACTOR IlulLDING PRIESStJitli 50 49 .
\___.-ar D 1 UI 30 --
t w
$ g tn g 20 - _. -
n. l0 - - - - - - - - - - - - - - - - 0 N.. 0.1 1 10 I(N) - l(MN) IINMMI IINNHN) l(HHHH10 TIMii(SliCONDS) _._ dot 1111.11liNDliD l'tJMP StJCrlON MIN SI
7
'f FIGURE 3.4.4-2 V C SUMMER NUCLEAR STATION ItliACIOlt lillli. DING VAPOR TiiMPlittATI1RIi 3(HI 250 - -- -- - - - - -
~*I E
5 $
$ d 21HI -- - - - - - - - -
m th 150 - - - - - - - - - - - - a
-am itH) 1 10 1(Mi ltMN) l(NWM) l(HHHN) IIHHHHN) 11.1 TIMil(SECONDS)
_ _ DOtJill.E ENDtiD PtJMPSUCI'lON MIN SI ______.-.m__..-._ __________.1_ _ _ _ _ _ _ . _ _ , , , , _ __
~
/ SLB MASS & ENERGY OUTSIDE CONTAINMENT !
, BREAK SIZES FROM 4.6 FT2 TO 0.1 l FT 2 ,
100 % & 75% POWER . 1 RANGES IDENTICAL TO THOSE PREVIOUSLY ANALYZED IN WCAP-10961, REV 1. t i
SLB P&T INSIDE CONTAINMENT SPECTRUM ANALYSIS PERFORMED TO DETERMINE THE MAXIMUM TEMPERATURE AND PRESSURE l CONDITIONS P&T CONDITIONS CALCULATED WITH CONTEMPT-LT26 INITIAL CONDITIONS AND ASSUMPTIONS CONSISTENT WITH CURRENT DESIGN BASIS EXCEPT RBCU CAPACITY REDUCED BY 60%. SUPERHEATED STEAM CONDITIONS RESULT DURING SLBs WITH DRY - BLOWDOWN AS ONLY 8% OF THE CONDENSATE IS ASSUMED TO BE REVAPORIZED
~
OPERATOR ACTION CREDITED AT 20 MINUTES TO TERMINATE EFW TO THE FAULTED SG WHEN WORST SINGLE FAILURE ASSUMED. Y l
. 1
SUMMARY
OF LIMITING SLB CONDITIONS C RB DESIGN PRESSURE = 57 PSIG , RB DESIGN TEMPERATURE = 283 F PARAMETER CURRENT RSG ANALYSES IMPACT OF RSG DESIGN BASIS WITH UPRATE AND UPRATE PEAK RB 45.96 PSIG FOR 53.5 PSIG FOR A PEAK PRESSURE ; PRESSURE A 1.4 FT2 DE 1.4 FT2 DE SLB INCREASES SLB @ 102% OF @ 25% OF 2900 APPROXIMATELY 2775 M.Wt MWt 7.5 PSI DUE TO LARGER SECONDARY I WATER MASS MARGIN TO RB DESIGN PRESSURE DESCREASES. PEAK RB 321.5 "F FOR A 379.2 *F FOR A PEAK ) TEMPERATURE 0.645 FT2 SPLIT 1.4 FT2 DE SLB TEMPERATURE i
@ 102% OF 2775 @ 102% OF 2900 INCREA.SES l MWt MWt APPROXIMATELY 57.5 *F DUE TO NO ENTRAINMENT FOR DER @ 102%-
POWER. I
FIGURE 3.4.4-10 V. C. SUMMER NUCLEAR STATION DESIGN BASIS MSLB ItB PitESSUltE 60 50 n y 40 - - - - -- --- - l
? $
30 . _ . _ _ _ _ . 19 2 ca 20 - - - - - - - -- - - - - - --- c4 10
/ .
