ML20196A069

Proposed Tech Specs Pages 2-6,B 2-4,3/4 3-2 & 3/4 3-5, Modifying Trip Setpoint & Allowable Value Units for Intermediate Range Nuclear Flux Detector & Revising Applicability Requirements for Source Range Flux Detector

ML20196A069
Person / Time
Site:	Sequoyah
Issue date:	12/02/1988
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20196A058	List: ML20196A053 ML20196A069
References
NUDOCS 8812050247
Download: ML20196A069 (80)

Text

s ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-88-42)

LIST OF AFFECTED PAGES Unit 1 2-6 B 2-4 3/4 3-2 3/4 3-5 Unit 2 2-6 B 2-4 3/4 3-2 3/4 3-5 e

e PDC

I TABLE 2.2-1 (Continued) mEg REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

t 1

1 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES c

13. Steam Generator Water 1 18% of narrow range instrument 3 17% of narrow range instrument R20 Level--Low-Low span each steam generator span-each steam generator y

14. Steam /Feedwater Flow

< 40% of full steam flow at

< 42.5% of full steam flow at Missatch and Low Steam RATED THERMAL POWER coincident RATED THERMAL POWER coincident Generator Water Level with steam generator water level with steam generator water level 1 25% of narrow range instru-1 24.0% of narrow range instru-ment span--cach steam generator ment span-each steam generator

15. Undervoltage-Reactor 2 5022 volts-each bus 14739 volts each bus R89 Coolant Pumps y

16. Underfrequency-Reactor 2 56.0 Hz each bus 1 55.9 Hz each bus Coolant Pumps o

17. Turbine Trip A.

Low Trip System 1 45 psig 1 43 psig Pressure B.

Turbine Stop Valve 1 1% open 1 1% open

. Closure

18. Safety Injectian Input Not Applicable Not Applicable frem ESF

-F 4d MTED Ttfstal4L p 3 4 a go-'$.-f MTEi> Ttf@est4L y

~

19. Int?rmediate Range Neutron

-10 NM

-11 FOWM d

Flex - (P-6) Enable Block

-' 1 ' 10 v

- " ; 10 c pr

%{

Source Range Reactor TripM a a h3

20. Power Range Neutron Flux

< 10% of RATED

< 11% of RATED

"[

(not P-10) ' Input to Low Pcwer THERMAL POWER THERMAL POWER

.% ?

Reactor Trips Block P-7 j

~*

p.4, 4 ~ ~ & % * " ' f dl

g

u SAFETY LIMXTS BASES WY Range Channels will initiate a reactor trip

= ~e"+

1-

' p cp rt h,el tv approximately 25 percent of RATED THERMAL PO unless manually blocked when P-10 becomes active.

No credit was taken for o n of the trips associat with either the Intermediate or Suurce Range Channels e

n s;

however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overtemperature Delta T The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the High ana Low Pressure reactor trips.

This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature and dynamic

• compensation for piping delays from the core to the loop temperature detec-tors. With normal axial pwer distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced l

according to the notations in Table 2.2-1.

,l Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature Delta T setpoint.

Three loop operation above the 4 loop P-8 setpoint is permissible after resetting the K1, K2 and K3 inputs to the Overtemperature Delta T channels and raising the P-8 setpoint to its 3 loop value.

In this mode of operation, the P-8 inter-lock and trip functions as a High Neutron Flux trip at the reduced power level.

Overpower Delta T The Overpower Delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all po'ssible overpower conditions, lhnits the required range for Overtemperature Delta T protection, and provides a backup to the liigh Neutron Flux trip.

The setpoint includes corrections for changes in density and heat capacity of water with temperature,.and. dynamic compensation for piping delays from the core to the loop temperature detectors.

No credit was taken for operation of this trip in the accident 0

SEQUOYAH - UNIT 1,

B 2-4 1

TABLE 3.3-1 y

. t l

REACTOR TRIP SYSTEM INSTRUMENTATION t

O 1

E MINIMUM TOTAL NO.

CilANNELS CilANNELS APPLICABLE E

SUNCTIONAL UNIT, OF CilAf81ELS 10 TRIP OPERABLE MODES ACTION U

r 1.

Mancal Reac' tor Trip 2

1 2

1, 2, and

• 1 2.

Powe'r Range *, Neutron Flux 4

2 3

1, 2 2#

4 3.

Power Range, Neutron Flux 4

2 3

1, 2 2'

liigh Positise Rate 4

Power Range, Neutron Flux, 4

2 3

1, 2 2#

liigh Negative Rate

{

5.

Intermediate Range, Neutron Flux 2

1 2

1, 2, and

• 3 Y

6.

Source Range, Heatron Flux b A.

Startup_

2 1

2 and

• 4 8.

Jhutdowa 2

0 1

3, 4 and 5 5

}'.

Overtemperature Delta T

.cour Loop Operation 4

2 3

1, 2 6,

S.

Overpower Delta T

~

lR45 Four Loop Operation 4

2 3

1, 2 6#

l R45 9.

Presstrizer Pressure-Low 4

2 3

1, 2 6#

10 Pressu:-izer Pressure--liigh 4

2 3

1, 2 6#

N E"

{n 11.

Pressur'izer Wcter Level--liigh 3

2 2

1, 2 7 #

oa Y.

p yy -j h v d ft 5I a.,

,; J + sec10 & d itlit. 2 9 2 O W Y$Y

?

4 og s

1 w wa

~he-

e 4

TABLE 3.3-1 (Continued)

TABLE NOTATION With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal, and fu11 in the reactor vessel.

==

The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

b pro ions pecif c on 3:

re no plic O ;' "c't:;: te deter +^- = y ha da-aaa T red 2: : th: " O (C'e:L ef 6 c-+

P=;: $r :t:r Trip) ::tp:! S.

CTI0t nTEMEN ACTION 1 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the reactor trip breakers.

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels, STAR 70P and POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

lR$1 b.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> lR51 for surveillance testing per Specification 4.3.1.1.1.

c.

Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL and the Fower Range, Neutron Flux high trip reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, d.

The QUADRANT POWER TILT RATIO, as indicated by the remaining three detectors is verified consistent with the normalized symmetric power distribution obtained by using the movable incare detectors in the four pairs of symmetric thimble locations at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when THERMAL POWER is greater than 75% of RATED THERMAL POWER.

. -s September 17, 1986 SEQUOYAH - UNIT 1 3/4 3 5 Amendment No. A

TABLE 2.2-1 (Continued) u, m

l!

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SEIPOINTS ii s

l I

FUNCTIONAL 04 TIT TRIP SETPOINT ALLOWABLE VALUES si

13. Steam Generatar Water 2 18% of narrow range instrument 3 17% of narrow range instrument R7 Level--!.ow-Low span-each steam generator span-each steam generator r.

14. Steae/Feedwater Flow

< 40% of full steam flow at

< 42.5% of full steam flow at Mismatcc aad Lee Steam RATED TilERMAL POWER coincident RATED THERMAL POWER coincident Generator Water Level with steam generator water level with steam generator water level

> 2S% of narrow range instru-1 24% of narrow range instru-4rnt span-cach steam generator ment span--each steam generator R76

15. Uncervoltage-Reactor 1 5022 volts each bus 1 4739 volts-each bus Coolant Pumps

16. Uncerfrequency-Reactor 1 56 Itz - each bus 1 55.9 Hz - each bus

?

Coolant Purps e

17. Turbine Trip A.

Low Trip System 1 45 psig 1 43 psig Pressure B.. Turbine Stop Valve 1 1% open

> 1% open

. Closure

18. Sa#3ty Injection Input Not Applicable Not Applicable

~

>\\pJOS$of RATED YttOUltL g gy70' of RAW W

19. Intermediate Range Neutron

-- I - 10 '"

1.,a-Mv c-Flux, P-6 Enable Block

~

u, 3g.

Source Range Reactor Trip

?y

< 10% of RATED

< ILE of RATED THERMAL POWER 3

't.

20. Power Range Neutron Flux (not P-10) Input to Low TilERMAL POWER Power Reactor Trips I'

r' u

!*j:

Block P-7 I'

2.~

~

O e

LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Nuclear Flux (Continued) d Range Channels will initiate a reactor trip ct c arrc.t hd prgutbrf-4+--

l approximately 25 percent of RATED THERMAL POWER unless manually blocked whsn P-10 becomes active.

