RC-95-0174, Application for Amend to License NPF-12,consisting of TS Change Request TSP 950001,permitting Plant to Operate at Uprate Power Level of 2,900 Mwt Core Power.Proprietary Best Estimate Analysis... Encl.Analysis Withheld
| ML20091L444 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 08/18/1995 |
| From: | Furstenberg S SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Stephen Dembek NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20091L452 | List: |
| References | |
| RC-95-0174, RC-95-174, NUDOCS 9508290202 | |
| Download: ML20091L444 (27) | |
Text
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South Cirolina El;ctric & GIs CompIny Virgit C. SummIr Nuctrir Stition S. J. Fur:unberg P. O. Box 88 MInager, Jenkinsville, SC 29065 Maintenance Services (803) 345 5209 ASCNntCwpany August 18,1995 RC-95-0174 Doctunent Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. Stephen Dembek Gentlemen: ,
Subject:
VIRGILC. SUMMER NUCLEAR STATION (VCSNS)
DOCKET NO.50/395 OPERATING LICENSE NO.NPF-12 TECHNICAL SPECIFICATION CHANGE REQUEST-TSP 950001 PLANT UPRATE South Carolina Electric & Gas Com a (SCE&G), acting for itself and as agent for South Carolina Public Service Au ,hereby requests an amendment to the Virgil C. Summer Nuclear Station (VC NS) Operating License and Technical Specifications (TS). This change is being submitted pursuant to 10 CFR 50.90. The proposed changes will permit VCSNS to operate at the uprate power level of 2900 MWt Core Power when the unit is restarted after the ninth refueling outage.
Amendment No.119 implemented Technical Specification changes to sup? ort VCSNS operation with replacement steam generators. These changes ref ect the impact of the rmosed power uprate since t ae majority of the supporting analyses and evaluatir '2bmitted were performed at a core power level up to 2900 MWt.
This submitte ' ravides additional analyses, evaluations and Technical Specifications c.anges to supplement those submitted for the replacement steam generators. When implemented, the proposed changes along with those previously approved for steam generator replacement will preserve the design, analytical methodology, and safety analysis assumptions outlined in this amendment request.
As a result of a comment in the Safety Evaluation Report approving steam generator replacement, additional reviews of Radiological Consequences were performed. It has been determined that a change to the maximum quanti,ty of radioactivity that !
can be stored in the Waste Gas Storage Decay Tank is required. Although not specifically an Uprate issue, a proposed change to 3.11.2.6 is included in this submittal to permit closure of this item. Operating License NPF-12, Section 2.C.1,
" Maximum Power Level," will require revision to reflect uprate power limits.
The analyses and evaluations within Attachments I, II, and VI provide the basis for the Safety Evaluation (Attachment IV) and No Significant Hazards Evaluation (Attachnient V) supporting the proposed Technical Specification changes outlined in AttachmentIII. maammanantVI5iinteisiWCT.%Infbrrstiour AttachmentI List of required analyses and evaluations for power uprate >
and those which have been reviewed and approved for Steam Generator Replacement f
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!!S n n a ,
9508290202 950828 PDR P
ADDCK 05000395 PDR f'
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Documont Contrsl Desk !
- TSP 950001 i RC-95-0174 7 Page 2 of 3 . ,
5 AttachmentII _ Evaluations Not Performed Previously For Power Uprate
.'A*-hmentIII DescriptionofChangesChart i Marked-up Technical Specification Pages Revised Technical Specification Pages g AttachmentIV Description of Amendment Request and the Supporting SafetyEvaluation 3 AttachmentV Description of Amendment Request and the Associated No Significant Hazards Evaluation ;
i AttachmentVI "VCSNS RSG/Uprating: Best Estimate Large Break LOCA - j Final Analysis Report' !
Specification chan!l
-SCE&G request (Tkuests NRC 15,1996, R) by January review and approval in order to support ofplanning this Technical and scheduling activities for the ninth refueling outage. ;
This proposed TSCR has been reviewed and approved by both the Plant Safety l Review Committee and the Nuclear Safety Review Committee.
l I declare that these statements and matters set forth herein are true and correct i to the best of my knowledge, information, and belief. !
Should you have'any questions, please call Mr. Philip Rose at (803) 345-4052 at ;
your convenience, i i
Ver trulyyours, l S. .
f rstenberg
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PAR /SJJ/dr
. Attachments (6) ;
c: J. L. Skolds ' (w/o Attachments) J. B. Knotts Jr. l O. W. Dixon M. K. Batavia R. R.~ Mahan (w/o Attachments) ' J. I. Byrd '
l' R. J. White CentralFile System !
S. D. Ebneter. . RTS (TSP 950001) j
, . NRC ResidentInspector ' File (813.20) ;
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NUCLEAR EXCELLENCE - A SUMMER TRADITION! ;
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STATE OF SOUTHCAROLINA :
- TO WIT : l COUNTY OF FAIRFIELD :
Ihereby certify that on the A da f [h3E19Thefore me, the subscriber, a
] Nota Public of the State of South Carolina, personally appeared S. J. Furstenberg, :
1 bein uly sworn, and states that he has signature authority as desi !
Vic resident, Nuclear Operations of the South Carolina Electric Gas Company, &gnated a by the t corporation of the State of South Carolina, that he provides the foregoing response for >
the purposes therein set forth, that the statements made are true and correct to the
, best of his knowledge, information, and belief, and that he was authorized to provide
- the response on behalf of said Corporation. , ;
WITNESS my Hand and NotarialSeal %bd O Notarf Public v
My Commission Expires A W A Date
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NUCLEAR EXCELLENCE - A SUMMER TRADITION!