0 0.1 1 10 100 1000 1(XX)O TlME(SECONDS)
,_1.4 SO IT Dell @ 25% POWEll
___,,m m.m___ _ ____,_______r. , _,,,s , _ , p. ,, ,,. , , , , . , _ , , , ,
FIGURE 3.4.4-17 V. C. SUMMER NUCLEAR STATION DBA MSLB RD TEMPERATURE VS TIME 400 350 -- -- - -- - - - --- - - - ---- - - - C 300 - -- - - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - e D 250
- a N
I* N g200 . -- - - -- - - - - - a 150 - - - - - 100 , 0.1 1 10 100 1000 10000 TlME (SECONDS) _o_ l.4 FT SO FT DER @ 102% POWER -MSIV FAILURE l. _ _ _ _ _ . ___________.___._.________m________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- _ _ _ _ _ . . _ - . - . _ _. . ._ ., . - . - . . . . _ . . , . _ . . . _ = _ . . . -_ . _
- 1
~
/ MAIN STEAM LINE BREAK !
M&E INSIDE CONTAINMENT I REACTOR POWER: 0, 25, 50, 75 & 100 %. i BREAK SIZES @ EACH POWER LEVEL
- i. FULL DOUBLE ENDED RUPTURE (DER OF 2
1.4 FT )
- 2. SMALLEST DER WITH ENTRAINMENT
- 3. LARGEST DER WITHOUT ENTRAINMENT
- 4. SMALL SPLIT BREAK WHICH REPRESENTS l THE LARGEST BREAK FOR WHICH THE :
ISOLATION SIGNALS ARE GENERATED FIRST BY THE HIGH CONTAINMENT PRESSURE SIGNALS AND RESULTS IN NO l ENTRAINMENT j SINGLE FAILURES CONSIDERED
- i. FW ISOLATION VALVE
- 2. MAIN STEAM ISOLATION VALVE
- 3. FAILURE OF AN EMERGENCY FEEDWATER ,
FLOW CONTROL VALVE TO CLOSE
- 4. ONE TRAIN OF SI i
l
l
)
SLB P&T OUTSIDE CONTAINMENT _ SPECTRUM ANALYSIS PERFORMED TO l DETERMINE THE MAXIMUM TEMPERATURES WITHIN THE COMPARTMENTS OUTSIDE CONTAINMENT. TEMPERATURE CONDITIONS CALCULATED WITH COM. PARE-MODl. INITIAL CONDITIONS AND ASSUMPTIONS CONSISTENT WITH CURRENT DESIGN BASIS ANALYSIS. TEMPERATURE PROFILES CHANGE DUE TO THE N2W M&E AND USE OF COMPARE. ORIGINAL ANALYSES REMAIN _ l APPLICABLE FROM A PRESSURE STANDPOINT. I i
*W' t i
FIGURE 3.4.5-2 COMPARE-MOD 1 MODEL FOR SLD 1.N ! WEST PENETRATION ACCESS AREA - 436 ft. ELEVATION i t i WPAA-463 ATMOSPHERE I l MASS i I ENERGY i P WPAA-436 IB-436 WPAA-412 3.4-77
Figure 3.4.5-5 Limiting Temperatures During an SLB West Penetration Access Area - 436' Elevation 500 . f.
.. / .
g-3; =* ~ ~ + % . - -n. ~ 400 -- j j I
~
f/
/
n 0' s o 00 1 '
+ + + + + 102s FP
/ + * * * *4.61.1f fI tl at ot 1025 F~ P (p
a- Limiting R .!sult s te ' ~ with D3 S is 'p }o : ,* E u _ (D ct 200 - E : (D p _ D Or W 100 : m 6 e e
=
0 200 400 -600 800 Time (seconds)
r Figure 3.4.5-9 Limiting Temperatures Within The West Penetration Access Area During on SLB 500 - -
- ,fW ~** " ' " * -s-,
r 1 400 -
~
e' l
- /
(T
~
m p lt 300 - - v :. 8 k F _
*; -~n a a a h ? : % '
e, o i g., o.200 - E : r
/*j ; ; - ; ; 0 0 g% ,
a) - g_ -.
