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' ENCLOSURE 1

PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE

SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2

DOCKET NOS. 50-327 AND 50-328

(TVA-SON-TS-95-04)

LIST OF AFFECTED PAGES

Unit 1

3/4 3-113/4 3-13B 3/4 3-1

Unit 2

3/4 3-113/4 3-138 3/4 3-1

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9504110243'950406:PDR' .ADOCK'05000327'P . . PDR_ .s

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! ,y TABLE 4.3-1.

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$ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS .x .'

CilANNEL MODES IN WifICifCilANNEL CilANNEL FUNCTIONAL- SURVEILtANCE _| E FUNCTIO!!AL UNIT CHECK CALIBRATION TESTy. REQUIRED-

w 1. .Hanual Reactor Trip N. A. N.A. S/U(1) and R(9) 1, 2, and *7. . Power Range, Neutron flux S D(2),ff(3) Q 1, 2

and Q(6)3. Power Range, Neutron Flux, N.A. R(6) _Q 1, 2liigh Positive Rate

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4. Power Range, Neutron Flux, N.A. R(6) Q 1, 2liigh Negative Rateto 5. Intermediate Range, S R(6) S/U(1) 1, 2, and *D Neutron FluxT 6. Source Range, Neutron Flux S(7) R(6) M and S/U(1) 2, 3, 4, 5, and *-e" 7. Overtemperature Delta T S R Q 1, 2 ' R1458. Overpower Delta T S R Q 1, 2

9. Pressurizer Pressure--Low - S R _Q 1, 210. Pressurizer Pressure--liigh S R Q 1, 2II. Pressurizer Water Level--liigh S R Q 1, 2 -12. Loss of Flow -' Single. Loop. 5 R Q 113. Loss of Flow - Two Loops S R H.A. 1

C $" 14. Main Steam Generator WaterJS ", Level--Lor Low

k A. Steam Generator Water Level'-- S R Q 1, 2 -RNw.cr, 3 Low-Low (Adverse)-55 g B. Steam Generator Water Level -- S Rg' Low-low (EAM) _Q 1, 2 -

? C. RCS Loop AT S R Q '1, 2{ 0. Containment Pressure (EAM) S R Q 1, 2

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. TABLE 4.3-1 (Continuedt

3.,JNOTATION* ' -

With the' reactor trip system breakers closed and the control rod' drivesystem capable of rod withdrawal. ..

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. (1) : -.If'not' performed in previous 31 days. R145 | Li

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(2)'

LHeatbalanceonly,above15%ofRATEDTHERMALPOWER. |-

(3)'

.if absolute difference greater than 2 percent.- Adjust channel.,

THERMAL POWER.Recalibrate if the absolute differenceCompare incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED'-

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to 3 De reater than*T5 Snvisu_act is No r ggngifts % se pptpooreb MWwueilsee is every Biff /A

a u(4). Delete .. -

THfhnel PoWit ts b is* % 9:r9. ru 96 Houts pflit? 145 a'(5)

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Each train or logic channel shall be tested at least every 62 days-

on a STAGGERED TEST BASIS.The test shall independently verify the .OPERABILITY. of the undervoltage and automatic shunt trip circuits.

(6) '

Neutron detactors may be excluded from CHANNEL CALIBRATION.-

(7) .Below' P-6 (Block of Source Range Reactor Trip)'setpoint. .

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(8)'

Deleted.-

R145-(9)-

.The CHANNEL' FUNCTIONAL TEST shall independently verify the operability'-

of the undervoltage and shunt trip circuits for the manual reactor+

;trip function.4

(10) -Local manual shunt trip prior to placing breaker in service.train shall be tested at least every 62 days on a. STAGGERED TESTEachBASIS.

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(11) - 1Automatic and manual undervoltage trip.

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SEQUOYAH - UNIT 1 3/4 3-13Amendment No. 54, 114 , 141Mby 1 # innn Mav 16. 199n i

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'3/4.3.17and 3/4'3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF).INSTRUMENTATION

, .The OPERABILITY.of'the protective.and ESF. instrumentation systems'and,

4

interlocks ensure that 1) the associated ESF action and/or reactor trip willbe , initiated when the parameter monitored by each channel or combination .

4'

thereof reaches its'setpoint, 2) the specified coincidence ~1ogic is maintained.3) sufficient redundancy is maintained to permit a channel to be out of servicefor testing or maintenance, and 4) sufficient system: functional capability.isavailable-for protective and ESF purposes from diverse parameters.:

The'0PERABILITY of these syst'ms is required to provide the overall.