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. Document Control Desk
- Attachment I
- ! Analysis / Evaluation Amendment 119 to Uprate Power Operatm, License Submittal
- No. N3F-12 NSSS Design Transients s 2900 MWt X
! Control System Setpoints s 2900 MWt X j Protection System Setpoint s 2900 MWt X Large Break LOCA s 2775 MWt using BASH X Large Break LOCA s 2900 MWt using X BELOCA Small Break LOCA s 2900 MWt X l
i Post-LOCA Long Term Core Cooling X j Suberiticality s 2900 MWt t
. Hot Le Switchover to Prevent Potential X
! Boron frecipitation s 2900 MWt 1
LOCA Hydraulic Forces s 2900 MWt X l
Non-LOCA Analyses s 2900 MWt X i LOCA M&E Releases (Long and Short Term) X
! s 2900 MWt i Short Term Containment s 2900 MWt X
! MSLB M&E s 2900 MWt X 1
l Long Term Containment s 2900 MWt X l
! Environmental Conditions s 2900 MWt X l SGTR s 2900 MWt X Reactor C. .vity Pressure s 2900 MWt X Radiological Consequences s 2900 MWt X X Source Terms s 2900 MWt X Primary Components s 2900 MWt X l
l Leak-Before-Break s 2900 MWt X Fluid Systems s 2775 MWt X Fluid Systems s 2900 MWt X i -
Document Control Desk AttachmentI TSP 950001 RC-95-0174 i
Page 2 of 2 RSG/UPRATE POWER ANALYSES & EVALUATIONS SER for Amendment 119 to Uprate Power i Analysis / Evaluation License Submittal Operatin.;F-12 No. NP Auxiliary Systems s 2900 MWt X X Fuel Structural s 2900 MWt X Impact on IPE s 2900 MWt X Technical Specification Changes for RSG X
, Technical Specification Changes for Uprate X 1 Power Operation l l
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Page 1 of 22 i TABLE OF CONTENTS l EVALUATIONS NOT PREVIOUSLY PERFORMED ,
FOR POWER UPRATE ;
i Title Page i 1.0 Introduction-Description of License Amendment Request 2 ;
2.0 Basis for Evaluations / Analyses Performed 3 3.0 Safety Evaluations / Analyses 6 3.1 Large Break Loss of Coolant Accident Analyses 6 I 3.2 Fluid and Auxiliary Systems Evaluations 7 3.3 NSSS/ BOP Interface 9 ,
3.4 Gas Storage Tanks 11 3.5 Balance ofPlant Systems 12 4.0 Impact on the Individual Plant Examination (IPE) 18 5.0 EnvironmentalEffects 19 !
6.0 Equipment Environmental Qualification and High Energy 20 Line Break Impact 7.0 Other Systems Evaluated But Not Affected by Uprate 21 j 8.0 Conclusions 21 9.0 References 22 i TABLES Table 2.1-1 Desi Performance Capability Parameters for 4 !
VCSkS Uprate Power Operation with A75 Steam Generators j Table 2.1-2 Best Estimate Operating Condition Parameters 5 For VCSNS Uprate Power Operation with A75 Steam Generators .
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1.0 INTRODUCTION
-DESCRIPTION OF LICENSE AMENDMENT REQUEST j l l This document contains the remaining safety analysis and evaluation results to ;
support uprate power operation of the Virgil C. Summer Nuclear Station !
4 (VCSNS). As noted in the Replacement Steam Generator (RSG) submittals (Reference 3), SCE&G decided to analyze the plant for a range of operating conditions to provide operational flexibility. Accident analyses were performed :
at a core power level up to 2900 MWt (a 4.5% increase). SCE&G received a l Safety Evaluation Report (SER) (Reference 1) from the NRC for the analyses '
and evaluations performed to support the RSG and utilized the Engineered Safeguards design rating ("uprate" power rating) of 2900 MWt core power. At .
that time, however, SCE&G did not seek approval for uprate power operation !
- since additional areas needed to be addressed at the 2900 MWt core power level. l There are three significant areas (not included in the RSG SER) that remain to i be addressed to implement the uprate power operation at VCSNS. These areas ,
i are Large Break LOCA using the Westinghouse Best Estimate methodology, :
NSSS Fluid Systems, and Technical Specification (TS) changes. Additionally, i the calculation for Waste Gas Decay Storage Tank Rupture was evaluated and
- determined to require a TS change, as well as reviewing the calculations for .
Spent Fuel Coolingcapability. j
, In order to permit more flexible plant operation, and support a coastdown at the
~3 end of the fuel cycle, a range of full power nominal Tavg values from a maximum value of 587.4*F to a minimum of 572.0*F was analyzed. Normal plant operation is expected to be 587.4*F (or slightly less). Thermal Design flow
- will be reduced to 92,600 gpm/ loop, to support up to 10% steam generator tube i
plugging. Minimum measured flow will be 283,500 gpm. Table 2.1-1 further 7 delineates the design performance capability parameters for VCSNS with the >
A75 SGs and uprate power operation. All other accident analysis results were submitted via Reference 3 and approved by the NRC in Reference 1. During the -
NRC review, the radiological consequences analysis for waste gas decay storage tank rupture was questioned, prompting SCE&G to evaluate the methodology -
used. As a result, the maximum quantity of radioactivity stored in any one tank is being reduced. l 1
This amendment request reflects the impact of the design, analytical methodology, and safety analysis assumptions on the VCSNS Technical ,
Specifications for the uprate power operation. l The proposed additional changes to the VCSNS Technical Specifications are addressed in Attachment III. These changes reflect the impact of the uprate
. power operation. When implemented, the proposed changes along with the
! previously approved Technical Specification changes (Reference 1) will preserve the design, analytical methodology, and safety analysis assumptions outlined in ,
this amendment request.
Documtnt Control D:sk Attachm:ntII TSP 950001 RC-95-0174 Page 3 of 22 2.0 BASIS FOR EVALUATIONS / ANALYSES PERFORMED 1
The analyses and evaluations performed to support the uprate power bound a !
range of operating conditions for VCSNS. Four cases are presented which i define a range of primary operating temperatures from 572 F to 587.4*F and a l range of steam generator tube plugging levels from 0% to 10%. This will j provide SCE&G with the flexibility to select the appropriate primary l temperatures on a cycle-by-cycle basis necessary to achieve full megawatt !
electric output and to adjust the temperature as necessary to perform end-of- I cycle Tavg coastdown.
DESIGN POWER CAPABILITY PARAMETERS ;
Design power capability parameters were developed for the VCSNS to l encompass both the A75 Replacement Steam Generators (RSGs) and the uprate power level (2900 MWt Core Power). The parameters developed are bounding for the lower power level of 2787 MWt NSSS (2775 MWt Reactor Core) that is _
the current licensed power for VCSNS. The safety analyses presented in this l submittal considered the case (s) which is most conservative for the specific analysis areas. The parameter cases are provided in Table 2.1-1 and are explained in detailbelow. t Cases 1 and 2, calculated for 0% and 10% Steam Generator Tube Plugging j (SGTP), respectively, incorporate the conservatively low Reactor Coolant ;
System (RCS) Thermal Design Flow (TDF) (92,600 gpm/ loop), as well as the value of 587.4*F. The TDF of 92,600 gpm/ loop was >
current licensed selected such that aTa7 equate margin (approximately 8%) exists between TDj and best estimate predictions of RCS flow, assuming the A75 steam generator -
with 10% SGTP. The RCS Best Estimate Flow (BEF) is based on A75 steam :
generator hydraulic characteristics, reactor coolant pump performance curves, and RCS pressure drop data. Cases 1 and 2 are used for those analyses [e.g.,
non LOCA, Departure from Nucleate Boiling (DNB)-related] where high RCS ,
temperatures and low RCS flow are bounding.
