ML20195H761

Addendum to 880520 Application for Amend to License NPF-12, Revising Tech Specs in Support of Refueling & Operating Plant W/Vantage 5 Improved Fuel Design.Addendum Reanalyzed T Average to 587.4 F

ML20195H761
Person / Time
Site:	Summer
Issue date:	06/20/1988
From:	Nauman D SOUTH CAROLINA ELECTRIC & GAS CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20195H766	List: ML20195H761 ML20195H770
References
NUDOCS 8806280375
Download: ML20195H761 (26)

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th Car fina Eloctric & Gas Company D n A. Na n Jenkinsville. SC 20065 Nuclear Operations SCE&G A 3* "

June 20, 1988 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Technical Specification Change Addendum - VANTAGE 5 Reload Gentlemen:

On May 20, 1988, South Carolina Electric & Gas Company (SCE&G) requested a revision to the Technical Specifications for the Virgil C. Summer Nuclear Station in support of refueling and operating the Virgil C. Summer plant with VANTAGE 5 improved fuel design. As previously provided, the attachments of the May 20, 1988 letter constituted a partial submittal for the transition to VANTAGE 5 fuel. The original analysis reduced Tave to 585.5'F from 587.4*F which would have resulted in a three megawatt electrical output penalty.

SCE&G elected to perform the reanalyses necessary to recover the electrical output.

This letter forwards an addendum to Attachment 1 of the original submittal which comprises the <scumentation necessary to complete the Technical Specifications change. request with Tave reanalyzed to 587.4*F. Enclosure 1 to the attached addendum contains the revised page changes to the original submittal. These should replace the pages from the original submittal on a one for one basis.

The enclosed Technical Specifications addendum has been reviewed and approved i by both the Plant Safety Review Committee and the Nuclear Safety Review Committee.

Should you have any questions concerning this Technical Specifications addendum, please call at your convenience.

/ rdtr[ly yours, l '\\ l

. . Nauman MDB/ DAN: led Attachment pc: See Page 2 na62mu 2$s 5 P

[.A q

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'l Document Control Desk June 20, 1988  !

Page 2 l pc: J. G. Connelly, Jr./0.~ W. Dixon, Jr./T. C. Nichols, Jr. I E. C. Roberts

0. S. Bradham W. A. Williams, Jr.

J. H. Grace J. J. Hayes, Jr.

1 General Managers C. A. Price R. B. Clary-W. R. Higgins R. M. Campbell, Jr.

K. E. Hodland J. C. Snelson G. O. Percival R. L. Prevatte J. B. Knotts, Jr.

H. G. Shealy M. D. Blue L. R. Cartin NSRC RTS (TSP 880013)

HPCF File (813.20) i l

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ADDENDUM TO ATTACHMENT 1 SUPPLEMENTAL SAFETY EVALUATION FOR THE VIRGIL C. SUMMER NUCLEAR STATION TRANSITION TO hTSTINGHOUSE 17 X 17 VANTAGE 5 FUEL WITH A TAVE OF 587.4 F SOUTH CAROLINA ELECTRIC & GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION i DOCKET No. 50-395 I

i JUNE, 1988 1

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1.0 INTRODUCTION

The safety evaluation supporting the Virgil C. Summer Nuclear Station (VCSNS) transition to VANTAGE 5 fuel was performed assuming a nominal average coolant temperature (Tave) of 585.5 F. This value was chosen in conjunction with other baseline parameters so that the RCS hot leg temperature remained at, or below, the original design value of 618.7 F with up to 15% steam generator tube plugging.

Bis limit on reactor coolant system (RCS) hot' leg temperature was originally imposed to minimize the corrosion rate of the steam generators wtilch is sensitive to high coolant temperatures. However, full power operation with a Tave of 585.5 F, instead of the current Tave of 587.4 F, would result in a decrease in the plant electrical output of approximately 3 megawatts due to a reduction in steam pressure.

Subsequent to the selection of the baseline parameters for the VANTAGE 5 transition evaluation, SCE&G established alternative programs to manage the impact of steam generator corrosion and thus desires to maintain the current Tave of 587.4 F for Cycle 5 and beyond. This adjustment in the VANTAGE 5 baseline Tave will not only prevent the loss of 3 megawatts electrical output but will also mininize the impact of the fuel transition on station operation (i.e. , Tave control, pressurizer level ,

control, steam generator level control, etc.) and operator training.

In order to minimize the impact of the 1.9 F increase in Tave on the VANTAGE 5 safety evaluation, existing margin in thennal design flow (TDF) has been utilized to offset the decrease in the departure from nucleate boiling ratio (DNBR) which would have occurred with an increase in Tave. Specifically, TDF will be increased by approximately 2.0%, from 277800 gallons per minute (gpm) to 283500 gpn. ne changes in Tave and TDF result in modifications to the VANTAGE 5 Technical Specifications revisions as previously forwarded to the Nuclear Regulatory Commission on May 20, 1988. Changes for these affected Technical Specifications are proposed herein.