No credit was taken for operation of the trips associ-ated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of tht Reactor Protection System.

t

'Overtemperature oT The 0vertemperature delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the High and Low Pressure reactor trips.

This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors.

With normal axial power distribution, this reactor trip limit-is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indir.ated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the i

notations in Table 2.2-1.

i I

Operation with a reactor coolant loop out of service below the 4 loop P-8 i

setpoint does not require reactor protection system set point modification because the P-8 setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature delta T setpoint.

Three loop operation above the 4 loop P-8 setpoint is permissible after resetting the K1, K2, and K3 inputs to the Overtemperature delta T channels and raising the P-8 setpoint to its 3 loop value.

In this mode of operation, the P-8 interlock and trip functions as a High Neutron Flux trip at the reduced power level.

Overpower AT The Overpower delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature delta T protection, and provides a backup to the High heutton Flux trip. 'The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors.

No credit was taker) for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

h N

SEQUOYAH - UNIT 2 8 2-4

}

s TABLE 3.2-1 4

v.

g REACTOR TRIP SYSTEM INSTRUMENTATION j

• 5 I

MINIMUM TOTAL NO.

CHK4NELS CHANNELS APPLICA8LE FUNCTIONAL ilNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1.

Manual React 5r Tr'ip 2

1 2

1, 2, and

• 1 2.

Power Range, ileutron Flux 4

2 3

1, 2 2#

3.

Power Range,i;eutron Flux 4

2 3

1, 2 2#

High Positive Rate i

4.

PowerRange,t[eutronFlux, 4

2 3

1, 2 2

I' HighNejativelate

{

5.

Intermediate Rpnge, Neutron Flux 2

1 2

1, 2, and

• 3

,Y 6.

Source Range, keutron Flux A.

Startty 2

1 2

2, and

• 4 8.

Sh,'utdowa.,

2 0

1 3, 4 and S 5

7.

Os.'riesperature AT F* ur Loop Operatfca 4

2 3

1, 2 6,

o 8.

Overpoweb AT R33 Fosr Loop Operation 4

2 3

1, 2 6#

,,".y lnn "s

9.

Pressurf ter Pressure-Low 4

2 3

1, 2 6

y a"

a3

!! j 10.

Pressurizer Pressure--High 4

2 3

1, 2 6-Y u -.&

11.

Pressurfrer Watzr level--High 3

2 2

1, 2 7

y O'

l Cu l

• 9 l

\\

}

b

4 TABLE 3.3-1 (Continued)

TABLE NOTATION s

With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel.

mA The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

[TheprovisionsofSpecification3.0.4arenotapplicable.

""Migh M t ;e : d;;;;th ::,7 b: d: :::r;t::d :b: : 15: "-S (Sh d eT me-

":n;: *:::t:r Trip) ::tp: ht.

ACTION STATEMENTS l

ACTIuN 1 - With the number of OPERABLE channels one less than required by 4

the Minimum Channels OPERABLE requirement, restore the inoperable i

channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the reactor trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total I

Number of Channels, STARTUP and/or POWER OPERATION may proceed l

provided the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition I

within 6 he';rs.

lR3,

b.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> lR3 i for surveillance testing per Specification 4.3.1.1.1.

Either, THERMAL POWER is restricted to less than or equal c.

to 75% of RATED THERMAL POWER and the Power Range, Neutron Flux trip setooint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 nours; or, the QUADRANT POWER TILT RATIO is monitored at le'..,t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, d.

The QUADRANT POWER TILT RATIO, as indicated by the i maining three detectors, is verified consistent with the normalized syr.?etric power distribution obtained by using the movable incere detectors in the four pairs of symmetric thimble locations at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when THERHAL POWER is creater than 75%.of RATED THERMAL POWER.

)

September 17, 1986 SEQUOYAH UNIT 2 3/4 3-5 Amendment No. 39

4 ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-;28 (TVA-SQN-TS-88-42)

DESCRIPTIOh AND JUSTIFICATION FOR MODIFICATION OF THE TRIP SETPOINT AND ALLOWABLE VALUE UNITS FOR THE INTERMEDIATE RANGE NCCLZAR FLUX DETECTOR AND CHANGES TO THE.UPLICABILITY REQUIREMENTS FOR THE SOURCE RANCE NUCLEAR FLUX DETECTOR f

l J

l l

I f

EiCLOSIAE 2 Description of Change Tennessee Valley Authority proposes to modify t'ae Sequoyah.Vuclear Plant (SQN) units 1 and 2 technical specifications to revise the trip setpoint and allowable value units for the intermediate-range (IR) nuclear flux detector and to revise the applicability requirements for the source range (SR) nuclear flux detector.

Reason for Change TVA is replacing the SR and IR neutron monitors as part of the equipment upgrade to comply with Regulatory Guide 1.97 as required by SQN license conditions 2.C.24 (unit 1) and 2.C.14 (unit 2).

The new SR/IR monitor is a fission chamber design manufactured by Gamma Metrics.

This design does not require high-voltage decaergization as part of the normal SR detector operatioa. Consequently, the applicability table 3.3-1 is being revised to delete an unnecessary note involvinr high-voltage deenergization.

The new IR monitor uses a signal that is in units of relative power.

Consequently, the trip setpoint and allowable value are being changed in table 2.2-1.

The bases to section 2.2 are also being revised to delete references to IR detector current signals that are proportional to power levels.

The changes to unit 1 also have appropriate footnotes added to indicate that the changes beccme effective for unit 1 after installation of the new detectors during the unit I cycle 4 refueling outage.

The unit 2 detectors will be installed during the unit 2 cycle 3 refueling outage, and the change will be effective at the time of startup following the outage.

Justification for Change The new Camma Metrics SR/IR detectors are being installed to achieve compliance with Regulatory Guide 1.97.

The new detectors are class-1E equipment that is seismically and environmentally qualified.

The new SR detector design is compatible with the current systemt however, it includes two improvements over the present detector design.

First, the electronic equipment automatically adjusts the high flux at shutdown alarm. Currently, this function is performed manually as described in the Final Safety Analysis Report, cection 15.2.4.2.

Second, the new SR detector does not have to be deenergized at higher power levels.

Above the P-6 setpoint, the SP, detector output signal is blocked from the reactor trip logic.

However, the SR/IR detector assemblies remain energized during the full range of power operation. As a result of this feature, the table notation in table 3.3-1 regarcing high-voltage deenergiration can be deleted iecause it is not applicable to the new design.

The new IR detector design is compatible with the current system except that the output signal is in units of relative power rather than amperes (amps). The P-6 setpoint and allowable value listed in table 2.2-1 are currently listed in units of amps.

TVA has performed a calculation to determine the relative power values corresponding to the present trip setpoint and allowable value. A relationship between reactor power 3

' and detector current was establisaed using startup test data from several power levels between 5-and 90-percent power.

This relationship was then used to convert the trip setpoint to a relative power value.

The computed

[

value was rounded to the next coeservative decade for ease of calibration. A correspond 8.ng allowable value was then calculated using the previously established setpoint and current-power relationship.

Finally the overlap between the SR/IR dstector ranges was checked to ensure sufficient margin betwsen the P-6 setpoint and the SR trip setpoint.

It is important to note that the actual setpoint is not changed; only the enginetring units have chenged.

A copy of the TVA calculation is incivded as an attachment to this enclosure.

Footnotes have been added to tables 2.2-1 and 3.3-1 of the unit 1 technical specifications that indicate that the proposed changes do not become effective until installation of the SR/IR detector assemblies during the unit 1 cycle 4 refueling outage scheduled for mid-1990.

This method of handling the unit 1 change will allow a single review cf the issue and avoid separate technical specification change reques:s for osch unit.

In stmaary, three administrative changes are proposed to support the installation of the Gamma Metries SR/IR assembly.

The first involves the deletion of a table note that is not applicable to the design of the new SR detectors. The second involves a change in engineering units for the P-6 setpoint that results from the difference in output signals from the IR detectors. The third involves the addition of certain footnotes to enoble the review and approval of the unit 1 changes to proceed independently of the unit 1 installation schedule.

Environmental Impact Evaluation The proposed revision involves an administrative change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements.

