ML20091L454

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Proposed Tech Specs for Uprate Power Operation
ML20091L454
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 08/18/1995
From:
SOUTH CAROLINA ELECTRIC & GAS CO.
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ML20091L452 List:
References
NUDOCS 9508290203
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Text

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l Document Control D:sk Attachment III i

TSP 950001 RC-95-0174 Ps.ge 1 of1 SCE&G -- Explanation of Technical Specification Changes for Uprate Power Operation PAGE

^If ted Bar Description of Change g Reason for Change eeti9n __

1-5 1.25 1 Rated Thermal Power definition is This change is necessary to support the revised to incorporate the increased uprate power condition.

power level.

3/4 11-5 3.11.2.6 1 Revise maximum quantity of Review of calculation for offsite doses duo radioactivity in each gas storage tank - to a gas tank rupture.

160,000 curies to 131,000 curies Noble gas.

B 3/4 2-1 3/4.2 1 Discussion of the 2200* F ECCS limit is This change is necessary to support the revised to reference the acceptance Best Estimate LOCA analysis, criteria provided in 10CFR50.46.

B 3/4 2-3 3/4.2.2 1 Discussion of the 2200'F ECCS limit is This change is necessary to support the and revised to reference the acceptance Best Estimate LOCA analysis.

3/4.2.3 criteria provided in 10CFR50.46.

, 616a 6.9.1.11.c 1 Methodology referenced by the COLR This change is necessary to support the that is used to determine the heat flux Best Estimate LOCA analysis.

hot channel factoris changed to reference Best Estimate LOCA analyses.

t OL page 4 2.C.1 1 Revising Maximum Power Level to 2900 This change is necessary to support the MWt Core Power Uprate Power Condition.

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9508290203 950818 PDR ADOCK 05000395

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' DEFINfTIONS *

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) i W ._PUEIS 3 i

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1.23 PUIE or PUIE!M is the controlled process of discharfing air or g 3

! or other operating condition, in such a manner t required to purify the confinemenc.

[; ouADRJurr POWER TILT RATfD

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i 1.24 )

i detector ca11breted output to the average of the up d brated outputs. or the rette of the mentaus laser encore detector calibra output to the is greater. Withaverage of the one essere lowerino detector encore detector calibrated outputs, whic s 4

2 shall be used for congsuting the average. perehlt, the remaining three detectors '

RATED THE10EL POWEA t

! the reactor coolant of p .1.25 RATED THEIBHL POWER shall be a atAcfoR TRfP sysTIM ens ** IIg ,

2 goo ht. ,

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1.26 the when The REACTOR monitored parameter TRIP SYSTDI ascends RESM315E its trip setpoint TIE shall at the channel sens I until loss of stationary gripper coil voltage.

4 mrPortAaLE Evfut i

j 1.27 i A REPORTABLE EVENT shall be ag of those conditions specified in I Section 50.73 to 10 CFR Part 50.

1 S W TDOWN MARRIP i 1.28 SWTDOMI M4lWIN shall be the instantaneous amount of reactivity by i the reactor is subcritical or usu14 he subtritical free its present condition

! assuming all full length red cluster assemblies (shutdeun and control) are fully inserted except for the single red cluster assembly of hghest reactivity worth whic$t is assumed to be fully withdream.

SLAVE RELAY TEIT 1.29 A SLAVE RELAY TEST shall be the ener j

verification of OptRASILITf of each relay.gitation of each slave relay and The SLAVE RELAY TE3T shall include a continuity check, as a miniasm, of associated testable actuation devices.

1.30 list Used f:

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E._E 1.31 A SOURCE OIECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

1 pyg ..g, g 1,3 den =ent N . nJN.1U 9

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i DEFINITIONS '

l i PURGE - PURGING 1.23 PURGE or PURGING is the controlled process ofdischarging air or gas from a confinement to maintain temperature, pressure, humidity, concentration '

or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. ,

QUADRANT POWER TILT RATIO j

^ 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-

' brated outputs, or the ratio of the maximum lower excore detector calibrated output to t whichever 4

me average is greater. With one excoreofdetector the lower excorethedetector inoperable, remainingcalibrated three outputs, detectors i

shall be used for computing the average.

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! RATED THERMAL POWER i

1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to I the reactor coolant of 2900 MWt.

