Text

Proposed Tech Specs 3/4.3.1 & 3/4.3.2 Re Quarterly Testing of ESFs
	ML20099E070
Person / Time
Site:	Byron, Braidwood
Issue date:	08/05/1992
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20099E054	List: ML20099E050; ML20099E070;
References
	NUDOCS 9208100010
	Download: ML20099E070 (62)
	v • d • e

l ATTACHMENT 2 Proposed changes te Technical Specifications of Facility Operating Licenses NPF 37, NPF 66, NPF 72 and NPF 77.

Revise' ' ages: B 2-8 3/4 3-15 B 2-9 3/4 3-17 3/4 3-2 3/4 3-19 3/4 3-3 3/4 3-20 3/4 3-5 3/4 3-21 3/4 3-6 3/4 3-22 3/4 3-9 3/4 3-34 3/4 3-10 3/4 3-35 3/4 3-11 3/4336 3/4 3-12 3/4 3-37 3/4 3-12a 3/4 3-38 3/4 3-13(Bw) B 3/4 3-1 NOTE: The attached markups reflect the incorporation of the ,

v pending amendment requests for Generic Letter 87-09 and the revision to the BDPS requirements. The following pages are affected:

G_eneric Letter 8LQ9 BQPJ 3/4 3-2 3/4 3-2 3/4 3-3 3/4 3-5 3/4 3 5 3/4 3-9 3/4312 3/4 3-12 3/4 3-12a 3/4 3-12a 3/4 3-15 -

3/4 3-17 3/4 3-19 3/4 3-21 9208100010 92080b PDR ADOCK 05000454 P PDR ZNLD/615/120

4 (lHITING StJETY SYSTEM SETTlH35 l

BASES Turbine Trtr M l A Turbine trip initiates a Reactor trip. On decreasing power the Turbine trip is automatically blocked by N-ee P-8 (a power level of opproximately de%-

4F7) ce-30% (P-8) of RATED THERMAL POWER with e turbine impulse chamber pres-i sure at approximately 40% (F7) cr 30% (P-8) of full power equivalent); and on increasing power, reinstated automatically by-p4e P-8.

Safety injection input from ESF If a Reactor trip has not already been generated by the Reactor 1 rip System instrumentation, the ESF automatic actuation logic channels will I initiate a Reactor trip upon any signal which initiates a safety injection.

The ESF instrumentation channels which bitiate a Safety injection signal are shown in Table 3.3 3. ,

Reactor Coolant Pump Breaker Position Trip The Reactor Coolant Pump Breaker Position trips are anticipatory trips which provide core protection against DNB. The Open/Close Position trips ,

assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. No credit was taken in the accident analyses for operation of these trips. Their u nctional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Trip System. Above P 7 (a power leve' of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber , ' aure at approximately 10% of full power equivalent) an

..--- .._._ _ . automatic Reactor trip will occur if more than one reactor coolant pump breaker is opened. Below P-7 the trip function is automatically blocked.

n

- 4A-*eatter-tritrcrr-Turtrine-trip-it enstrit+3boveH-fdO%) entil-the-1nedifi -

estion-41Hmp-lemented,whictr enab1w-Reactor-trip otr-Turbine-trlTr above+P

-t3Mt BYRON - UNITS 1 & 2 B 2-8 AmendmentNo.%

. . - . , . . . . - . . - , _ _ . . . . . . . . . _ . . . . . . . . . . . - - . . - . _ - . .m._.. . __. ._ _.

llMITING SAFETY SYSTEM SETTINGS i

BASES Reactor Trip $ystem Interlocks The Reactor Trip System Interlocks perform the following functions:

P-6 On increasing Reactor (power, trip i.e., P-6 premcture prevents allows the block manual block ofRange of Source the Source trip), Range '

provides an automatic backup block for Source Range Neutron Flux doubling, and the manual block that de-energizes the high voltage to the Source Range detectors. On decreasing power, Source Range Level trips and Neutron Flux doubling circuits are automatically reacti-vated and high voltage restored.