TVA has determined that the proposed change involves no significant increase in the amounts, and no sig'ificant change in the types, of any effluents that may be released rffsite and that there is no significant increase in individual or cumulatise occupational radiation exposure. Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b). no environmental impact statement nor environ ental assessment needs to be prepared in connection with tra issuance of the amendment.

l

ATTACHMENT 1 TVA CALCUI.ATION, "!NTEU1EDIATE RANGE NEUTRON FLUX P-6 SETPOINT," REVISION 1 (B25 881117 808) l

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Just; fica [lon(expl.,ln below):

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t Noth' d 2:

In the alternate calculation cethod, identify the pages where the alternata calculation has been included in the calculation pa:kage and esp 1., ln why this nethod is adequate.

Method 3:' In the qualification test method, identify the QA docu..ented source (s where testing adequately demonstrates the adequacy of this calculat on and explain.

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Method of design verification (independent review) used (check method used):

1 1.

Design Review 2,

Alternate Calculation 3.

Qualification Test i

Justific6 tion (esplain below):

Method 1: In the design review method, justify the technical adequacy of the calculation and explain how the adequacy was verified (calculation is similar to another, based on accepted handbook methods, appropriate

]

sensitivity studies included for confidence, etc.).

]

Method 2: In the alternate calculation method, identify the pages where the alternate calculation has been included in the calculation package and explain why this method is adequate.

Method 3: In the qualification test method, identify the QA documented source (s) where testing adequately demonstrates the adequacy of this calculation and explain.

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Calculation No.

Revision Method of design verification (independent review) used (chect.iethoc used):

1.

Design Review 2.

Alternate Calculation 3.

Qualification Test Justification (esplain below):

Method 1:

In the design review reethod. justify the technical adequacy of the calculation and explain how the adequacy was verified (calculation is strellar to another, based on accepted handbook, methods, appropriate sensitivity studies included for confidence, etc.).

Method 2:

In the alternate calculation method. identify the pages where the alternate calculation has been laciuded in the calculation paciste and explain why this method is adequate.

Method 3:

In the qualification test teethod. Identify the QA doew entert source (s) where testing adequately demonstrates the scequacy of this calculation and esplain.

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1.

Oc:!&S Review V

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Alternate Calculation 3.

Qualificatiss Test Justification (espi,$1n below):

Method 1:

In the des'gn review method, justify the technical adequacy of the calculation and explain how the adequacy was verified (calculation is similar to another, based on accepted handboot methods, appropriate sensitivity studies included for confidence, etc.).

Method 2:

In the alternate calculation method, identify the pages where the alternate calculation has been included in the calculation pactate and esplain why this method is adequate.

Method 3:

In the qualification test method, identify the QA documented source (s) where testing adequately demonstrates the adequacy of this calculation and explain.

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C TENNESSEE VAlt.EY AUTHORITY SIGUOYAH NUCLEAR PLANT UNIT NUMBERS 1 AND 2

(

PAECAUT!0NS, LIMITATIONS AND SETPOINTS

~

FOR NVCLEAR STEAM SUPPLY SYSTEMS

(

REVISION 9 PAY,1981 1

.... :.t (7 as te.hed p.qrs ) ' ',:. i :

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WESTINGHOUSE El.ECTAIC CORPORATION

\\** MJ7 .T /

Nuclear Energy Sys m s t

' " 'o' '.'.'*' ' 'a ' e',' "f-' * * *

..,.., -..,.. e,..

P* 0* Box 355 1

3'i 2;.'.'ff,i.'.....'.'l,'...,'."s..

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SP.EET REV C

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33% of rated stasn generator

! 3.

turb'ne trip stear generator Hf level signsi for C

fcedwiter valve closure, turbine trip and feedwater pump trip (LS 517A, LS-527A LB-537A, LB 547A, 75 of level s; n LB-218A, L! 523A, LB-53SA, LB-548A, C

(5 519A L3 520A, LB 539A, LB 549A) to f DT h!, Pemissive_ ard intpelock circuits 4b )X lC I i

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B.

7 (aut: atically blocks various "at P

pow 6r tripsat1cwpower) 1.

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2.

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C.fPB(411c <s one loop loss of flow belew jsetpoint)

'(NC41N.

'iC 42N, NC-43N, NC 44N) 35: of full ecoer D.

P-9 lblec ;s reacter trip en turbir.e

, trip bele < set;oint nuclear power level)

• (t:C-415 9C425,NC435,NC-445)

SCt of full p:Wer

E. 'P 10 (411 :ws r.anual bicek of pewer rative

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'(lewsetpoint) trip, intermediate ~ range

\\.

trip, and C 11 blecks source range trip,.

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end preyt les a portion ' f 'P-7 signal) o I(NC-41H,4C-42M, NC 43H, WC, HM)-

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F.

I 10% of full ;cwer

'j P-11 (allows eenval block 6f f afety in. '"

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(TB L120. T3 4220, 73 4320 T8 4423) 540'F 4

Autenatic reset of manual block on highpressurtzerpretsure(P11)

(P3 1558, PR 4563 P5 4578) 1970 Mig 5.

Contoineent high pressure (PB 1348 FB 9353. P3 9365) 5 1.!4 pst;

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(

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Steam Lite Isolatten 1

Highsteamline. flow (SetI.l.A.3atye) 2.

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Nuclear (n.strumentation 1.

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'Where 4;pi ttsble.

equirtrent sh:uld be met with all lead, lag, a,d filter tiec constanti set to 0FF.

,\\.14 Centre 11er Tesisfer Fun:tions yet Appliesble 1.15 Le t:sints

• 7# I.

Ia 5E 4 toeds 1%s hai lble Y

R nce of Setting 4

Intermediate Range High Neutren 5 to 30". full power riux'Reacter I ip Bource Range H gh'Neuteen Flux

-10*b 's -10 t

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1 (11 settings w th the exception of time c:nstants shall

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idjustablewithin their range an'd all time c:nstants shall te 4:ntinu usly a:justable or adjustable in in:rements su:n tha any

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Let;oint can be obtained within ?,10% of the set:oint value.

i Jer the P 10 si teeint see Nuclear Pcwer Raage Protection (Oe:

est 2f.

1.16 locub e ents f: e Test and Calibratien 1

(

11 protecticr.

hannels.should be su; plied with suff uient redundancy i

c provi,de tF3 ceMM14 ty,foi' tch,.. ' r.slibrnti

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NEP-6.6

• • • • . v8 Page 1 of @

M-SAFE 1Y EVALUATION FORM 39 s aatn e 1.

To

_ Sheet i 3.

.USQ?

4 Safety Evaluation Number Sequoych Nuc!Oor PIcnt Cab /, TN a Yes Eca tg/94 2.

Fecm F.16 SNP.

VN

9. RIMS Accession Number R

SQP B5 of f 7 503 Rev Tot 0

Date R

No.

DD

5. P' spared 6. Reviewed 7. Approved

8. Apod JB25 881114 5 3I O

3 A. A. GNoJ 94 %,Lat(vic t,;.,/VAB //Wl8S' O25 081117 h 7?;

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10. Project pnd Affected enit(s)

11. PMP or DCN Number PMP or DCN Revision

.re.g u a ye l U.,, /s I i *t.

e c.A )

t tt94

12. FCR, SCR, MCR, C DCN, or CAQR Number Date of Document (s)

D C A'

//S 3////W

13. Other Document Identifier A) ** <-

Date of Docurent A)/A

'14 Special Requirements?

