ML20082E490

Proposed Tech Specs,Changing SRs for Power Range Neutron Flux Channel Calibr Frequency from Monthly to Every 31 EFPD & Delaying First Performance of Surveillance After Reaching 15 Percent Power for 96 H

ML20082E490
Person / Time
Site:	Sequoyah
Issue date:	04/06/1995
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20082E461	List: ML20082E459 ML20082E490
References
NUDOCS 9504110243
Download: ML20082E490 (15)

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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 3-11 3/4 3-13 B 3/4 3-1 Unit 2 3/4 3-11 3/4 3-13 8 3/4 3-1 i'

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TABLE 4.3-1 8

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x REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CilANNEL MODES IN WifICif CilANNEL CilANNEL FUNCTIONAL-l E FUNCTIO!!AL UNIT CHECK SURVEILtANCE _

y CALIBRATION TEST 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.

liigh Positive Rate R(6) _Q 1, 2

4. Power Range, Neutron Flux, N.A.

liigh Negative Rate R(6) Q 1, 2 to 5. Intermediate Range, S D Neutron Flux R(6) S/U(1) 1, 2, and

• T 6. Source Range, Neutron Flux S(7) e

" R(6) M and S/U(1) 2, 3, 4, 5, and *-

7. Overtemperature Delta T S R Q 1, 2 '

R145

8. Overpower Delta T S R Q 1, 2

9. Pressurizer Pressure--Low - S R _Q 1, 2

10. Pressurizer Pressure--liigh S R Q 1, 2 II. Pressurizer Water Level--liigh S R Q 1, 2 -

12. Loss of Flow -' Single. Loop. 5 R Q 1

13. Loss of Flow - Two Loops S R H.A. 1 C $" 14. Main Steam Generator Water JS ", Level--Lor Low

w. k A. Steam Generator Water Level'-- S R Q 1, 2 -RN cr, 3 Low-Low (Adverse)-

55 g B. Steam Generator Water Level -- S R 1, 2 -

g' Low-low (EAM)

_Q

? C. RCS Loop AT S R Q '1, 2

0. Containment Pressure (EAM) S

{ R Q 1, 2

._-___m_____..._______..__.m_ _ _ _ _ _ _ _ - _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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. TABLE 4.3-1 (Continuedt 3.,JNOTATION With systemthe' capable reactor trip system breakers closed and the control of rod withdrawal. .

. (1) : - .

.If'not' performed in previous 31 days. R145 l Li (2) -

l LHeatbalanceonly,above15%ofRATEDTHERMALPOWER.

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

(3) - '

Recalibrate THERMAL POWER.

to 3 De if the absolute differenceCompare reater than inc,

• T a u (4).. - 5 Snvisu_act Delete . is No r ggngifts % se pptpooreb MWwueilsee is every Biff /A THfhnel PoWit ts b is* % 9:r9. ru 96 Houts pflit?

' 145 a (5) -

Each on train orTEST a STAGGERED logic BASIS. channel shall be tested at least every 62 d The test shall independently verify the .

OPERABILITY. of the undervoltage and automatic shunt trip circuit .

(6)

Neutron detactors may be excluded from CHANNEL CALIBRATION (7) - .

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

(8) -

Deleted.

R145

-(9) -

+ .The CHANNEL' FUNCTIONAL TEST shall independently ve of the trip undervoltage and shunt trip circuits for the manual reactor 4 function.  ;

(10) -

Local manual shunt trip prior to placing breaker in service. Each BASIS. shall be tested at least every 62 days on a. STAGGERED TEST train .

(11) - 1 Automatic and manual undervoltage trip.

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

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F a, c 3N. 32 INSTRUMENTATIO'N I

}* BASES i'

'3/4.3.17and INSTRUMENTATION3/4'3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF 4

.The OPERABILITY.of'the protective.and ESF. instrumentation systems'and interlocks ensure that 1) the associated ESF action and/or reactor trip will 4 be , initiated when the parameter monitored by each channel or combination . '

thereof reaches its'setpoint, 2) the specified coincidence ~1ogic is maintained.

3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system: functional capability.is available-for protective and ESF purposes from diverse parameters.