PM/Onincreasingpower,p*7automaticallyenablesReactortripsonlow l flow in more than one reactor coolant loop, more than one reactor

, coolant pump breaker open, reactor coolant pump bus undervoltage and underfrequency, hbiwt% pressurizer low pressure and pressurizer high level. On decreasing power, the above listed trips are auto-matica11y blocked.

P-BE0n incrosing power, P-8 automatir. ally enables Reactor trips on low flow in one or more reactor coolant loops and Turbine trip. 'On de- O creasing power, the P-B automatically blocks the single loop low flow o trip and Turbine trip.

P-10 On increasing power, P-10 allows the manual block of the Intermediate Te ge Reactor trip and the Low Setpoint Power Range Reactor trip; and automatics 11y blocks the Source Range Reactor trip and provides an

~

automatic backup function to de-energize the Source Range high voltage poweri- 9n-decreasing power, the Intermediate Range Reactor trip and the Low S(tpoint Power Range Reactor trip are automatically reactivated and Source Range high voltage to the detectors is restored if power decreases below the P-6 setpoint. Provides input to P-7.

P-13 Provides input to P-7. -

IXher-trip-on-1vrbine trip -i3 enaMed-ebove-P-7-(iO%) untii-the-tnodifi-

-cetien is implementeenhieh-eneMes-Reactor trip- en ivrMne-trip ebeve P 0-

-60%h-BYRON - UNITS 1 & 2 B 2-9 AmendmentNo.K

TABLE 3.3-1 REACTORTRIPSYSTUMINSTRUMENTATION l

MINIMUM TOTAL NO. CHANNELS CHANNcLS APPLICABLE c OP RABLE MODES ACTION

0F CHANNELS TO TRIP FUNCTIONAL UNIT' w 1 Manual Reactor Trip 2 1 2 1, 2 w 1. 2 3*, 4*, 5* 10 2' 1 e

m

2. Power Range, Neutron Flux

a. High Setpoint 4 2 3 1, 2 28

b. Low Setpoint 4 2 3 I##f, 2 28

3. Power Range, Neutron Flux 4 2 3 1, 2 2[

High Positive Rate

.R 4. Power Range, Neutron' Flux, 4 2 3 1, 2 [

High Negative Rate 1 2 1###, 2 3 I

5. Intermedia'te Range, Neutron Flux 2

6. Source Range, Neutron Flux

a. Startup 2 1 2. 2##F 4 1 2 3,4,5 5

b. Shutdown 2

~

7. Overtemperature AT 4 2 3 1, 2 6f

8. Overpower AT 4 2 3 . 1, 2 6[

^

. Pressurizer Pressure-Low (Above P-7) 4 2 3 1 6b nn gyv

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_ TABLE 3.3-1 (Continued) )

TABLE NOTATIONS

With the Reactor Trip System breakers in the closed position and the ;

Control Rod Drive System capable of rod withdrawal. i

-hibe-borca etartun dhtien fhx d:eling ign:h :y be bleded-dw4ng reactor !

aAdUseMannel s-also-prov4 de-inpu ts -to-E S FA SrThe-Aet4 en-Sta temen tr-fee-the--

--ch:nnch in-T bh 3.3-3 h more-eenservat4ve-and;-therefore, controliing. :

-#The-peev4+40n: ef Specificatter 3.0.' :re not applicab h. ,

-. ##Behw the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1. 1. Below the P-10 (Low Setpoint Power h'inge Neutron Flux Interlock) Setpoint. !

@Whenever the Reactor Trip Bypass Breakers are rackad in and closed for bypass-Ing a Reactor Trip Breaker. 7

/

ACTION STATEMENTS ACTION 1 - 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 be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - With the number of OPERABLE channels one less than the Total 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. 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 Specificat. ion

. 4.3.1.1; and

c. Either, THERMAL POWER is restricted to less than or equal 3 to 75% of RATED THERMAL POWER and the; Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to t 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the

-QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.2.