See t) Yes (No SheetNo.MM

15. Potential Tech Spec See Ch gr,e ( Yes a No

16. References (include system number and name as appropriate)

Sheeu No

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s&F c S&fyEvalu/,5 a>. Eco us e s a u t. c AonJ7.toA>Al. JAJroRHA TJoD I Y e~ tion (s) of System (s) Affected F - s* ~ aty functions of the systems affected by this ECN are described e POST ACCIDENT MONITORING (PAM) The safety function of the post accident monitoring system is to provide information on plant variables required by control roora operating personnel during accident situations tot i 1. permit the operator to take preplanned manual actions to accomplish safe plant shutdown. 2. determine whether safety systems or systems importsnt to safety are performing their intended functions. 3. . determine 't'he potential for causing gross breach of the barriers to radioactivity release and to deternine if a gross breach of a barrier has occurred. 4 assess the'cperation of plant systems to make appropriate decisions as to their use. allo'v for early indication of release of radioactive materials 5. in order to initiate action necessary to protect the public and estimate the magnitude of any impending threat. Tle.,cA h ya,, b lfs, r Q & j f h.s M i.s n e uVo"n y mons}*<-oh. A)<uh~ ~ /semo/eri fm, A-s i&~.o /a 4 [ / < $ " "Sev<-. 4,,. p u a. ifade/rehalsa/kly M/uan l .s e-s f 4 r..4 /, n a vo, /alk ik Ac. 4. 4.n u_.s

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4 s v.h/ hk u :q <. n e.u f r.*, ~-, o toe o3 sa Ny Intermediate range high neutron flur trip - The intemediate range high neutron flur trip circuit shall trip the reactor when one out of the two intermediate range channels exceed the trip setpoint. This trip, which provides protection during reactor startup, can be manually blocked if two out of four power Three out of the four power range channels below this valueran automatically reinstates the intemediate range high neutron flur trip. be separate from the power range channels.The intemediate range channe shall The intemediate range channels can be individually bypassed at the nuclear 'nstru entation racks to permit channel testing at any time under prescribed administrative procedures and only under the direction of authorized supervision. This bypass action shall be annunciated on the control board. Source range high neutron flur trip - The source range high neutron fluz trip circuit shall trip the reactor when one of the two source range channels ezeeeds the trip setpoint. This trip, which provides protection during reactor startup and plant shutdown, can be manually bypassed when one of the two intemediate range channels roads above.the P-6 setpoint value (source range cutoff power level) and shall be automatically reinstated when both intermediate range channels decrease below the P-6 value. four logic from the power range permissive (P-10).This trip shall be autom This trip function shall also be reinstated below P-10 by

• .n board mounted switches. administrative action requiring manual actuation of two cont fynction in une of the two protection logic trains.Each switch will reinstate the trip The source range trip shall be cet between the P-6 setpoint and the maximun nource range level, the channels can be individually blocked at the nuclear instrumentation racks to permit channel testing at any time un.ter prescribed adainistrative procedures and only under the direction of authorized supervision.

anauociated on the control botrd. This blocking action shall be l 7h ** u eu..,./.5 /< e ~,ede a k. ro.y e. n eo,e o v w, - n.

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• from eteessive roactor system cooldown and to atold unnecessary sa.t.attlen

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21. Would the proposed activity increase the probability of a evaluated in the SAR?

n accident previously O Yes

1. No J_us t i fic a t lon:

., ~f~la .s.uasa. .,J r-k 4 - Ce. <* y< ne v /r, n. niAe sq sysL s/ <a,,.+ pr.vad<. s fus<.fs h ro <<dese 14< te-La Lal /y.f ca.,d,j,,;, ~ g o e. G {*

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22. Would the proposed activity increase the c.a y aylu <.,

a,,f, egu,peani, *e ryJM s<ssss**y Lasokiv. _an ac=ide. f evaluated in the SAR7 onsequences of an accident previously, 0 Yes il No Justificatlon_: 7~le s. u em., a/ < $ 4e..e /im /<. ra a)e adv/e**1

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23. Would the proposed activity increase the probability of a malfuncti important to safety previously evaluated in the SARf on of equipment f) Yes

1. No Justification:

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24. Would the proposed activity increase the consequences of a malfunct.lon of egulpm important to safety previously evaluated in the SARf O Yes V No Justification:

As 4sw n</*n fb tm 4*N 13 f.o.sf l'<.a /> A, f L s d<lo & o .J C.. jua.( i L. n e.u /~ m , Ms ,s n. + a., e u u./. N<.- w d,_ p.., m x., m /* a. e<1c.4.- p- ~ <xeeneL.f leu fu w pf,k u,'l sai L*.,,, < <<a d so-e./L. n <.v /m l s waj, w,)/ 4,,,,j!,JL /. ,,,f /,4 %.. r e4 < M / r @.

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i ,SI.1 /I {' A 2-Lt S* A'Y E**luai%',, A)o. E c o z g i g g' I SAFETY GVALuA~r.toAl

25. Would the proposed activity create a possibility for an accident of a different type than any evaluated previously in the SAR7 0 Yes W No

.Justificatient .a rl<..~p.s 4.a/..,s /, v e, /. fia w.'t/ p.uf.-- vie ( ,d -/,h/ fa- <.fi A 4t/-~.Q rle. ~. A Sa /,,:, o.s ',o co fL ~ /AAlia wld a 4: i 7 - l/aLh. -./ nt,:4;/,l. rl. in b ~<4.t'<. y \\ t.;l,,,. n, //- /-.)a. ! de. s~~. ~y i it ,,w t~.- -(lax /c.p w,/*/ a,li l.,< I o rC.,<.fa k a c/,.s e,& \\ i '~ M /t u.f.~.s

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26. Would the proposed activity create a possibility for a malfunction of equipment of a different type than any evaluated previously in the SAR7 0 Yes

(( No Justification: Z~ple~e.-/k 1E pN aud a dwoly n ne coy & p e.-p/y w.il i La uwa c 's 4,, a. <;4. i. 4 7. A//.w,y jl;s ~./,Aea/a n, /4..fL4of >Ls lev ~uklik w,'y ^

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w tc s <4.ly &a /ve.lii n.. se-n teiec ,v e e c. c i .5 A FETy' EL/ Alt.)A *TZoAJ. 27'. Would the proposed activity reduce any margin of safety as defined in t any technical specification? or O Yes / No 4 Justifleation: ~ 7'la ., 4. - -- 1.., 4 n. V., < d p <,ef-s /i. y ,.u. - a L, A G m a,'.-<.-h ~- addm.r<d,., rad s,a.e. sa.a,.:. 2/y.c./. .r..,. /<-.-r /ik il< f r*pd a c/,0,17 's%<s t 'A 3.) A' t UNI r<1u.h vP14*3 /4. -tw elsele, Aa / pev.?c a li k "sy s 4-s, si~.1,,

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.a 3 d

p u d. 4 S l a.-f 1 2.,j s J a4.fy Eve / val,.a TABLE 3.3-1 (Continued) TABLE NOTATION tor trip system breakers ~in the closed position tea capable of rod withdrawal, and fuel in the r and the control ) associated with the protective functions deriv d f eactor vessel. actor Coolant Loop shall be placed in the tripped conditi e rom the out t .0.4 ar'e not applicable, on, to detector may be de energized above the P-6 (Bl Jr!al atPoiat,._ ~ ock of Source ACTION STATEMENTS Oth the number of channels OPERABLE one less than required b ne Minimum Chunnels OPERABLE requirement, restore the inoperabl 1annel to OPERABLE status within 48 hours y ithin the next 6 hours and/or open the reactor trip br e or be in HOT STANDBY

eakers, th the number of OPERABLE channels one less than the T aber -of Channels STARTUP and POWER OPERATION may proceed oa evided the follow,ing cond',tions are satisfied:

The inoperable channel is placed in the tripped condition ~ within 6 hours. lR51 The Minimum Channels OPERABLE requirement is met; howev one additional channel may be bypassed for up to 4 hours, er for surveillance testing per Specification 4.3 1 1 1 lR51 to 75% of RATED THERMAL and the Power RangeEither, high trip reduced to less than or equal to 85% of RATED , :.eutron Flux THsRMAL POWER within 4 hours; or RnTIO is monitored at least once,per 12 hoursthe QUADRANT POWER TILT

• The QUADRANT POWER "ILT RATIO, as indicated by the remai i three detectors is serified consistent with the normali m

n ng symmetric power dist.aibution obtained by using the movable zed incore detectors in ti,e four pairs of 'ymmetric thimble locatiens at least once per 12 hours when THERMAL POWER i s a; greater than 75% of RATED THERMAL POWER. s EE211LE l' sj W , S,* 0,*.e wb e r 1/, 19 8 6 3/4 3;5 Amendment No. 47 'g,, 7 ,, ~j i

AWcf S Lu.t 1.1j ~ ~ J. A.G &c fue G i />.. Eca &tt9d TABLE 3.3-1 (Continued) TABLE NOTATION = With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel. The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

1. he revisions of Specification 3.0.4 are not applicable.

