Text

Proposed Tech Specs,Consisting of Change Request 95-01, Permitting VCSNS to Operate at Uprate Power Level of 2900 Mwt Core Power When Unit Restarted After Ninth Refueling Outage
	ML20094E411
Person / Time
Site:	Summer
Issue date:	11/01/1995
From:	; SOUTH CAROLINA ELECTRIC & GAS CO.
To:	;
Shared Package
ML20094E405	List: ML20094E411; RC-95-0258, Application for Amend to License NPF-12,consisting of Change Request 95-01,permitting VCSNS to Operate at Uprate Power Level of 2900 Mwt Core Power When Unit Restarted After Ninth Refueling Outage;
References
	NUDOCS 9511070229
	Download: ML20094E411 (16)
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i 1.23 free a confinement to maintain temperaturePungE or PUMIM ts th i

i required to purify the confinement.er other operating condition, in QUADRANT scWER TILT RATIO 1 1.24 1

detector calibrated output to the average of the ugg -

i brated outputs. or the ratie of the maxima lamer encerg detector calibra output is greater. to the Withaverage of the one encore detector lower inoencore detector calibrated outputs, w sna11 De used for coguting the average. perette, the runntning three detectors a

! RATED THERMAL POWER

'4 i 1.25 the reactor coolant of C" ^^1 RATED THERMAL POWER s i atAcTOR TRIP rysTEM REsM Ttw C l

1.28 i

whenThe the monitored REACTOR T1tIPescaeds parameter SY3 TEM RESPt315E its trip setpoint at theTIE shallse channel until loss of stationary gripper coil voltage.

f REPORTABLE EVtWT 4

! 1.27 Section 50.73 to 10 CFR Part 50.A REPORTABLE EVENT shall be any SHUTOOWN MARGIN 1.28 SNilTDOWN M4 MIN shall be the instantaneous amount of reactivity b i the reactor is suscritical or would be suberitical from its present condition i assuming all full length red clustar assemblies (shutdown and control) are

fully inserted worth which is assumed except for thefully to be single red cluster asseely of hghest reactivity withdraus.

i SLAvt RftAY TEIT 1.29 A s u YE RE MY TEST shall be the ener i

verification of OPERASILITY of each relay.gitation of each slave relay and The SLAVE RELAY TEST shall include

! a continuity check, as a sinism, of associated tastable actuation devices.

l 1.30 Mot Used 4

i 500tti CHECK 1.31 i A SOURCE CHECK shall be the qualitative assessment of channel response wnen the channel sensor is assosed to a radioactive source.

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8 DEFINITIONS 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 operatin condition, in such a manner that replacement air or gas is required to puri the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shallbe 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 whichever output to With is greater. the average one excoreofdetector the lower excore the inoperable, detector calibrated remaining three outputs, detectors shall be used for computing the average.

RATED THER$1AL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2900 MWt. [

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

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shallbe any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shallbe theinstantaneousamountofreactivityby which the reactor is suberitical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SLAVE RELAY TEST 1.29 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shallinclude a continuity check, as a minimum, of associated testable actuation devices.

1.30 Not Used SOURCE CHECK 1.31 A SOURCE CHECK shallbe the qualitative assessment ofchannel response when the channel sensor is exposed to a radioactive source.

SUMMER - UNIT 1 15 Amendment No. 35,104,117, l

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i SUMER - UNIT 1 3/4 4-32 Amendment 33, 113 i

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SUMMER - UNIT.1 3/4 4-32 Amenament flo. 33,11 f % o m-

RADI0 ACTIVE EFFLUENTS GASSTORAGETANkS LIMITING CONDITION FOR OPERATION l 1

3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to; Ukes 200 curies noble gases (considered as Xe-133). f 3 jt o.o APPLICABILITY: At all times.

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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit.

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

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.

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

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1 RADIOACTIVE EFFLUENTS I GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION l 1

3.11.2.6 The quantity of radioactivity contained in each gas storage tank l

' shall be limited to less than or equal to 131,000 curies noble gases l l

{ (considered as Xe-133).

} l APPLICABILITY: . At all times.

ACTION:

a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit.

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

! 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 one per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.

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

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient 3.1.3.5 - Shutdown Bank Insertion Limit. 3.1.3.6 -

Control Bank Insertion Limit. 3.2.1 - Axial Flux Differe~nce,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 SPECIFICATION", February 1994 (W Proprietary).

(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (FQ Methodology for W(Z) surveillance requirements).)

c. WCAP-10266-P-A.Rev. 2. "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 198pWProprietary).