Cases 3 and 4 (0% and 10% SGTP) incorporate TDF and the lowest reactor vessel Tavg considered,572'F. The reduced temperature conditions allow for ;
constant operation at reduced temperatures, or end-of cycle Tavg coastdown t capability. These cases are used for analyses where low vessel inlet >
temperature is bounding (NSSS design transients for the cold leg) or where low ,
steam pressure is bounding (e.g., consideration of pressure drop across the steam generator tubes).
l Table 2.1-2 provides the Best Estimate Operating Condition Parameters. These l are the parameters that will be expected following startup from RF-9. VCSNS ;
intends to operate with Tavg = 587.4'F. The steam parameters have been :
calculated as best estimate for 100% power and will be used to predict actual performance.
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. 1 l DESIGN PERFORMANCE CAPABILITY PARAMETERS FOR VCSNS !
- UPRATE POWER OPERATION WITII A75 STEAM GENERATORS !
I For all parameter cases:
Parameter i
! NSSS Power, MWt 2912 Non-Nuclear NSSS Power MWt* 12 Core Power,MWt 2900 Core Bypass Flow, % 8.9 -
RCS Desirn Pressure, psia 2250 Thermal Design Flow, gpm/ loop 92,600 '
Minimum Measured Flow, gpm total 283,500 '
! Best Estimate Flow,gpm/ loop 102,600 j Mechanical Design Flow, gpm/ loop 107,100
- Fuel Design VANTAGE + (V +) t FuelPeaking Factors Fo = 2.45, Fan = 1.62 -
Positive Moc terator Temp. Coef., pcm/ F +7 j Includes heat input from RCP and other non nuclear sources. ,
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b High Tave Cases: Low Tavg Cases:
1 Parameter Case 1 Case 2 Case 3 Case 4 CoolantTemperatures, *F Core Outlet 627.7 627.7 613.5 613.5 Vessel Outlet 621.9 621.9 607.4 607.4
- Core Average 592.8 592.8 577.1 577.1 Vessel Average 587.4 587.4 572.0 572.0
- Vessel / Core Inlet 552.9 552.9 536.6 536.6 Zero Load 557.0 557.0 557.0 557.0 Steam Generator Feedwater Temperature,'F 440.0 440.0 440.0 440.0 Moisture Carryover, % 0.1 0.1 0.1 0.1 Steam Temperature,*F 540.4 538.4 523.7 521.7 SteamPressure psia 966 950 839 824 Steam Flow, miilionIb/hr. 12.84 12.83 12.77 12.7 Tube Plugging, % 0 10 0 10 t
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1 BEST ESTIMATE OPERATING CONDITION PARAMETERS FOR VCSNS UPRATE POWER OPERATION WITH A75 STEAM GENERATORS Parameter Reactor Vessel Outlet Temperature (TIIOT) F 618.8 Reactor Vessel Inlet Temperature (Tc0LD)*F 556.1 Average Temperature (Tavg) F 587.4 Steam Temperature *F 544.6 Steam Pressure PSIA 1000 Steam Flow MillionIb/hr 12.86 4
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- i 3.0 ; SAFETY EVALUATIONS / ANALYSES j j 3.1 LARGE BREAK LOCA J
Introduction:
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- SCE&G replaced their Model D3 steam generators at the VCSNS with Delta 75 ;
j steam generators. The A75 SGs have been addressed previously for the LOCA- ,
- related accident analyses in Reference 3. Except for the large break LOCA '
analysis performed with the Westinghouse 1981 Evaluation Model with BASH, the remaining analyses (small break LOCA, loop / vessel hydraulic forcing :
, functions, post-LOCA long term suberiticality/ minimum flow, and hot leg .
l switchover) addressed a core power of 2900 MWt. - As such, the purpose of this 1' 3
licensing submittal is to address the large break LOCA analysis performed at
- i. uprate power with the Westinghouse Best Estimate LOCA (BELOCA) !
i Evaluation Model (EM) with WCOBRA/ TRAC (Reference 9).
AnalysisInnut:
f l To address the VCSNS core power of 2900 MWt, the large break LOCA analysis ,
[ was performed with the Westinghouse BELOCA Evaluation Model. This :
i analysis is documented in Reference 10 and, in addition to the uprate power level, includes: ;
V A75 RSGs with up to10% tube plugging 1- Totalpeaking factor (Fg) of 2.5 Hot channel enthalpy nse factor (Fan) of 1.7 -
17x17 VANTAGE + fuel (ZIRLOm) / once-burned VANTAGE 5 fuel (Zirc-4) !
Vessel average temperature range of 572.0-587.4*F 15.3*F uncertainty ;
.; Maximum Accumulator /RWST water temperature of110*F Results: i f ;
The BELOCA analysis demonstrates compliance with the peak clad temperature (PCT), local metal-water reaction and core-wide hydrogen !
> generation criteria of 10 CFR 50.46 by confirming, with a high probability, that
! these Acceptance Criteria will not be exceeded. This requires calculation of a 95th-percentile PCT which is a combination of four categories of uncertainty: ;
- 1. Code bias and uncertainty accounts for differences between code predictions of tests which simulate the LOCA transient and measured :
data. This is independent of the plant-specific application. ,
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- 2. Power distribution' bias and uncertainty are determined from plant- ,
, specific sensitivity analyses of fifteen power shapes which bound all ,
! possible power distributions which may occur during normal plant :
operations. .
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- 3. Break flow bias and uncertainty are determined from plant-specific !
sensitivities to break disdarge coefficient and broken loop resistance. A >
- total of nine calculations are performed for discharge coefficients of 0.8, !
1.0, and 1.2 for a range of broken loop resistances. A split break is also l'
. analyzed to confirm whether the double ended cold leg guillotine break 1 . remains limiting. t l 4. Initial condition bias and uncertainty are determined from plant-specific ;
j sensitivities to key LOCA parameters. ,
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- All of the plant-specific sensitivities performed as part of the 95th percentile i PCT calculation and the combination of the four categories of uncertainty are ;
detailed in Reference 9. The resultant 95th percentile PCT of 2068'F, l maximum local cladding oxidation less than 10%, and maximum total cladding oxidation less than 0.25%, demonstrate that there is a high probability that the ,
acceptance criteria of 10 CFR 50.46 would not be exceeded in the event of a large break LOCA. This high probability is based on a 95% probability with a ,
i 95% confidence level.