++~ 4 . 6 f t a t 1025 FP, 4 36'
* ** + + 1.1 fl al 1025 F P, 4 36'
;, -+- 4.6 f t at 75% FP, 463'
.< = + ~ 1.1 fI at 75m FP, 463' '
100 -- --+ + + + + -4 :6-ft et 4025-F P--44 2' -
' 00000 4.6 f t at 75% FP, 412' e
e e O
~
O 200 400 600 800 Time (seconds) _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ - __ _ 3
I l EQUIPMENT QUALIFICATION ) P & T EFFECTS DUE TO HELBs ARE MORE SEVERE LICENSING REPORT INCLUDES : COMMITMENT TO RECONCILE THESE ' IMPACTS ON EQ PRIOR TO THE REPLACEMENT OUTAGE. i l f
I OTHER EVALUATIONS T NSSS COMPONENTS RV RI SG PRZ . RCP'S & MOTORS CRDM RC PIPING & SUPPORTS AUXILIARY & FLUID SYSTEMS CHG/SI RHR ' BOP INTERFACES l MAIN STEAM l MAIN & EMERGENCY FEEDWATER f FUEL , i l l NO ADVERSEIMPACTS i 1 l l l
3.11
SUMMARY
OF TECHNICAL SPECIFICATIONS CHANGES The proposed changes to the Virgil C. Summer Nuclear Station Techmcal Specifications are summarized in Table 3.11-1. Dese changes reflect the impact of the design, analytical methodology, and safety analysis assumptions outlined in this amendment request and are provided in the proposed Technical Specification page changes (See Appendix 3 of this report). A brief overview of the significant changes follows. Core Safety Limits Core safety limits and associated bases for 3-loop operation during modes 1 & 2 (Figure 2.1-1) are revised to reflect the impact of the A75 RSGs and a core power level up to 2900 MWt through
- 1. He increase in the Reactor Core Heat Transfer Rate up to 2900 MWt.
- 2. He reduced Steam Generator Tube Plugging,0% to 10%.
- 3. The application of the Revised Hermal Design Procedure.
RCS Mow The revision proposed for Table 2.2-1 corresponds to the mmtmum measured flow value used as input to the RTDP DNBR analyses for the loss of flow eve.nt. He RCS Total Flow Rate as shown in the Figure 12 of the COLR has been reduced. Indicated RCS flow is derived from the thermal design flow based on the flow measurement uncertamry of 2.1 % including a 0.1 % uncertamty for Feedwater Venturi fouling. Rese changes to the RCS flow rate are proposed to accommodate the following:
- 1. The differences between the old and the new steam generators.
- 2. Up to 10% Steam Generator tube plugging in all threa Steam Generators.
Nerative Htrr Rate Trin The deletion of the Power Range, Neutron Flux High Negative Rate is proposed since this reactor trip
~
function is currently not credited in the dropped rod analysis. The dropped rod analysis is consistent with the WOG program developed in WCAP-11394,* Methodology for the Analysis of the Dropped Rod Event" and the Technical Specification changes are consistent with WCAP-12282,
" Implementation Guidelines for WCAP-11394 (Methodology for the Analysis for the Dropped Rod Event)". -
3.11-1
l t OPAT/OTAT Setnoints
~
Revisions to the limiting safety system settings for the Thermal Overpower AT and Thermal Overtemperature AT trip functions are proposed to maintain consistency with the non-LOCA Accident Analysis. These trip functions provide primary protection against departure from nucleate boiling and fuel centerline melting (excessive kw/ft) during postulated transients. He proposed settings have been based on the new core safety limits anc account for instrument uncertainties. The reference temperatures are now indicated values and the temperature range that was analyzed is specified. Steam Generator Water Isvel Setnoints Revisions to the Steam Generator Water Level Low-Low setpoints are proposed to incorporate the differences between the current Model D3 Steam Generators and the A75 Replacement Steam Generators (e.g., constant level control). He setpoints are given for both Barton transmitters and Rosemount transmitters. The bases also shows an increase to the steam /feedwater flow mismatch activation setpoint. Similar changes are also proposed for the Steam Generator Water Level High-High for Turbine Trip and Feedwater Isolation with Engineered Safety Features. Setpoint changes have been provided for both Barton and Rosemount transminers. Shutdown Marrin for Modes 3. 4. and 5 Figure 3.1-3 of the Technical Specifications defines shutdown margin requirements as a function of average RCS boron concentration during Modes 3,4, and 5. The proposed revisions are driven by both the steam generator replacement and the increase in the NSSS power limit and are required to > maintain the current bases of the Technical Specifications. DNB Parameters The proposed changes to the DNB related parameters (T , and Pressurizer Pressure) assure that each parameter is maintamed within the normal steady state envelope of operation assumed in both the transient and the accident analysis. He proposed changes are consistent with the accident analysis. Both parameters now represent indicated values by including allowance for reading and averaging three control board indications. Pressurizer pressure has been changed to gauge pressure since these units are read from the control boards. , Steam Generator Surveillance He replacement of steam generators requires that a first inservice inspection of the steam generators be performed again as required in Technical Specification 4.4.5.3. To add clarity to this surveillance, the first sentence of the surveillance which describes when the "first inservice inspection" should take place has been modified. The surveillance must take place after 6 Effective Full Power Months 3.11-2 i
s f i (EFPM) from the time of the replacement but within 24 calendar months of initial criticality after the steam generator replacement. The F*and the L* criteria have been deleted from the Technical Specifications since these criteria are nat applicable to the A75 replacement steam generators. He Model D3 SG sleeving process reports have also been removed since these reports are not applicable to the A75 replacement steam generators.