R e purpose of this addendum to Attachment 1 of SCE&G's licensing submittal is to (1) justify the continued applicability of the results and conclusions of the VCSNS VANTAGE 5 licensing submittal with a revised Tave of 587.4 F and (2) provide the Technical Specifications page markups, reflecting the adjustment in Tave and TDF, to replace those previously proposed.

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2.0 BASELINE PARAMETER CHANGES Two primary modifications to the VCSNS VANTAGE 5 baseline parameters are proposed:

1. Increase the 100% full power Tave from 585.5 F to 587.4 F.

2. Increase TDF from 277800 gpm to 283500 gpm.

The increase in TEF provides an offset to the resulting decrease in DtiBR due to the increase in Tave. The increased TDF, however, remains adequate to allow up to 15%

unifom steam generator tube plugging without the need for additional safety analysis.

The above modifications affect other plant design parameters such as RCS hot leg and cold leg tenperatures. These secondary changes are illustrated in the change pages (see Table 5.1) in Enclosure 1.

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I 30 PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGES l

. Five modif'. cations to the previously proposed Technical Specifications changes, supporting the transition to VANTAGE 5 fuel, are required to reflect the adjustments in Tave and TDF. These modifications are described below:

1. Core Safety Limits (TS Figure 2.1-1)

The core safety limits, which are the loci of points of terminal power, RCS pre.ssure and Tave, below which the calculated DNBR is no less than the design DNBR or below which the core average enthalpy at the vessel exit is less than the s enthalpy of saturatri liquid, are revised to reflect the increase in RCS flow.

The revised limits continue to be those for the low parasitic (LOPAR) fuel which are limiting during the transition period.

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2. Loop Design Flow (TS Table 2.2-1) l l

The loop design flow, derived from the TDP and by the flow measurement j uncertainty (2.1%) for use in the reactor protection system low flow trip, is increased from 96200 to 96500 gpm. This change insures that the consequences of a loss of flow transient would be bounded by those predicted in the VANTAGE 5 safety analysis with a nominal Tave of 587.4 F.

3 Overpower-delta-T/0vertemperature-delta-T Trip Setpoints (TS Table 2.1-1, Notes 1

& 3)  !

The overpower and overtemperature delta-temperature (delta-T) trip setpoints are revised since they are a function of Tave and are derived from the core safety l limits. The reference Tave at rated thermal power (RTP) is modified from 585.5 F to 587.4 F; therefore, no changes in the current values for the Overtemperature-delta-T and Overpower-delta-T trips will be required to support the transition to VANTAGE 5. In addition, a minor slope change for the change in  ;

axial flux penalty tem in the Overtemperature-delta-T trip (item iii in note 1) is required due to the change in the core safety limits.

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4. RCS Total Flow Rate (TS Figure 3.2-2)-

The increase in system flow has been factored into the limiting conditions of operation defined by RCS total flow rate and the nuclear enthalpy rise hot channel factor within Technical Specifications Figure 3 2-2. The 100% RTP value corresponds to the minimum measured flow assumed in the VANTAGE 5 transition safety analysis.

5. DNB Parameters (TS Table 3.2-1)

The RCS Tave is modified to reflect the nominal value of 587.4 F plus the uncertainty for Tave (4.3 F) asstrned in the safety analysis utilizing the Improved Thermal Design Procedure.

No other modification to the proposed Technical Specifications changes to support transition to VANTAGE 5 fuel are required. Technical Specifications markup pages for the above modification are provided in Enclosure 1.

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4.0 SAFETY EVALUATION 4.1 Scope of Evaluation The impact of the proposed changes to the VANTAGE 5 baseline parameters on the results and conclusions of the original licensing submittal dated May 20, 1988 have been assessed with the fb11owing conclusions relative to the major technical topics addressed in Attachments 1, 3, 4, 5, 6, and 7 of the original submittal.

Mechanical Evaluation ( Attachment 1, Section 3.0)

The changes in baseline parameters do not impact the Mechanical Evaluation to the VCSNS license submittal since the calculations conducted to justify the mechanical integrity of the fuel assembly and core components utilized design basis parameters which are bounding for VCSNS. The results are ,

within the allowable design limits. The Tave change also results in a negligible increase in fuel pellet temperatures and fission gas release.

Therefore, the fuel rod design evaluation perfomed for the VCSNS licensing submittal remains valid.