The'0PERABILITY of these syst'ms e is required to provide the overall 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 assumptions used in the accident analyses.

The Engineered Safety' Features System interlocks perform the functions J indicated below on increasing the required parameter, consistent with.the

.setpoints listed in Table 3.3-4:

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

l. P-14 R145l Trip of all feedwater pumps, turbine trip, closure of feedwater isolation valves and inhibits feedwater control valve modulation. ,

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

R145

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

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i SEQUOYAH - UNIT 1 td/O l d l@}9 11 B 3/4 3-1 Amendment No. 141

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The surv;ill

nca for tha ctmpirison of th2 incora to tha excore Axial Flux Diff:renca is rsquir:d only when reactor power is 115 percent. The 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 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.

m TABLE 4.3-1 E

E REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS .

SE -

3 -

i CHANNEL MODES FOR WHICH e CHANNEL CHANNEL FUNCTIONAL y FUNCTIONAL UNIT CHECK CALIBRATION TEST SURVEILLANCE IS REQUIRED to 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, 2 and Q(6)

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

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High Positive Rate

4. Power Range, Neutron Flux, N.A. R(6) Q 1, 2 High Negative Rate

5. Intermediate Range, Neutron Flux S R(6) S/U(1) 1, 2, and *

6. Source Range, Neutron Flux S(7) R(6)

T M and S/U(1) 2, 3, 4, U 5, and *

7. Overtemperature AT S. R Q 1,2

8. Overpower AT S R R132 Q 1, 2

9. Pressurizer Pressure--Low S R Q 1, 2 :

10. Pressurizer Pressure--High S' R Q ' 1, 2 R16

11. Pressurizer Water Level--High S R Q 1, 2

12. Loss of Flow - Single Loop S R Q 1

13. Loss of Flow - Two Loops S. R N.A. -

l' gg 14. Steam Generator Water Level-- . h16 o *, g low-Low

>gg A. Steam Generator Water Level-- S R Hgn ~ Low-Low (Adverse)

Q 1, 2 3$ B. Steam Generator Water Level-- S R sF-Low-Low (EAM)

Q 1, 2 R132 4 C. RCS Loop AT- S R

.;$;g D. Q 1, 2

.;- Containment Pressure (EAM) S R Q -1, 2 U

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h. w - w s + - + - - - , . - - -r- -________.____.r -

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

(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; RATEDE . lgt'04 -

THERMAL POWER. Recalibr te f he absolute difference reater- a or e usi to 3 per_ cent. Tjff fffQllfNCYof f'//tSS0lV6/44MCf 15 (Vff93lEfPD (4) -

De //15 e e3VA16lWMF

. 15 Powrt_

~ 7#fAMAf. /Ver AfaVtRCb is lt 15 % Td

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RTP,86 #ARRtw MTit 9611000.5 AF R132 (5) -

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. 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 operability of the undervoltage and shunt trip circuits for the lR104 manual 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.

gio4 (11) - Automatic and manual undervoltage trip.

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.SEQUOYAH - UNIT 2 3/4 3-13 Amendment No. 46, 104 , 132 007

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3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION '

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

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action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, 2) the specified ~ '

coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and

4) suffi.cient system functional capability is available from diverse parameters'.

The OPERABILITY of these systems is' required to provide the overall 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 assumptions used in the accident analyses. The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. ,

.)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 ,

only when reactor power is AL15 percent.' The 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 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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, w ENCLOSURE 2

.. PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE -

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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' I

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Descriotion of Chance TVA proposes to modify the Sequoyah Nuclear Plant (SON) Units 1 and 2 Technical Specifications (TSs) to change the surveillance requirements (SRs) for the power range l

neutron flux channel calibration frequency from monthly to every 31 effective full power. 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 accurately reflect 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 axial power and the excore detectors response. Also, at reduced power levels the effectiveness of the monitoring activity is reduced. The reason for delaying the first performance of the SR, until 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after reaching 15 percent rated thermal power - ';

(RTP), is to allow for the unit to be in a more stable condition. Typically, the unit will be above 30 percent RTP, preferably near 50 percent RTP. This will ensure that the results of the evaluation are more accurate and the adjustments more reliable. ,

i Justification for Chance l The frequency of every 31 EFPD allows slow changes in neutron flux during the fuel j cycle to be more accurately detected and evaluated. The relationship between the i incore axial power and the excore detectors response is dependent on fuel burn-up. i At reduced power levels, fuel consumption is reduced and the effectiveness of the I 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 effectiveness of the monitoring activity. TVA's proposed change is consistent with the bases for STS SR 3.3.1.3.