ACTION 3 - With the number of channels OPERABLE o'ne less than the Minimum-Channels. OPERABLE requirement and with the THERMAL POWER level: '

a._. Below the_P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint, rostore the inoperable channel to OPERABLE status-prior-to' increasing THERMAL POWER above the P-6 Setpoint; and

~

b. -Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to tnereasing THERMAL POWER above 10% of RATED THERMAL POWER.

3 . BYRON - UNITS _1 & 2 3/4-3-S Amendment- No. )M(

_,, - a.u .

. ----_u . _ _ ___ __ _ _..a

r j

TABLE 3.3-1 (Continued) .

ACTION STATEMEN15 (Continued)

ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving i positive reactivity changes.

ACTION With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement restore the inoperable channel to OPEP,ABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or within the next hour open the reactor trip breakers, suspend all operations involving positive reactivity changes, and verify valves CV-111B, CV-8428 CV-8441 and CV-8435 are closed and secured in position., WithCV-8439, no channels OPERABl.E verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable. '

and take the actions stated above within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and verify _

compliance at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed 1 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 ; and-

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.

ACTION 7 - Deleted

-ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within I hour determine by-observation-of the associated permissive annunciator window (s) that the interlock is in its required state ,

for the existing plant condition, or a Ndort he nasta% chune to OPERA 6LE stdvs W%pply Specification 3.0.3.

L, hors,or_

ACTION-9_._ - With_the number of OPERABLE hannels one less than the Minimum Channels OPERABLE requirement be in at least. HOT STANDBY j/hoursforsurveillancetestingperSpecification4.3.11withinl6 V .,

hours; provided the other channel is OPERABLE.

ACTION 10 - 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 l ;

, breakers within the next hour.

ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped ~ condition within / hour 5.

6 ACTION 12 - a. 'With one of the diverse trip features (Undervoltage or Shunt' Trip ~ attachment) inoperable, restore it to OPERABLE /7

. BYRON - UNITS 1 & 2 3/4 3 AmendmentNo.p :

~o -!- TABLE 4.3-1 1 '

j '

REACTOR TRIP SYSTEM;INSTRUNENTATION SURVEILLANCE REQUIREMENTS !

TRIP E.

. ANALOG ACTUATING MODES FOR j Z CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE l 7 0NCTIONAL UNIT ' CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED ,

, w i

~

"1. Manual Reactor Trip N.A. N.A.- N.A. .R(14) N.A. 1, 2,-3*, 4*, 5* ,!

2.. Power Range, Neutron Flux

, a. High Setpoint- S- D(2, 4), Q- N.A. N.A. 1, 2 l M(3,4),. -

i Q(4, 6) [,

4 R(4

b. Low Setpoint S R(4) ,a Q N.A. M.A. 1###, 2 :

R'

^3. Power Range, Neutron Flux, N.A. . R(4)g -

Q N.A. N.A. 1, 2 .

-Y High Positive Rate -

i-

,' 4. Power Range, Neutron Flux. N.A. ~R(4)p Q N A. M.A. 1, 2

[

High Negative Rate i

5. -Intermediate Range, S R(4,Sa Q N.A. N'. A. 1###, 2 Neutron Flux j Source Range, Neutron Flux

' 6. S - R(4, Sb Q(9) N.A. N.A. 2##, 3, 4, 5

7. Overtemperature AT S R(13 Q N.A. N.A. 1, 2

- k8. Overpower AT S' Q W.A. N.A. 1, 2

{ 9. Pressurizer Pressure-Low S QE N.A. N.A. 1

5 (Above P-7)

.E10. Pressurizer Pressure-High 5 R -Q N.A. M.A. 1, 2 I

11. Pressurizer Water Level-High 5 'Q N.A. N.A. 1 l (Above P-7) i i

, t L_ . _ _ . _ _- .. ___ _ _

^

~ a

r. s. -

4 ,

t r 1

i I

TABLE 4.3-1 (Continaed)

E ,

y REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REQUIREMENTS .