Nuclear Evaluation ( Attachment 1, Section 4.0) l The change in Tave from 585.5 F to 587.4 F and increase in TDF from 277800 gpm to 283500 gpm will not adversely impact Attachment 1, Section 4.0 of the VCSNS licensing submittal. The nuclear parameters analyzed are i i

relatively insensitive to small changes in vessel temperature and flow so j any changes in the parameters can be accomodated with the uncertainty that was applied during the analysis. The VCSNS licensinS submittal conclusions remain valid. ,

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I Thermal Hydraulic Evaluation (Attachment 1, Section 5.0)

I For the Thermal Hydraulic Evaluation, Attachment 1, Section 5.0 of the VCSNS l

licensing submittal, the effects of a Tave change from 585.5 F to

) 587.40F and a TDF change from 277800 spn to 283500 gpm were evaluated.

The nct change resulted in a maximum DNBR penalty of 1%. There is sufficient margin between the safety analysis and design DNBR values to accomodate this 1% INBR penalty.

Accident Evaluation ( Attachment 1, Section 6.0 and Attacnments 3 and 4) i

'Ihe proposed changes in baseline parameters are key initial conditions to  ;

the accident analyses. Each accident analysis has been assessed for impact and a detailed overview is provided in Section 4.2. Overall, all l

conclusions in the VCSNS licensing sutnittal relative to accident analysis were confirmed to remain valid with no degradation in margin to safety.

Thimble Plug Removal (Attachment 5) j i

The Thimble Plug Removal Evaluation, Attachment 5 of the VCSN.5 licensing j submittal is not affected by the baseline parameter changes, because the l reactor flow resistances do not change. I l

Significant Hazards Evaluation ( Attachment 6)

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The engineering evaluation and additional accident analyses performed show that the technical bases for the Significant Hazards Evaluation is maintained. The proposed changes in baseline parameters do not invalidate the conclusions of the VCSNS licensing submittal, including those in the l Significant Hazards Evaluation. With changes to Table 1 of Attachment 6 (see Enclosure 1) to reflect changes in baseline parameters, it is concluded that the Significant Hazards Evaluation is appropriate for the VCSNS VANTAGE 5 Technical Specifications as modified by this addendum.

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Radiological Impact Assessment ( Attachment 7)

The discussion of Radiological Impact Assessment provided in Attachment 7 of the VCSNS licensing subaittal addresses the impact of extended fuel burnup on core fission product inventories and on the radiological consequences of accidents. The increase in Tave from 585.5 F to 587.4 F does not affect this discussion.

4.2 Accident Evaluation The impact of the proposed changes in baseline parameters has been assessed for each of the non-loss-of-coolant-accident (LOCA) and LOCA safety analyses reported in Attachments 1, 3, and 4 of the VCSNS licensing submittal, a summary of this input assessment is provided below. Each section title provides a cross-reference to the corresponding section of the previous -

licensing submittal.

Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Suberitical Condition (Attachment 3, Section 15.2.1)

An increase in the nominal at power Tave from 585.5 F to 587.4 F will not have an effect on this transient. This transient is analyzed under hot zero power conditions and is not impacted by changes in the at power reactor  !

coolant conditions. In addition, a 2% increase in flow would have a beneficial impact on the calculated minimtn DNBR. Bus, for this event, the conclusions in the VCSNS licensing submittal remain valid.

Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (Attachment 3, Section 15.2.2) i Eis analysis is perfonned at various power levels, asstning minimun and maximum feedback, for a range of reactivity insertion rates. Protection is provided by the overtemperature delta-T and high neutron flux-high setting reactor trips. Revised overtemperature delta-T setpoints have been calculated based on a nominal Tave of S87.4 F and a minimum measured flow of 289500 gpm. These setpoints ensure that departure from nucleate boiling (IEB) will not occur over the range of reactivity insertion rates analyzed.

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Proper calibration of the excore neutron flux detectors at the revised RCS ,

average temperature will ensure that the high neutron flux trip function  !

will provide adequate protection. For those reactivity insertion rates which rely upon the high neutron flux trip for protection, the INB penalty associated with the temperature increase is offset by the increase in RCS flow.

Thus, the revised overtemperature delta-T setpoints ensure that the conclusions in the VCSNS licensing submittal remain valid.

Rod Cluster Control Assembly Misoperation ( Attschment 3, Section 15.2.3)

Three rod cluster control assembly (RCCA) misalignment accidents are presented in VCSNS licensing submittal Section 15.2.3: one or more dropped RCCAs within the same group, a dropped bank, and a statically misaligned RCCA. The analysis presented in the VCSNS licensing submittal demonstrates i that for all cases, the DNB design basis is met. Evaluations have sMwn that, with an increase in Tave to 587.4 F and an increase in the minimum measured flow to 289500 gpm, the DNB design basis in still met for each of '

these events.