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reliability, redundancy.and diversity. assumed available in the facility design..for the protection and mitigation of accident and transient conditions.~ The.Integrated operation of each of these systems is. consistent with the assumptionsused in the accident analyses.

The Engineered Safety' Features System interlocks perform the functionsindicated below on increasing the required parameter, consistent with.the J..setpoints listed in Table 3.3-4:

;g .P-11 Defeats the manual block of safety injection actuation on lowpressurizer pressure.w

l. P-14 Trip of all feedwater pumps, turbine trip, closure of feedwater R145|,

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isolation valves and inhibits feedwater control valve modulation.,

On decreasing the required parameter the-opposite function is performed atreset setpoints.R145

-The surveillance requirements specified for these-systems ensure that the'overall system functional capability is maintained comparable to the originaldesign standards. The periodic surveillance tests performed at;the minimumfrequencies are sufficient to demonstrate this capability.

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td/O l d l@}9SEQUOYAH - UNIT 1 B 3/4 3-1 Amendment No. 141

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a; The surv;ill:nca for tha ctmpirison of th2 incora to tha excore Axial Flux Diff:renca is rsquir:d

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only when reactor power is 115 percent. The 96 hour delay in the first performance of the'

, f surveillance after reaching 15 percent reactor thermal power (RTP) , following a refueling outage,.

is- to achieve a higher power level and approach Xenon stability. The surveillance is' typically<

performed when RTP is A 30 percent to ensure the results of the evaluation are more accurate -

and the adjustments more reliable. The frequency of 31 EFPD is to allow slow changes in neutron

flux to be better detected during the fuel cycle.

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TABLE 4.3-1E -E

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS .

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CHANNEL MODES FOR WHICHe CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE ISy FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIREDto 1. Manual Reactor Trip N.A. N A. S/U(1)and R(9) 1, 2, and * -

2. Power Range, Neutron Flux 5 0(2),)I(3) Q 1, 2and Q(6)

3. Power Range, Neutron Flux, N.A. R(6) Q 1, 2~

High Positive Rate4. Power Range, Neutron Flux, N.A. R(6) Q 1, 2High Negative Rate5. Intermediate Range, Neutron Flux S R(6) S/U(1) 1, 2, and ** 6. Source Range, Neutron Flux S(7) R(6) M and S/U(1) 2, 3, 4,T

U 5, and * .7. Overtemperature AT S. R Q 1,28. Overpower AT S R Q 1, 2 R1329. Pressurizer Pressure--Low S R Q 1, 2 :10. Pressurizer Pressure--High S' R Q ' 1, 2 R1611. Pressurizer Water Level--High S R Q 1, 212. Loss of Flow - Single Loop S R Q 113. Loss of Flow - Two Loops S. R N.A. - l'gg 14. Steam Generator Water Level-- . h16o *, g low-Low

>gg A. Steam Generator Water Level-- S R Q 1, 2Hgn ~ Low-Low (Adverse)3$ B. Steam Generator Water Level-- S R Q 1, 2 R132sF- Low-Low (EAM)*

C. RCS Loop AT-. S R Q 1, 2

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.;$;g D. Containment Pressure (EAM) S R Q -1, 2.;-UT5 .

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Table 4.3-1 (Continued).

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~-ANOTATION

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* With the reactor trip system breakers closed and the control rod-drive system capable of rod withdrawal. '.

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(1) -If not performed in previous'31 days.-R132

(2) Heat balance only, above 15% of RATED THERMAL POWER. . Adjust channel'-if absolute difference greater than 2 percent..- '

(3) Compare incore to excore XIAL FLUX DIFFERENCE 'above 15% of; RATED . |gt'04 --

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THERMAL POWER. Recalibr te f he absolute difference reater- aor e usi to 3 per_ cent. Tjff fffQllfNCYof f'//tSS0lV6/44MCf 15 (Vff93|EfPD

//15 3VA16|WMF 15 /Ver AfaVtRCb Td 86 #ARRtw MTit 9611000.5 AFTF2.

(4) De e e . ~ 7#fAMAf. Powrt_ is |t 15 % RTP, R132- ~

(5) Each train or logic channel shall be tested at'least every 62 days-on a STAGGERED TEST BASIS. The test shall independently verifythe OPERABILITY of the undervoltage and automatic shunt trip -circuits. R104

(6) Neutron detectors may be excluded from CHANNEL CALIBRATION.-

(7) Below P-6 (Block of Source Range-Reactor Trip) setpoint.- -

(8) Deleted.- -,

(9) The CHANNEL FUNCTIONAL TEST shall independently verify the|R104 - <

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operability of the undervoltage and shunt trip circuits for themanual reactor trip function.

R46(10) - Local manual shunt trip prior to placing breaker in' service.