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

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and a'ecident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid cvele revisions or l 1 supplements there to shall be provided upon issuance, for each relo'ad cycle, to the NRC Document Control Desk with copies to the Regional Admmistrator and i Resident inspector. ;

["%mincluding Addendum 2-A, w

7

" BASH METHODOLOGY IMPROVEMENTS AND b' RELIABILITY ENHANCEMENTS," MAY 1988, 1

/

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

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I ADMINISTRATIVE CONTROLS CORE OPERATING LI511TS REPORT (Continued)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient 3.1.3.5 - Shutdown Bank Insertion Limit,3.1.3.6 -

Control Bank Insertion Limit,3.2.1 Axial Flux DifTerence,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

. SPECIFICATION" February 1994 (W Proprietary).

(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (FQ Methodology for W(Z) surveillance requirements).)

c. WC AP-10266-P-A, Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987; Including Addendum 2-A," BASH METHODOLOGY 151 PROVE 51ENTS AND RELIABILITY ENHANCE 51ENTS," MAY 1988, (W Proprietary).

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

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements there to shall be provided upon issuance, for each reload cycle, i to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector, l

d

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

- - -= _ _ . - .- . - - - __ - -

Document Control Desk AttachmentIV TSP 950001 RC-95-0258 Page 1 of 3 SAFETY EVALUATION l

FOR REVISING THE SPECIFICATION FOR UPRATE VIRGIL C. SUMMER NUCLEAR STATION i TECHNICAL SPECIFICATIONS 3

Description of Amendment Reauest I South Carolina Electric & Gas Company (SCE&G) proposes to revise the following-

Virgil C. Summer Nuclear Station (VCSNS) Technical Specifications (TS) pages: 1-5, e 3/4 4-31,3/4 4-32, 3/4.11-5, and 6-16a. These s:hanges support the Uprate project and

! provide the following: ,

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

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

This new value is based on the methodology in NUREG 0133 and only affects the j

maximum quantity stored.

e a new reference to the Core Operating Limits Report (COLR) which is based on the BASH /B ART methodology for Large Break Loss of Coolant Accident analysis.

l

! e revision to the Pressure Temperature Limitations Curves due to effects ofincreased

! neutron fluence at 2900 MWt.

1

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 i plant Uprate were included in the SGR submittal. Most evaluations performed for SGR

utilized 2900 MWt core power as an initial condition. i i

This Technical Specification Change Request (TSCR) primarily revises those areas in

! TS which were not included in Reference 3. The primary supporting analyses 3 performed for uprate are: Large Break Loss of Cooling Accident (LOCA) utilizing the Westinghouse 1981 Evaluation Model with B ASH, 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 4

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.

I Increased neutron fluence resulting from uprate core conditions has an effect on the

reactor vessel Pressure Temperature Curves. Their applicability will change from 14 Effective Full Power Years (EFPY) to 13 EFPY with no other changes at this time. 1 l

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Document Control D:sk i t

AttachmentIV TSP 950001 1 RC-95-0258 l l Page 2 0f 3 i

Safety Evaluation The conditions that result from uprate power are increased heat transferred from the

Reactor core, increased steam flow, increased feedwater flow, and increased electrical

! output. The additional heat load of approximately 4.5 percent can be met with the existing capacities of all NSSS and interfacing systems.

4 Modifications such as Closed Cycle Cooling are being planned to improve the capability

of secondary systems to meet the additional load.

i The increase in the secondary mass flow rates has been evaluated and does not present i any concerns. The A75 steam generators are rated for this condition and comply with all ASME Code requirements. The condenser, piping, and valves have allbeen i 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 RF-8 to reduce l the effects of erosion / corrosion.

f The additional heat produced will generate additional electricity. The turbine- f

generator has been evaluated and is capable, with a modification to the Stator Water j 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 j Inside and Outside Containment, etc.) are acceptable and remain within applicable regulatory acceptance criteria. The results further show that the integrity of the primary / secondary / containment pressure boundary is not challenged and that the 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 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 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 and has never been exceeded. These gas tanks are sampled daily when adding to the tank to assure this limit is not exceeded.

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 limiting 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.

l

Document Control D:sk

, Attachmt.nt IV TSP 950001

- RC-95-0258 Page 3 of 3

The Pressure Temperature Limitations Curves are derived using NRC Approved Methodology to comply with 10 CFR 50, Appendix G. These curves provide an acceptable range of operating temperatures and pressures for heatup, cooldown, low temperature overpressure, criticality, and inservice leak and hydrostatic testing conditions. The reduction in applicability for these curves has no effect on the curves themselves. Only the amount of time between the next scheduled specimen capusle analysis and the next revision to these curves will be effected.

Uprate power will not adversely affect the operation of the Reactor Protection System, Engineering Safety Features, or other systems or components that are required for 4

accident mitigation. The revised operating conditions will not affect these systems' performance or qualification for either normal operation or accident conditions. The 4

calculated results to VCSNS FSAR Chapter 15 Analyses demonstrate that there are no challenges to the integrity of the primary / secondary / containment pressure boundaries i and that the plant remains within the regulatory acceptance criteria applied to the VCSNS currentlicensing basis.