Thus, the conclusions in the VCSNS licensing submittal remain valid. ,

Uncontrolled Boron Dilution ( Attachment 3, Section 15.2.4)

The VCSNS licensing submittal presents the resul;s of analyses performed in Modes 1 through 6. D e analyses demonstrate that during a dilution event, adequate operator action time is available to detect and tenninate the dilution before shutdown margin is lost. Verification that an increase in Tave to 587.4 F will not impact the core physics characteristics has been '

made. Since the reactor coolant flow has no impact upon this safety ,

analysis, the behavior of the reactor during the dilution event and the time at which shutdown margin is lost has not been impacted.

R us, the conclusions in the VCSNS licensing submittal remain valid.

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Partial Loss of Forced Coo] ant Flow ( Attachment 3, Section 15.2.5)

The evaluation for the partial loss of ferced coolant flow analysis presented in Attachment 3, Section 15.2.5 is discussed with the complete loss of flow evaluation.

Startup of e 1 Inactiv? R(actoc Coolant Loop ( Attachment ?, Section 15.2.6)

The analysie presented in the VCSNS licensing submittal is initiated from 60% power. The initi.al Tave corresponding to this power is 574.1 F (without uncertainties) for the analysis prevnted in the VCSNS licensing cubmittal. An increase in nominal Tave to 587.4 F would represent a Tave of 575.2 F at 60% power, a 1.1 F increase from the VCSNS licensing submittal case.

he transient is termit ated by a reactor trip on low coolant loop flow when the power range neutron flux exceeds the 1-8 setpoint which is conservativly assumed to be 74% power. he nuclear ;cwer transient is driven by the effect of the moderator tecperature ccefficient and the reduction in the RC3 temperature caused by the forced flow of colder water (cold leg temperature) to the core frcr 2 he inactive loop.

A 2% increase in the TEf' to 233500 gpm will result in a decrebe in the hot  ;

leg La cold leg delta-T, Thus, the driving force of the nuclear power transient discuss +d above will be slightly dimir.ished, resulting in a slower i transient. Howe'rer, reactor trip will still occur upon reaching the P-8 setpoint. The small increase in the initial average temperature would not have a significant effect on the transient benavior of this event.

An exarnination was made of the available Df5 margin. Due to tha large margin cvailable, a 1.1 F increase ir Tave would not have a significant adverse impact upon the available DfE margin. In addition, a 2% increase in flow would have a baneficie.1 impaat u on the minimum DriBR.

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Additionally, an examination was made of the available margin to the overpressurization limit of 110% design pressure. The maximum RCS pressure calculated in the VCSNS licensing submittal is 2247 psia with a limit of 2750 psia. On the secondary side, the maximum steam pressure was calculated to be 1084 psia with a limit of 1320 psia. Due to the margins available, the changes in RCS average temperature and flow will not have a significant '

adverse impact on the values presented in the VCSNS licensing submittal.

Rus, for this event, the conclusions in the VCSNS licensing submittal remain valid.

Mss of External Electrical Load and/or Turbine Trip ( Attachment 3, section 15.2.7)

A reanalysis of the loss of load / turbine trip event presented in the. VCSNS licensing submittal has been parformed. TM revised analysis assumes initial conditions consistent '- a nominal Tave of 587.4 F; however, no credit was taken for any increase ;n RCS flow over the value assmed in the VCSNS licensing submittal analysis. Bis is a conservative assmption for this event. The analysis demonstrates that the DNB design basis is met and that overpressurization of the primary and secondary systems does not occur.

Rus, the conclusions in the VCSNS licensing submittal remain valid.

Loss of Nomal Feedwater ( Attachment 3, section 15.2.8)

A reanalysis of the loss of normal feedwater event presented in the VCSNS licensing submittal has been perfomed. Tne revised analysis assmes initial conditions consistent with a nominal Tave of 587.4 F. However, no credit was taken for any increase in RCS flow over the value assmed in the VCSNS licensing submittal analysis. This is a conservative assmption fbr l this event. The analysis demonstrates that the core is not adversly ,

affected and that no water relief occurs through the pressurizer safety or l relief valves.

l Rus, the conclusions in the VCSNS licensing submittal remain valid.

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Loss of Offsite_ Power to the Station Auxiliaries (Station Blackout)

( Attachn;ent 3, Section 15.2.9)

The results of this event which could be impacted by the subject increases in Tave and minimum measured flow are bounded by the results for the loss of nonnal feedwater evaluation discussed above. .

Excessive Heat Removal Due to Feedwater System Malfunctions ( Attcchment 3, Section 15.2.10)

The analysis presented in the VCSNS licensing submittal simulates the opening of one feedwater control valve. Two cases are analyzed: one with the reactor just critical at zero power with a moderator density coefficient characteristic of end-of-life conditions and second with the reactor in manual control at full power.