N High voltage to detector may be de-energize 6 (81'ock of Source ~ ~ Range Reactor Trip) setpoint. f&lETG ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than re utred by the Minimum Channels OPERABLE requirem nt, restore th inoperable channel to OPERABLE status within 48 hours or be in HOT STANOBY within the next 6 hours and/or open the reactor trip breakers. ACTION 2 - With the number of OPERABLE channels one less than the Tot Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: The inoperable channel is placed in the trioped condition a. within 6 hours. lR35 b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 4 hours lR39 for surveillance testing per Specification 4.3.1.1.1. Either, !!iERMAL POWER is restricted to less than or equal c. to 75% of RATED THERMAL POWER and the Power Range, Neustron Flux trip setpoint is reduced to less than or e 85% of RATED THERMAL.e0WER within 4 hours; or, qual to the QUADRANT POWER TILT RATIO is monitored at least once per i 12 hours. d. The QUADRANT POWER TILT RATIO, as indicated by the remaining three detectors is verified consistent with i.he normalized symmetric power, distribution obtained by using the movable incore detectors in the four pairs of symmetric thimbic locations at least once per 12 hours when THERMAL POWER is greator than 75% of RATED THERMAL POWER. .e 4 l September 17. 1986 SEQUOYAH - UNIT 2 3/4 3-5 t b Amendment No. 3? - e. i , g. .. gg6 j 2 .c., '6

p ie>* c. c 3 l " 'f *.lW S Aly EvoNal'A w,

a. sce t.//W TABLE 3.3-1 E

i REACTOR TRIP SYSTEM INSTRUNENTAM i o-<$ ) "i

• ^

FUNCTIONAL UtsIT TOTAL NO. CHANNELS CHANNELS APPLICA8tE MINIMUM OF CHANNELS Manual Re.setor Trip -TO TRIP OPERA 8tE MODES ACTION 1. 2 1 2. 2 1, 2, and

• Poe.er Range, Neutron Flux 1

4 2 3. 3 1, 2 R Power Ran;e, Neutron Flux 2 High Positive Eate 4 2 3 ~

1. 2 2

4 Power Range, Neutron Flux, High Negative Rate 4 2 3 1, 2 R 5. 2 Inter. mediate Range, deutron Flux g. ~ 2 1 Y 6. Sourc,e Range, Neutron Flux 2 1, 2, and a 3 A. Startep j)gcETE 8. .ShutdoJn~ 2 ~ 1 2 2 0 , and

• 4 7.

Overtemperatere Delta T 3, 4 and 5 1 5 Four Loo ~p Operation 4 2 i,. 8. Overpower Delta T 3 1, 2 6, 3, Four Loop Operation 2 lR45 4 3 1, 2 6, 9. Pressu izer Pressure-Low h D. 2 l R45 4 10. 3 1, 2 II" Pressu.-izer Pressere.'Ifgh 6 ? 4 11. 3 1, 2 9 2 .E.E. Pressurizer Weter Leve!--High 6 3 i 2 De 2 1, 2 7, f* M T ba = y-i .G. i e.* e '} 2,- }

• O.
• 1~

4

v.. -

,rp em ,.s> >e , *$~*>**' N;

• 'o*'

Q

\\ nGf Li S$sef.25 l ~ Safedy too/w.bl, p.. Eco c cret TABLE 3.3-1 v. Mg REACTOR TRIP SYSTEM INSTRUMENTATION g. ~^ MINIMUM TOTAL tiO. CHANNELS CHANNELS APPLICAB'.E FUNCTIONAL (; NIT OF CHANNELS TO TRIP OPERABLE MODES ACTION U 1. Manual Reactor Trip 2 1 2 1, 2, and

• 1 u

2. Power Range, Neutron Flux 4 2 3 ,2 2# 3. Power Range, Neutron Flux 4 2 3 1, 2 2# High Positive Rate i. 4. Power Range, Neutron Flux, 4 2 3 1, 2 2 High Negative Rate ug 5. Intermediate Range, Neutron Flux 2 1 2 1, 2, and

• 3.

' ~ "4 6. Source Range, Neutron Flux d A., Startup 2 1 2 , and

• 4 B.

Shutdown 2 0 1 3, 4 and 5 5 7. Overtemperature AT ' Four Loop Operation 4 2 5 1, 2 6, 3. Overpswer si R33 c. Four Loop Operation 4 2 3 1, 2 6 ). 4g. M i e - E "s ?. Pressurizer *ressure-Low l R33 ? a 4 2 3 1, 2 6 C. u aa [,, 4 $ 10. Presstrizer Pressure--High 4 2 3 1, 2 6 u 2 o 11. Pressu-izer.'ater Level--High 3 2 2 1, 2 7 y G. '. c %Je y f f h y . l ** - m[_. V d 5

Alu 1./ JSea f 1$ 'f A% Safe.ly Evabal$ n.. e<.a atai m _ TABLE 3.3-10 ACCIDENT MONITORING INSTRt#tENTATION MINIM 0M JNSTRUPENT REQUIRED NO. CHANi4ELS g OF CllANNELS OPERABLE 1. Reactor C:olsac T g (Wide Range) 2 2. Reactor Cao'iang T 1 '"9'} Cold 3. Containzer.'. Pre.ssure (Wide Range) 2 4. Refyling Water Storage Tank Level lR5L 1 2 5. Rear.ror Cc,olant Pressure (Wide Range) 1 2

6. ' Pressurizer, Level (Wide Range) l RSI 1

7. Steaa Line Pressure 1 2 2/ steam line u 8. Steam Generator Level - (Ulde Range) 1/ steam line N 1/ steam generator 1/ steam generator 9. S tear.: Generator Level - (Narrow Range)

10. Aux 111ary feedwa.ter Flow Rate 1/ steam generator 1/ steam generator m

1/ pump

11. Reactor Cooltnt System Subcooling Margin Monitor 1/ pump 1

12. Pressurizer PORV, Position. Indicator

• 0 2/ valve #

1/ valve

13. Pressurizer p RV. Block Valve Position Indicator **

2/ valve

14. Safety Valve ~Posi, tion. Indicator 1/ valve 2/ valve #

1/ valve

• 15. Conta(neent hyter Leve! (Wide Range) 2

16. In Cora Therm > couples 1

4/ core quadrant 2/ core quadrant

17. Reacts Vesstl level Instrumentation System ***

.N 5 2 1 faas lR50 ~- - r L &g TNot applicable if tt'e associate;$ olock valve is in th 1 / 2 cr

• "Not applicable ff the block valve is verified in the closed position with e closed position.

34

• • "This Technical 5pecification and surveillance requirement will not be imple power to tne valve cperator removed.

z- . Instructbns are developed for the use of this system as committed to in the TVA res ~ P* PUREG-0737. R50 } -$U o Supplement 1 of ? ? At least ene channel shat! be the acoustic monitors. ( .v L-w.-. l .> ~

w n ~ p n G-Uf Algf f.f S TABLE 3.3-10 O* AJ E"'!/# T N Ecs) t. h8 C -2 ACCIDENT MONITORING INSTRUMENTATION o ^*-E MINIMUM. INSTRUMENT REQUIRED NO. CHANNELS OF CHANNELS OPERA 8LE e {

1. Reactor Coolaret THot ( Ide Range) 2 1

L e. '2. Reactor Coolant T I I'"9') Cold

3. Containsect Pressure (Wide Range) 2 1

l1

4. Refueling Water Storage Tank Level 2

1

5. Reactor Coolant Pressure (Wide Range) 2 1

l3 1

6. Pressurizer tevel (Wide Range) 2 1

7. Steam Line Pressure 2/ steam line 3/ steam ifne

8. Steam Generator Level - (Wide Range) 1/ steam generator 1/ steam generator

,1

9. Steam Generator Level - (Narrow Range) 1/ steam generator 1/ steam generator

~; Y

10. Auxiliary Feedwater Flow Rate 1/ pump 1/ pump u.

",q 11. Reactor Coolant System Secooling Margin Monitor 1 0

12. Pressurizer PORV Positio.~.14dicator*

2/ valve # 1/ valve

13. Pressurizer PORV a, 1 Vaive Position Indicator **

2/ valve 1/vaiv2

14. Safety Valve Po,-

im Indicator 2/ valve # 1/ valve

• 15. Cantafrurent Wa4 + i.s A (Wide Range) 2 1

16. In Core Thernoct... a 4/ core quadrant 2/ core quadrz.

17. Reactor 'lessel Level Instrumentation System "a 2

1 a R F.m /8 Sos **.Rl 7-% cof.aJ Aa g_ N.oc1w. j,,,j,,,4 1,;., x U m- + l ? $3

• Net applicable if the associated block valve is in the closed position.
  • Not applicable if the block valve is verified in the closed position with power to the valve operator removed.