Each train shall be tested at-least every 62 days on a'

STAGGERED TEST BASIS.'

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(11) - Automatic and manual undervoltage trip..

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.SEQUOYAH - UNIT 2 3/4 3-13 Amendment No. 46, 104 , 132;

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3/4.3 INSTRUMENTATION-

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BASES

3/4.3.1 and 3/4.3.2 REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEMINSTRUMENTATION ' '

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The OPERABILITY of the Reactor Trip and Engineered Safety FeaturesActuation Systems instrumentation and interlocks ensure that 1) the associated

{action and/or reactor trip will be initiated when the parameter monitored byeach channel or combination thereof reaches its setpoint, 2) the specified ~ 'coincidence logic is maintained, 3) sufficient redundancy is maintained topermit a channel to be out of service for testing or maintenance, and4) suffi.cient system functional capability is available from diverse parameters'.

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The OPERABILITY of these systems is' required to provide the overallreliability, redundancy and diversity assumed available in the facility designfor the protection and mitigation of accident and transient conditions. Theintegrated operation of each of these systems is consistent with the assumptionsused in the accident analyses. The surveillance requirements specified forthese systems ensure that the overall system functional capability is maintainedcomparable to the original design standards. The periodic surveillance testsperformed at the minimum frequencies are sufficient to demonstrate thiscapability. ',

.)1 The Engineered Safety Feature Actuation System interlocks perform the"functions indicated below on increasing the required parameter, consistent. :with the setpoints listed in Table 3.3-4:

P-11 Defeats the manual block of safety injection actuation on low .pressurizer pressure.

jR132:P-14 Trip of all feedwater pumps, turbine trip, closure of feedwater 'isolation valves and inhibits feedwater conttol valve modulation.

On decreasing the required parameter the opposite function is performed at.reset setpoints.

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SEQUOYAH - UNIT 2 B 3/4 3-1 Amendment No.132

00T 311990. _. ..

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$. . .. . . . . .# ' Tha surv:ill:nca for tha comparis:n of tha incore to tho excora Axirl Flux Differancs is requir;;d

,

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only when reactor power is AL15 percent.' The 96 hour delay in the first pe'rformance of the,

surveillance after reaching 15 percent reactor thermal power (RTP) , folloviing a refueling outage,,

is ~ to achieve a higher power level and approach Xenon stability. The surveillance is typically.,

- performed when RTP is 1 30 percent to ensure the results of the evaluation are more accurate ,,

and the adjustments more reliable. The frequency of 31 EFPD is to allow slow changes in neutron ;-

flux to be better detected during the fuel cycle.,

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ENCLOSURE 2

.. PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE -c. ,

SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2

DOCKET NOS. 50-327 AND 50-328:

-(TVA-SON-TS-95-04)

DESCRIPTION AND JUSTIFICATION FOR

' SPECIFICATION REQUIREMENT 3/4.3.1'

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'''Descriotion of Chance

TVA proposes to modify the Sequoyah Nuclear Plant (SON) Units 1 and 2 Technical lSpecifications (TSs) to change the surveillance requirements (SRs) for the power rangeneutron flux channel calibration frequency from monthly to every 31 effective fullpower. days (EFPD). Also, the period for the first performance of the surveillance after ~reaching 15 percent power is being changed. '

The proposed TS change is modeled after the Westinghouse Electric Corporation"

Standard TS (STS), NUREG-1431. A bases change is also made'to more accuratelyreflect the assumptions for the performance frequency.

Reason for Chance:

This change is needed to more accurately monitor changes in the condition of the. ,core. Fuel burn-up is necessary to change the relationship between the incore axialpower and the excore detectors response. Also, at reduced power levels theeffectiveness of the monitoring activity is reduced. The reason for delaying the firstperformance of the SR, until 96 hours after reaching 15 percent rated thermal power - ' ;(RTP), is to allow for the unit to be in a more stable condition. Typically, the unit willbe above 30 percent RTP, preferably near 50 percent RTP. This will ensure that theresults of the evaluation are more accurate and the adjustments more reliable.

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Justification for Chance |

The frequency of every 31 EFPD allows slow changes in neutron flux during the fuel jcycle to be more accurately detected and evaluated. The relationship between the iincore axial power and the excore detectors response is dependent on fuel burn-up. iAt reduced power levels, fuel consumption is reduced and the effectiveness of the I

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monitoring ac'tivity is reduced when performing the r4 3 ring based on time.Therefore, changing the frequency to be based on EW willimprove the effectivenessof the monitoring activity. TVA's proposed change is consistent with the bases forSTS SR 3.3.1.3.