An increase in the nominal Tave to 587.4 F will have no impact on the zero power case. Thus, the maxLmum reactivity insertion rate analyzed in Section 15.2.1 of the VCSNS licensing submittal, Uncontrolled RCCA Bank Withdrawal frce a Suberitical Condition, will still bound that calculated for the feedwater system malfunction at zero power. In addition, a 2% increase in the minimum measured flow will provide a DNB benefit for this analysis.

For the full power case, excessive feedwater flow results in an increase in core heat flux, pressurizer pressure and RCS delta-T ano a decrease in Tave. Reactor trip ccours on high-high steam generator level. An increase '

in Tave to 587.4 F will not have an adverse impact on the transient behavior. The time of reactor trip would be essentially unchanged since i

this is primarily a function of the feedwater flow increase. In addition, a 25 increase in fle's would provide a DNB benefit.

t Thus, the conclusions in the VCSNS licensing submittal will remain valid.

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Excessive Load Increase Incident ( Attachment 3, Section 15.2.11)

The analysis presented in the VCSNS licensing submittal demonstrates plant behavior following a 10% step load increase from full power. Four cases are analyzed: at the beginning and end-of-core life, each considered with the reactor in manual and automatic control. In all four cases, reactor trip does not occur and the plant reaches a stabilized condition following the load increase.

An increase in Tave to 587.4 F will not have a significant adverse impact upon the transient behavior calculated in the VCSNS licensing submittal. As can be seen in the VCSNS licet. Ing submittal, the excessive load increase events without automatic control result in a reduction in Tsve due to increased heat extraction by the secondary side. Should these events be initiated with a 1.9 F increase in Tave, the cate and magnitude of the Tave reduction would be unchanged. The result would merely be a 1.9 F increase in the stable temperature when it is finally reached.

Conversely, for those cases in which automatic control is assumed to be ,

available, Tave is essentially maintained at the initial value throughout the transient. Thus, a 1.9 F increase in the initial Tave will result in a 1.9 F increase in the stable temperature calculated in the VCSNS licensing submittal.

. t For all of the above cases, a ?% increase in the RCS flow would tend to reduce the time required for the primary system to achieve a new stable condition. This is due to the improved heat transfer characteristics which result from the higher flow. In addition, the higher RCS flow would serve ,

to increase the margin to DNB calculated in the VCSNS licensing submittal. <

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L Accidental Depressurization of the Reactor Coolant System ( Attachment 3, Section 15.2.12)

Re' analysis presented in the VCSNS licensing submittal simulates the inadvertent opening of a pressurizer safety valve. Bis results in a rapid .

depressurization of the RCS until saturation occurs in the hot leg at which  !

time depressurizations continues, but at a-slower rate. Be reactor is .

tripped on an overtemperature delta-T signal.

An increase in the nominal Tave to 587.4 F and a minimum measured flow of-289500 gpn will have no significant impact upon the transient. The transient behavior (e.g. , the RCS depressurization) is driven by the safety valve relief rate which will not be significantly impacted by the change in

, RCS flow or the small increase in temperature.

As discussed above, revised overtemperature delta *T setpoints have been calcuated based on a nominal Tave of 587.4 F and a minimum measured RCS ficw of 289500 ggn. - These setpoints will maintain the initial margin to trip and provide a resotor trip signal with the same margin to DNB as in the VCSNS licensing ..u mittal analysis. .

i Thus, the revised overtemperature delta-T setpoints will ensure that the DtB .

design basis is met and the conclusions in the VCSNS licensing submittal remain valid.

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Accidental Ibpressurization of the Main Steam System ( Attachment 3, Section 15.2.13)

An increase in the nominal at power Tave from 585.5 F to 587.4 F will not have an effect on this transient. This transient is analyzed under hot zero power conditions and is not impacted by enanges in the at power reactor coolant conditions. In addition, sensitivity studies (WCAP 9227) show that a 2% increase in flow would have essentially no impact on the calculated minimum DiBR. Thus, for this event, the conclusions in the VCSNS licensing sutraittal remain valid.

Spurious Operation of the Safety Injection System at Power ( Attachment 3, Section 15.2.14)

Two cases are discussed in the VCSNS licensing submittal. In Case A the reactor trip signal occurs at the same time as the spurious safety injection-begins. For this case the operator would tenninate safety injection and I

bring the plant to a cold shutdown condition. The > asults for this case are unchanged for an increase in the nominal Tave to 587.4 F and an increase in the minimum measured flow to 289500 gpn.

j In Case B, reactor trip is not ir.itiated by a safety injection signal, but I

occurs later on low pressurizer pressure. Throughout the Case B transient, the average temperature, nucleac power, pressurizer water level and pressure decrease due to the negative reactivity provided by the injection of borated l water. The minimtra DiBR is never less than the initial value.