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REACTOR TRIP SYSTEM RESPONSE TIME 1.26 TheREACTOR TRIP SYSTEM RESPONSE TIME shallbe the timeintervalfrom when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage. .

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shallbeanyofthoseconditionsspecifiedin Section 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount ofreactivity by which the reactor is suberitical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are i fully inserted except for the single rod cluster assembly of highest reactivity -

worth which is assumed to be fully withdrawn. t SLAVE RELAY TEST 1.29 A SLAVE RELAY TEST shallbe the energization ofeach slave relayand verification of OPERABILITY ofeach relay. The SLAVE RELAY TEST shallinclude a continuity check, as a minimum, of associated testable actuation devices.

1.30 NotUsed  :

SOURCE CHECK 1.31 A SOURCE CHECK shallbethequalitativeassessmentofchannelresponse -

when the channel sensor is exposed to a radioactive source.

i SUMMER -UNIT 1 1-5 Amendment No. 35,104,117, i

_RADI0 ACTIVE EFFLUENTS GAS STORAGE TAN'KS i

LIMITING CONDITION FOR OPERATION 4

, 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to.16Gr000 curies noble gases (considered as Xe-133). f3fj o.o APPLICABILITY: At all times. l ACTION:

a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend al1 additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

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. SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materials are being added to the tank.

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SUMMER - UNIT 1 3/4 11-5 Amendment No. 104 l

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RADIOACTIVE EFFLUENTS  :

GAS STORAGE'I'ANKS >

LIMITING CONDITION FOR OPERATION l

3.11.2.6 The quantity ofradioactivity contained in each gas storage tank shall be limited to less than or equal to 131,000 curies noble gases [

(considered as Xe-133).

APPLICABILITY: Atalltimes.

ACTION:

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a. With the quantity of radioactive material in an gas storage tank exceeding the above limit, immediately sus alladditions of radioactive material to the tank and withi 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.  ;

3 j b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

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SURVEILLANCE REQUIREMENTS 4

4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least one 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materials are being added to the tank. per l

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l SUMMER - UNIT 1 3/4 11-5 Amendment No.104,

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{ 3/a.2 POWER OISTRTS1/ff0N LIMITS 8ASES i

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during Condition I (Normal Operation) and events Dy:

rate Frequency) 11 (Incto egrity i

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m design limit during normal operation and in (2) limiting the fission gas release , and short-term e

3Jff .

i . J mechanical properties to within assum,ed design criteria. fuel pellet tempera l ( '~ j ,"the initial peak senditions linear power density durine Condition I events provides assuIn a

used to s are rance that

[acce nee crilecla limit 200*F he LOCA and tTs- 7 t exc a gee t

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1 these specifications are as follows:The definitions of certain het ch ,

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! Fg (z) l i

Heat Flux Hot Channel Factor, is defined as the maximum local j

i by the average fuel rod heat flux, aliowing for tolerances on fuel pellets and rods; l

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F i # t Nuclear Enthalpy Rise Not Channel Factor, is defined as the ratio the integral power of linear to the average rod power power. along the rod with the highest integra 3/4.2.1 AXIAL FLUX O!FFERENCE i

envelopeTheoflimits the F on AX!AL FLUX O!FFERENCE (AFD)gassure that the F

} 3 Itait specified in the CORE OPERATING LIMITS REPORT (C times the normalized axial peaking factor is not exceeded during either operation or in the event of xenon redistribution following power changes.

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1 6.9.1.11.The lief ts on AFD will be provided in the COLA per Technical Specific j Target flux difference is determined at equilibrium menen conditions. The

full-length rods aey be positioned within the core in accordance with their

! respective insertion Itaits and should be inserted near their neraal position

! for steady-state operation at high power levels.

j difference obtained under these condittens divided by the fraction of RATE j- THERMAL POWER is the target flux difference at RATED THENEL POWER for the associated core burnup conditions.

] Target flux differences for other THESEL POWER levels are obtained by euttiplying the RATED THENEL POWER value by the appropriate fractions) THENEL POWER 1evel. The perledic updating of the target flux difference value is necessary te reflect core burne considerations.

