Text

Amend 141 to License DPR-77,revising Tech Specs 2.2.1, 3/4.3.1.1 & 3/4.3.2.1 & Associated Bases for Revised Specs to Reflect Reactor Protection Sys Upgrades & Enhancements Implemented on Unit 1 During Cycle 4 Outage
	ML20043H903
Person / Time
Site:	Sequoyah
Issue date:	05/16/1990
From:	Black S; Office of Nuclear Reactor Regulation
To:	;
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ML20043H904	List: ML20043H903; ML20043H905;
References
	NUDOCS 9006270055
	Download: ML20043H903 (60)
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UNITE D STATES

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NUCLEAR REGULATORY COMMISSION g

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i Anendment No.141 l

License No. DPR 77 1.

The Nuclear Regulatory Cora.ission (.the Comission) has found thatt A.

The application for amendtrxnt by Tennessee Valley Authority (the licensee)datedJanuary24, April 25,andMay 15, 1990, complies with the standards and requirernents of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; D.

The facility will operate in cerformity with the application, the provisions of the Act, and the rules'and regulations of the Conr"sssien; C.

Thereisreasonableassurance(1)thattheactivitiesauthorizedby this amendment can be cenducted without ender. Sering the health ana safety of the public, ard (ii) that such activities will be conducted in cceplianct with the Comission's regulations; O.

The issuance of this artndment will ret be inimical to the comon defense and security or to the health end sefety of the public; and E.

The issuance of this anndrent is in 6ccordu.ce with 10 CFR Part 51 of the Comission's regulatiens ard all applicable requirements beve bu.r. satitfiec'.

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Accordingly, the Itcense is an. ended by changes to the Technical Specifications as indicated in the attachnent to this license anendment I

and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby anended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as i

revised through Amendment No.141, are hereby incorporated in the license.- The licensee shall operatte the facility in accordance with the Technical Specifications.

This license amendnent is effective as of its date of issuance.

3.

4 FOR THE NUCLEAR REGULATORY COMMIS$10N Suzann ' Black, Assistant Director for Projects TVA Projects Division Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of !ssuar.ce: May 16, 1990

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ATTACHMENT TO LICENSE AMENDMENT NO.141 FAClt.lTY OPERATING LICENSE NO. DPR-77 l

DOCKET NO. 50-327 l

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages e

are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages, marked with an "*".

are provided to maintain document completeness.

REMOVE INSERT 1-2 1-2 1-5 1-5 1-6 1-6*

2-5 2-5 2-6 2-6 2-6a 2-7 2-7 2-8 2-8 2-9 2-9 2-10 2-10 2-11 B 2-3 B 2-3*

B 2-4 8 2-4 B 2-5 B 2-5 B 2-6 8 2-6 B 2-6a B 2-7 B 2-7*

l 3/4 3-2 3/4.',-2 3/4 3-3 3/4 3 3 t

3/4 3-4 3/4 3 4*

3/4 3 5 3/43-5 3/4 3-6

?/4 3-6 3/4 3-7 3/4 3-7 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-13 3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12

-3/4 3-13 3/4 3-13 3/4 3-15 3/4 3-15 3/4 3-16 3/4 3-16 3/4 3-17 3/7 3-17*

3/4 3-18 3/4 3 3/4 3-19 3/4 3-19 3/4 3-19a-3/4 3-20 3/4 3-20*

3/4 3-21 3/4 3-21 3/4 3-21a 3/4 3-21a*

3/4 3-22 3/4 3-22 3/4 3-23 3/4 3-23 3/4 3-23a l

3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25

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2-REMOVE INSERT

-3/4 3-26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-27a 3/4 3-27a 3/4 3-27b' 3/4 3-28 3/4 3-28 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32*

3/4 3-33 3/4 3 33*

3/4 3-33a 3/4 3-33a 3/4 3-34 3/4 3-34 3/4 3-35 3/4 3-35*

3/4 3-36 3/4 3-36 3/4 3-37 3/4 3-37 3/4 3-37a 3/4 3-37a 3/4 3-38 3/4 3-38*

B 3/4 3-1 B 3/4 3-1 i

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TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES S

E

1. Manual Reactor Trip Not Applicable Not Applicable

2. Power Range, Neutron Flux Low Setpoint - < 25% of RATED Low Setpoint

$ 27.4% of RATED l

c5 THERMAL POWER ~

THERMAL P0tER High Setpoint

$ 109% of RATED High Setpoint

$ 111.4% of RATED l

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THERMAL POWER THERMAL POWER

3. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with

$ 6.3% of RATED THERMAL POWER High Positive Rate a time constant 3 2 second with a time constant 3 2 second

4. Power. Range, Neutron Flux, 5 5% of RATED THERMAL POWER with

$ 6.3% of RATED THERMAL POWER High Negative Rate a time constant 3 2 second with a time constant 1 2 second

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5. Intermediate Range, Neutron

$ 25% of RATED THERMAL POWER

$ 30% of RATED THERMAL POWER Flux 5

5

6. Source Range, Neutron Flux

$ 10 counts per second 5 1.3 x 10 counts per second

7. Overtemperature aT See Note 1 See Note 3 g

8. Overpower ST See Note 2 See Note 4 n>

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9. Pressurizer Pressure--Low 3 1970 psig 3 1964.8 psig e

5

10. Pressurizer Pressure--High

'$ 2385 psig

$ 2390.2 psig

11. Pressurizer Water Level--High 5 92% of instrument span

$ 92 7% of instrument span

12. Loss of Flow

> 90% of design flow g 89.4% of design flow per loop

per loop

g

Design flow is 91,400 gpa per loop.

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

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mE FUNCTIONAL UNIT TRIP SETFOINT ALLOWABLE VALUES E

13. Steam Generator Water Level--Low-Low c

a.

RCS Loop AT Equivalent to RCS Loop AT variable RCS Loop AT variable Power i 50% RTP input i 50% RTP input i trip setpoint + 2.5% RTP Coincident with-Steam Generator Water

> 19.0% of narrow range

> 18.4% of narrow range Level -- Low-Low (Adverse) instrument span instrument span and Containment Pressure - EAM

$ 0.5 psig 1 0.6 psig or

?

Steam Generator Water

> 13.0% of narrow range

> 12.4% of narrow range Level -- Low-Low (EAM) instrument span instrument span With A time delay (T ) if one

<T (Note 5) 1 (1.01) TS (Note 5)

S S

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Steam Generator is affec*ad or 1,(Mete 5) 1 (1.01) T,(Note 5)

T A time delay (T ) if two or-more Steam Gene,rators are k

affected ny b.

RCS Loop aT Equivalent to Power > 50% RTP e

Coincident with Steam Generator Water

> 19.0% of narrow range

> 18.4% of narrow range z

Level -- Low-Low (Adverse) instrument span instrument span and Containment Pressure (EAM) 1 0.5 psig 1 0.6 psig or Steam Generator Water

-> 13.0% of narrow range

-> 12.4% of narrow range instrument Level -- Low-Low (EAM) instrument span

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TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ALLOWABLE VALUES TRIP SETPOINT S

FUNCTIONAL UNIT N

14. Deleted b

15. Undervoltage-Reactor 1 5022 volts-each bus 1 4739 volts-each bus

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Coolant Pumps

16. Underfrequency-Reactor 1 56.0 Hz - each bus t 55.9 Hz - each bus Coolant Pumps 1 45 psig 1 43 psig

17. Turbine Trip A.

Low Trip System B.

Turbine Stop Valve 3 1% open 3 I% open Pressu6e 7

Closure Not Applicable

?

Not Applicable

18. Safety Injection Input from ESF 1 1 x 10-5% of RATED 2 6 x 10-6% of RATED THERMAL POWER

19. Intermediate Range Neutron THERMAL POWER Flux - (P-6) Enable Block Source Range Reactor Trip

< 10% of RATED

$ 12.4% of RATED THERMAL POWER

20. Power Range Neutron Flux THERMAL POWER (not P-10) Input to Low Power

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Reactor Trips Block P-7 E

R E

er.

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TABLE 2.2-1 (Continued)

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E REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS

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C TRIP SETPOINT ALLOWABLE VALUES f FUNCTIONAL UNIT E 21.

Turbine Impulse Chamber Pressure -

5 10% Turbine Impulse

$ 12.4% Turbine Impulse l

Q (P-13) Input to Low Power Reactor Trips Pressure Equivalent Pressure Equivalent Block P-7 22.

Power Range Neutron Flux - (P-8) Low

$ 35% of RATED

$ 37.4% of RATED l

g Reactor Coolant Loop Flow, and Reactor THERMAL POWER THERMAL POWER Trip 23.

Power Range Neutron Flux - (P-10) -

> 10% of RATED

> 7.6% of RATED Enable Block of Source, Intermediate, THERMAL POWER THERMAL POWER and Power Range (low setpoint) Reactor Trips 7 24.

Reactor Trip P-4 Not Applicable Not Applicable l

25.