~
Enactor coolant system volume , The combined water and steam volume of the Reactor Coolant System at an indicated Tw condition has been increased to address the differences between the old and new steam generators. Containment Pimm e ne limiting conditions for operation and surveillance requirements for the contamment systems are dependent in part on the manmum peak pressure during an accident. The proposed values for P, and P reflect the maximum containment pressure calculated for the SLB accident with the A75 SGs. l l 3.11-3 l l j 1 i
2
~
(2 % 663) l- l l. Unacceptacle 660 l Ooeration 4 N 2450
\ PSIA ,
[2 % 651) 55 N N l . 640 - (2%,634) - l 22 pq 630
? N X \
i e20 l X l 20 N \
%616)l l 3
.(120 % 618) -
1' N N I \ i
< N '
'a N \
(120 % 605) M l o 600
\ !
\
590 N' (120 % 590) f 580 l Acceptabb I Operaton T
\
i > 570 l (120 % 573) a 560 ' ' O 20 40 60 80 100 120 -i Power (Percent) When operating in the reduced RTP region of Technical l Specification 3.2.3 the restricted power level ! rnust be considered 100% RTP for this figure. Figure 2.1-1 Reactor Core Safety Lunits - Three Loop Operaton .
- S M R - UNIT 1 2-2 Amendmant No. 45, 75 f
- e. .;
r
. 1910Md BL80!LHLl!!E_11531!Ll!!!guntg!AIlLm IglP .5t!r0lgis m
I Irlp Setpoint Alloweble Value
- Functional Unit gygo,,, ,gygg Z S c
5 .. 1. Manual Reactor irly llot Applicable 18 4 NA NA NA s
-4
-- 2. Power Re . Ileutron Flum High 5 Int 7.5 4.56 0 1109% of RTP $111.21 of RIP tow 5etpoint; 8.3 4.56 0 $25% of RIP $27.2% of RTP
- 3. Power Range Neutron Flum 1.6 0.5 0 251 of RTP with a Llee 2635 of RIP with a time l'* k t
- A Nigh Positive pate constant 12 seconds constant 22 seconds 7
\ \ 5
- 5. Intermediate Range. 17.0 8.4 0 1255 of RIP 5315 of RIP Neutron Flun
- 6. Source Range. lieutron Fluu 17.0 10.0 0 5805 cps $ 1.4 m 105 cps 3
~
- 7. .Overtemperature ai . . 87 L See note 1 fl. See note 2- l Q'l.7 2i &W e. 3 i g 8. Overpower ai M M My See note 3 Q See note 4 11870 psig 11859 psig
- 9. Presseriter Pressure-Lou 3.l @ 0.71 1.5 4dij $239fpsig g 10. Presseriter Pressure-Nigh M , @ $2300 psig' x II. Pressurlier lister Level-High 5.0 2.18 .5- <gt1 of instrument <93.8% of instrument e cii;;;/ 9.., ;,.n i,.n l 100.g% of loop design 1.40 3905 of loop design I h 12. toss of Flow - 2.5 .6 flow
- flow *
= s
.b *toop'de' sign fleu = h l4- o SP" I
? Alf . RAIEt ilIEN14L POIER g * $ .98 span for Bette-T (Ales) and @ for Pressurlier Pressure.