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SupO9ER - UNIT 1 8 3/a 2-1 Amendment No. 55. 15, E ,

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TSP 950001 -

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INSERTA in the event of a LOCA, there is a high level ofprobability that the acceptance criteria of10CFR50.46 would not be exceeded.

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3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance offuel integrity during Condition I(Normal Operation) and II (Incidents of Modera events by: (1) maintaining the calculated DNBR in the core at or above the design limit during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and claddin mechanical properties to within assumed design criteria. In addition, the peak linear power density during Condition I events provides assura in the event of a LOCA, there is a high level of probability that the acce criteria of10CFR50.46 would not be exceeded.

these specifications are as follows:The definitions ofcertain hot chan FQ(z) heat flux on the surface of a fuel rod at core eleva tolerances on fuelpellets and rods;by the average fuel rod F$*n theintegral of1mear poweralonNuclear Enthalpy Rise Hot Cha power to the average rod power. g the rod with the highest integrated 3/4.2.1 AXIAL FLUX DIFFERENCE envelope of the Fq limit specified in the Q CORE OP times the normalized axial peaking factor is not exceeded during either norm operation or in the event ofxenon redistribution following power changes.

The 6.9.1.11. limits on AFD will be provided in the COLR per Technical Specificat full length rods may be positioned within the core in ac for steady-state operation at high power levels. The v difference obtained under these conditions divided by the fraction of RATED associated core burnup conditions. Target flux diffe the appropriate fractional THERMAL POWER lev target flux difference value is necessary to reflect core burnup considerations 4

SUMMER - UNIT 1 B 3/4 21 Amendment No. 66,75,88,

J POWER DISTRIBUTION LIMIT BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE ANO l NUCLEAR 5.NTHALPY RISE n0T CHANNEL FACTOR The limits on neat flux het channel factor, RCS flowrate, and nuclear i entnaloy rise not enannel factor ensure that 1) the design limits on peak r

/ NM"m local power censity and minimum Ch3R are not exceeoed and 2) in the event of3 LOCAIt peak fue < lad temo- ure will exceeG ZZCD'F E accepta J

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Each of these is measuraelelut will normally only be determined periodica11 as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is

sufficient to insure that the limits are maintained provided

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a. Control rods in a single group move together with no individual rod insertion differing by more than
  • 13 steps, indicated, from the j group demand position. ,
b. Control rod groups are secuenced with overlapping groups as described  !

in Specification 3.1.3.6.

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c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.

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d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

F g will be maintained within its limits provided conditions a. througn

*d. above are maintained. As noted on the RCS Total Flow Rate Versus R figure j in the CORE OPERATING LIMITS REPORT (COLR), RCS flow rate and power-may be l l " traced off" against one another (i.e. , a low measured RCS flow rate is l
acceptable if core power is also low) to ensure that the calculated DN8R

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l will not be below the casign DN8R value. The relaxation of Fh as a function I of THERMAL POWER allows changes in the radial power shape for all permissible red insertion limits.

R, as calculated in 3.2.3 and used in the RCS Total Flow Rate Versus R l figure in the COLR, accounts for F" less than or equal to the F limit specifigdintheCOLR. This value is used in tne various accident analyses where F j influences parameters other than DN8R, e.g. , peak clad temperature l

j and thus is the maximum "as measured" value a110wed.

Margin is maintained between the safety analysis limit ON8R and the j design limit DNSR. This margin is more than sufficient to offset any rod bow cenalty and transition core penalty. The remaining margin is available for

plant design flexibility.

When an gF measurement is taken, an allowance for coth experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with tne incore detector flux mapping systes and a 3% allowance is appropriate for manufacturing tolerance.

LMMER .1IT *. S 3/4 -3 Menc ent NL/3.

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  • Page 1 of1 INSERT B thereis a highlevelof would not be exceeded. probability that the acceptance criteria of10CFR50.46 l

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POWER DISTRIBUTION LIMIT l l

BASES J  !

3/4.2.2 and ~ 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and  !

l NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ,

The limits on heat flux hot channel factor, RCS flowrate, s.nd nuclear  !

I enthalpy rise hot channel factor ensure that 1) the design limits on peak l local power density and minimum DNBR are not exceeded and 2) in the event of a  !

LOCA there is a high level of probability that the acceptance criteria of 10GFR50.46 ,

would not be exceec ed.