Power Range Neutron Flux - (P-9) -

$ 50% of RATED 5 52.4% of RATED u

Blocks Reactor Trip for Turbine THERMAL POWER THERMAL POWER Trip Below 50% Rated Power NOTATION b

b Overtemperature ai (I * '4 ) $ AT, g - g (1 + 1 bI + 'I HT-T'] + K (P-P') - f (aI)}

3 7

NOTE 1:

1+T S 2

S I * '4b lead-lag ccupensator on measured AT h

=

Where:

1+Tb 5

=3su.

= Time constants utilized in the lead-lag controller for AT,t4 = 12 secs, 15

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'4S

= Indicated AT at RATED THERMAL POWER AT, K

5 1.15 y

gw K

= 0.011 2

m' TABLE 2.2-1 (Continued)

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8 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

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NOTATION (Continued)

E NOTE 1:

(Continued)

G 1+15 g

7 y.

3 The function generated by the lead-lag controller for T,,9 dynamic compensation

=

Time constants utilized in the lead-lag controller for T,,g, ty = 33 secs.,

y, & t2

=

t 1 * # S'

2 Average temperature *F T

=

T' 5 578.2*F (Nominal T,,g at RATED THERMAL POWER)

0. 0%

't K

=

3 a>

Pressurizer pressure, psig P

=

2235 psig (Nominal RCS operating pressure)

P'

=

Laplace transform operator (sec-1)

S

=

and f (AI) is a function of the indicated difference between top and bottom detectors y

of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

en - 29 penent aM + 5 penent f (aI) = 0 (Wre qt and g (i) for qt'9b y

are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, y

al' n Penent of M MN NRL and 4 +%s t

n a

E M -

. - ~ -,.

- - -. ~.. -

-,~..

,. - ~ -,,

,..,, ~., - -.

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS wE NOTATION (Continued)-

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NOTE 1:

(Continued)

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(ii) for each percent that the magnitude of (q g) exceeds -29 percent, the AT trip set-point shall be automatically reduced by 1.50 percent of its value at RATED THERMAL POWER.

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(iii) for each percent that the magnitude of (q g ) exceeds +5 percent, the AT trip set-b point shall be automatically reduced by 0.86 percent of its value at RATED THERMAL POWER.

5 Overtemperature AT (1 * '4 ) < AT,{K

) T - K (T-P) - f IOIII 5 (1 * '3

-K NOTE 2:

6 2

4 5

I*Ib5 7

Where:

'4 as defined in Note 1

=

1+T 5 as defined in Note 1

=

T4,TS as defined in Note 1 AT

=

g K

5 1.087-4 K

= 0.02Pf for increasing average Wrature aM 0 for &creas% average f

S temperature i

o S

'3S The function generated by the rate-lag controller for T,yg dynamic

=

z 1+T S o

3 compensation M-e e

e-*

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TABLE 2.2-1 (Continued)

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'h REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 8g NOTATION (Continued) x NOTE 2:

(Continued)-

E

" "' C "St*"' "ti i i" th' r***-'*9 ' "tr " '" ' " 7 avg * 'a " 18 5'*5-

=

}

'3 OM11 for T > V aM K = 0 for T $ V l

K

=

g L

as defined in Note 1 T

=

Indicated T, g at RATED THERMAL POWER (Calibration temperature for T"

=

AT instrumentation, 1 578.2*F) as defined in Note 1 S

=

0 for.all AI f (al)

=

7 2

6 NOTE 3:

The channel's maximum trip setpoint shall not exceed its computed trip point by more than 1.9 percent AT span.

NOTE 4:

The channel's maximum trip setpoint shall not. ex *ed its computed trip point by more than 1.7 percent AT span.

g R

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TABLE 2.2-1 (Continued)

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REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 8

NOTATION (Continued) g NOTE 5:

Tri.a Ti_-- p=Tay - Steam Generator Water Level -- Low-Low T, =

-0.00583)(P)3 + (0.735)(P)2 - (33.560)(P) + 649.5 0.99 i

=

(-0.00532)(P)a + (0.678)(P)2 - (31.340)(P) + 589.5 l 0.99 T =

J Where:

j RCS Loop AT Equivalent to Power (% RTP), P < 50% RTP P =

Time delay for Steam Generator Water level -- Low-Low T

=

Reactor Trip, one Steam Generator affected.

Time delay for Steam Generator Water Level -- Low-Low I

T

=

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Reactor Trip, two or more Steam Generators affected.

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s SAFETY LIMITS

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BASES j

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Range, Neutron Flux l

l The Power Range, Neutron Flux chantiel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected I

by temperature and pressure protective cir:uitry.

The low set point provides redundant protection in the power range ',oe a power excursion beginning from l

low power.

The trip associated with tN ltw setpoint may be manually bypassed when P-10 is active (two of the four rower range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER) and is auto-l matically reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approx'mately 9 percent of RATED THERMAL POWER).

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

The Power Range Negative Rate trip provides protection to ensure that the oinimum DNBR is mair ained above 1.30 for control rod drop accidents.

At high powei a single or N itiple rod drop accident could cause local flux peaking which, I

when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconserva-tive local DNBR to exist.

The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor for all single or multiple dropped rods.

1 Intermediate and Source Range, Nuclear Flux The Intermediate and Source Ranga, Nuclear Flux trips provide reactor core protection during reactor starti.p.

These trips provide redundant protec-tion to the low setpoint trip of the Power Range, Neutron Flux channels.

The Source Range Channels will initiate a reactor trip at about 10+5 counts per second unless manually blocked when 3-6 becomes active.

The Intermediate SEQUOYAH - UNIT 1 B 2-3 Revised 08/18/87 ML ChWt0j:

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SAFETY LIMITS BASES Range Channels will initiate a reactor trip at approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

No credit was taken for operation of the trips associated 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 the Reactor Protection System.

Overtemperature Delta T The Overtemperature Delta T trip provides core protection to prevent ONB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to transit, thermowell, and RTD response time delays from the core to the temperature detectors (about 8 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 tempera-ture and dynamic compensation for transport, thermowell, and RTD response time l

delays from the core to RTO output indication.

With normal axial power distri-l bution, 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 according to the notations in Table 2.2-1.

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

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.

Delta-T,, as used in the Overtemperature and Overpower AT trips, repre-sents the 100% RTP value as measured by the plant for each loop.

This normalizes each loop's aT trips to the actual operating conditions existing at i

i the time of measurement, thus forcing the trip to reflect the equivalent full l

power conditions as assumed in the accident analyses.

These differences in RCS loop AT can be due to several factors, e.g., measured RCS loop flows greater than thermal design flow, and slightly asymmetric power distributions between quadrants.

While RCS loop flows are not expected to change with cycle i

life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific AT values.

Accurate determination of the loop specific AT value should be made when performing Incore/Excore quarterly l

recalibration and under steady state conditions (i.e., power distributions not affected by Xenon or other transient conditions).

Overpower Delta T 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 Neutron Flux trip.

The setpoint includes corrections for changes in SEQUOYAH - UNIT 1 B 2-4 Amendment No. 136. 141

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SAFETY LIMITS BASES i

density and heat capacity of water with temperature, and dynamic compensation for transport, thermowell, and RTD response time delays from the core to RTD output indication.

The Overpower Delta-T trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."

Delta T,, as used in the Overtemperature and Overpower AT trips, repre-sents the 100% RTP value as measured by the plant for each loop.

This normalizes each loop's AT trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident analyses.

These differences in RCS loop ai can be due to several factors, e.g., measured RCS loop flows greater than thermal design flow, and slightly asymmetric power distributions between quadrants.

While RCS loop flows are not expectec to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific AT values.

Accurate determiiation of the loop specific AT value should be made when performing Incore/Excoce quarterly recalibration and under steady state conditons (i.e., power d'stributions not affected by Xenon or other transient conditions).

Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted.

The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig).

The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolint pressure.

Pressurizer Water Level i

The Pressurizer High Water Level ; rip ensures protection against Reactor Coolant System overpressurization by 1 miting the water level to a volume sufficient to retain a steam bubble anil prevent water relief through the l

pressurizer safety valves, No credit was taken for operation of this trip in i

the accident analyses; hnwever, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

MssofFlow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drops below 90% of nominal full loop flow.

Above the P-8 interlock, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow.

This latter trip will prevent the minimum value of the DNBR from going below the safety analysis l

DNBR limit during normal operational transients and anticipated transients l

when 3 loops are in operation and the Overtemperature Delta T trip set point I

is adjusted to the value specified for all loops in operation.

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SEQUOYAH - UNIT 1 B 2-5 Amendmant No. 138, 141

SAFETY LIMITS 4

BASES i

Steam Generator Water Level J

The Steam Generater Water Level low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch i

resulting from loss of normal feedwater or a feedwater system pipe break,

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inside or outside of containment.

This function also provides input to the steam generator level control system.

IEEE 279 requirements are satisfied by 2/3 logic for protection function actuation, thus 611owing for a single failure of a channel and still performing the protection function.

Control / protection interaction is addressed by the use of the Median Signal Selector which prevents a single failure of a channel providing input to the control system requiring protection function action.

That is, a single failure of a channel providing input to the control system does not result in the control system initiating a condition requiring protection function action.

The Median Signal Selector performs this by not selecting the channels indicating the highest or lowest steam generator levels as input to the control system.