TRIP [

,E ANALOG ACTUATING MDDES FOR C

CHANNEL DEVICE WHICH CHANNEL. CHANNEL OPERATIONAL ^ OPERATIONAL ACTUATION SURVEILLANCE [

" FUNCTIONAL' UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED !

e- :

l " 12. Reactor Coolani. Flow-Low 5 R Q N.A. N.A. I f 1

, 4 ,

13. Steam Generator Water Level- S R* C*r- N.A. N. A. 1, 2 i i

. Low-Low . t t

14 Undervoltage-Reactor Coolant Pumps (Above P-7)

N.A. R N.A. 9) N.A. 1 T ,

. t R 15. Underfrequency-Reactor M.A. R M. A. Q(10) N.A. I 4

Coolant Pumps (Above P-7) f '

Y j o 16. Turbine Trip (Above ? 7 e'P-8)O l 1

, a. Emergency Trip Header M.A. R N.A. S/U(1,10) N.A. I r l Pressure -

l b .' Turbine Throttle Valve N.A. R N.A. S/U(1,10) N.A. I t Closure !

I i l 17. Safety Injection Input from .N.A. M.A. N.A. R N.A. 1, 2 i l ESF l

18. Reacto'r Coolant Pump Breaker N.A. M.A. N.A. R N.A. 1

E Position Trip (Above P-7) I I E f E 19. Reactor Trip System Interlocks -

E

a. Intermediate Range ,

}

1 Neutron Ficx, P-6 N.A. R(4)h -Q- R N.A. M.A. 2ff I

. b .. Low Power Reactor !

Trips Block, P-7 N.A. R(4) -

-Q-(S) R. M.A. M.A. I f

c. Power Range Neutron & i

! Flux, P-8 :M.A. R(4)U MR N.A. M.A. 1 j

TABLE 4.3-1 (Continued) y ,

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS f

TRIP E ANALOG ACTUATING MODES FOR O CHANNEL DEVICE

" WHICH CHANNEL. CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE

[FbHCTIONALUNIT CHECK CALIBRATION TEST TEST LOGIC TEST 15 REQUIRED "19. Reactor Trip System Interlocks (Continued)

~ d. Low Setpoint Power Range Neutron Flux, F-10 N.A. R(4 Q (S) R N.A. N.A. 1, 2

e. Turbine Impulse Chamber Pressure, P-13 N.A. W Q (0) R. .N.A. N.A. 1

$0. Reactor Trip Breaker N.A. N.A. N.A. M (11) N.A. 1, 2, 3*,'4*, 5* h

[21. Automatic Trip and Interlock N. A. N.A. N.A. N.A. M (7) 1, 2, 3* , 4 * , 5 *

- Logic i

22. Reactor Trip Bypass Breakers n N.A. N. A. N.A. (15),R(16) N. A. 1, 2, 3*, 4*, 5*J;f 4

1 e

.O

1hc in 4"d ungk Pdnt h pm W ci mccre. 4o oco re'. AM L flu y. hg F FG CC NrE Gilm.n R ATE bkn. reb hn . P ouWER,00thcMw hu c. shn.ll be. perGrme d precc t.o exceelim8 7 s % of I HERMAL W i

_ TABLE 4.3 1 (Continued) I TABLE NOTATIONS

With the Reactor Trip System breakers closed and the control Rod Drive System capable of rod withdrawal.

"These-channels-also provide-inputs-to-ESFASr-The-Operational Test 4requency-

-foothese-channelt-in-Table 4r3-2-is-more-<onservative-andr-thereforer --

1

-contro)4 ingc-1*AJeactor-tr4p-on Turtiae trip 4sanabled-aboveJ27-010%)-unt44-the-modif4ce- t 4 lea 4 !aplemented-which-enablesJeactor4r4p-on-Turbine-trip-ebove-P-8-(404 L Mi he-speef fi ed-10-mon t h-i nt erv a l - m ay4e-ex t e nded - to-32-mon t hs-for-Cyc4 e--l-

-oMyr

1. Below P-6 (Intermedia *e Range Neutron Flux Interlock) Setpoint.

)

1. 1. Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint. '

(1) Ifnotperformedinprevious(days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of De_c_ - ,

f. cation 4.0.4 are not applicable for entry into H0DE 2 or 1. 6. be terb 2 '

(3)4/inglepoin'.,comparisonofincoretoexcoreAXIALFLUXO!FFERENCE}above i 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are

._.. not applicable for entry into MODE 2 or lg l (4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(Sa) Initial plateau curves shall be measured for each detector. Subsequent plateau curves shall be obtained, evaluated and compared to the initial curves. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(5b) With the high voltage setting varied as recommended by th's manufacturer, an initial discriminator bias curve shall be measured for each detector.