Should this event be initiated with a higher Tave of 587.4 F and a 21 increase in minimtzn measured flow, the time of reactor tr$p would renain essentially unchanged since the depressurization of the RCS is driven by the negative reactivity provided by the safety injcetion flow. Thus , t he behavior of this transient would be unchanged and the minimum DiBR will increase throughout the transient.

Thus, the conclusiors in the VCSNS licensing submittal remain valid.

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Minor Secondary System Pipe Breaks ( Attachment 3, Section 15 3.2)

An increase in the nominal at power Tave from 585.5 F to 587.4 F will not have an effect on this transient. This transient is analyzed under hot zero power conditions and is not impacted by changes in the at power reactor coolant conditions. In addition, sensitivity studies (WCAP 9227) show that a 2% increase in flow would have essentially no impact on the calculated minimum DiBR. Thus, for this event, the conclusions in the VCSNS licensing submittal remain valid.

, Inadvertent Loading of a Fuel Assembly Into An Improper Position ( Attachment 1 Section 15 3 3)

An evaluation has been perfonted for this event. An increase in the nominal Tave to 587.4 F, as well as an increase in the RCS flow rate will not have an dverse Ltpact upon the analysis presented in the VCSNS licensing submittal. Thus, the conclusions in the VCSNS licensing submittal renain valid.

Complete Loss of Forced Reactor Coolant Flow ( Attachment 3, Section 15.3.4)

A reanalysis of the most limiting of the 1 of 3 loops loss of flow (partial loss of flow) and 3 of 3 loops loss of flow (c splete loss of flow) cases has been perforr.ed. This analysis assmed nominal conditions consistent with a vessel average temperature of 587.4 F and a minimum measured flow of 289500 gpn. All other assmptions were identical to those made in the VCSNS licensing submittal analysis. The results demonstrate that the DJB design basis is net for both the partial and complete loss of flow events.

Thus, the conclusions in sections 15.2.5 and 15.3.4 of Attachment 3 of the VCSNS licensing submittal remain valid.

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Single Rod Cluster Control Assembly Withdrawal at Full Power ( Attachment 3, Section 15 3.6)

An evaluation has been performed for this event. An increase in the nominal-Tave to 587.40 F, as well as an increase in the RCS flow rate will not have an adverse impact upon the analysis presented in the VCSNS licensing submittal. Thus, the conclusions in the VCSNS licensing submittal remain valid.

Major Secondary System Pipe Ruptures ( Attachment 3, Section 15.4.2)

Two analyses are presented in Attachment 3, Section 15.4.2: rupture of a main steam line and a major rupture of a main feedwater pipe.

Rupture of a Pain Steam Line ( Attachment 3, Section 15.4.2.1)

An increase in the nominal at power Tave from 585.5 F to 587.4 F would not have an effect on this transient. This transient is analyzed under hot zero power conditions and is not impacted by changes in the at power reactor coolant conditions. In addition, sensitivity studies (WCAP 9227) show that a 2% increase in flow would have essentially no impact on the calculated minimum NBR.

Thus, for this event, the conclusions in the VCSNS licensing submittal remain valid.

Major Rupture of a Main Feedwater Pipe ( Attachment 3, Section 15.4.2.2)

A reanalysis of the feedline break event has been perfomed. The revised analysis was perfomed at engineered safeguards design power assming initial conditions consistent with a nominal Tave of 587.4 F. However, no credit was taken for any increase in RCS flow over the value assmed in the VCSNS licensing submittal analysis, which is a conservative assmption for this event. The analysis demon =trates that the core retuains covered and  ;

that overpressurization of the RCS does not occur.  ;

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Single Reactor Coolant Pump Locked Rotor ( Attachment ?, Section 15.4.4)

A reanalysis of the locked rotor event has been performed. This analysis assumed nominal conditions consistent with a Tave of 587.4 F and a 2%

increase in the minimm measured flow, all other assumptions were identical to those made in the VCSNS licensing sutmittal analysis. The maximm RCS pressure was calculated to be 2605 psia, a reduction of 1 psi from the analysis presented in the VCSNS licensing submittal. The maximum clad temperature at the hot spot was calculated to be 19740 F, 9 F less than that presented in the VCSNS licensing submittal, and the amount of zirconium-water reacting at the hot spot is 0.55% by weight. The results also demt tstrate that no more than 15% of the fuel rods will experience DNB.

Bus, the conclusions in the VCSNS licensing submittal remain valid.