3 7 ***This Techrical Specification and surveillance req. irement will not be implemented until Sequoyah Specific Instructio?s are deve*oped for the use of this system as committed to in the TVA response to Supple. ment 1 of R 2 . ~ f.~. NUREG-0737. ,At l 4 east ont channel shall be the, acoustic monitors. ~ s 4-p S*, y f. ,c u..:

A )E P l. t J lss i ;L t l.t L~ .raSQ evale:/4 4 Et*-Al lfild E TABLE 4.3-7 M ^ ACCIDENT MONIT0 KING IMSTRUMENTATION SURVEILLANCE R 8 { .g -INSTRUAENT CilANNEL CilANNEL 'U CilECK E 1. Rea,ctor Cqolar.t T,gg (Wide Range) CALIBRATION ] 2. Reactor Cqolact T R M Cold d* N'"9*) ^ 3. Containeer:t Pressure (Wide Range) R l M 4 Refjeling 'Jate-Storage Tank Level l R T R M J. Ree: tor Coolant Pressure (Wide Range) R M t. Prefsurizer Level R lR2 M 7 Stease Lire, Pressure R M 8. Steam Generator Level - Wide R M { 9. Swam Generator Level - Narrow L M g.,

10. Aux 1,11ary Feedwater Flowrate R

M d '.1., Reac*cr Cochat System Subcooling Margin Monitor R = M

32. Pressurizer PORV Position Indicator R

M

13. ;Fressurizer PORV Block Valve Position Indicator R

K ~ i4. Safety Valve Pos1 tion Indicator d {.S. Containment'WaterLevel(WideRange) n R i M 26.InCe[eTrer;nocouples R x s IR50

• t M

g !.

17. Reactbr Vessel L2 vel Instrumentation"*

0p R e M 3

t..

3..taara- {f k edenk-sieye JJac.$w hsfro* <nla/*

R ~ ~ 14 lg$o y-g, ~ 3 } "*"r.'s Tecanical Specification and surveillance requirement will not be inga~nented until S y i S s. Irntr:;tions are developed for the use of this system as concsitted to in *he TVA res $= pecific Ni'4EO-0737. g3o pcase in Suppiement 1 of ~ e. y . g '. (Pt 4* .g., "se

• A O

e

~ AE/ t.( 5f Y. 2E .s 4 9 E d < h e.' ~ A.se. sc.n unc : TA8tE 4.3-7 w ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS h CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION _E 1. Oesetcr Coolant Tg (Wide Range) M R ] 2. Keactor Cool,snt T e av) M R Cold 3. Q ntainment Pressure (W!de Range) M R 4. Refueling Water Stc. age Tank Level ht38 M R 5. Reactor Coolant Pressum (Wide Range) M R 6. Pressurizer Level ht38 M Q 7. Steam Line Pressure M R J. Steam Generator Level - (Wide) M R Y 9-Steam Generator Level - (Marrow) s M R 10,. Auxf 7.f ary Feedweter Flowr-te <a M R E

11. Reactor Coolant. System Subcooling Margin Monitor M

.R

12. Press rizer PORV Position Indicator s

M R

13. Press'arizer PORY Block Ys1ve Position Indicator M

R

14. Safety Valve Position Ind!cator M

R

15. Contajnsent Water Level (Wide Range)

M R 16.,In Co;e ^re-mocouples h38 M R [ ~ _171 Reacto 1 Instrumentation System

• M R

r n-Q /9 1**ns : f-t<~ A4. % nd l n _n

s4,m f.1,an n

p3s y y? -~ ~ e w _- w -g f $ 3 "This Techr.ical Specification and survefilance requirement will not be impienented until Sequoyah Specific i d g. Instructions are developed for the use of this system as committed to in the TVA response to Supplement I of <3g 3a NUREG-0727 i a 'S e o b 5 -,/.

,1

..s.

I A)U J.4 JLea.] 30" g .sm% s./,sa, [$1. Ala. ECAIt.//8f' TABLE 2.2-1 (Continued) 5 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOIMS _FUNCsIONAL JNIT _ TRIP SETPOINT g

13. Eteam Generator Water All0WABLE VALUES level--Low-Low 118% of narrow range instrument y

span each steam generator 1 17% of narrow range instrument R20

14. Steam /feedwater Flow span each steam generator Mismatch.and Low Steam

< 40% of full steam flow at Generator Water level RATED TitERR'.i. POWER coincider.t< 42.5% of full steam flow at with steam generator water level RATED Tt!ERMAL POWER coincident 1 25% of narrow range instru-with steam generator water level ment span-each steam generator 1 24.0% of narrow range instru-

15. Ur.dervoltsge-Reactor ment span-each steam generator Coolant P xps 1 5022 volts each bus 14739 volts each bus ha9 y

16; Underfregeency-Reactor I Coalant Pumps 1 56.0 Itz each bus e 1 55.9 Hz each bus

17. Ttc.bine Trip A.- Low Tr;p System Pressure 1 45 psig B..Turbir.e Step Valve 1 43 psig

-> 1% open Closure -~> 1% open i3

18. Safety Injection Input h{

from ESF Not Ap ic3hle pplicable '9 4 > txto -s,I*

19. Inte.rmediai,3 Raage Neutron it 4 x to -s %

1m 10' v3 'S: 10 re' 5 e Rang a r r W^ * " " ## N e, g' s,, 20, Power Range Neutron Flux M [ "[ (not ?-10) Input to Low Power < 10% of RATED .%,o Reactor Trips Block P-7 THERMAL POWER < 11% of RATED THERMAL POWER N qD M i co O3 *. tis'. P = e., f o ,o 4'8 4,, ((A,.N4 -M#- O - %:1

' S W 3. f '- /JE9 C S Satf faLa 4-:. p t a.. m au TABLE 2.2-1 (Continued) g .o REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS _FUNs.T5ONALUNIT TRIP SETPOINI E 13. Steam Generator Water All0WABLE VALUES L, O Level--Lov-Loi 118% of narrow range instrument span each steam generator 1 17% of narrow range instrument 0*J i,

14. Stram/Feedwater Flow span each steam generator Mi watch aed low Steam

< 40% of full ster.2 flew at t Generater.iater Level RATED THERMAL POWER coincident < 42.5% of fu11 steam flow at with steam generator water level RATED THERMAL POWER ceincidut 1 25% of narrow range instru-with steam generator water level ment span-each steam generator 124% of narrow range instru-

15. Under: voltage-Reactor ment span-each steam generator Coolaat Pumps 1 5022 volts each bus l

1 4739 volts each bus K to

16. Underfrequency-Reactor i

4 7 Coolant. Pumps 1 56 Hz each bus 1 55.9 Hz each bus os

17. -Turbine Trip A.