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I Document Control Desk Attachmsnt V TSP 950001 RC-95-0258 Page 1 of 3 SIGNIFICANT HAZARDS EVALUATION FOR REVISING THE SPECIFICATION FOR UPRATE 1 VIRGIL C. SUMMER NUCLEAR STATION TECHNICAL SPECIFICATIONS

Description of Amendment Request l South Carolina Electric & Gas Company (SCE&G) proposes to revise the following Virgil C. Summer Nuclear Station (VCSN S) Technical Specifications (TS) pages: 1-5,

3/4 4-31,3/4 4-32, 3/4.11-5, and 6-16a. These changes support the Uprate project and i 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 j 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 in NUREG 0133 and only affects the

,4 maximum quantity stored.

! e a new reference to the Core Operating Limits Report (COLR) which is based on the B ASH /BART methodology for Large Break Loss of Coolant Accident analysis.

I e revision to the Pressure Temperature Limitations Curves due to effects ofincreased neutron fluence at 2900 MWt.

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.

l This Technical Specification Change Request (TSCR) primarily revises those areas in 4 TS which were not included in Reference 3. The primary supporting analyses performed for uprate are: Large Break Loss of Cooling Accident (LOCA) utilizing the l Westinghouse 1981 Evaluation Model with BASH, 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.

Increased neutron fluence resulting from uprate core conditions has an effect on the reactor vessel Pressure Temperature Curves. Their applicability will change from 14 Effective Full Power Years (EFPY) to 13 EFPY with no other changes at this time.

Docum:nt Control Desk Attachment V TSP 950001 RC-95-0258 Page 2 0f 3 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 of10 CFR

, 50.92 and has determined that the changes do not involve any significant hazard for the following reasons:

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

i Implementation of uprate power operation does not contribute to any accident evaluated in the FSAR. The NSSS Components (RV, RCPs, CRDMs, SGs, and 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 L.icensing Basis.

. Interfacing Systems which are important to safety are not adversely impacted and will continue to perform their design function. Overall secondary plant performance is not significantly altered by the proposed changes.

The revision to the Pressure Temperature Limits will not adversely impact the RCS Pressure Boundary. The length of time these curves will be applicable, due to increased neutron fiuence,is being reduced. Before the 13 Effective Full Power Years have elapsed, new curves will be generated to reflect the analysis of the specimen capsule and will be derived utilizing NRC approved methodology.

Therefore, since the Reactor Coolant pressure boundary integrity and system functions are not adversely im? acted, the probability of occurrence of an accident evaluated in the VCSNS FSAE will be no greater than the original design basis of the plant.

An extensive analysis has been performed to evaluate the consec uences of the following accident types currently evaluated in the VCSNS FSA3:

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

DNBR, Primary and Secondary System Pressure, Peak Clad Temaerature, Metal Water Reaction, Challenge to Long Term Cooling, Environmenta Conditions Inside and Outside containment, etc.) for the accidents are similar to those currently 4

reported in the VCSNS FSAR and remain within applicable Regulatory Acceptance Criteria. Select results (i.e., Containment Pressure during a Steam Line Break, Minimum DNBR for Rod Withdrawal from Subcritical, etc.) are slightly more limiting than those currently reported in the FSAR due to the use of the assumed operating conditions with the A75 SGs and in some cases, use of an uprated core power of 2900 MWt. However, in all cases, the calculated results do not challenge the integrity of the primary / secondary / containment pressure boundary and remain within the regulatory acceptance criteria applied to VCSNS's current licensing basis.

Document Control Desk Attachment V TSP 950001

. RC-95-0258

Page 3 of 3 Given that calculated radiological consequences are not significantly higher than

! current FSAR results and remain well within 10CFR100 limits, it is concluded that the consequences of an accident previously evaluated in the FSAR are not significantly increased.

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

I Uprate power operation will not introduce any new accident initiator mechanisms.

Structural integrity of the RCS is maintained during all plant conditions through 1

compliance with the ASME code and 10 CFR 50 Appendix G requirements. 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 i

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

i 2 1

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

Although uprate power operation will require changes to the VCSNS Technical Specifications, the proposed changes are supported by extensive LOCA, NON-LOCA and SGTR analyses. These analyses show acceptable consequences with margin to the applicable regulatory limits. All equipment required to function during

accident conditions has been shown to remain qualified and thus will perform their i 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).

Low Temperature Overpressure transients which could challenge RCS structural integrity are not impacted by the revision to the Pressure Temperature Limitations Curves. The curves are not directly impacted, the changes do not reduce any margin ofsafety.

Based on the above, it is concluded that there is no significant reduction in a margin of safety.

i J i

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ML20094E411

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