Each of these is measurable but will normally only be determined periodically p as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is j sufficient to insure that the limits are maintained provided: .

a. Control rods in a single group move together with no individual rod insertion differing by more than i 13 steps, indicated, from the I group demand position.

! b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6.  ;

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c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.
d. The axial power distribution, expressed in terms of AXIAL FLUX  ;
DIFFERENCE, is maintained within the limits. l NEu will be maintained within its limits arovided conditions a. through  ;
d. above are maintained. As noted on the RCS Cotal Flow Rate Versus R figure  ;

j in the CORE OPERATING LIMITS REPORT (COLR), RCS flow rate and power may be  !

" traded off" against one another (i.e., a low measured RCS flow rate is acceptable if core power is also low) to ensure that the calculated DNBR '

, will not be below the desi gn DNBR value. The relaxation of F[s as a function of THERMAL POWER al ows changes in the radial power shape ,for all permissible rod insertion limits,

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! R, as calculated in 3.2.3 and used in the RCS Total Flow Rate Versus R

! figure in the COLR accounts for F5s less than or equal to the FElimit specified in the CObR. This value is used in the various accident analyses l where FEu influences parameters other than DNBR, e.g., peak clad temperature

and thus is the maximum "as measured" value allowed.

I Margin is maintained between the safety analysis limit DNBR and the I

, design limit DNBR. This margin is more than sufficient to offset any rod bow penalty and transition core penalty. The remaining margin is available for

plant design flexibility.

When an Fq measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate I for a full core map taken with the incore detector flux map

' 3% allowance is appropriate for manufacturing tolerance. ping system and a i

SUMMER -UNIT 1 B 3/4 2-3 Amendment No. 75,88,

! ADMINISTRATIVE CONTROIA l

. t l- , CORE OPERATING LTvfTS REPORT (Continnad) 1 l (Methodology for Sa-ame=

' tion 3.1.1.3 - Moderator Temperature

{ Coenicient 3.1.3.5 - Shutdown Bank Insertion Limat,3.1.3.6 -

i Control Bank Insertion Limit,3.2.1 - Axial Flux Difere' ace,3.2.2 - '

Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot

{ Channel Factor).

b. WCAP 10216-P.A Rev.1A." RELAXATION OF CONSTANT AXIAL  ;

OFFSET CONTROL FQ SURVEILLANCE TECHNICAL l j SPECIFICATION", February 1994 (W Proprietary).

l (Methodology for S;+N==Hons 3.2.1 - Axial Flux Diference

! (Relaxed Azaal Odbet Control) and 3.2.2 - Heat Flux Hot Channel

Factor (FQ Methodology for W(Z) surveillanos requarements).)

insEA.T March f E NA OD C ).

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l (Methodology for S;+ ~r#- 3.2.2 Heat Flux Hot Channel Factor).

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The core operating limits shall be determined so that all applicable limits j (e.g., fuel thermal. mach ==ie=1 limits, core thermal-hydraulle limits, nuclear
limits such as shutdown marsm, and transient and accident analysis limits) of l the safety analysis are met.

i TheCORE OPERATING LIMITS REPORT,includa'nganymid-cyclerevisionsor i l supolaments there to shall be provided upon issuance, for each reload cycle, .

I to de NRC Document Contro, Desk with copies to the Regional Ad=ini% tor and ResidentInspector.

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SUMMER-UNIT 1 6-16a Amandment No.M,121

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c. WCAP-12945-P " CODE QUALIFICATION DOCUMENT FOR BEST ESTIMATE LOOA ANALYSES", (W Proprietary).

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ADMINISTRATIVE CONTROLS

. CORE OPERATING LIMITS REPORT (Continued)

I (Methodology for Specification 3.1.1.3 - Moderator Temperature 4

Coefficient 3.1.3.5 - Shutdown Bank Insertion Limit,3.1.3.6 -

Control Bank Insertion Limit,3.2.1 - Axial Flux Difference,3.2.2 - <

Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot j ChannelFactor). I

b. WCAP-10216-P-A,Rev.1A," RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL- '

SPECIFICATION", February 1994 (W Proprietary). -

i (Methodology for Specifications 3.2.1 - Axial Flux Difference i (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel