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Conclusions:
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! The VCSNS uprate core power level of 2900 MWt with the A75 Steam ;
. Generators has been evaluated for the Large Break LOCA analysis indicated i i above. It was determined that there is a high probability that the acceptance I criteria of 10 CFR 50.46 would not be exceeded in the event of a large break LOCA and that the acceptance criteria of 10 CFR 50.46 would not be exceeded as demonstrated by the remaining LOCA related analyses.
3.2 FLUID AND AUXILIARY SYSTEMS EVALUATIONS 3.2.1 Introduction
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The impact of the RSG/Uprate Power program upon the Nuclear Steam
- Supply System (NSSS) Fluid Systems, NSSS auxiliary equipment and the NSSS/ Balance-of-Plant interface systems was performed for the 4
- bounding plant operation cases presented in Table 2.1-1. The NSSS Fluid
- Systems are comprised of the Residual Heat Removal System (RHR) and 4
Chemical and Volume Control System (CVCS). For the initial licensing submittal regarding the RSG program (Reference 3), the auxiliary
- equipment evaluation was performed for an NSSS power level of 2912 i M W t.
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i 3.2.2 Discussion of Evaluations Performed )
- i i 3.2.2.1 Fluid Systems Evaluations ;
i Residual Heat Removal System i
The RHR System is designed to remove residual and sensible -
j heat from the core during the second phase of plant cooldown. ;
The RHR System was designed to reduce the temperature of the '
Reactor Coolant System from 350*F to 140*F within 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> assuming two RHR heat exchangers and two RHR pumps are in- ;
, service. The cooldown calculation performed for 2775 MWt j j demonstrated that the RHR System could achieve this cooldown :
in 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. The calculation performed for 2900 MWt indicated ;
the RHR System required 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> to cool the Reactor Coolant ;
i System to 140*F. This increase in cooldown time is an ;
operational issue (ie, outage scheduling) but does not impact any '
licensingbases for the plant.
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. The VCSNS Technical Specifications contain action statements which require plant cooldown from Hot Standby to Cold j Shutdown within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> following a reactor trip from full
! power. The cooldown calculation performed for 2900 MWt demonstrated that this cooldown could be achieved with two RHR ;
heat exchangers and two RHR pumps available. j Safety Iniection System l
i The minimum emergency core cooling flow requirement for the
- Charging / Safety Injection System is dictated by Small Break LOCA (SBLOCA) requirements. The SBLOCA safety analyses for the uprating used the original emergency core cooling flow rates based on a minimum delivered flow of 321 gpm through two of three Changing / Safety Injection branch lines and a maximum
. pump runout flow rate of 680 gpm. Acceptable SBLOCA analysis results were obtained with these minimum core cooling flow i
rates. The NRC has reviewed and accepted the SBLOCA analysis, as listed in the NRC SER (Reference 1). As part of the uprating program, the Technical Specification limits were changed to require a minimum flow rate of 338 gpm through two of three branch lines and a maximum pump runout flow rate of 688 gpm. Thb nrovides margin with respect to the analysis limits.
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As stated in Section 3.2.1, the evaluation of the auxiliary 2 equipment supporting the RSG program was performed assuming core thermal power of 2900 MWt and NSSS power of 2912 MWt.
These plant conditions were determined not to have an adverse effect upon these systems, and therefore, no further evaluation of the NSSS auxiliary equipment is necessary.
3.3 NSSS/ BALANCE-OF-PLANT INTERFACE ,.
The interfacing systems are relied on for heat rejection to the Ultimate Heat I
- Sink (Service Water Pond). These systems ar. Component Cooling Water System (CCW) and Service Water System (SW). The equipment evaluation was performed for an NSSS power level of 2912 MWt.
< 3.3.1 Component Coolinst Water System ,
- l. This system's major function is to transfer heat from the Residual Heat i i Removal System (RHR) to the Service Water System. The heat rejection 3
load from RHR will increase with the plant uprate for cooling i requirements at 4 and 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after plant shutdown. A secondary i function is to remove decay heat from the Spent Fuel Pool Cooling !
! Systems during both power and refueling operations. The projected heat loads for both concerns have been evaluated and are within the capacity of i the existing system. Along with the increases in required heat removal f rates, we are changing the Design Basis requirements for the heat j exchanger to reflect conservative regulatory requirements for cooldown j j time frames rather than the component suppliers' contractual !
! requirements. This may result in changes to minimum required flow rates for the CCW and/or SW systems. Changes as necessary will be reflected in design modification packages.
1 3.3.2 Service Water System !
This system's major function is to transfer heat from the various heat :
- exchangers connected to the Ultimate Heat Sink (Service Water Pond). '
For plant uprate, the heat load of primary concern is the CCW Heat Exchanger, with the increase in heat removal requirements which l cascade from the RHR system. Other heat loads on the SW System do not :
significantly change from current design requirements. Monitoring of the )
performance of this heat exchanger will ensure that design basis '
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capabilities are maintained. Calculations indicate that sufficient margin .
remains for the SW System. l l
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- 3.3.3 Spent Fuel Pool Coolinst j The plant uprate causes an increase in Spent Fuel Pool (SFP) Cooling heat ;
loads. The maximum Normal and Abnormal heat loads are affected by the unit uprate. The increases are ~15% and 10% respectively, which are ,
wi.hin the capability of the system. These increases in SFP temperatures !
j have been determined using a more conservative calculation for SFP decay heat loads prior to fuel movement including an assumption of the pool being full with fuel burned at 2900 MWt, core power. A comparison of j the heatloadsis asfollows:
Maximum Normal heatload (single train cooling)-Partial Core Off-load L Heat Load (MBTU/HR) Peak SFP Temperature
- - Current 16.605 140 F @ 227 hrs after Rx shutdown i Uprate 19.213 146*F @ 216 hrs after Rx shutdown i
! Maximum Abnormal heat load (both trains available-no failures)-Full 2
Off-load l Heat Load (MBTU/HR) Peak SFP Temperature e Current 31.647 139*F @ 202 hrs after Rx shutdown j Uprate 34.9139 142*F @ 202 hrs after Rx shutdown For the maximum normal heat load, the calculated peak pool temperature !
- increased approximately 5% from 140 F to 146*F when using the conservative assumptions stated above. Although this exceeds the Standard Review Plan, NUREG-0800, Section 9.1.3, limit of 140*F, the i
! calculated peak temperature of 146*F is acceptable since, as discussed j' below, local boiling at the surface of the fuel cladding will not occur and since it is well within the structural capability of the pool and its liner
, (i.e., analyzed for bulk boiling at 250*F).