4 With the transmitters located inside containment and thus possibly experiencing adverse environmental conditions (due to a feedline break), the Environmental Allowance Modifier (EAM) was devised.

The EAM function (Containment Pressure (EAM) with a setpoint of, 0.5 psig) senses the 5

presence of adverse containment conditions (elevated pressure) and enables the Steam Generator Water Level - Low-Low trip setpoint (Adverse) which reflects the increased transmitter uncertainties due to this environment.

The EAM allows the use of a lower Steam Generator Water Level - Low-Low (EAM) trip setpoint when these conditions are not present, thus allowing more margin to trip for normal operating conditions.

l The Trip Time Delay (TTD) creates additional operational margin when the plant needs it most, during early escalation to power, by allowing the operator time to recover level when the primary side load is sufficiently small to allow such action.

The TTD is based on continuous monitoring of primary side power through the use of RCS loop AT.

Two time delays are calculated, based on the number of steam generators indicating less than the Low-Low Level trip setpoint and the primary side power level.

The magnitude of the delays decreases with increasing primary side power level, up to 50%

RTP.

Above 50% RTP there are no time delays for the Low-Low level trips.

In the event of failure of a Steam Generator Water Level channel, it is placed in the trip condition as input to the Solid State Protection System and does not affect either the EAM or TTD setpoint calculations for the remaining operable channels.

It is then necesscry for the operator to force the use of the shorter TTD time delay by adjustment of the single steam generator time delay calculation (T ) to match the multiple steam generator 3

time delay calculation (T ) for the affected protection set, through the M

Eagle-21 System Man-Machine-Interface (MMI) test cart.

Failure of the Containment Pressure (EAM) channel to a protection set also does not affect SEQUOYAH - UNIT 1 B 2-6 Amendment No.141

i SAFETY LIMITS BASES Steam Generator Water Level the EAM setpoint calculations.

This results in the requirement that the operator adjust the affected Steam Generator Water Level - Low-Low (EAM) trip setpoints to the same value as the Steam Generator Water Level - Low-Low (Adverse) trip setpoints.

Failure of the RCS loop AT channel input (failure of more than one T RTD or failure of a T RTO) does not affect the TTO calcu-H C

lation for a protection set.

This results in the requirement that the operator adjust the threshold power level for zero seconds time delay from 50% RTP to i

0% RTP, through the MMI.

Undervoltage and Underfrequency - Reactor Coolant Pupp Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of voltage or under-frequency to more than one reactor coolant pump.

The specified set points assure a reactor trip signal is generated before the low flow trip set point is reached.

Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients.

For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds.

For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip set point is reached shall not exceed 0.6 seconds, i

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SEQUOYAH - UNIT 1 B 2-6a Amendment No.141 Revised 03/18/87'

SAFETY LIMITS BASES Turbine Trip A Turbine Trip causes a direct reactor trip when operating above P-9.

i Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient.

No credit was taken in the accident analyses for opera-tion of these trips.

Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.

Safety Injection Input from ESF If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a safety injection.

This trip is provided to protect the core in the event of a LOCA.

The ESF instrumentation channels which initiate a safety injection signal are shown in Table 3.3-3.

Reactor Trip System Interlocks The Reactor Trip System Interlocks perform the following functions on increasing power:

1 P-6 Enables the manual block of the source range reactor trip (i.e.,

prevents premature block of source range trip).

P-7 Defeats the automatic block of reactor trip on:

Low flow in more l

P-13 than nne primary coolant loop, reactor coolant pump undervoltage and underfrequency, pressurizer low pressure, and pressurizer high level.

P-8 Defeats the automatic block of reactor trip on low RCS coolant flow in a single loop.

P-9 Defeats the automatic block of Reactor Trip on Turbine Trip.

P-10 Enables the manual block of reactor trip on power range (low setpoint),

interrediate range, as a backup block for source range, and intermediate range iod stops (i.e., prevents premature block of the noted functions).

On decreasing p wer, the opposite function is performed at reset setpoints.

P-4 Reactor tripped - Actuates turbine trip, closes main feedwater valves on T below sei. point, prevents the opening of the main feedwater valvd9which were closed by a safety injection or high steam generator l

water level signal, allows manual block of the automatic reactuation of safety injection.

Reactor not tripped - defeats manual block preventing automatic reactuation of safety injection.

SEQUOYAH - UNIT 1 B 2-7 Amendment No. 7 Revised 08/18/87

- v.

TABLE 3.3-1 Eg REACTOR TRIP SYSTEM INSTRLaqENTATION L

5 1

MINIMUM g

TOTAL NO.

CHANNELS CHANNELS APPLICABLE q FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 8LE MODES ACTION w

1.

Manual Reactor Trip 2

1 2

1, 2, and

1 2.

Power Range, Neutron Flux 4

2 3

1, 2 2

t 3.

Power Range, Neutron Flux 4

2 3

1, 2 2

High Positive Rate 4.

Power Range, Neutron Flux, 4

2 3

1, 2 2

High Negative Rate 5.

Intermediate Range, Neutron Flux 2

1 2

1, 2, and

3 6.

Source Range, Neutron Flux g

A.

Startup 2

1 2

2, and

4 B.

Shutdown 2

0 1

3, 4 and 5 5

7.

Overtemperature Delta T Four Loop Operation 4

2' 3

1, 2 6,

8.

Overpower Delta T Four Loop Operation 4

2 3

1, 2 6,

9.

Pressurizer Pressure-Low 4

2 3

1, 2 6

+

[ 10.

Pressurizer Pressure--High 4

2 3

1, 2 6

s

{ 11.

Pressurizer Water Level--High 3

2 2

1, 2 6#

E.

Mw'

~ - -

..m

..~

7

.., - ~.,,

~~

~.

g TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUNENTATION Y

x e

MINIMUM c

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE N00ES_

ACTION 12.

Loss of Flow - Single Loop 3/ loop 2/ loop in 2/ loop in 1

6 (Above P-8) any oper-each oper-ating loop ating loop 13.

Loss of Flow - Two Loops 3/ loop 2/ loop in 2/ loop 1

6 (Above P-7 and below P-8) two oper-each oper-ating loops ating loop 14.

Main Steam Generator Water Level--Low-Low w

D A.

Steam Generator Water 3/Sta. Gen.

2/Stm. Gen.

2/ Sts. Gen.

1,2 9

w Level -- Low-Low in any oper-in each oper-(Adverse) ating Stm.

ating Sta.

Gen.

B.

Steam Generator Water 3/Stm. Gen.

2/Stm. Gen.

2/Sta. Gen.

1,2 9

Level -- Low-Low in any oper-in each oper-(EAM) ating Sta.

ating Stm.

Gen.

C.

RCS Loop aT 4(1/ loop) 2 3

1,2 10 D.

Containment Pressure 4

2 3

1,2 11 (EAM)

I& 15.

Deleted e5 16.

Undervoltage-Reactor Coolant 6,

Pumps 4-1/ bus 2

3 1

4 z

O 17.-

Underfrequency-Reactor Coolant

{

Pumps 4-1/ bus 2

3 1

6 18.

Turbine Trip 6,#

A.

Low Fluid Oil Pressure 3

2 2

1 B.

Turbine Stop Valve Closure 4

4 4

1 6

g TABLE 3.3-1 (Continued)

~

REACTOR TRIP SYSTEM INSTRUMENTATION 5

~

1 MINIMUM E

TOTAL NO.

CHANNELS CHANNELS APPLICABLE Z FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION w

19.

Safety Injection Input from ESF 2

1 2

1, 2 12 20.

Reactor Trip Breakers A.

Startup and Power Operation 2

1 2

1, 2 12, 15 B.

Shutdown 2

1 2

3*, 4* and 5*

16 21.

Automatic Trip Logic R

A.

Startup and Power Operation 2

1 2

1, 2 12 B.

Shutdown 2

1 2

3*, 4* and 5*

16 22.

Reactor Trip System Interlocks 4

A.

Intermediate Range Neutron Flux P-6 2

1 2

2, and*

Ba B.

Power Range Neutron Flux - P-7 4

2 3

1 8b C.

Power Range Neutron Flux - P-8 4

2 3

1 8c 3

D.

Power Range Neutron g

Flux - P-10 4

2 3

1, 2 8d 3

E.

Turbine Impulse Chamber g

Pressure - P-13 2

1 2

1 8b F.

Power Range Neutron o

Flux - P-9 4

2 3

1 8e z

G.

Reactor Trip - P-4 2

1 2

1, 2, and* 14

o TABLE 3.3-1 (Continued)

TABLE NOTATION AWith the reactor trip system breakers in the closed position and 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.

The provisions of Specification 3.0.4 are not applicable.

1. ource Range outputs may be disabled above the P-6 (Block of Source S

Range Reactor Trip) setpoint.

ACTION STATEMENTS 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STAN0BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 2 -

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

i a.

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

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.1.

c.

The QUADRANT POWER TILT RATIO is monitored in accordance with Technical Specification 3.2.4.

i

\\

l 1

l l

l SEQUOYAH - UNIT 1 3/4 3-5 Amendment No. 47,135,136, 141 i

TABLE 3.3-1 (Continued)

ACTION 3 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER 1evel:

l a.

Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE statur, prior to increasing THERMAL POWER above the P-6 Setpoint, b.

Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below 5% of' RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

c.

Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.

d.

Above 10% of RATED THERMAL POWER, the provisions of Specification 3.0.3 are not applicable.

ACTION 4 -

With the number of channels 0PERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a.

Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b.

Above the P-6 (Block of Source Range R9attor Trip) setpoint, operation may continue.

ACTION 5 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within I hour and at least once per i

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 6 -

With the number of OPERABLE channele one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed j

provided the following conditions are watisfied:

l a.

The inoperable channel is placeo in the tripped-condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.1.

ACTION 7 -

Deleted.

l SEQUOYAH - UNIT 1 3/4 3-6 Amendment No. 47, 141

TABLE 3.3-1 (Continued) l e

ACTION 8 -

With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the appro-priate ACTION statement (s) for these functions.

Functions to be evaluated are:

a.

Source Range Reactor Trip b.

Reactor Trip Low Reactor Coolant Loop Flow (2 loops)

Undervoltage Underfrequency i

4 Pressurizer Low Pressure Pressurizer H10h Level c.

Reactor Trip Low Reactor Coolant Loop Flow (1-loop) d.

Reactor Trip Intermediate Range Low Power Range Source Range e.

Reactor Trip Turbine Trip With the number of OPERABLE channels one less'than the Total ACTION 9 Number of Channels, STARTUP and/or 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

For the affected protection set, the Trip Time Delay for i

one affected steam generator (T ) is adjusted to match the 3

Trip Time Delay for multiple affected steam generators (T ) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , M

c.

The Minumum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for.up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for l

surveillance testing of other channels per Specification 4.3.1.1.1.

ACTION 10 -

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , for the affected protection set, the Trip Time Delays (T and T ) threshold power level for 3

M zero seconds time delay is adjusted to 0 % RTP.

l SEQUOYAH - UNIT 1 3/4 3-7 Amendment No. 54, 141 4

i TABLE 3.3-1 (Continued)

ACTION 11 -

With the number of OPERABLE channels one less than 'the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed i

i provided that within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , for the affected protection set, the Steam Generator Water Level - Low-Low (EAM) channels trip setpoint is adjusted to the same value as Steam Generator Water Level - Low-Low (Adverse)

ACTION 12 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1 i

provided the other channel is OPERABLE.

ACTION 13 -

With the number of OPERABLE channels one less than the Total Humber of Channels and with the THERMAL POWER level above the P 7 (Block of Low Power Reactor Trips) setpoint, place the inoperable channel in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

i ACTION 14 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 15 -

With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 12.

The breaker shall not be bypassed while one of the diverse trip features is inoperable except for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for performing maintenance to restore the breaker to OPERABLE status.

ACTION 16 -

With the number of OPERABLE channels one less than the minimum channels operable requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour, t

i i

SEQUOYAH - UNIT 1 3/4 3-8 Amendment No. 54, 141

~

TABLE 3.3-2 v.

15 E5 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E

FUNCTIONAL UNIT RESPONSE TIME E

El 1.

Manual Reactor Trip NOT APPLICABLE w

2.

Power Range, Neutron Flux

$ 0.5 seconds

3.

Power Range, Neutron Flux, NOT APPLICABLE High Positive Rate 4.

Power Range, Neutron Flux, High Negative Rate 1 0.5 seconds

NOT APPLICABLE 5.

Intermediate Range, Neutron Flux SL NOT APPLICABLE 6.

Source Range, Neutron Flux Y

7.

Overtemperature Delta T

$ 8.0 seconds

8.

Overpower Delta T i 8.0 seconds 9.

Pressurizer Pressure--Low

$ 2.0 seconds 10.

Pressurizer Pressure--High 1 2.0 seconds NOT APPLICABLE 11.

Pressurizer Water Level--High 12.

Loss of Flow - Single Loop

$ 1.0 seconds 3,

3 (Above P-8)

E 2

Neutron detectors are exempt from response time testing.

Response time of the 3

A 8

neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

l2

g; TABLE 3.3-2 (Continued) x>

E5 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES Y

1 FUNCTIONAL UNIT RESPONSE TIME E

El 13.

Loss of Flow - Two Loops (Above P-7 and below P-8) i 1.0 seconds p.

14.

Main Steam Generator Water Level--

Low-Low II)

A.

RCS Loop AT 1 8.0 seconds (P< 50% RTP: P> 50% RTP)

II)

B.

Steam Generator Water i 2.0 seconds Level -- Low-Low (Adverse, EAM) u, II)

C.

Containment Pressure i 2.0 seconds (EAM) 0 15.

Deleted 16.

Undervoltage-Reactor Coolant Pumps i 1.2 seconds 17.

Underfrequency-Reactor Coolant Pumps 1 0.6 seconds 18.

Turbine Trip A.

Low Fluid Oil Pressure NOT APPLICABLE B.

Turbine Stop Valve NOT APPLICABLE g{

19.

Safety Injection Input from ts.--

NOT APPLICABLE NOT APPLICA8LE il 20.

Reactor Trip Breakers a

NOT APPLICA8LE 21.

Automatic' Trip Logic NOT APPLICA8LE

[e 22.

Reactor Trip System Interlocks U

Response times noted' include the transmitters, Eagle-21 process Does not include Trip Time Delays.

(1) protection cabinets, solid state protection cabinets, and actuation devices..This reflects the response dS time necessary for THERMAL POWER in excess of 50% RTP.

l-TABLE 4.3-1

.c5 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS I

CHANNEL MODES IN WHICH CHANNEL CHANNEL" FUNCTIONAL SURVEILLANCE l

E FUNCTIONAL UNIT CHECK CALIBRATION-TEST REQUIRED.

l 1.

Manual Reactor Trip N.A.

N.A.

S/U(1) and R(9) 1, 2, and *

~

l 2.

Power Range, Neutron Flux S

D(2), M(3)

Q 1, 2 and Q(6) 3.

Power Range, Neutron Flux, N.A.

R(6)

Q 1, 2 High Positive Rate 4.

Power Range, Neutron Flux, N.A.

R(6)

Q 1, 2 High Negative. Rate S.

Ini.ermediate Range, S

R(6)

S/U(1) 1, 2, and

w l-Neutron Flux Y

~6.

Source: Range, Neutron Flux S(7)

R(6)

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

7.

Overtemperature Delta T S

R Q'

1, 2-8.

Overpower Delta T.

S R

Q 1, 2 9.

Pressurizer Pressure--Low 5

R Q

1, 2 10.

Pressurizer Pressure--High S

R Q-1, 2 11.

Pressurizer Water Level--High S

R Q

1, 2 12.

Loss of Flow--- Single Loop 5

R Q-1 j

13.

Loss of Flow - Two Loops S

R N.A.

1

$" 14.

Main Steam Generator Water.

Level--Low-Low A.

Steam Generator Water Level --

S R

Q 1, 7 k5 Low-Low (Adverse)-

F B.

Steam Generator Water Level --

S R

Q 1, 2 Low-Low (EAM) f C.

RCS Loop AT S

R Q

1, 2 w

D.

Containment Pressure (EAM)

S R

Q 1, 2

~

,.y s

l TABLE 4.3-1 (Cont hued) m.E REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

8 CHANNEL NODES IN WHICH y

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

'I Z-FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 15.

Deleted c.

h 16.

Undervoltage - Reactor Coolant H.A.

R Q

1 Pumps

~

17.

Underfrequency - Reactor Coolant N.A.

R Q

1 Pumps 18.

Turbine Trip A.

Low Fluid Oil Pressure N.A.

N.A.

S/U(1)

I B.

Turbine.Stop Valve Closure N.A.

N.A.

S/U(1) 1 l

19.

Safety Injection Input from ESF N.A.

N.A.

R

.- 1, 2 20.

Reactor Trip Breaker N.A.

N.A.

M(5) and S/U(1)-

1, 2, and

w

21.

Automatic Trip Logic N.A.

N.A.

M(5) 1, 2, and *

.s q

$ 22.

Reactor Trip System. Interlocks to A.

Intermediate Range N.A.

R N.A.

2, and

Neutron Flux, P-6 B.

Power Range' Neutron N.A.

N.A.

I Flux, P C.

Power Range Neutron N.A.

R N.A.

1 Flux, P D.

Power Range Neutron

-N.A.

R N.A.

1, 2 Flux, P-10 E.

Turbine Impulse Chamber

.N.A.

R N.A.

I

,a Pressure, P-13 g

F.

Power Range Neutron:

N.A.

R N.A.

I a

Flux, P-9 G.

Reactor - Tri p,. P. N.A.

N.A.

R' 1, 2, ano -

e 5 23.

Reactor Trip Bypass Breaker N.A.

N.A.

M(10)R(11) 1, 2, and

g MW t

.s.-

.O..

..h,.~.

,y..,

c-..,..

. i,

.r m_

-_~;

TABLE 4.3-1 (Continued)

NOTATION With the reactor trip system breakers closed and the control rod drive l

system capable of rod withdrawal.

If not performed in previous 31 days.