' Factor (FQ Methodology for W(Z) surveillance requirements).) {

c. WCAP-12945-P, " CODE QUALIFICATION DOCUMENT FOR BEST -

ESTIMATE LOCA ANALYSES", (WProprietary).

i (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

i The core operating limits shall be determined so that all applicable limits '

(e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear i limits such as shutdown margin, and transient and accident analysis limits) of  :

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the safety analysis are met. l l

4 TheCORE OPERATING LIMITS REPORT,includinganymid-cyclerevisionsor j supplements there to shall be provided upon issuance, for each reload cycle,  ;

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to the NRC Document Controa Desk with copies to the Regional Administrator and l

ResidentInspector. l l I r

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SUMMER -UNIT 1 6-16a Amendment No. 88,121,

l Document Centr:1 Desk '

Attachm:ntIV TSP 950001 RC-95-0174 l Page 1 of 3 '

SAFETY EVALUATION i FOR REVISING THE SPECIFICATION FOR '

UPRATE VIRGIL C. SUMMER NUCLEAR STATION TECHNICAL SPECIFICATIONS Description of AmendmentReauest  !

South Carolina Electric & Gas Company (SCE&G) proposes to revise the Virgil C.

Summer Nuclear Station (VCSNS) Technical Specifications (TS) pages 1.5,3/4.11-5, 6-16a, and Bases pages B3/4.2-1 and B3/4.2-3. These changes support the Uprate project and provide the following:  :

i e a new definition of Rated Thermal Power (RTP) to incorporate the uprate power condition of 2900 MWt. This value represents the total heat transfer rate from the reactor core to the reactor coolant and does not include heat generated by the reactor coolant pumps.  ;

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e a revised limit for the quantity of radioactivity stored in any one gas storage tank. l This new value is based on the methodology in NUREG 0133 and only affects the ,

maximum quantity stored.  !

j e a new reference to the Core Operating Limits Report (COLR) which is based on  :

the Best Estimate Loss of Cooling Accident methodology, and a deletion of the >

reference to the B ASH /B ART methodology in the COLR specification. ,

s revised bases information to indicate that VCSNS meets the generic acceptance

criteria of10 CFR 50.46, rather than only using the ECCS acceptance criteria of )

L 2200'F. This change is appropriate since the peak cladding temperature may not j be the most limiting criteria associated with the ECCS evaluations.

d Many TS changes were required to support the Steam Generator Replacement (SGR), .

which were approved and issued via reference 1. Many of the TS changes expected for a plant Uprate were included in the SGR submittal. Most evaluations performed for SGR utilized 2900 MWt core power as an initial condition.

This Technical Specification Change Request (TSCR) primarily revises those areas in l TS which were not included in Reference 3. The primary supporting analyses  :

l performed for uprate are: Large Break Loss of Cooling Accident (LOCA) utilizing the I Westinghouse Best Estimate LOCA methodology, spent fuel pool cooling capacity analysis resulting from our outage practices, and Waste Gas Decay Tank Rupture analysis resulting from a comment included in the SER for SGR (Reference 1.). Other analyses and evaluations were performed to assess the capability of other systems and components to support Uprate, with the results indicating that both the Nuclear ,

. Steam Supply System (NSSS) and the Balance of Plant systems are capable of '

supporting uprate power operation assuming modifications to several balance of plant systems.

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Document Centrol D:sk l AttachmentIV l TSP 950001 1 RC-95-0174

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j Safety Evaluation j The conditions that result from uprate power are increased heat transferred from the Reactor core, increased steam flow, increased feedwater flow, and increased electrical 4

output. The additional heat load of approximately 4.5 percent can be met with the

existing capacities of all NSSS and interfacing systems. j Modifications such as Closed Cycle Cooling are being planned to improve the capability of secondary systems to meet the additiona load.

!' mass flow rates has been evaluated and does not The present increase in theThe any concerns. seconda[75 steam generators are rated for this condition and  :

comply with all ASME Code requirements. The condenser, piping, and valves have  !

4 . all been evaluated and have adequate margin to support uprate conditions. The same  :

' is true for Feedwater and Emergency Feedwater Systems. In addition to the code ,

requirements, chrome-moly steel has been used in feedwater piping replaced during l RF-8 to reduce the effects of erosion / corrosion. i i

The additional heat produced will generate additional electricity. The turbine-generator has been evaluated and is capable, with a modification to the Stator Water Cooling System to adequately meet the demands of uprate.