SCE&G has chosen to perform a full core off-load each refueling for outage f
. flexibility and safety. Therefore, in order to bound the worst case for SFP 1
Cooling, an additional analysis was performed assuming the maximum -
Abnormal heat load occurs with only one train of cooling available (i.e.
single failure). This results in a peak heat load of 34.9139 MBTU/hr with a resultant peak SFP temperature of ~178*F at 208 hrs after reactor ,
shutdown. The SFP was also reanalyzed from a structural standpoint and has been found acceptable for SFP bulk boiling conditions up to 250*F. ,
The time to boiling and boil-off rates for a loss of cooling in the SFP have also been evaluated for the uprate conditions with the following results:
Maximum Normalheatload time to boiling ~7.7 Hrs, Rate 40 gpm Maximum Abnormalheatload time to boiling ~4.5 Hrs, Rate 72 gpm ,
Worst Case Abnormalheatload time to boiling ~2.2 Hrs, Rate 72 gpm >
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4 Based on the indicated time to boiling, sufficient time exists to restore SFP !
] cooling before the onset of boiling. Given the various makeup water
- sources available on site, it is reasonable to assume that there is l
. reasonable assurance that the fuel would not be uncovered should boiling !
occur. ,
e The licensing report on the High Density Spent Fuel Storage Racks )
previously submitted (Reference 4) showed that Local SFP water
- temperatures could reach a temperature of ~170*F. Based on the increases .
i in bulk water temperatures and increased heat loads due to uprate, the i conclusions in the licensing report remain valid and margin exists against j the initiation oflocalized boiling.
. It should also be noted that the SER issued with the licensing Amendment l
! 116 on August 23, 1994, provided approval of the recent SFP Rack 4- Criticality Analysis. The racks have been approved for 5 weight percent enriched uranium fuel, and the results remain bounding for the uprated ;
j conditions of 2900 MWt. l
}
4 )
l 3.4 GAS STORAGE TANKS I i l j To support proposed technical specification changes for VCSNS's replacement i Steam Generators, the offsite dose calculations for Chapter 15 of the FSAR were j- performed using revised source terms which reflected plant operation at the I Engineered Safeguards Power Level. The licensing submittal (Reference 3)
{ included an evaluation on the consequence of a Waste Gas Decay Tank rupture.
- NRC approval of the technical specifications supporting SG replacement was l issued via Reference 1. With regard to the dose resulting from the rupture of a
. Waste Gas Decay Tank, the NRC Staff's Safety Evaluation was as follows:
l' j "The licensee reevaluated the consequences of a waste gas decay
! tank rupture. The licensee's submittal stated that the analysis was performed not because of the RSG or due to changes in the design
, power capability, but rather to reflect TS limits on decay tank i radioactivity. The licensee assumed the release of 160,000 Ci of ,
188Xe Gamma and beta doses were calculated at the Exclusion Area !
4 Boundary and tho Low Population Zone. The staff independently l assessed the potential consequences of the release of the contents of a waste gas decay tank. The acceptance criterion for the release of the ,
l contents of a waste gas decay tank is 0.5 rem total body. Based upon I
. this criterion, the staff determined that the allowable waste gas tank i inventory would be approximately 131,000 Ci of 188Xe. While this particular issue is not associated with the replacement of the D3 steam generators, the licensee should reevaluate the determination of the allowable TS qu'antity of188Xe in the waste gas decay tank."
I
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' Attachm:ntII TSP 950001
, RC-95-0174 Page 12 of 22 The recommended reevaluation has lead to this proposed change in the Limiting Condition For Operation for Specification 3.11.2.6. It should be noted, however, that the VCSNS has never exceeded an administrative limit of 90,000 Ci of tasXe in a gas storage tank.
3.5 BALANCE OF PLANT SYSTEMS 3.5.1 Main Steam System 1
i The Main Steam System has been evaluated relative to the Replacement
- Steam Generator conditions of higher normal steam pressures and the
- _ increased steam flow conditions required for uprate when compared to the
! initial thermal design for the plant. The ASME Design Specifications for l the system identify the " Normal" operating pressure for the system to be 4 1185 psig which bounds the increase in thermal design pressure from 940 to 1000 psia.
! Note: The " thermal design pressure "is the pressure identified in the i thermal performance kits prepared by the turbine manufacturer.
i The mass flow rates for the system remain within acceptable fluid 4 velocities for the piping systems. No load steam conditions remain j constant for the NSSS uprate conditions at 2912 MWt.
l The capacity of the Main Steam Code Safety valves has been evaluated
! and adequate margin exists with the existing setpoints and tolerances as 4
identified in the Technical Specifications for the VCSNS.
l The rapid (7 seconds or less) closure requirements for Main Steam i Isolation Valves (MSIVs) will be retained for the RSG/Uprate power
- program. The rapid closure of these valves during a large downstream steam line break causes a significant differential pressure across the valve
' seats and a thrust load on the main steam piping supports in the area of the Main Steam Isolation Valves. During the evaluation of the RSG J impact upon NSSS/ BOP systems, it was determined that the worst case I loading on the MSIVs is generated by rapid closure from no load conditions in response to a Steam Line Break. Since no load temperature and pressure are unaffected by normal operation at a core power of 2900 MWt, the MSIV loadings will not be affected.
The Turbine bypass system is designed to bypass main steam to the main condenser and/or atmosphere, to provide an artificial steam load for the steam generators. The system was originally designed to accommodate a load reduction from 100 percent of rated turbine power to plant auxiliary load without a reactor trip. The capacity of the system was ~85 percent of main steam flow at full load temperature and pressure. With the uprate
Document Centrol D:sk i i
Attachm::ntII l TSP 950001 1 i RC-95-0174 l Page 13 of 22 l increase in steam flow of approximately 650,000 lbm/hr, the steam dump -
capacity of S5 percent is not maintained. The projected capacity at uprate conditions is ~81% of uprate steam flow out of the steam generators. This i reduction in capacity would indicate that a large load reduction would !
result in a reactor trip. The response of the system is not changed due to l 4 the plant uprate, only a slight reduction in capacity.