Rupture of a Control Bod Wive Mechanism Housing (Rod Cluster Control Assembly Ejection) (Attachment 3, Section 15.4.6)

Four cases are analyzed in the VCSNS licensing submittal, at the beginning and end-of-life, each considered under hot zero power and full power conditions. As can be seen in the VCSNS licensing submittal, those analyses perforned under hot zero power conditions approach the peak clad tmperature limit of 2700 0F, while the hot full power cases approach the fuel melt limit of 10% at the hot spot, i

l An increase in tha nominal at power Tave from 585.5 F to 587.4 F will

, not have an effect on the hot zero power transient cases. These cases are analyzed under hot zero power conditions and are not impacted by changes in the at power reactor coolant conditions. In addition, a 2% increase in the reactor coolant flow would provio) improved fuel to moderator heat transfer, servirg to reduce the calculated peak clad temperatures presented in the VCSNS licensing sutmittal.

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1 The hot full power analyses will not be adversely impacted by an increase in  !

Tave to 587.4 F. Ibe to the inherent thennal lag in the fuel pellet, fuel melt occurs prior to effective heat tranfer to the clad and coolant. Thus, the amount of energy deposition in the pellet following the ejection of a  ;

control assembly and the degree of fuel melt is not sensitive to Tave. In addition, the analysis presented in the VCSNS licensing submittal conservatively assumes that DNB occurs immediately following ejection of t.he control assembly. This conservatively degrades heat transfer capability to the coolant, thereby reducing any heat transfer temperature dependency.

Note that a 1.9 F increase in Tave would slightly increase the initial pellet average temperature. However, this effect is very small and will not significantly impact the results. Similarly, due to the thermal lag in the pellet, a 2% increase in the reactor coolant flow would have no significant beneficial impact on the degree of fuel melt.

Thus, for this event, the conclusions in the VCSNS licensing submittal remain valid.

Steam Line Break Mass & Energy Releases (Attachment 1, Section 6.1.4)

An examination was made of the steam line break mass and energy (WE) l release analyses performed for the VCSNS plant with standard fuel as well as VANTAGE 5 fuel. Both of these analyses conservatively assume a Tave of 587.4 F and a TDF of 294000 gpm. An increase in Tave from 585.5 F (as ,

assumed in the VCSNS licensing submittal) to 587.4 F will not have an adverse impact on the steam line break WE release analysis since this condition is already accommodated.

In addition, the revised TDF of 283500 gpm is lower than the value of 294000 gpm asstrned in the WE analysis. Reduced RCS flows reduce the primary to secondary heat transfer characteristics which will impede the ability to transfer heat to the effluent. Thus, the enthalpy of the fluid released to containment will be reduced.

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Therefore, the steam line break WE releases previously calculated for V. C.

Summer, will remain bounding fbr VANTAGE 5 fuel with a Tave of 587.4 F and a TEF of 283500 spn.

Large Break LOCA (Attachment 4, Section 15.4.1) ne licensing basis for the large break LOCA analysis for VCSNS was perfomed to bound a range of operating temperature conditions.

Sensitivities to the range of tmperature conditions detemined that the higher operating temperature conditions yielo'ed higher peak clad temperatures. Thus the break spectrum was analyzed at the higher operating temperature conditions. The Tave used in the LOCA analysis for the upper range of plant temperatures bounds the revised Tave of 587.4o f. Thus it can be concluded that the revised Tave for VCSNS will not have an adverse _

effect on the large break LOCA results. ,

l Small Break LOCA (Attachment 4, Section 15 3 1) l The small break LOCA licensing analysis for VCSNS included calculations to investigate the influence of loop operating temperatures on peak clad temperature results. The results of these calculations showed that peak clad temperatures were predicted to decrease with decreasing initial loop temperatures in the range of several degrees from ncrainal. Based upon this behavior the actual licensing calculations were performed at the higher, nominal operating tmperatures. The Tave used in the LOCA analysis for the upper range of plant temperatures bounds the revised Tave of 587.4 F; therefore, the calculations reflect the proper conditions for the licensing basis, hus, it can be concluded that the revised Tave fbr VCSNS will not have an adverse effect on the small break LOCA results.

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Mass and Energy Releases (Attachment 1, Section 6.2.4)

Evaluations were perfomed for VCSNS for reduced temperature and TDF. These evaluations were perfomed to support the transition to VANTAGE 5 fuel.-

Since VCSNS requested a return to a Tave of 587.4 F, a discussion of the -

effects on the long-tem WE releases follows.  ;

For the long-tem WE release calculations, an evaluation was perfomed for the effects of a Tave of 587.4 F. Bis evaluation shows that the results of the current FSAR analysis would remain valid and unaffected. The FSAR containment integrity analysis would still be applicable since the WE releases remain unchanged.