The proposed TS change to delay the period for the first performance of thesurveillance until 96 hours after reaching 15 percent power is based on operatinghistory. The delay has no safety impact on the plant because of the lowering of thepower range trip setpoints before start-up from a refueling outage. The setpoints areset conservatively low to ensure that the protection interlocks will be initiated tomitigate the consequences of an accident or transient. The basis for the 96 hours isas follows:

4 hours - Allowance for power escalation from 15 to 30 percent RTP..

40 hours - Allowance for Xenon to reach equilibrium conditions.

10 hours - Allowance for obtaining, transferring, processing, analyzing,completion and review of the incore flux mapping data.

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LM 4 hours - Allowance for is'suance and review ~of the calibration data calculation.' :[| . f procedure and notification of the instrument maintenance and issuance . .I

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; of the work order and/or procedure for calibration.

- 32 hours - Allowance to perform the calibration on the four nuclear instrument ,p . system channels. '

~6 hours -| Contigency allowance for the above activities.' j;

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Environmental imoact Evalua' iont

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The proposed change does not involve an unreviewed environmental question because- - ioperation of SON Units 1 and 2 in accordance with this change would not: j1. Result in a significant increase in any adverse environmental. impact previously 1 <

evaluated in the Final Environmental Statement (FES) as modified by NRC'stestimony to the Atomic Safety and, Licensing Board, supplements to the FES,environmentalimpact appraisals, or decisions of the Atornic Safety and Licensing .Board.

n2. Result in a significant change in effluents or power levels.

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. . .i3. Result in matters not previously reviewed in the licensing basis for SON that may' '

have a significant environmentalimpact.

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Significant Hazards Evaluation

TVA has evaluated the proposed technical specification (TS) change and hasdetermined that it does not represent a significant hazards consideration based oncriteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SON) inaccordance with the proposed amendment will not:

1. Involve a significant increase in the probability or consequences of an accidentpreviously evaluated.

The likelihood that an accident will occur is neither increased or decreased by thisTS change, which only affects when the first surveillance is performed followingan outage and changes the frequency of performance of the surveillance. Beforestart-up following a refueling outage, the power range high trip setpoint is setbelow 85 percent power, typically 60 percent, for conservatism. The powerrange low trip setpoint is set at 22 percent power, TS requires the setpoint to beless than or equal to 25 percent power. These settings are in addition to theconservatism built into start-up following a refueling outage. Therefore, delayingthe first performance for 96 hours will not impact on the operation of the plantsince the setpoints are set conservatively. Also, the change of the frequency toevery 31 effective full power days (EFPD) only delays the surveillance when theplant is operated at reduced power. During operation at reduced power changesin the neutron flux are also reduced. Therefore, changing the frequency frommonthly to every 31 EFPD allows slow changes in neutron flux during the fuelcycle to be more accurately detected and evaluated.

This TS change will not impact the function or method of operation of plantequipment. Thus, there is not a significant increase in the probability of apreviously analyzed accident due to this change. No systems, equipment, orcomponents are affected by the proposed change. Thus, the consequences of amalfunction of equipment important to safety previously evaluated in the UpdatedFinal Safety Analysis Report are not increased by this change.

The proposed changes provide TS improvements that ensure the system operateswithin the bounds of SON's accident analysis as contained in the Final SafetyAnalysis Report (FSAR) and only affects when a surveillance is performed. Thischange has no impact on accident initiators and does not involve a physicalmodification to to the plant. Accordingly, the proposed changes do not involve )an increase in the probability or consequences of an accident previouslyevaluated.

2. Create the possibility of a new or different kind of accident from any previouslyanalyzed.

This revision will not change any plant equipment, system configurations, or ;accident assumptions. This change will more accurately monitor changes in thecondition of the core.

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Fuel burn-up is necessary to change the relationship between the incore axialpower and the excore detectors response. At reduced power levels theeffectiveness of the monitoring activity is reduced. Therefore, changing thefrequency to 31 EFPD allows slow changes in neutron flux during the fuel cycle tobe more accurately detected and evaluated. Delaying the first performance of thesurveillance requirement, until 96 hours after reaching 15 percent rated thermalpower, will allow the unit to be in a more stable condition. Therefore, this changewill not affect the safety function of any components and will not create thepossibility of a new or different kind of accident.

3. Involve a significant reduction in a margin of safety.

The proposed changes provide TS improvements for SON's power rangemonitoring system that ensure the system operates within the bounds of SON'saccident analysis as contained in the FSAR since only the time interval betweenperformances of the surveillance is being extended. This change does not involvea physical modification to SON's power range monitoring system. Accordingly,the margin of safety has not been reduced.