3.5.2 Feedwater System
, 3.5.2.1 Pumos ,
The Feedwater system was evaluated relative to the overall
! capability to provide adequate Feedwater not only for normal
- steady state operation, but to also operate through expected transients without causing a plant trip. The capability of the j pumping systems via the Feedwater Pumps and Feedwater Booster Pumps has been found adequate to provide reliable j uprated Feedwater flow. The flow requirements utilized to evaluate these components is normal 100% uprate flow plus,1%
! blowdown from the steam generators plus, a 5% flow margin for
- surges. This results in a required capacity of 106% of normal uprate flow which is within the capability of the pumping i components.
a ,
1 3.5.2.2 Feedwater Isolation Valve Closure Piping Loads Actual Feedwater Isolation Valve closure times are dependent on j the pressure difference across the valve as it closes. The power l uprate results in higher Feedwater flows thereby changing the l pressure differential across the valves.
1 l These new conditions were evaluated via computer model and j
resulted in acceptable closure times and piping loads. The '
evaluation was performed by comparing the rate of decrease in feedwater flowrate at uprate conditions to the previous evaluation. Three different cases were run, with all values :
bounded by previous analysis results.
)
! 3.5.2.3 Feedwater Heaters and Drains 1 Feedwater heaters and drains were evaluated for the uprate
- conditions. The Feedwater heaters are of adequate size and have the capacity to support plant uprate conditions in an efficient manner. Several of the feedwater heater drain valves have been found to be somewhat undersized and will require minor modifications to increase their capacity to an acceptable value with appropriate margins.
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i Page 14 of 22 i
3.5.3 Condensate -
The Condensate System was evaluated relative to the capability to pass l the required flow, condense uprate steam flows, and fulfill design l requirements under transient conditions. The system will fulfill the ;
e necessary requirements for uprate conditions with a minor modification.
. A summary of the evaluation for key components and a description of the ,
modificationisincluded. ,
i 3.5.3.1 Condensate Pumps !
The system has the pumpmg capacity to support uprate .
operations. Typical 100% power operation requires two ;
condensate pumps to supply flow to the Feedwater System. These pumps had an original design capacity of 9360 gpm @ 561 ft f TDH. The pumps supplied capacity is 10,015 gpm @ 590 ft TDH. i j The uprate flow requirements have increased to ~9635 gpm per
- pump and a total flow of 19268 gpm. The performance of the ;
pumps has been evaluated and found adequate to support uprate
, power operation. For Enhanced System Reliability, two of the ,
pump motors have been rewound with a higher class ofinsulation f
l to preclude any potential for excessive condensate motor stator l
- temperatures during high ambient temperature conditions. The i third motor will be rewound during RF-9 and will be available
!- prior to plant startup following the outage.
3.5.3.2 Condenser i i i The Main and Auxiliary Condensers have been evaluated i
relative to their capability to support uprate steam flows, maintain adequate back pressures for turbine performance, and ;
I protect maximum exit temperatures based on approximate ;
1 Circulating Water flows. Uprate steam rejection conditions to l the condenser have been evaluated based on input from the ,
i turbine vendor. While the steam mass flow rates through the :
Low Pressure Turbine rotors increase, the exit / exhaust :
enthalpies will be lower due to higher turbine efliciencies. This l results in an net increased heat load to the condenser of ~4%.
This heat load increase was evaluated and found to have minimal i affect on the back pressures of the condensers. Exit temperatures !
of the Chculating Water through the condensers are projected to increase ~0.8*F above current operating conditions, which has !
- been found to be acceptable for the uprate conditions. The ,
temperature effects and environmental concerns will be discussed !
furtherin the submittal. i
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TSP 950001 ;
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Page 15 of 22 l 3.5.3.3 Condensate System Modifications ,
One modification to the Condensate System has been identified !
- as necessary to support uprate flows. The Condensate pumps !
supply water to the Deaerator through two, parallel cage trim l ball valves. Through evaluation and analysis, the capacity of this j configuration was determined to be marginal. The configuration j will be modified by replacing one of the valves with a higher i capacity valve. The valves are sequentially opened, thus the l
! summation of the capacity is the important component. This i modification will have minimal if any apparent impact on the >
operation of the system other than the increased capacity of the flowloop. l 3.5.4 Circulatinst Water System t The Circulating Water System (CW) was evaluated relative to the ability to remove heat from not only the Main Condenser, but also from the ;
- Auxiliary Condensers and the various turbine auxiliary systems cooled by
- this system. A major part of the CW System evaluation included a system L flow measurement, utilizing a fluorescent dye dilution method. From this SCE&G confirmed that the design flow of the CW System was still ,
appropriate, and that the flow rates to the Main Condensers were slightly ;
s higher than design which would negate some of the effects of uprate heat loads. These higher CW flow rates through the condenser are within
- design allowables and present an advantage in heat removal capabilities ,
relative to projected uprate heat loads. l L ,
3.5.4.1 Main Condenser t
The Main Condensers were evaluated relative to the uprate steam conditions and found acceptable. Heat rejection to the CW ,
System is projected to increase ~4% above existing thermal i' design. This results in a projected increase in the differential l temperature across the condenser of 1.0*F. With the Main !
Condensers being the major heat load on the system, the effects !
on maximum lake return temperatures were a major concern.
The original and current discharge temperature limit of 113*F
,~
to Monticello lake was based on a maximum projected lake supply temperature of 88'F, plus a temperature rise of 25'F across the condensers. With a higher projected temperature rise at uprate ;
conditions, the maximum allowable inlet temperature will be
~86*F which should be acceptable for operation at uprate power conditions. The exit temperature limit of113*F currently in place is not being changed. Operational strategies have been developed
, in the event that we challenge the exit temperature limit.
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i TSP 950001 ;
- RC-95-0174 !
. Page 16 of 22 ,
3.5.4.2 Auxiliary Condensers t
i .
The Auxiliary Condensers were evaluated and found acceptable. ;
, The condensers were initially designed for 50% load (i.e. two out ;
of three FW Pumps operating at 100% power). The condensers [
j have inherent margin in that at 100% power, three Feedwater (
j pumps are operated at approximately 331/3% each. l j
3.5.4.3 Turbine Auxiliary Systems j
- Initial design of the plant had various turbine auxiliary systems ;
{ cooled by an Open Cycle Cooling Water System supplied by the !
[. Circulating Water system. These auxiliary systems exhibited accelerated corrosion problems due to lower flow velocities and 4
fouling. In order to support the plant uprate operation and to ,
arrest the problem with corrosion and fouling, this open cycle ;
i' system will be converted to a closed system cooled with a modular l forced draftcooling tower. l 1
i Not only will this modification solve the fouling problem, !
( enhance performance, and increase reliability, but it will also !
- l. take a heat / flow load off of the CW System. This will remove a !