M - __ ___ __ _ __
i J ( ] n0VER DISTRIBUTION LIMITS ! 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHAt:PY RISE HOT CHANNEL FACTOR t LIMITING CONDITION FOR OPERATION i 3.2.3 The concination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowanle operation as specified in the CORE OPERATING LIMITS REPORT (COLR) figure nntitled RCS Total Flow Rate Versus R For Three Loop Operation. i p ! Where: ' ( mcivde3 o, y,
- a. R= '
f u b y % , m ,*, Fg [1.0 + PFj (1.0 - P)] '
- b. P = THERMAL POWER ,
y, J) , RAlw THERMAL POWER
- c. F
= Measured values of Fh obtained by using the novable incere -
detectors to obtain a power distribution map. The esasured values of F shall be used to calculata R since the RCS Total , Flow Rata Versus R figure in the COLR includes esasurement ' uncagtainties of 2.1% for flow and 4% for incore measurement' , of Fg , and T
- d. F =TheFhlimitatRATEDTHERMALPOWERspecifiedintheCOLR.
- e. PFg = The Power Factor Multiplier specified in the COLA.
APPLICABILITY: MODE 1. ACTION: With the combination of RCS total flow rata and R outside the region of accept-able operation specified in the COLR: ,
- a. Within 2 hours either: t
- 1. Restore the combination of RCS total flow rate and R to within the above limits, or ,
- 2. Reduce THERMAL POWER to less than 505 of RATED THERMAL POWER i and reduce the Power Range Neutron Flux - High trip setpoint to i less than or equal to 55% of RATED
- THERMAL POWER within the next 4 hours. '
- b. Within 24 hours of initially being outside the above limits, verify through incere-flux espping and RC. total flow rate comparison that the coenination of R and RCS total flow rata are restored to within the above limits, or reduce THElWAL POWER to less than 5% of RATED ,
THERMAL POWER within the next 2 hours. i SUPMER - UNIT 1 3/4 2-8 Amenchment No. f), gg, 75, 88 e P
POWER DISTRIBUTION LIMITS ! f LIMITING CONDITIdN FOR OPERATION ACTION. (Continued) c. i to increasing THERMAL POWER above the red required by ACTION items a.2. and/or b. above; subseouent POWER OPERATION may proceed provided that the combination of R anc ! indicated RCS total flow rate are demonstrated, through incore flux' - mapping and RCS total flow rata coeparison, to I following THERMAL POWER levels: l ,
- 1. !
A nominal 50% of RATED THERMAL POWER,
- 2. ;
A nominal 75% of RATED THERMAL POWER, and 3. Within 24 hours of attainik greater than or equal to 95% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 deterimined Theto combination of indicated RCS total flow rate'and be within the region of acceptable operatica specified in R shall be ! the COLR. ' a. Prior loading,to operation and above 75% of RATED THERMAL POWER af ter each fuel 1 i b. At least once per 31 Effective Full Power Days. a.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptabic $wration.specified in the COLR at least once per 12 hours . [ when the most is assumed recen:' T obtained value of R obtained per Specification 4.2.3.2 to exisi I
]
4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. l l 4.2.3.5 once perThe RCS total flow rate shall be determined b esasurement t least 18 months. I
%*.mA.