With a RATED CORE POWER level of 2900 MWt, the calculated results (i.e., DNBR, l

Pressure, Peak Clad Temperature, Metal Water Reaction, Environmental Conditions l Inside and Outside Containment, etc.) are acceptable and remain within applicable

, regulatory acceptance criteria. The results further show that the integrity of the l primary / secondary / containment pressure boundary is not challenged and that the l

, extent of fuel failures during Condition III and IV events remains bounded by assumptions within the dose analyses. The calculated radiological consequences remain well within applicable regulatory limits.

Offsite Dose Limits will be maintained with the revision to the gas storage i specification. Although this is not specifically an uprate concern, it affects the
radiological consequences section in the SGR submittal (Ref. 3). The TS 3.11.2.6 j limit will decrease from 160,000 curies Noble Gas to 131,000 curies Noble Gas.
However, the station administrative limit of 90,000 curies Noble Gas is unchanged 4

and has never been exceeded. These gas tanks are sampled daily when adding to the tank to assure this limit is not exceed ed.

The uprate conditions will produce additional heat loads on the Spent Fuel Cooling System due to increased decay heat. Analyses indicate that the system has sufficient capacity to limit the pool temperature to less than 150*F during hmiting Normal heat loads and to less than bulk boiling during limiting Abnormal heat loads. In the event of a loss of spent fuel cooling, adequate time remains available to restore spent fuel cooling to preclude the onset of boiling. For the postulated condition of an extended loss of normal cooling, various makeup water sources are available on site with sufficient capacity to match the pool boiloff rate, thus precluding fuel uncovery.

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Docum:nt Centrol D:sk Attachm:ntIV TSP 950001 RC-95-0174 Page 3 of 3 Uprate power will not adversely afTect the operation of the Reactor Protection System, Engineering Safety Features, or other systems or components that are required for accident mitigation. The revised operating conditions will not affect these systems' performance or qualification for either normal operation or accident conditions. All calculated results to VCSNS FSAR Chapter 15 Analyses demonstrate that there are no challenges to the integrity of the fission product boundaries and the plant remains within the regulatory acceptance criteria applied to the VCSNS currentlicensing basis.

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' SIGNIFICANT HAZARDS EVALUATION

. FOR REVISING THE SPECIFICATION FOR - ,

! UPRATE VIRGIL C.-SUMMER NUCLEAR STATION TECHNICAL SPECIFICATIONS j i  !

Description of AmendmentReauest  !

i i South Carolina Electric & Gas Company (SCE&G) proposes to revise the Virgil C. i Summer Nuclear Station (VCSNS) Technical Specifications (TS) pages 1.5, 3/4.11-5, 6-16a, and Bases pages B3/4.2-1 and B3/4.2-3. These changes support the Uprate  :

, project and provide the following:  :

e a new definition of Rated Thermal Power (RTP) to incorporate the uprate power '

condition of 2900 MWt. This value represents the total heat transfer rate from the reactor core to the reactor coolant and does not include heat generated by the  ;

reactor coolant pumps.

e a revised limit for the quantity of radioactivity stored in any one gas storage tank.

This new value is based on the methodology, provided in NUREG 0133, and only 1 affects the quantity stored.  ;
e a new reference to the Core Operating Limits Report (COLR) which is based on <

l the Best Estimate Loss of Cooling Accident methodology, and a deletion of the reference to the BASH /BART methodology in the COLR specification. i e revised bases information to indicate that VCSNS meets the generic acceptance criteria of 10 CFR 50.46, rather than only using the ECCS acceptance criteria of ,

2200*F. This change is appropriate since the peak cladding temperature may not l

) be the most limiting criteria associated with the ECCS evaluations.

i 1 Many TS changes were required to support the Steam Generator Replacement (SGR), ,

i which were approved and issued via Reference 1. Many of the TS changes expected i 4 for a plant Uprate were included in the SGR submittal. Most evaluations performed l 4

for the SGR utilized 2900 MWt core power as an initial condition. l This Technical Specification Change Request (TSCR) primarily revises those areas in TS which were not included in Reference 3. The primary supporting analyses performed for uprata are: Large Break Loss of Cooling Accident (LOCA) utilizing the Westinghouse Best Estimate LOCA methodology, spent fuel pool cooling capacity

. analysis resulting from our outage practices, and Waste Gas Decay Tank Rupture analysis resulting from a comment included in the SER for SGR (Reference 1.). Other analyses and evaluations were performed to assess the capability of other systems and components to support Uprate, with the results indicating that both the Nuclear Steam Supply System (NSSS) and the Balance of Plant systems are capable of supporting uprate power operations, assuming several modifications to balance of plant systems.