-(1)

Heat balance only, above 15% of RATED THERMAL POWER.

Adjust channel (2) if absolute difference greater than 2 percent.

Compare incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED (3)

THERMAL POWER, Recalibrate if the absolute difference grecter than or equal to 3 percent.

Deleted.

(4)

'Each train or logic channel shall be tested at.least every 62 days-(5) on a STAGGERED TEST BASIS.

The test shall independently verify the OPERABILITY of the undervoltage and automatic shunt trip circuits.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(6)

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

(7)

Deleted.

(8)

The CHANNEL FUNCTIONAL TEST shall independently' verify the operability (9) of the undervoltage and shunt trip circuits for the manual reactor trip function.

Local manual shunt trip prior to placing breakeriin service.

Each (10) -

train shall be tested at least every 62 days on a STAGGERED TEST 7

BASIS.

(11) -

' Automatic and manual undervoltage trip.

l l

l t

t l

SEQUOYAH - UNIT 1 3/4 3-13 Amendment No. 54, 114, 141

~.

TABLE 3.3-3 u.

.E8 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION x

MINIMUM-E TOTAL NO.

CHANNELS CHANNELS APPLICABLE Z FUNCTIONAL UNIT td CHANNELS TO TRIP OPERABLE MODES ACTION w

1.

SAFETY INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION a.

Manual Initiation 2

l' 2

1,2,3,4 20 b.

Automatic Actuation 2

1 2

1,2,3,4 15 Logic c.

Containment 3

2 2

.1, 2, 3 17*

w1 Pressure-High 4

d.

Pressurizer-3 2

2 1, 2, 3#

17*

.w u'

Pressure - Low e.

Deleted E

lit M

e-*

.~

__.i_.

~ _.

._:..J_._.,..__.

g TABLE 3.3-3 (Continued)

E ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION g

x MINIMUM-E TOTAL NO.

CHANNELS CHANNELS APPLICA8LE-Z FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION g

f.

Steam Line Pressure-Low 3/ steam line 2/ steam line 2/ steam line.

1,'2, 3

'17

in any steam line 2.

CONTAINMENT SPRAY a.

Manual 2

1*

2 1,2,3,4-20 b.

Automatic Actuation 2

1 2

1, 2,.3, 4 15 w)

Logic b

c.

~ Containment Pressure--

4 2

3 1, 2,' 3 18 High-High 3.

CONTAINMENT ISOLATION a.

Phase "A" Isolation 1)

Manual 2

1 2

1, 2,-3, 4 20 2)

From Safety Injection 2

1 2

1,2,3,4

-15 Automatic Actuation.

Logic E

2

Two switches must be' operated simultaneously for actuation.

?* -

I

?

M 4

-,r.c...

..m~

, ~.

,L...

1 TABLE 3.3-3 (Continued)

-- m ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 35 r

MINIMUM EE TOTAL NO.

CHANNELS CHANNELS APPLICABLE 53 FUNCTIONAL UNIT OF CHANNELS.

TO TRIP OPERABLE MODES ACTION s.

b.

Phase "B" Isolation l

1)

Manual 2

1*

2 1, 2,-3, 4

. 20 2)

Automatic 2

1 2

1,2,3,4 15 Actuation Logic 3)

Containment 4

2 3

1,2,3 18 Pressure-High-High us 3:

c.

Containment Ventilation Isolation u,

Ja 1)

Manual 2

1 2

1,2,3,4 19

'd 2)

Automatic Isolation 2

-1 2

1,2,3,4 15

(

Logic 3)

Containment Gas 2

1 1

1,'2, 3, 4' 19 Monitor Radioactivity-High 4)

Containment Purge 2

1 l~

1,2,3,4 19 3,

2

' Air Exhaust Monitor l

EL Radioactivity-High' i

5

- 3 5)

Containment Particu-2 1

1 1,2,3,4 19 late' Activity _High 2e

?

Two switches must.be operated simultaneously for actuation.

)

P

.,..$h,

...,y

.. +,.,

.iw.

m.

._c y_

m.w,

. TABLE 3.3-3 (Continued) v>-

m E

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E

MINIMUM.

E TOTAL NO.

CHANNELS CHANNELS APPLICABLE M FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 8LE MCOES ACTION

~

4.

STEAM LINE ISOLATION a.

Manual 1/ steam line 1/ steam line 1/ operating 1, 2, 3 25 steam line b.

Automatic 2

1 2

1,2,3 23_

Actuation Logic c.

Containment Pressure--

4 2

3 1,2,3 18 High-High w

d.

Steam Line Pressure-3/ steam line.2/ steam line 2/ steam line 1, 2, 3 17' w

in any steam O

Low line Negative Steam Line-3/ steam-line.

2/ steam line 2/ steam line 3

17*-

e.

Pressure Rate-High in any steam line 8-

?

M

l TABLE 3.3-3 (Continued) u, E8 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION r

MINIMUM E

TOTAL NO.

CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 5.

TURBINE TRIP &

FEEDWATER ISOLATION a.

Steam Generator 3/ loop.

2/ loop in 2/ loop in' 1, 2, 3 17*

Water Level--

any oper-each oper-

.High-High ating loop ating loop b.

Automatic Actuation 2

1 2

1,2,3 23 Logic 1:'

6.

AUXILIARY FEEDWATER a.

Manual Initiation 2

l' 2

1,2,3

.24

[

b.

Automatic Actuation 2

-1 2

1,2,3 23 h

Logic c.

Main Stm. Gen. Water Level-Low-Low i.

. Start Motor-Driven Purps a.

Steam Generator 3/Stm. Gen.

2/Sta. Gen.

2/Ste. Gen.

1, 2, 3, 36*

Water Level ---

in any oper-in each oper-Low-Low ating Sta.

ating Sta.

g Gen.

Gen.

(Adverse) g b.

Steam Generator

.3/Sta. Gen.

2/Sta. Gen.

2/Stm. Gen.

' 1, 2, 3 36*

l Water Level --

in any oper-in each oper-a Low-Low (EAM) ating Sta.

.ating.Sts.

y Gen.

Gen.

c.

.RCS Loop AT 4 (1/ loop). 3 1,2,3 37*

~

[

'd.

Containment

-4 2

'3 1,. 2, 3 38*

3 Pressure (EAM)

~

_ _ -. _ _ =

_=

=

TABLE 3.3-3 (Continued) on l

'E.

l E5 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION i

55 l

MINIMUM l

TOTAL NO.

CHANNELS CHANNELS APPLICABLE SE FUNCTIONAL UNIT OF CHANNELS T0 TRIP OPERABLE MODES ACTION Z

ii.

Start Turbine-Driven Pump e.

a.

Steam Generator 3/Stm. Gen.

2/Stm. Gen.

2/Stc. Gen.

1,2,3 36*

Water Level --

in any 2 in each Low-Low Stm. Gen.

operating (Adverse)

Ste. Gen.

b.

Steam Generator 3/Stm. Gen.

2/Stm. Gen.

1,.2, 3 36*

Water Level --

in any 2 in each oper-Low-Low Stm. Gen.

ating Sta. Gen.

(EAM) u, 4

c.

RCS Loop AT 4/(1/ loop) 2' 3

1,2,3 37*

)$

d.

Containment 4

2 3

1,2,3 38*'

i

%?

Pressure (EAM) d.

S.I.

Start Motor-Driven Pumps and Turbir.e Driven Pump See 1 above (all S.I. initiating functions and requirements)

E it

.J c-*

I a,

=-

- ~

-~--;--- -

. ~ -

-,-~.-

TABLE 3.3-3 (Continued)

- u, -

g 8

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

r.

MINIMUM g

TOTAL NO.

CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION e

e.

Station Blackout Start Motor-Driven Pump' associated.

2/ shutdown 1/ shutdown 2/ shutdown with the shutdown board board board 1, 2, 3 20 board and Turbine Driven Pump f.

Trip of Main Feedwater Pumps m}

Start Motor-Driven Pumps and. Turbine-4 Driven Pump 1/ pump 1/ pump 1/ pump 1, 2 20*

y o

g.

Auxiliary Feedwater Suction Pressure-Low 3/ pump 2/ pump 3/ pump 1,2,3 21*-

h.

Auxiliary Feedwater Suction Transfer Time Delays 1/ pump 1/ pump 1,2,3 21*

1.

Motor-Driven Pump 1/ pump.

y 2.

Turbine-Driven Pump 2/ pump 1/ pump 2/ pump 1,2,3 21*

a E

e

~~

1 g

TABLE 3.3-3 (Continued) o ENGINEERED SAFETY FEATURE-ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE E FUNCTIONAL UNIT OF-CHANNELS TO TRIP OPERABLE MODES ACTION

-e 7.

LOSS OF POWER a.

6.9 kv Shutdown Board

--Loss of Voltage

'1.

Start Diesel 2/ shutdown 1 loss of 2/ shutdown 1,2,3,4 20*

Generators board voltage on board any shutdown board 2.

Load Shedding 2/ shutdown 1/ shutdown 2/ shutdown 1, 2, 3,'4 20*

R board board board b.

6.9 kv Shutdown Board

{

Degraded. Voltage 1.