Lov Trip System Pressure 1 45 psig 1 43 psig E. Turpine Stop Valve C1csure 3 1% open > 1% open

18. Safety 'njection Input from E57 Not Applicable Not Applicable 2 i x. e c g

19. Intersec.iate Range Neutron 2 c, s io M-x-10 ' O Flux, P-5 Enable Block

~ m u. ' g: ."y Source Penge deactor Trip 2g eo n.e = % We u. ' S x 10 ow. og Jr Aw me. Re p Pr ; n,

• 'g-

20. Power Range Neutron Flux

< 10% of RATED e , t (not P-N) Input to tow THERMAL POWER < 11% of RATED THERMAL POWER F-Pc.*er Reactor Trips u i, '- " y Bisck P-7 4 .M ll' J m )) b "5 y 1 %. ?- ~ y & ye lpo qp %. ~ m 1

- i g y(,{,_ O NeP(.,4 Cwser # 31 At r: [ '... %ssvy Evat.unsou ..y..,. ..y g* MO ECn LC18(a r'. '~ 4 m s' TABLE 3.3-9 3, ~ - I o i. -> { '( REMOTE SHUTDOWN MONITORING INSTRUMENTATION ) hh, %- i 1 .g. ~ T. 78 INSTRUMENT READOUT MINIMUM MEASUREMENT CHANNELS _ LOCATION f ji T 7 eoveerrscATE RANGE OPERABLE Tl U > 1. Soucce Ran w uclear Flux d y NOTE 1 6 1 to 1 x 10 7 2. Reactor Trip Breakr Indication /si' w zoo % CTP at trip switchgear ' OPEN-CLOSE 1/ trip D aker 3. R'eactor Coo! ant Temperature - NOTE 1 , Hot. Leg 0-650'F 1/ loop lR80 ,4 Pressurizer Pressure NOTE 1 0-3000 psig 1 g 5. Pressurizer Level NOTE 1 0-100% 1 3 Steam Generator. Pressure lR80 6 y NOTE 1 w. 0-1200 psig 1/ steam generator 7. "Steam Generator Level NOTE 2 or near Auxilary F. W. 0-100% Pump 1/ste'an generator 8. Full Length Control Rod Position Limit Switches Auxilary Instrument 1 insertion limit Room: Racks R41-44 On-off switch / rod 9.. RHR Flow Rate .N0TE 1 0-4500 gpm 1

10. RHR Temperature NOTE 1 50-400*F 1

11. n c h, Acxiliary Feedsater Flow Rate NOTE 1 0-440 gpm 1/ steam generator R80 3 U = p . *~ 25 ^2 2. g .: a 1 y . ~. l f, 3,. .,a u... - 3'.

3 h

.cb

[j, MEP Cc6 Susur$f-gf- ~.,)_ '%* ' : :y EW U"T'*# ni-s .z ....p; jf..,I I' p oa EC.H LLibG TABLE 4.3-6 m.,,. s<.

-<.i.i 5.,'

-.-.vnn..! HONITORING INSTRUMENTATION ~5f c SURVEILLANCE REQUIREMENIS g IJ ~ INSTRUMENT '. CHANNEL CHANNEL CHECK CALIBRATION L. 1. h.T& M2ca1G" T e Soerce Rang Juclear Flux -e n . ~. s H R ~ 2. Reactor Trip Breaker Indication u H H.A. 3. Reactor Coolant Temperature - Hot Leg H R 4. P'ressurizer Pressure H R 5. Pressurizar Level H R 6. Steara Generator. Pressure M H R 7. Steam Generator Level T H R 8. Fulh length Con.rol Rod Position Li iit Switches M R 9. RHR J10w Rn e H R

10. RHR icepera<ure H

R

11. Auxi?iary Ferdwer Flow Rate H

R

12. Press.rrizer.telf 2f Tank Pressure H

R 17 ntainment Pressure yg

e..

H R C 3 et 3 l e. ."w I u, .- y a. e 1,..... ;. h,. a -4 .~

r .=~ ',* {

• ~

Ntt Ce(o SHEET 3b $f l .i. ~ SA.amTy E.VAtoATswo ,q.. f t kb E m LG(6s _ TABLE 3.3-9 ~ i REMOTE $HUTDOWN MONITORING INSTRUMENTATION C hbb ~~ ez READOUT MINIMUM INSTR EENT ',' MEASUREMENT ti e tMd. LOCATION J CHANNELS -4

1. ~four-f ange Nuclear Flux

_. RANGE t OPERABLE NOTE I 6 1 to 1 x 10 cp 2. React / Trip Breaker Indication \\ 4.3@7 3T gP i 30 at trip switchgear OPEN-CLOSE

3. ' Reactor' Coolant Temperature -

1/ trip breaker Fpt. Leg NOTE 1 0-650'F 1/ loop 067 4. Pressurizer Fressure NOTE 1 0-3000 psig g - 5. Pressurizer lavel 1 NOTE 1 0-100% y 6. Steam Generator Pressure 1 NOTE 1 a67 0-1200 psig 7. Steam Generator level 1/ steam generator NOTE 2 or near Auxilary F. W. 0-100% Pop 8 Full Length Control Rod

• 1/ steam generatcr

,Positipn Limit Switches Auxilary Instrument Room: Racks R41-44 On off 1 insertion limit 9' RHR Flow Rate switch / rod NOTE 1 I, 0-4500 Spa 1 10. 1 xHR Temperature NOTE 1 50-400*F 11. 1 c$ 7.uxilitry Feedwater Flow h.te u NOTE 1 G L*. 0-440 gpa 1/ steam generator a67 n.G. -

• 3 w ",

..O 0 y.. V V )

.t

i.... ',i~ :q e

f t }...d,. 4~ m c <. su

r. g.n y _r i

s.=w e ,,so.- ~ ~ I '" 5 ~ "*IW TABLE 4.3-6 I, . G.. Ip ~ RDm)TE SHUT 00WN MONITORING INSTRUMENT"ATION 6 SURVElttANCE REQUIRENEN15 5 \\~ i e .g n s.. ~ c i CHANNEL ClANNEL I-3 INSTRUMENT ) CilECK Call 8AAT10N w m ue**sra m. 1. Source Aas uclear Fium M R \\; ; 2. Reactor Tr' Breaker Indication M N.A. 3. Ce.rcter Cor,] ant, Temperature - Het Leg. M~ R ~ ~ 4. Pressurizer,. Pressure M R ~ 5. Preswrizer Leve) M R u A. N 6. Steae,Genert, tor Pressure M R e u , 7. Steam.Gener, tor Level ~ = M R s ~ 8. Full Length i:ontrol Rod Posttien Limit Switches M R* l l 9. RHR Flow Rate M R

10. RHR.Tegeratar e :

M A ,g li. Auxili.ary fet,Asat:r F*ow Rati M A ~. .t ~ c ,f gp ?- N di' 10 Presse.-izer R 11e* Tad Pressure M R ,~ a. n ,a, ?

13. Contal.ieent Pressure M

R a e !6 o **

_ ~

• Fo: cycle 1, this surveillance is to be completed before the next cooIJwn or by A6 gust 5.1983 whichever is earlier.

R20 i[I \\ ,~ 4 lh j .N R, '

k. i..

i.....

... u.

• U

~~ =

UNHEVIEWED SAFETY QUF.STION DETERMINATION Skui 3 6 gi TV.A 10551 (EN DES 2 81) SHEET #1 TO: Secuevch Nuclear Plant. DOisy, TN 061%Ro

nou *.

El6 SNP. MEDS ACCESSION NO. oT pp PREPARED R EVIEYvEO APPROVED D^TE DP S@ 'M 0117 503 APPO q E M AP l i[, 3 <l 1 \\ R b k\\ Y \\ \\ JJj \\

1. FA 5

R 'R 'INITI AL l$ PAOJECT__6 G N AFFECT NIT (5) ld ".1 ECN NO. Md.b ECN DATE -} Al' b 7g ggg gj YE S/NO SHEET NO. McM1DCR NO 1 15 c. _.DATE t. 2VIREMENT(5) Mo OTHER DATE _ i POTENT CHANGE [f5 O /" 1ErERENCES Llb S Sup nrI -6bre bbt CnnjfirmNe n oi+ L derr[H GEP e le c x - n N DES (.RIPTION OF CHANGE Cib N e r* 1 el b e41c_~ en v, t $.LL rade. f.'ume < < e tab . ut e hie d.h bond ceb:.b e_<j o $ me Chu_.1F $ g Guide i.95,brd re t e. e t o ' meme ' ~ & F M.mer.e re U dle pmfuc;clent men. 'n.R J etm+ie. d. _ [ c n. I $ Vi rnt u>i'l l e ui r.. 130-V ClamM d u;e l l dra (dfRfpf 4 50 b f_rdiPorsmEnf4ll .d[Pf_O MR 4 } Mp M # ta k t h e r . Q t! ftr ff l ......... - ~. ~ C ( ATT AC HM f N f $n 3'0 Yr5 CHIEF NUCLEAR ENGINttR.Wl0Cl24 I' CHIEF. ARCHITECTURAL DESIGN BR ANCH.W4Cl26 C K C Hit F, MECH ANIC Al EN DE S B R ANCH.102 FT 3 Chit F. CIVIL ENGINEE RING git ANCH, wt0324 C.K CHitF. QUALif V ASSU't A.NCE PRANCH.WitCl26 C K .(. CHitF. Civil EN CES OR ANCH.W 3Cl26 C K M AN AGE R OF LONST RUCTION. E7824 C K