! heat load of ~54.087 MBTU/hr along with a reduction in flow l
- demand from the CW System of ~10,000 gpm. This heat load and i
- flow reduction should lessen the impact of the uprate loads on the j Main Condensers. This reduction in flow and heat load has not i been included in evaluations of the condensers which maintains )
the original conservatism designed in the plant.
i ,
3.5.5 Turbine Generator l While not directly tied to nuclear safety, an important aspect of the uprate l' i project is the turbine / generator systems. The LP Turbine rotors are being replaced due to stress corrosion cracking issues and to incorporate technological improvements in steam path design which will increase the j efficiency of the turbine and increase the output of the plant. The new LP i Turbine rotors are of a monoblock design which eliminates the shrunk on ;
- wheels.
3.5.5.1 Missile Elimination The existing LP Turbine rotors are of a design which incorporates a wheel which is " shrunk" onto the shaft of the turbine rotor.
These wheels then had the turbine blades attached to the wheels.
This design had several disadvantages in that at any point where l
l
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-Docum:nt Centrol D:sk l Attachm:ntII !
TSP 950001-RC-95-0174 Page 17 of 22 l l
, l a machined joint between two pieces taists, a susceptibility to l stress corrosion cracking also exic's. Cracking cas identified on i these rotors in the area of the keyways on ne respective wheels 1
- and also in the dovetail blade attachment areas of the wheels. ;
Another negative aspect of this design ! that the rotor / wheels
. could reach a critical speed such that a - 1. could burst (due to
- centrifugal stresses). This condition m ir.e criteria requiring consideration of missile ~ generation, a detailed analysis was required to evaluate the probability el missile generation based on crack growth rates identified - at ;nspection intervals prescribed by the manufacturer.
The LP Rotors will be replaced ir Refueling 9 (RF-9) due to I reduced service life from stress u asion cracking previously 4
identified. The new rotor design . corporates enhancements to minimize stress corrosion crackine The new rotors' monoblock :
i design incorporates integral wh- ; and couplings as part of a
- solid shaft to eliminate the maf nal surfaces that promote the 2
initiation of Stress Corrosion Cracking. The monoblock rotor
. design decreases the probability for missile generation / wheel burst to an annual probability in the range of 10-8, as evaluated by the turbine manufacturer.
[
The rotor replacement is being implemented under the design change program, with all necessary analysis and calculation j changes being made under these programs. i 3.5.5.2 Turbine Stop/ Control Valves Another change as a result of the LP Rotor replacement is the [
l relaxation of the testing frequency of the Turbine Stop/ Control :
valves. The proposed change is to extend the valve testing l l
frequency to quarterly. Since the capability to generate a missile l is reduced to less than the annual guideline of 10-5 with the new i rotor design, control system reliability is no longer a significant i input to missile probability analysis.
The turbine manufacturer recommends extensions to the testing i frequency interval based on experience with the current control i
, system design, no reportable inservice failures of this design, and the annual probability of a complete control system failure being less than 10-8 Post monoblock rotor installation valve testing !
will be performed for system reliability only. l e
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j Page 18 of 22 )
j i
- 3.5.5.3 Generator Stator Water Cooline l Uprate evaluations have concluded that the generator stator !
1 cooling system capacity will need to increase to ensure adequate l i heat removal. The stator cooling flow rates are being increased to i fulfill this requirement and changes are being made to jl j instrumentation in the system to enhance operability of the system. The generator stator was rewound in Refuel 8 due to l 1 increasing Stator Bar cooling gas inleakage and to enhance !
performance and margin capacity at uprated power. As a result, j the generator capacity rating was upgraded to 1,162,300 KVA. ;
- a 4.0 - IMPACT ON THE INDIVIDUAL PLANT EXAMINATION (IPE) ;
- The Individual Plant Examination (IPE) was based on a core power level of j j 2785 MWt. The effect ofincreasing core power to 2900 MWt was investigated to !
- determine if any changes in the results of the original Probabilistic Safety l
. Assessment (PSA) would occur. Additionally, the review " sought out" areas of l
! - the PSA where a decrease in margin in equipment or human performance might !
4 occur. The analysis of the impact of uprate on the PSA serves to enhance the more traditional deterministic analysis described in the body of this document.
! The evaluation methodology was based on the following criteria which was applied to the individual success criteria and then to the overall IPE results:
i e Will the power uprating impact any of the numbers that directly link to
- ' the core damage frequency, the frequency of fission product release
. categories, or change any of the top 100 core damage sequences reported in l the IPE?
)
e Does the power uprating cause a change to any of the assumptions or
, models used in the IPE analysis, but does not impact the IPE results
- reported in the submittalreport?
- e Does the power uprating cause a significant loss of margin in the success criteria such that the success criteria may change if other modeling sensitivities are considered?
4 e- Is there an issue related to the original IPE success criteria that was i 1 identified during this review? I 4
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, TSP 950001 RC-95-0174 Page 19 of 22 4-During the review, several issues were identified as being impacted by the power uprate. For example, the time available for successful operator action in
, areas dependent on core power, such as the time.to steam generator dry out
. during a loss of heat sink event, have been slightly effected. The various issues
. were reviewed for impact on the IPE results and conclusions. In all cases, the effects were determined to be minor and bounded by the original IPE analysis.
The impact of the uprate on plant response will be incorporated into the "living" PSA program.
5.0 ENVIRONMENTAL EFFECTS i
The plant uprate effects on the environment are considered to be minor. The power uprate does increase the heat rejection rate to the environment 3
approximately 4% via the Main Condenser to the Circulating Water System.
Original environmental assessments identified that the system had a capability to transfer ~6.67 MBTU/hr based on a total Circulating Water system flow rate
~
of ~534,000 gpm with a temperature rise of ~25'F as noted in the Operating Licen 3 Environmental Report (OLER). The conversion of the open and closed i cycle auxiliary system heat loads will remove ~54.087 MBTU/hr from the
- system. The following summarizes the design, current, and projected uprate i heat loads:
HEATSOURCE DESIGN DUTY CURRENT UPRATE MBTU MBTU MBTU d
Main Condensers 6000.00 6006.3 6240.0 Auxiliary Condensers 249.75* 161.1 164.4 Other Auxiliary Loads 54.10 54.1 ---
TOTALS 6303.85 6221.5 6404.4
{
4 l The uprate heatload is:
~1.6% > Design
~2.9% > Current
~96% ofSystem Capability
- The load in the OLER reflects only the design duty of one condenser.