Voltage Sensors 3/ shutdown 2/ shutdown.

2/ shutdown 1,2,3,4 20*

board board board i

2.

Diesel Generator 2/ shutdown 1/ shutdown 1/ shutdown 1,2,3,4 20*

Start and Load.

' board board board Shedding Timer 3.

SI/ Degraded 2/ shutdown 1/ shutdown' 1/shutdowa

' 1,- 2, 3, 4 -

20*

Voltage Enable board board boa:J y

. Timer-

' k 8.

ENGINEERED SAFETY FEATURE 2

ACTUATION SYSTEM INTERLOCKS a.

Pressurizer Pressure -'

3 2-

.2

.1, 2, 3 22a

=

l P-11/Not P-11 l

b.

. Deleted

~

3/ loop 2/ loop

'3/ loop 1, 2-22c c.

-Steam Generator M

Level P-14 any-loop e

l i

~

l l

..u..,._

.. _.. ~

__._.-.._am..,

m

... ~.

TABLE 3.3-3 (Continued) y, E8 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION Y.

I MINIMUM.

TOTAL NO.

CHANNELS CHANNELS APPLICABLE g

FUNCTIONAL UNIT.

Of CHANNELS TO TRIP OPERABLE MODES

- ACTION Q

9.

AUTOMATIC SWITCHOVER TO g

CONTAINMENT SUMP-a.

RWST Level - Low 4

2 3

1.2,3,4 18 i

COINCIDENT WITH Containment Sump Level - High 4

2 3

1, 2, 3, 4 18 AND Safety Injection (See 1 above for Safety Injection Requirements) b.

A't+.amatic Actuation 2

- 1 2.

1,2,3,4 15

,,o

)

Logic w

4 3

R n

Z

,o

[

C$

~

J

+mn

-+

+,--c e.-

- -, ~

- - ~

,_e-_,

-c

_ _. =_ -,

m,__..

TABLE 3.3-3 (Continued)

TABLE NOTATION--

rip function may be bypassed in this MODE below P-11 (Pressurizer Pressure T

g, Block of Safety Injection) setpoint, Trip function automatically blocked above P-11 and may be blocked below P.-11 when Safety Injection Steam Line Pressure-Low is-not blocked.

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

nThe provisions of Specification 3.0.4 are not applicable.

4 ACTION STATEMENTS ACTION 15 -

With the number of OPERABLE Channels one less than the Total Number of Channels, be in at least HOT STAN0BY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one cha.nnel may be bypassed foi up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .for surveillance 1

i testing per Specification 4 3.2,1.1 provided the other channel-I is OPERABLE.

ACTION 16 -

Deleted.

ACTION 17 -

With the number of 0PERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed y

provided the following conditions are satisfied:

a.

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

The Minimum Channels OPERABLE requiremen',s is met; h_owever, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specifica-i tion 4.3.2.1'.1.

ACTION 18 -

With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable.

channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met; one additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.1.

ACTION 19 -

With less than the Minimum Channels OPERABLE, operation may continue provided the containment ventilation isolation valves are maintained closed.

ACTION 20 -

With the number of OPERABLE Channels one less than the Total.

Number of Channels, restore the inoperable channel to OPERABLE i

status within 4B hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SEQUOYAH - UNIT 1 3/4 3-22 Amendment No. 63,141 l

TABLE 3.3-3 (Continued) t t.CTION 21 --

With less than the Minimum Number of. Channels OPERABLE, declare the associated auxiliary feedwater pump. inoperable, and comply with tht ACTION requirements of Specification 3.7.1.2.

ACTION 22 -

With less than the Minimtw Number of Channels OPERABLE, declare t

the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the appropriate ACTION statement (s) for those functions.

Functions to be evaluated are:

s.

Safety Injection Pressurizer' Pressure Steam Line Pressure Negative Steam Line Pressure Rate b.

Deleted.

c.

Turbine. Trip Steam Generator Level High-High Feedwater Isolation Steam Generator Level High-High ACTION 23 -

With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUT 00WN within tho following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed t up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.

ACTION 24 -

With the number of OPERABLE channels one less than the Total-Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTOOWN-within-the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 25 -

With the number of OPERABLE channels one less than the Total Number of Channels, restore'the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the associated valve inoperable and take the ACTION required by Specification 3.7.1.5.

ACTION 36 -

With the number of OPERABLE chai.nels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions:are <atisfied:

a.

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

b.

For the affected protection set, the Trip Time Delay for-one affected steam generator (T ) is adjusted to match'the 3

Trip Time Delay for multiple af fected steam generators (T ) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

g SEQUOYAH -~ UNIT 1

-3/4 3-23 Amendment No. 12, 63, 129, 141

i t.-

H TABLE 3.3-3 (Continued) c.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specifica-tion 4.3.2.1.1.

ACTION 37 -

With the number of OPERABLE channels one less than the Total-Number of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , for the affected protection set,-

the Trip Time Delays (T and T ) threshold power level,for zero 3

M seconds time delay is adjusted to 0% RTP.

ACTION 38 -

With the number of OPERABLE channels one less than the' Total Number of Channels, STARTUP and/or POWER OPERATION maj. proceed' provided that within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , for the affected protection set, the Steam Generator Water Level - Low-Low (EAM) channels; trip:

setpoint is adjusted to the same value as Steam Generator Water-Level - Low-Low (Adverse),

i I

?

L i

]

SEQUOYAH - UNIT 1 3/4 3-23a Amendment No; 141 1

m' TABLE 3.3-4 E8 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS s

x FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E'

Q l.-

SAFETY _ INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION g

a.

Manual Initiation Not Applicable Not Applicable b.

Automatic Actuation Logic Not Applicable Not Applicable c.

Containment Pressure--High 1 1.54 psig 5 1.6 psig d.

Pressurizcc Pressure--Low

> 1870 psig 1 1864.8 psig w}

e.

Deleted

[

f.

Steam'Line Pressure-Low 1 600 psig steam line

> 592.2 psig steam line pressure (Note 1)'

pressure (Note 1)

'l 1

i n

,)

l l

i.~._

c.. _,

-. =.

.c g-TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS l

l tUNCTIONAL UNIT TRIP'SETPOINT ALLOWABLE VALUES l

E-Z 2.

CONTAINMENT SPRAY w

a.

Manual Initiation Not Applicable Not Applicab'le b.

Aut.omatic Actuation Logic Not Applicable Not' Applicable c.

Contaiment Pressure--High-High 5 2.81 psig i 2.9 psig 3.

CONTAINMENT ISOL?. TION I

w a.

Phase "A" Isolation 1

1.

Manual Not Applicable Not Applicable w.

lo Not Applicable Not Applicable 2.

From Safety Injection Automatic Actuation logic b.

Phase "B" Isolation t

1.

Manual Not Applicable Not Applicable 2.

Automatic Actuation Logic Not. Applicable Not Applicable 3.

Containment Pressure--High-High 1 2.81 psig i 2.9 psig 2>

2 c.

Containment Ventilation Isolation E

2 1.

' Manual Not Applicable Not Applicable 5

2.

Automatic Isolation. Logic Not Applicable Not Appl.icable 2

P R

M

=

- < - ~ -

~

v-x-

- -- - -~

e--

,~- -- + ~

w-

- - - < ~

r-,-

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

.g

';l!

I FUNCTIONAL: UNIT TRIP SETPOINT ALLOWABLE VALUES

-3

-3 g

3.

Containment Gas Monitor 1 8.5 x 10 Ci/cc 1 8.5 x 10 Ci/cc p

Radioactivity-High

-3

-3 4.

Containment Purge Air Exhaust i 8.5 x 10 Ci/cc i 8.5 x 10 Ci/cc Monitor Radioactivity-High

-5

-5 5.

Containment Particulate i 1.5 x 10 Ci/cc 1 1.5 x 10 Ci/cc-Activity-High 4.

STEAM LINE ISOLATION a.

Manual Not Applicable Not Applicable wS b.

Automatic Actuation Logic Not Applicable Not Applicable Containment Pressure--High-High.

5 2.81 psig i 2.9 psig c.

d.

Steam Line Pressure--Low

> 600 psig steam

> 592.2 psig steam Tine pressure (Note 1)

' Tir.e pressure (Note 1) e.

Negative Steam Line Pressure Rate--High

-1 100.0 psi (Note 2)

$ 107.8 psi (Note 2) 5.

TURBINE TRIP-AND FEEDWATER ISOLATION Steam Generator' Water-level -

< 81.0% of narrow range

< 81.7% of narrow range

~

Instrument span each steam Instrument span each steam a.

2 High-High generator generator 3

N.A.

. co

. Automatic _ Actuation Logic

.a b.

i TABLE 3.~3-4 (Continued) nE8 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS j

s x

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES EQ 6.

AUXILIARY FEEDWATER a.

Manual Not Applicable Not Applicable s-b.

Automatic Actuation Logic Not Applicable Not Applicable c.

Main Steam Generator Water Level-low-low i.