• • e' Chit F. ELECT RIC AL ENGINEERING BR ANCH.W8Cl24 C K CHitP.N$T PLANNING AND CONTROL STAFF,W12C74 C K Chit', ELE CT RIC AL E N cts SR ANCH. W20224 C.K PLAN 4 $UPE RINTENotNT Chit F. MECH ANIC AL ENGINEE RING B R ANCH.W3 CO225 C K OIRECTOR
• ILE AR NWE R DIVill0N,736 CS#

MEDS, E4837 C K A

v\\5 % RMSBA (EN DES 7 80) % 3,7. \\ Sheet UN AEVIEWED SAFETY QUESTION DETE AMINATION } LGIS&R.o ) (. lDENTIFIER Unreviewed Safety Questiont

1. l$ the probability of occurrence of t i

onsequencel of an accident or malfunction of equipment

15 fety previously evaluated in the $4fety Anal)

Report inc d? ...........,...........Yes _No _ )c. 9 Justification: e Aeuf f ^M d' t4 A Ye m ek 0 e.S not ocekor en w Canek;es\\. ne ee s s, - O, conelifieh The _ LNa k;n +ha _ u n ; +-

r a

-is bei ra de d to ensure. _16 co en h' ll f'el lw [n%ee;dkn_u a +. nd %s ocevid e. d'b t. era *fo r s rJi Add e ornac [o m._ reActedi Yk.c. _cond [ h[o A ok

• f b t f e.

ps' O AY tke. (9) c s a

• prw enY w'll e t e e t d.

f h e. de 'A n . r*e o Yb e, eo 'o e nt' beim re olete e d. +\\n Neo hit I s, eurren e s % e. _ Cons o en kAI maeTS,,re3 e, Alvaf ed Io. "fb nm4 :

d.._.

\\

2. Is the possibihty for an accident or malf' ie of a iferent ce than any evaluated previously in 13e 5.ety A aly-Re por. c rea d?............

Ye _No1 Justificatio n t Me rg n i g rd _wjll .,,, ge m m

• C C et

^. 4 d _C1nu 16. T'h e r e-/o' _. + 'll _e u er.d + 1 s Sb dtme i ;E t.ttnLPLf4._ i O r\\ re Q ib5

• S l Y b r~

m o! nn mp eliffe'r, & i8nd create.d. pe .a.Mi ia Nj,An _. d_dupYio r. Lo e' l 1 re a cle. O ensure tbd fbe. n d d i fiana._t loa.d Fke CNs IE_paat staan _vi l l ,n on h wc _. nnpdvcr;u _od^Srfs - l 3. t* the' rayon of safety as denned in the basis for any technical s r ed u c ed ?.....................................p e c i f ic a t i o n ...................Yes. ,,, N o. D lw^n o t_$ I ' C t1 hlO c\\ $9PVer YO ImprOVt s. +,'< - nu w a w p. Act ch m.x w.e _stfe N _ a9 de Iined inO-feeh rpe c. 3/f. 3. 3.~1 hceiden __ . hte'e,'S 9er.tsd~Xra-leueenf So, ) 12 _I % : nck redoee d. Pe - tk e. n&a e % k, - p.; w.* e t _:re c t_ _. '"^d'~- ~ ~ ~d ' +

a

..x es a 4 Les x+ i ..A

TVA 105510 (EN DES 7 80) I

• ~

Sheet O UNREVIEWED SA.TETY QUESTION DETERMINATION \\ LL,I8(o R c) (- p,,g $(Q d l Specl61 equirement(s) or Precaution (s).. ( , Marks Subject to Which { \\ Safety Eva a t io n.......... 7....... Preparer MA 6 Additional in rma tion......... Reviewer _ 8% (initial) l- _ I%tenficI Te\\ S oe e_ % rdUC P/t -shewlel_ eview af. isc h soces on ace;de n f if or ;nn imte am hti el 'F _m on eebu.0e d. ,\\ os. _t e anw ch ne ne s } U a r e_ \\ \\ /_Z/ 1 g \\ ' 7J

\\

V/ A ) 1 /y\\ s ~ -h a y; ', f ~ I 1 ..***F* e- 'y e (._ ,,.o..:- ? - ~ ~ - - e' ....,e e ..y,,.,-., 'W..I ..a u b

REVISION LOG, Sa A.h evaloa -lish Ac Gco t. tt9C r n., n.,i. i.,. DESCRIPTION OF REVISION $',, I n.. i, O .7., i yr a'l.1s$ o 4-. I ke V ' dCN 05'SS U.SQb f /fftt.5 $Cp 9 $~oti7 $*Q 7'o *n 5*et or* k NEP S d / My ui 9 -s, do of ,:, a b a.la) L.- 4.- ~ f. a..,/, m 4 t, k usoi) 5lu./:.3 4, T, l -: E ,z ~3 4, 37, f38 2 fn e.

• o p o e r s'<- plani /s <. ~ s o n) <-, -a., y

.y,, ,., < k,)z

4.,~. - /, / fr. L -A /t s a A /y e v. /v a / L.. i

.n ce eer./a k gm } *n # t7 / 4 m.s /o o%flc~e.nfo J vlty45 / r w t n et. f }ef ncw jbaL& P m **# /Y, a vl s't',v! u. Jo& a . a < <, x, i u -. a,:,i,. L I ( I P l

l l

b i .o r,'. - 3 TVA 94434 (EN Otl 4 70) r ,g-1.- "th

== 4

I o ENCLCSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (IVA-SQN-TS-L8-42) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS r i O .S f = ~. 4 k .ua. .i s ] i a.n a d 65-4 .m ,av. ~

e ENCLOSURE 3 Significant Hazards Evaluation TVA has evaluated the preposed technical specification 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 SQN in accordance with the proposed amendment will nots (1) Involve a significant increase in the probability or consequence of an accident previously evaluated. The three administrative changes are proposed to support the installation of the Camma Metrics source range (SR) and ir,termediate range (IR) detector assemblies. The first involven the deletion of a table note that is not applicable to the design of the new SR detectors. The second involves a change in engineering units for the P-6 setpoint that results from the difference in output signals from the IR detectors. The third involves the addition of a certain footnote to enable the review and approval of the unit 1 change to proceed independently of the unit 1 installation schedule. The new SR/IR detectors are class-1E equipment that is seismically and environmentally qualified and compatible with the present design requirements. Because the new hardware is compatible with the present design requirements and the proposed technical specification changes are administrative in nature, the proposed amendment will not involve a significant increase in the probability or consequences of an accider t previous 1/ evaluated. (2) Create the possibility of a new or different kind of accident from any previously analyzed. The three administrative changes are proposed to support the installation of the Gamma Metrics SR and IR detector assemblies. The first involves the deletion of a table note that is not appilcable to the design of the new SR detectors. The second involves a change in engineering units for the P-6 setpoint that results from the difference in output algnals from the IR de tec tors. The third involves the addition of a certain footnote to i enable the review and approval of the unit 1 change to proceed independently of the unit 1 installation schedule. The new SR/IR detectors are class-1E equipment that is seismically and environmentally qualified and compatible with the present design l requirements. Because the new hardware is compatible with the present design requirements and the proposed technical specification I changes are administrative in nature, the proposed amendment will not 3 create the possibility of a new or different hinJ of accident from 4 any.previously cnalyzed. (3) Involve a significant reduction in a margin of safety. The three administentive changes arv proposed to support the installation of the Carna Metrics SR and IR detector assemblies. The first involves the deletion of a table note that is not applicable to the design .t 1 I i f ~..~. -w . s.4

se o. w -2 of the new SR detectors. The second involves a change in engineering units for the P-6 setpoint that results from the difference in output signals from the IR detectors. The third involves the addition of a certain footnote to enable the review and approval of the unit I change to proceed independently of the unit 1 installation schedule. The new SR/IR detactors are class-1E equipment that is seismically and environmentally qualified and compatible with the present design requirements. Because the new hardware is compatible with the present design requirements and the proposed technical specification changes are administrative in nature, the proposed amendmcnt will not involve a significant reduction in a margin of safety. 4 t a s ,}}