bO o.cc e,
- d. 3 .90 %
RATE b THer.mA Pow iE R StBMER - UNIT 1 3/4 2-$r Amenment No. A5, 75, 88
^
4
~
l i OVERTEMPERATURE DELTA-T TRIP SETPOINTS . 4 ATsAT, Ks-K2l, (T- T] +Ks(P-P] -f1(AI) Trip Setpoint Allowable Value Current Value Proposed Value Current Value Proposed Value K i 1.1.195 S_1.23 12.2% Delta-T same.as current 1 Span value K, 10.03006/*F 10.0292/*F
- K, 10.00147/ psi 10.00161/ psi T' 1587.4 *F 572
- FAT"1587.4
*F ,
Cunent Value Proposed Value i) 0 for Delta I between -24% & 4% i) 0 for Delta I between -35% and 6% , ii) Decrease the setpoint by 2.27% ii) Decrease the setpoint by 2.46% f-delm I Penaly for each pcreent that the magnitude for each percent that the magnitude of Delta I exceeds -24% of Delta I exceeds -35% iii) Decrease the setpoint by 2.34% iii) Decrease the setpoint by 3.29% for each percent that the magnitude for each percent that the magnitude i of Delta I exceed 4% of Delta I exceed 6% 4 m W
OTDT F-del ll'A I PENAL!l'Y 100 - 90 NEW @ 1297. PER %Di 80 . ' CURRENI @ 2.34% PER %DI 70 G " 60 w [ w 50 CURRENT @ -2.27% PER %DI 40 NEW @ -2.46% PER 7. DI ]7 . 30 20 10 -
-56 -46 -Id -I6' '.'6
-1 l'0 ' 2'O 30' '40~ 50
-% del!I'A I
OVERPOWER DELTA-T TRIP SETPOINTS i 1 ATSATo Ka - Ksp* T-Ks(T- f) Trip Setpoint Allowable Value . Current Value Proposed Value Current Value Proposed Value K. I 1.0875 1 1.078 12.4% Delta-T 12.3% Delta-T Span Span , K5 10.02/*F same as current value K. 10.00156/ psi -10.00198/ psi ; T" < 587.4 "F 572 *F$T"5587.4 *F i i C I I l i
MARGIN TO TRIP MARGIN TO TRIP ON OTDT EVENT @ 572 F @ 587.4 F , 100% LOAD -13 % -4 % REJECTION 50% LOAD 6% 11 % REIECTION LOSS OF A FW 9% 16 % PUMP i
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4 J INSERT F The maximum indicated Tavg limit of 589.2*F and ihe minimum indicated pressure limit of 2206 psig correspond to analytical limits of' 591.4*F and 2185 psig respectively, read from control board indications. l 4 en O I l l l l
0 DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core small contain 157 fuel assemolies with ea g.' din , except that limited substitution of fuel rods by ofkircaley-4,IIRLDalloy,stainlessstet) o justified by a cycle specific reload analysI.s.r by vacancies, may be made if -l nominal active fuel length of 144 inches. Eacn fuel roa shall have a a maxinua enrichment of 3.2 weight percert U-235.The initial core loaoino shall have ment of 4.25 weight percent U-235.in pnysical design to the initial c CONTROL R00 ASSEE LIES 5.3.2 The reactor core shall contain 48 full length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches o absorber material. silver, 15 percent indfue and 5 percent ca mium.The nominal values of 3 clad with stainless steel tubing. All control rods shall be * ' 5.4 REACTOR CO0lANT SYSTEM ' DESIGN PRES $URE AND TEMPERATURE 5.4.1 The reactor coolant systas is designed and shall be maintained: a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the appifcable Surveillance Requirements,
- b. Fer a pressure of 2485 psig, and ,
c. For a temperature of 650*F, except for the pressurizar which is
%N 680*F.
VOLUME ,
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5.4.2 4W The total water and staan volume of the reactor coolant system is 100 cubic feet at a ---'--' ' avg """ " '
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- 5. 5 METEOROLOGICAL TOWER LOCATION 5.5.1 The seteorological tower shall be located as shown er. Figure 5.1-1. '
~ :ndic2W T 3 j of SBT M '
SU M ER - UNIT 1 5-6 Amenoment No. 27. H , f2
~ 75, 105 ,
I 1 l
l Startup . Testing Program r Mark Mi ner l W l i
. \
l START-UP FUNCTIONS 1 e COORDINATE STARTUP ACTIVITIES 1 INTERFACES l LICENSING WESTINGHOUSE i OUTAGE SCHEDULING. I e FACILITATE PLANT OPERATIONS STARTUP PREPARATION l PLANT OPERATIONS NUCLEAR TRAINING i COMPUTER SERVICES e IDENTIFY AND DESIGNATE POST MODIFICATION TESTING REACTOR ENGINEERING TEST UNIT I&C GROUP DESIGN ENGINEERING ! BECHTEL ENGINEERING '