RCS Loop AT Equivalent to RCS Loop AT variable RCS Loop AT variable Power 1 50% RTP input 5 50% RTP input i trip setpoint +2.5% RTP wD Coincident with 4

Steam Generator Water Level --

2 19.0% of narrow range 1 18.4% of narrow range w

Low-Low (Adverse) instrument. span instrument span and Containment Pressure - EAM i 0.5 psig 1 0.6 psig or Steam Generator Water Level --

2 13 0% of narrow range 2 12.4% of narrow range _

Low-Low (EAM) instrument span instrument span E

with A time' delay.(T ) if one

.5 T '(Note 5, Table 2.2-1) 5 (1.01) Ts (N te 5, Table 2.'2-1) 3 3

k Steam Generator is affected

[

or U$

A. time delay (T ) if two or i T,-(Note 5, Table 2.2-1) 5 (1.01) T,(Note 5, Table 2.2-1) more' Steam Gene,rators are affected

,m._

.., _,..._.. __.._..__.:=_;._.

.-.;-~.-..-,,...-.........

,,n.____..,

s wi"are-i.

TABLE 3.3-4 (Continued)

~

jrn ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 8

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 3

ii.

RCS Loop AT Equivalent to a

Power > 50% RTP

~

Coincident with Steam Generator-Water 1 19.0% of narrow range 1 18.4% of narrow range Level -- Low-Low (Adverse) instrument span instrument span and l

Containment Pressure (EAM) 5 0.5 psig i.0.6 psig or Steam _' Generator Water 3 13.0% of narrow range 2 12.4% of narrow range level -- Low-Low (EAM) instrument span instrument span d.

S.I.

See 1 above (all SI Setpoints)-

ru e.

Station Blackout-0 volts with a 5.0 second 0 velts with a 5.0 i 1.0 second 2

time. delay time delay f.

Trip'of Main Feedwater N.A.

N.A.

Pumps-g.

Auxiliary Feedwater Suction' 1 2 psig (motor-driven pump) 1 1 psig.(motor driven pump)

Pressure-Low 2 13.9 psig-(turbine driven 2 12 psig (turbine driven

,g pump) pump) 2 h.

Auxiliary Feedwater Suction.

14 seconds (motor driven

.4 seconds i 0.4 seconds kg Transfer Time Delays pump)~

(motor' driven pump) 5.5 seconds-(turbine driven 5.5 seconds 0.55 seconds 5

.. pump)

(turbine driven pump)

N e

. - = -

u-r m*'-

r

--~-

e.

~

g TABLE 3.'3-4 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS s

1 FUNCTIONAL UNIT

-TRIP SETPOINT ALLOWABLE VALUES' EZ 7.

LOSS OF-POWER a.

6.9 kv Shutdown Board Undervoltage j

Loss of Voltage 1.

Start of Diesel Generators a.

Nominal Voltage Setpoint 4860 volts 4860 volts +97.2 v'oits

'b.

Relay Response Time for 0 volts with a 1.5 second 0 volts with a 1.5 10.5 Loss of Voltage time delay second time delay 2.

Load Shedding a.

Nominal Voltage Setpoint 4860 volts 4860 vol'., - 9/.2 vo't=

w i

b.

Relay Response Tirae for 0 volts with a 5.0 second 0 volts with a 5.0 +1.0 Loss of Voltage time' delay-second time delay ~

o h

b.

6.9 kv Shutdown Board-Degraded Voltage 1.

Voltage Sensors.

6560 volts 6560 volts 1 33 volts 2.

Diesel Generator Start and Load Shed Timer 300 seconds 300 seconds 1 30 seconds 3.

SI/ Degraded Voltage Logic Enable Timer 10 seconds 10 seconds 1 0.5 seconds R

8.

ENGINEERED SAFETY FEATURE' ACTUATION SYSTEM INTERLOCKS-

{

a.

Pressurizer Pressure.

e g

1.

Not-P-11,-' Automatic Unblock of Safety. Injection on Increasing Pressure 1 1970 psig-

..$ 1975.2 psig C

2..P-11, Enable Manual

-Block of Safety Injection on Decreasing Pressure

>.1962 psig

> 1956.8 psig I-B

- ~

TABLE 3.3-4 (Continued) l.

m m

h ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS t

i '~

35 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E

Q 8.

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS (Continued) g b.

Deleted c.

Deleted d.

Steam Generator Level Turbine Trip, Feedwater Isolation P-14 (See 5. above) 9.

AUTOMATIC SWITCHOVER To CONTAINMENT SUMP w

ro' m

a.

RWST Level - Low 130" from tank base 130" i 2.71" from tank base COINCIDENT'WITH Containment Sump Level - High 30" above elev. 680' 30" i 1.68" above elev. 680' AND Safety Injee. tion (See 1 above for-all Safety Injection Setpoints/A710wable Valves) b.

Automatic Actuation Logic N.A.

N.A.

, Note 1:

Time constants utilized in the lead-lag controller for Steam Pressure - Low are 11 1 50 seconds and 3

1 I 5 seconds.

2 E

Time constant utilized in the. rate-lag controller for Negative Steam Line Pressure Rate - High is g

Note 2:

g 11 50 seconds.

E

_mm.

_r_

. ~.

s g

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECON05 3.

Pressurizer Pressure-Low a.

Safety Injection (ECCS) 1 32.0(1)/28,0(7) b.

Reactor Trip (from SI) 13.0 c.

Feedwater Isolation

< 8.0(2) d.

Containment Isolation-Phase "A"(3) 18.0(8) e.

Containment Ventilation Isolation-5.5(8)(13) f.

Auxiliary Feedwater Pumps

< 60(11) g.

Essential Raw Cooling Water System 65.0(8)/75.0(9) h.

Emergency Gas Treatment System 1 28.0(8) 4.

Deleted 5.

Negative Steam Line Pressure Rate - High a.

Steam Line Isolation s 8.0

.SEQUOYAH - UNIT 1 3/4 3-30 Amendment No. 55, 77, 106, 141 1

?

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES-INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS 6.

Steam Line Pressure-Low a.

Safety Injection (ECCS) 1 28.0(7)/28.0(1) b.

Reactor Trip (from SI) 1 3.0 c.

Feedwater Isolation

< 8.0(2) d.

Containment Isolation-Phase "A"(3)-

h18.0(8)/28.0(9) e.

Containment Ventilation' Isolation Not Applicable f.

Auxiliary Feedwater Pumps

< 60(11) g.

Essential Raw Cooling Water System 65.0(8)/75.0(9) h.

Steam Line Isolation

< 8.0 i.

Emergency Gas Treatment System 38.0(9) 7.

Containment Pressure--High-High-a.

Containment Spray

< 208(9) b.

Containment Isolation-Phase "B"(12) 65(9)/75(9) c.

Steam Line Isolation 1 7,0 i

d.

Containment Air Return Fan

> 540.0 and 1 660-8.

Steam Generator Water Level--High-High a.

Turbine Trip 1 2.5 b.

Feedwater Isolation 1 11.0(2) 9.

Main Steam Generator Water Level -

Low-Low 60.0(14) a.

Motor-driven Auxiliary 1

Feedwater Pumps (4) b.

Turbine-driven Auxiliary 1 60.0(14)

Feedwater Pumps (5)(11)

SEQUOYAH - UNIT 1 3/4 3-31 Amendment.No. 55, 59, 63, 77,-82, 141

TABLE 3.3-5 (Continued) 3 l

ENGINEERED SAFETY FEATURES RESPONSE TIMES y

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS-10.

Station Blackout a.

-Auxiliary Feedwater Pumps 1 60(11) i 11.

Trip of Main Feedwater Pumps a.

Auxiliary Feedwater Pumps 1 60(11) l 12.

Loss of Power a.

6.9 kv Shutdown Board - Degraded i 10(10)

Voltage or loss of Voltage 13.

RWST Level-Low Coincident with Containment Sump Level-High and Safety injecttnn a.

Automatic Switchover to Containment Sump i 250 14.

Containment Purge Air Exhaust i

Radioactivity - High a.

Containment Ventilation Isolation s 10(0) 15.

Containment Gas Monitor Radioactivity High a.

Containment Ventilation Isolation i 10(6) 16.

Containment Particulate Activity High a.

Containment Ventilation Isolation i 10(6)

SEQUOYAH - UNIT 1 3/4 3-32 Amendment No. 29, 77

^

.1 TABLE 3.3-5 (Continued)

TABLE NOTATION (1) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps, SI and RHR. pumps.

(2) Using air operated valve (3) The following valves are exceptions to the response times shown in the table and will have the values listed in seccnds for the initiating sig-nals and function indicated:

Valves:

FCV-26-240. -243 3.d. 22(0)/ 31(9) 2.d. 21(8)

Response times:'

4.d.-21(0)/ 31(9) 1

/3 Valves:

FCV-61-96, -97, -110, -122, -191, -192, -193, -194 Response times:

2.d. 31(0) 8)

3.d. 32(0) 4.d. 31(0) 5.d. 34(

6.d. 31(0)

Valve:

FCV-70-143 Response times:

2.d. 61(0)/71(9) 3.d 62(8) 4.d. 61(0)j 8)

(9)-

5.d. 64(

(9) 6.d. 61(8)j/

(9)

(4) On 2/3 any Steam Generator (5) On 2/3 in 2/4 Steam Generator (6)

Radiation detectors for Containment Ventilation Isolation'may be excluded from Response Time Testing.