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Basis for No Significant Hazards Consideration Determination  ;

South Carolina Electric & Gas Company (SCE&G) has evaluated the proposed changes to the VCSNS TS described above against the Significant Hazards Criteria of 10 CFR 50.92 and has determined that the changes do not involve any significant hazard for the followingreasons:

, 1. The probability or consequences of an accident previously evaluated is not significantly increased.

l Implementation of uprate power operation does not contribute to any accident  !

evaluated in the FSAR. The NSSS Components (RV, RCPs, CRDMs, SGs, and 4 piping) are compatible with the revised operating conditions. These components have been reanalyzed and the results show that ASME Code requirements remain satisfied and are within the current Licensing Basis.

3 Interfacing Systems which are important to safety are not adversely impacted .

t and will continue to perform their design function. Overall secondary plant performance is not significantly altered by the proposed changes.

Therefore, since the Reactor Coolant pressure boundary integrity and system functions are not adversely im3 acted, the probability of occurrence of an accident i 2

evaluated in the VCSNS FSAR will be no greater than the original design basis of

the plant.

s An extensive analysis has been performed to evaluate the consec uences of the  !

following accident types currently evaluated in the VCSNS FSA1:

I e Non-LOCA Events  !

i e Large Break and Small Break LOCA t l e Steam Generator Tube Rupture l With the A75 SGs and revised operating conditions, the calculated results (i.e.,

DNBR, Pri. mary and Secondary System Pressure, Peak Clad Tem?erature, Metal Water Reaction, Challenge to Long Term Cooling, Environmenta Conditions

. Inside and Outside containment, etc.) for the accidents are similar to those currently reported in the VCSNS FSAR and remain within applicable Regulatory  ;

4 Acceptance Criteria. Select results (i.e., Containment Pressure during a Steam  ;

Line Break, Minimum DNBR for Rod Withdrawal from Suberitical, etc.) are slightly more limiting than those currently reported in the FSAR due to the use of I the assumed operating conditions with the A75 SGs and in some cases use of an I uprated core power of 2900 MWt. However, in all cases, the calculated results do l j not challenge the integrity of the primary / secondary / containment pressure l boundary and remain within the regulatory acceptance criteria applied to i

VCSNS's currentlicensing basis. 1 1

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Given that calculated radiological consequences are not significantly higher than current FSAR results and remain well within 10CFR100 lunits, it is concluded i that the consequences of an accident previously evaluated in the FSAR are not significantly increased.

2. The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.  ;

Uprate power operation will not introduce any new accident initiator t mechanisms. Structural integrity of the RCS is maintained during all plant conditions through compliance with the ASME code. Design requirements of auxiliary systems are met with the RSGs and uprate power operation. No new

, failure modes or limiting single failures have been identified. Since the safety and design requirements continue to be met and the integrity of the reactor coolant system pressure boundary is not challenged, no new accident scenarios have been created. Therefore, the types of accidents defined in the FSAR

- continue to represent the credible spectrum of events to be analyzed which j determine safe plant operation.

3. The proposed license amendment does not involve a significant reduction in a j
margin of safety.

Although uprate power operation will require changes to the VCSNS Technical  !

Specifications, the proposed changes are supported by extensive LOCA, NON- l

LOCA and SGTR analyses. These analyses show acceptable consequences with
margin to the applicable regulatory limits. All equipment required to function i during accident conditions has bec n shown to remain qualified and thus will perform their design function, and all components remain in compliance with the codes and standards in effect when VCSNS was originally licensed (with the exception of the replacement steam generators which use the 1986 ASME Code Section III Edition). Based on the above,it is concluded that there is no .

significant reduction in a margin of safety. ,

)

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