The proposed TS change to delay the period for the first performance of the surveillance until 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after reaching 15 percent power is based on operating history. The delay has no safety impact on the plant because of the lowering of the power range trip setpoints before start-up from a refueling outage. The setpoints are set conservatively low to ensure that the protection interlocks will be initiated to mitigate the consequences of an accident or transient. The basis for the 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> is as follows:

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Allowance for power escalation from 15 to 30 percent RTP.

40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> - Allowance for Xenon to reach equilibrium conditions.

10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> - Allowance for obtaining, transferring, processing, analyzing, completion and review of the incore flux mapping data.

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LM 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Allowance for is'suance and review ~of the calibration data calculation.'  :

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f procedure and notification of the instrument maintenance and issuance . .I

of the work order and/or procedure for calibration.

- 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> - Allowance to perform the calibration on the four nuclear instrument ,

p . system channels.

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

j Environmental imoact Evalua't ion The proposed change does not involve an unreviewed environmental question because- -i operation of SON Units 1 and 2 in accordance with this change would not: j

1. Result in a significant increase in any adverse environmental. impact previously 1 <

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

Board.

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2. Result in a significant change in effluents or power levels. ,

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

have a significant environmentalimpact.

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ENCLOSURE 3 .

PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 'AND 2 DOCKET NOS. 50-327 AND 50-328 - ,

(TVA-SON-TS-95-04)

-. DETERMINATION OF.NO SIGNIFICANT HAZARDS CONSIDERATION ',

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4 Significant Hazards Evaluation TVA has evaluated the proposed technical specification (TS) change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SON) in accordance with the proposed amendment will not:

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

The likelihood that an accident will occur is neither increased or decreased by this TS change, which only affects when the first surveillance is performed following an outage and changes the frequency of performance of the surveillance. Before start-up following a refueling outage, the power range high trip setpoint is set below 85 percent power, typically 60 percent, for conservatism. The power range low trip setpoint is set at 22 percent power, TS requires the setpoint to be less than or equal to 25 percent power. These settings are in addition to the conservatism built into start-up following a refueling outage. Therefore, delaying the first performance for 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will not impact on the operation of the plant since the setpoints are set conservatively. Also, the change of the frequency to every 31 effective full power days (EFPD) only delays the surveillance when the plant is operated at reduced power. During operation at reduced power changes in the neutron flux are also reduced. Therefore, changing the frequency from monthly to every 31 EFPD allows slow changes in neutron flux during the fuel cycle to be more accurately detected and evaluated.

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

The proposed changes provide TS improvements that ensure the system operates within the bounds of SON's accident analysis as contained in the Final Safety Analysis Report (FSAR) and only affects when a surveillance is performed. This change has no impact on accident initiators and does not involve a physical modification to to the plant. Accordingly, the proposed changes do not involve )

an increase in the probability or consequences of an accident previously evaluated.

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

This revision will not change any plant equipment, system configurations, or  ;

accident assumptions. This change will more accurately monitor changes in the condition of the core.

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0 Fuel burn-up is necessary to change the relationship between the incore axial power and the excore detectors response. At reduced power levels the effectiveness of the monitoring activity is reduced. Therefore, changing the frequency to 31 EFPD allows slow changes in neutron flux during the fuel cycle to be more accurately detected and evaluated. Delaying the first performance of the surveillance requirement, until 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after reaching 15 percent rated thermal power, will allow the unit to be in a more stable condition. Therefore, this change will not affect the safety function of any components and will not create the possibility 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 range monitoring system that ensure the system operates within the bounds of SON's accident analysis as contained in the FSAR since only the time interval between performances of the surveillance is being extended. This change does not involve a physical modification to SON's power range monitoring system. Accordingly, the margin of safety has not been reduced.