Steam Generator Tube Rupture - (Attachment 1, Section 6.2.5)

For the steam generator tube rupture (SGTR) event, an analysis was performed to evaluate the radiological consequences of this accident. We major factors that affect the radiological doses of an SGTR event are the amount of primary coolant transferred to the secondary side of the ruptured steam generator through the ruptured tube after reactor trip and the steam released from the ruptured steam generator to the atmosphere. The impact on these factors of operating with nominal Tave of 587.4 F,15% steam generator tube plugging, and VANTAGE 5 fuel has been detemined.

A sensitivity analysis was completed to assess the effect of these parameters using the original VCSNS SGTR design basis analysis asstrnptions.

The results of the sensitivity analysis which incorporated a nominal Tave  ;

of 587.4 F, 15% steam generator tube plugging and VANTAGE 5 fuel parameters indicate that the primary to secondary break flow will increase by less than 1% while the steam released via the ruptured steam generator will decrease by 4.9%. In addition, the offsite radiation doses considering the net effect of the above parameters have been calculated to be less than that reported for the VCSNS SGTR analysis. Berefore, the SGTR radiological consequences reported in the VCSNS FSAR are valid for this scenario and remain well within the 10CFR100 limits.

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l LOCA Hydraulic Forces (Attachment 1, Section 6.2.6)

The LOCA hydraulic forcin6 functions are considered as part of Attachment 1, Section 6.2.6 of the VCSNS licensing submittal. For LOCA hydraulic forces considerations, the RCS fluid temperatures are used as analysis input. Two of the major factors in detemining the magnitude of the calculated LOCA hydraulic forcing functions from a postulated LOCA is the RCS temperature and pressure. It has been detemined that increasing the Tave from .

585.5 F to 5t'7.4 F decreases the magnitude of the peak LOCA hydraulic forcing functions. Since, Tave is being increased from 585.5 F to 587.40 F no design or regulatory limits will be violated with respect to the LOCA hydraulic forcing functions.

l Post-LOCA kng-Tem Core Cooling ( Attachment 1, Section 6.2.7)

Post-LOCA long-term core cooling is discussed in Attachment 1, Section 6.2.7 of the VCSNS licensing submittal. The Westinghouse licensing position for .

satisfying the requirements of 10CFR50.46(b)(5) "Wng Tem Cooling" concludes that the reactor will renain shutdown by borated mergency core cooling system (ECCS) water residing in the RCS/ containment sump after a LOCA. Since credit for the control rods is not taken for a large break LOCA, the borated ECCS water provided by the accumulators and the refueling water storage tank (RWST) must have a boron concentration that, when mixed with other water sources, will result in the reactor core remaining suberitical assuming all control rods out. The calculation is based upon the reactor steady state conditions at the initiation of a LOCA and l considers sources of both borated and unborated fluid in the containment l sump post-LOCA. It has been determined that increasing Tave from 585.5 F to 587.4 F will resCt in small beneficial increases in the RCS/ containment sump boron concentration. On this basis, an evaltation of the post-LOCA long-term core cooling calculation has been demonstrated that there will be no decrease in the calculated RCS/ containment stnp boron concentration after a postulated LOCA for VCSNS with no violations of design or regulatory limits, i

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Hot Leg Switchover To Prevent Boron Precipitation ( Attachment 1, Section 6.2.7)

Post-LOCA hot leg recirculation switchover time is detennined for inclusion in emergency procedures to ensure no boron precipitation in the reactor vessel following boiling in the core. This time is dependent on power level, and the RCS, BWST and accumulator water volumes and boron concentrations. Since the revised Tave for VCSNS will not affect the maximum baron concentrations or volumes associated with the RCS, BWST and accumulators, there is no effect on the post-LOCA hot leg switchover time, i

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i 5.0 ' SUWARY AND CONCLUSION  ;

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Revisions to the VANTAGE 5 baseline parameters to reflect an increase in nominal Tave to 587.4 F and in TDF to 283500 gpa have been shown by engineering evaluation and additional safety analysis to have a minor impact on the results of SCE&G's licensing submittal supporting the transition to VANTAGE 5 fuel. All conclusions in the licensing submittal, including those in the Significant Hazards Evaluation, remain applicable for the proposed changes in baseline parameters and resu2 ting Technical Specifications changes. Thus, it is concluded that VCSNS can be safety operated at 100% full power while maintaining the current design Tave of 587.4 F during the transition through mixed fuel cores to an all VANTAGE 5 core. ,

i Enclosure 1 contains change pages which reflect the impact of the baseline parameter changes on Attachment 1, 2, 5, 6, and 7. Changes to Attachment 3 and 4 are not deemed necessary because of minor impact on the calculated results of ,

the accident analysis, i

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