l i V. C. SUMMER START-UP ! FOR STEAM GENERATOR REPLACEMENT j e V. C. SUMMER LICENSED OPERATING : FACILITY ) L e STEAM GENERATOR REPLACEMENTIS A LARGE MODIFICATION OF EXISTING EQUIPMENT e MAJORITY OF TESTING 15 IDENTIFIED AND PERFORMED AS POST MODIFICATION TESTING e FSAR AND REG GUIDE 1.68 REV. 0 . WERE REVIEWED FOR FUNCTIONAL TESTING REQUIREMENTS . e SOME TESTING WILL BE REQUIRED FOR WESTINGHOUSE WARRANTY REQUIREMENTS F * - - ~ ' - n+ - m - v -
I TEST SELECTION METHODOLOGY l e REVIEWED ALL STARTUP TESTS DESCRIBED IN CHAPTER 14 FSAR e MANY TESTS ARE ROUTINELY I PERFORMED POST REFUELING I e IDENTIFIED THOSE INTEGRATED TESTS j WHICH MIGHT NEED TO BE 1 RE-VERIFIED AFTER STEAM GENERATOR REPLACEMENT l e THE NEED FOR SOME OF THESE TESTS WERE ELIMINATED THROUGH ANALYSIS e- FINAllZED THE LIST OF TESTS REQUIRED FOR POST REPLACEMENT STARTUP F V V - - - - -
o i 4 START-UP TESTING Modes 3, 4 Instrument Tube Gap Measurement Hot Gap Measurement S/G Supports RCP Tie Rods RCS Hydro RB Ambient Heat Load RVLIS Calibration EFW Functional Testing Mode 1 Calorimetric & Precision RCS Flow Measurement Data Collection Instrumentation Verification Power Ascension (FW Functional S/G Level Control Steam Dump Ops Moisture Carryover 1
O Other Issues April Rice ; 1 l l I l l l
I r V. C. Summer Nuclear- Station . Steam Generator Replacement Project j [M$$M1Middd$$$$N$$d u l O Piping Analysis for RCS, MS, FW, EFW j and BD
/ SCE&G is Interested in Feedback o' n the l March 12,1993 Piping Analysis Letter ;
i
/ March 12 Letter Describes. Piping Analysis .
Performed to Support S/G Replacement and : 4 Snubber & Whip Restraint Reduction l i
/ This Activity is Being Performed.under ;
10 CFR 50.59. ,
/ Utilizing Leak before' Break, Generic Letter. !
87-11.(Elimination 'of ArbitraryJn'termediate Breaks) and' Code Case N .411'(RCS Piping) l l RCS Piping Analysis'is also Described in--the
~
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October 29,1993 Licensing ' Submittal i r h ~
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s i V. C. Summer Nuclear. Station l Steam Generator Replacement Project NkbbMN$ N$N$$h5h[$hNd t 9 Retired Steam Generator Recycle Facility l t
/ SCE&G has Decided to Store-the Retired Steam Generators on Site
/ S/G Recycle Facility will be Similar to Other Plants
/ Key Features . . . l l
Outside the Protected area but Within the ! Exclusion Zone Reinforced Concrete Structure (ACI/ ASTM Codes) Single Locked Entry i Doses External to the Building Meet 10 CFR 20 and
- and 40 CFR 190 Requirements ,
Sump Provided for Water Collection and Sampling
/ SCE&G is Working with South Carolina .
Department of Health & Environmental-Control on Permit / Licensing Requirements
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SURFACE SEE NOTE 3 , 1 Figure 2 Building Layout I 1 l l
0 P n t Plans for : Power Uprate l 9 I Ron C ary
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V. C. Summer Nuclear Station Uprate Project [ [i.
. j r g p: Q yt g g d h g Q E s A g . . J ,; ) j Modifications to Support Uprate . . .
9 NSSS Evaluations Majority were Completed for S/G Replacement Best Estimate LOCA S Balance of Plant Evaluations Non-Safety Related Hardware Changes Only g Environmental Evaluations f i 9 Core Design 9 Turbine Building Open Cycle Cooling 9 Skimmer Wall Addition P S Heater Drain Valve Modification 9 MSR Forth Pass Drain Line Support 9 Condensate System Upgrade 9 High Pressure Turbine Upgrade 9 Turbine Building HVAC Upgrade e Low Pressure Rotor Replacement l
Steam Generator Replacement Project Uprate Summary Schedule 1993 1994 1995 1996 1997 Uprote Feasibility Studies y ' Uprate Modifications Conceptual Design Phase I Uprate Modifications Design Phase l I Licensing Submittal Development l I DHEC/NRC Approval of Uprate I ] On-Line implementation of Secondary Side I improvements Ref uel 9 Modification Implementation O , implementation of HP Turbine Modification 0
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