The heat rejection increase is relatively insignificant and is within the system capability values presented in the Operating License Environmental Report. -
Document Centrol D:sk AttachmentII TSP 950001
-- RC-95-0174 Page 20 of 22 i Slight changes to overall system flow will be made when the Auxiliary System loads are removed from the CW system. This will result in a corresponding i reduction of flow demand on the system of ~7000 to 10,000 gpm. A majority of
- this flow will re-distribute itself within the Main and Auxiliary Condensers, but i the overall CW system flow rate should decrease to something less than the
- 534,000 gpm identified in the OLER. Therefore, the velocities and flow rates identified in the OLER will remain bounding for the uprate conditions.
1 A key aspect of this effort, is that the existing discharge temperature limit for the Circulating Water return stream to the lake will not change from the existing 113 F limit. .
i 4
- This information has been presented to the South Carolina Department of
- Health and Environmental Control (SCDHEC) and has been found acceptable.
i
! Environmental monitoring will continue along with trending of the CW system discharge temperatures to ensure that the environmental impact is benign.
I CLOSED CYCLE COOLING SYSTEM MODIFICATION
! The removal of the closed and open cycle cooling systems from the CW system are being performed under the modification program. Environmental assessments of the heat loads, atmospheric affects, anel operational considerations are (oordinated within the design control program. Submittals for environmental considerations have been coordinated through the SCDHEC. ,
- Environmental monitoring is maintained by SCE&G within prescribed requirements, t
' 6.0 EQUIPMENT ENVIRONMENTAL QUALIFICATION AND HIGH ENERGY LINE BREAK (HELB) IMPACT l The Equipment Environmental Qualifications were evaluated under the Replacement Steam Generator Project which was implemented in Refueling 8.
Due to power uprate and changes in mass inventory of the Steam Generators, 4
the environmental consequences during Loss-of-Coolant Accidents and Steam Line Breaks, both inside and outside containment, were evaluated. The results of these evaluations were then used to evaluate the effects on the respective Equipment Qualifications and other components as applicable. The uprate conditions were utilized to perform the evaluations to avoid repeating the effort.
i All analysis was completed under the Replacement Steam Generator Project and were submitted for review as appropriate.
4 4
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TSP 950001 RC-95-0174
! Page 21 of 22 7.0 OTHER SYSTEMS EVALUATED BUT NOT AFFECTED BY UPRATE Various systems were evaluated and found not effected by the uprate project
! over the course of several review phases. Those systems were evaluated for respective capacities, heat removal capabilities, and in many cases no direct
, connection to plant uprating was found. The following is a summary listing of i
major plant systems that were not affected by the uprate:
. Auxiliary Steam Condenser Air Removal Chemical Feed Systems i Emergency Diesel Generators and Auxiliaries !
I Solid and Liquid Waste Systems <
i Fire Service !
- Stati;n / Instrument Air i i Reactor Building Cooling I
- Generator Gas and Vents '
l Non-Nuclear Drains Plant Waste
. RB Spray Demineralized Water System Nuclear and Secondary Sampling
- HVAC s
8.0 CONCLUSION
S l l
2 The power uprate of the VCSNS will increase the RATED THERMAL POWER of the reactor core from a nominal 2775 MWt to 2900 MWt, with a corresponding increase in net electrical output of approximately 64 MWe. l I Amendment No.119 to the Facility Operating License No. NPF-12 for the i VCSNS was recently issued by the Nuclear Regulatory Commission via i Reference 1. This amendment changed the Technical Specification in response
- to SCE&G's requests in Reference 3 to support VCSNS operation with replacement steam generators at its current RATED THERMAL POWER level
, of 2775 MWt. Where possible, the analyses and evaluations within Reference 3 (see Attachment I) were, however, performed at a core power level up to 2900 2
MWt, which corresponds to the plant's original Engineered Safeguards design j rating.
! This amendment request seeks approval to operate the VCSNS at a RATED THERMAL POWER level of 2900 MWt with replacement steam generators. In support, additional analyses, evaluations, and Technical Specifications changes are provided to supplement those outlined in Attachment 1. When 4
implemented, the proposed changes along with those previously approved via
- Reference 1 will preserve the design, analytical methodology, and safety analysis assumptions outlined in this amendment request.
4
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' AttachmentII .
TSP 950001
. RC-95-0174 Page 22 0f 22 The safety analyses, evaluations and supporting documentation provided in or referenced by this submittal demonstrate that the VCSNS can be safety operated at a core power level of up to 2900 MWt without undue risk to the health and safety of the public.
4
9.0 REFERENCES
i
- 1. USNRC Safety Evaluation Report (SER) Related to Amendment No.119 i to Facility Operating License NPF-12, South Carolina Electric and Gas Company, South Carolina Public Service Authority, Virgil C. Summer
! Nuclear Station, Unit No.1, Docket No. 50/395, November 18,1994.
j 2. USNRC Safety Evaluation Report (SER) Related to Amendment No. 27 to j
Facility Operating License NPF-12, South Carolina Electric and Gas Company, South Carolina Public Service Authority, Virgil C. Summer i Nuclear Station, Unit No.1, Docket No. 50/395, September 27,1994.
! 3. SCE&G Letter from J. L. Skolds to the USNRC Document Control Desk,
- " Virgil C. Summer Nuclear Station, Docket 50/395, Operating Licensing
- No. NPF-12, Steam Generator Replacement Technical Specification Change Request (REM 6000-8, TSP 930015)", October 29, 1993.
(Additional submittals in support of this were made September 3,1992; March 12,1993; April 30,1993; March 11,1994; and October 20,1994.)
3
~
- 4. SCE&G Letter from O. W. Dixon to the USNRC Document Control Desk,
! " Virgil C. Summer Nuclear Station, Docket 50/395, Operating Licensing No. NPF-12, High Density Spent Fuel Storage Racks", January 23,1984.
i
- 5. Virgil C. Summer Nuclear Station Final Safety Analysis Report l 6. Virgil C. Summer Nuclear Station Technical Specifications l 7. SECL-93-036 (dated October 1993) and SECL-93-036, Rev.1 (dated
- February 1994): Licensing Submittal to Support Replacement Steam 1
Generator Technical Specification Changes for the VCSNS.
! 8. WCAP-13714 (Proprietary Class 20), " Virgil C. Summer Nuclear Station
- Replacement Steam Generator /Uprating Engineering Report", June 1994.
i
- 9. WCAP-12945-P (Proprietary), " Code Qualification Document for Best Estimate LOCA Analyses", June 1993 (unapproved - currently under
, NRC review).
- 10. CGE-94-0019SGUL,"VCSNS RSG/Uprating: Best Estimate Large Break LOCA - Final Analysis Report", from D.D. Molitoris (Westinghouse) to
- W.W. Turner (SCE&G), December 2,1994.
5
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