(7) Diesel generator starting and sequence loading delays not included.

Offsite power available.

Response time limit includes opening and closing of valves to establish SI path and attainment of discharge pressure for centrifugal i

charging pumps.

(8) Diesel generator starting and sequence loading delays not included.-

Response

time limit includes operating time of valves.

(9) Diesel Generator starting and sequence loading delays included.

Response

. time limit includes operating time of valves.

l SEQUOYAH - UNIT 1 3/4 3-33 Amendment No. 17, 55-I.

TABLE 3.3-5 (Continued)

TABLE NOTATION-(10) The response time for loss of voltage is measured from the time voltage is lost until the time full voltage is restored by the diesel.

The-response time for degraded voltage is measured from the time the load shedding signal is generated, either from the degraded voltage or the SI enable timer, to the time full voltage is restored by the diesel.

The response time of the timers is covered by the requirements on their setpoints.

(11) The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 for the turbine-driven Auxiliary Feedwater Pump.

(12) The following valves are exceptions to the response times shown in the Table and will have the values listed in seconds for the initiating-signals and the function indicated:

Valves:

FCV-67-89,-90,-10gg)/85pg)

-10 Response times:

7.b, 75 Valve:

FCV-70-141 Response times:

7.b, 70(8)/80(9)

(13) Containment purge valves only.

Containment radiation monitor valves have a response time of 6.5 seconds or less.

(14) Does not include Trip Time Delays.

Response times noted include the transmitters, Eagle-21 process protection cabinets, solid state protection cabinets, and actuation devices (up to and including pumps).

This reflects the response times necessary for THERMAL POWER in excess of 50% RTP.

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l SEQUOYAH - UNIT 1 3/4 3-33a Amendment No. 29, 77, 82, 106, 141

3/4.3 INSTRUMENTATION BASES

\\

3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and interlocks ensure that 1) the associated ESF 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) sufficient system functional capability is available for protective and ESF purposes from diverse parameters, t

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 systemt is consistent with the assumptions used in the accident analyses.

The Engineered Safet.y Features 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.

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

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

l 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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SEQUOYAH - UNIT 1 B 3/4 3-1 Amendment No. 141 l

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TABLE 4.3_2 v.

x>

E ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INST'20NENTATION 5

SURVEILLANCE REQUIREMENTS x

e E

CHANNEL M00ES IN WHICH U

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE e

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 1.

SAFETY INJECTION AND FEEDWATER ISOLATION a.

Manual Initiation M.A.

N.A.

R 1, 2, 3, 4 b.

Automatic Actuation Logic n.A.

N.A.

M(1) i, 2. 3, 4 w

c.

Contairment Piessure-High 5

R Q

1,2,3 k

d.

Pressurizer Pressure--Low 5

R Q

1, 2, 3 w

w e.

Deleted f.

Steam Line Pressure--Low 5

R Q

1, 2, 3 2.

CONTAINMENT SPRAY a.

Manual Initiation N.A.

N.A.

R 1, 2, 3, 4 b.

Automatic Actuation Lcgic N..^..

M.A.

M(1) 1, 2, 3, 4 c.

Containment Pressure--High*High S R

Q 1, 2, 3 y

5

- 8 2a

.E w

w

.-y.

._.s.

..v

,., +.

w -

TABLE 4.3-2 (Continued)

S8 ENGINEERED SAFETY FEAIUAE ACTUATION SYSTEM INSTRUMENTATION SURVEIlf.ANCE REQUIREENTS g

9 CHANNEL MODES IN WHICH CMNNEL CHANNEL FUNCTIONAL SURVEILLANCE Q

FUNCTIONAL UNIT CHECK C.ALIBRATION TEST REQUIRED g

3.

CONTAINMENT ISOLATION a.

Phase "A" Isolation

1) Manual N.A.

N.A.

R 1, 2, 3, 4

2) From Safety Injection M.A.

N.A.

M(1) 1, 2, 3, 4 Automatic Actuation Logic U

b.

Phase "B" Isolation J,

1) Manual N.A.

N.A.

R 1, 2, 3, 4 w

w

2) Automatic Actuation N.A.

N.A.

M(1) 1, 2, 3, 4 Logic

3) Containment Pressure--

S R

Q 1, 2, 3 High-High Containment Ventilation Isolation c.

1) Mar:ua1 M.A.

N.A.

R 1, 2, 3, 4 g

2) Automatic Isolation Logic.

M.A.

M(1) 1, 2, 3, 4 h

3) Containment Gas Monitor S

R M

1,2,3,4 Radioactivity-High g

e

TABLE 4.3-2 (Continued) u, E8 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION g

SURVEILLANCE REQUIREMENTS CHANNEL M00'c5 IN WHICH E

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

i FUNCTIONAL tJNIT CHECK CALIBRATION TEST REQUIRED

4) Containment Purge Air 5

R M

1,2,3,4 Exhaust Monitor Radio-activity-High

5) Containment Particulate S

R M

1, 2, 3, 4 Activity-High i

4.

STEAM LINE ISOLATION a.

Manual N.A.

N.A.

R 1, 2, 3 b.

Automatic Actuation Logic N.A.

N.A.

M(1) 1, 2, 3

'Ig c.

Containment Pressure--

S R

Q 1,2,3 High-High d.

Steam Line Pressure--Low S

R Q

1, 2, 3

.e.

Negative Steam Line 5

P Q

3 Pressure Rate--High S.

TURBINE TRIP AND FEEDWATER y

ISOLATION g

a.

Steam Generator Water S

R Q

1, 2, 3 3

Level--High-High 5

b.

Automatic Actuation Logic N.A.

N.A.

M(1) 1, 2, 3 z

O 6.

AUXILIARY FEEDWATER 7

a.

Manual N.A.

N.A.

R 1,2,3

+

b.

Automatic Actuation Logic N.A.

N.A.

M(1) 1, 2, 3

u e.*

r v~r-,,

..---y

-3

-..m

-4.-

-(

'{.g

~

.4

/

g; TABLE 4.3-2 (Continued) 8 ENGINEERED SAFET'r FEATURE ACTUATION SYSTEM INSTRUMENTATION

~

j{

SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH E

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE El FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

a c.

Main Steam Generator Water Level-Low-Low 1.

Steam Generator Water 5

R Q

1, 2, 3 Level -- Low-Low (Adverse) 2.

Steam Generator Water S

R Q

1,2,3 Level -- Low-Low R

(EAM) 3.

RCS Loop aT S

R Q

1,2,3 ti 4.

Containcent Pressure 5

R Q

1, 2, 3 (EAM) d.

S.I.

See 1 above (all SI surveillance requirements) e.

Station Blackout N.A.

R M.A.

1, 2, 3 f.

Trip of Main Feedwater M.A.

N.A.

R 1, 2.

Pumps y

g.

Auxiliary Feedwater Suction M.A.

R M

1, 2, 3 3

Pressure-Low h

h.

Auxiliary Feedwater Suction M.A.

R N.A.

1, 2, 3 Transfer Time Delays 7.

LOSS OF POWER g

a+

z O

a.

6.9 kv Shutdown Board -

O!

Loss of. Voltage 1.

Start Diesel Generators S

R M

1, 2, 3, 4 O!

2.

Load Shedding 5

R N.A.

1, 2, 3, 4

TABLE 4.3-2 (Continued) w5 E

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUNENTATION h

SURVEILLANCE REQUIRENENTS CHANNEL NODES IN WHICH E

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

] FUNCTIONAL UNIT

_ TEST REQUIRED CHECK CALIBRATION l

b.

6.9 kw Shutdown Board -

s Degraded Voltage 1.

Voltage sensors S

R M

1, 2, 3, 4 2.

Diesel Generators N.A.

R N.A.

1, 2, 3, 4 Start and Load Shedding Timer 3.

SI/ Degraded Voltage N.A.

R N.A.

1, 2, 3, 4 w1 Logic Timer 8.

ENGINEERED SAFETY FEATURE 7

' ACTUATION SYSTEM INTERLOCKS a.

Pressurizer Pressure, N.A.

R(2)

N.A.

1,2,3 P-11/Not P-11 b.

Deleted F

c.

Steam Generator N.A.

R(2)

N.A.

1, 2 Level P-14 9.

AUTOMATIC SWITCHOVER TO CONTAIMPCNT SUMP 5

~

a.

RSWT Level - Low-S R

Q 1, 2, 3, 4 O

COINCIDENT WITH Containment Sump Level - High 5

R Q.

1, 2, 3, 4

J AND (See 1 above for all Safety Injection Surveillance Requirements)

Safety Injection U

b.

Automatic Actuation Logic N.A.

N.A.

'M(1)

I, 2, 3, 4

o-TABLE 4.3-2 (Continued)

TAB (E NOTATION (1) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BAS!$.

(2) The total interlock function shall be demonstrated OPERABLE during CHANNEL CALIBRATION testing of each channel af fected by interlock operation.

)

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SEQUOYAH - UNIT 1 3/4 3-38 Amendment No. 47 1

ML20043H903

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