Subsequent discriminator bias curves shall be obtained, evaluated and compared to the initial curves.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provi-I sions of Specification 4.0.4 are not applicable (7) Not vied. train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

Each for entry into MODE 2 (8) AMi t h-powe r-grea te r-tha n-oe-eque l-to- the-i n t e rl o c k-Se t po i n t-t he-re qu i re d -

ANAt004HANNEL-OPERATIONAt-TEST shall-consist-of verifying-that--the-inter-4 ec k-i s-in-t h e-required-s t a t e4y-obs erving-the9e rmi s s i ve-e nnuncle to& window .

(9) Surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P 6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Se r v e ! !--- I Janse-shal' include-ver!'icatter of the Borer O!!utier ^!:r: Setpoint --

-ef-4ess-than-or-equal-to-4n-incr+ase-of-twice - the count-rate-withh-

4ba kute-pee 4+4,-

r wmp r guemi%c.S rnen%L sd rnen d W e ('

r at creb. $ c.e,,cf W 3a4 houc emplh he pres \ ens of SpoGeh 4.o.3 acumcuw e.

BYRON - UNITS 1 & 2 3/4 3-12 AMEN 0MENTNO.14 r-w w e

w w %% pq gy , a isog.,. 92 me.

- - . - - - - - . . - . , . , , . - . - , - c . ,, - -

TABLE 4.3-1 (Continued)

_ TABLE NOTATIONS (10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICE.0PERATIONAL TEST shall be performed such that each train is tested at least every 62 days on a STAGGERED TEST BASIS and (

following maintenance or adjustment of the Reactor Trip Breakers and )

shall include independent verification of the OPERABILITY of the ( ,

Undervoltage and Shunt Trip Attachments of the Reactor Trip Breakers. '

Hot us,.J .

(12) At-least-enc: per 18-eenthr.-during thutdom "^"fy that er-:- cimulatod-Boron-044ut4en-OctM4ng-test 449na4--GVGG-valves-1-120-and4-epen-and- ,

412B-and-C-olose-within-304econds-(13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.

(14) Verify that the appropriate signals reach the Undervoltage and Shunt Trip [

Relays, for both the Reactor Trip and Bypass Breakers from the Manual )

Trip Switches. -Initial-penformance-of-thisaur+e(444nce-requ4r4msnt. for- '

the-Reactor-TrJp-Sypass-Br4akers is te48-completed-pr4cr-te the4tartup-

-fe44 ewing-the-thi rd-re fue44 ng-ou ta ge-4o n-Un i t-1-a nd-the4econd-re fuel 4 ng.

, outage-ton Un44-3, -

c (15) Manual Shunt Trip 3rior to the Reactor Trip Bypass Breaker being racked )

in and closed for )ypassing a Reactor Trip Breaker. [

c .

(16) Automatic Undervoltage trip. 4n44441--performance of- this-sur-ve444ancc-.

-requirement to be completed-prtor--te the4tartup-following-the thir&-

-refueling-outage-fo444-1-and-the4econd-refueling-outage--for Unit 2, BYRON - UNITS 1 & 2 3/4 3-12a AMENDMENTNO.[

TABLE 3,3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM c- TOTAL NO. CHANNELS . CHANNELS APPLICABLE w

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION w 1. Safety Injection (Reactor e- Trip, Feedwater Isolation, re Start Diesel Generators, "

Containment Cooling Fans, Cor. trol Room Isolation, Phase "A" Isolation Turbine Trip, Auxiliary Feedwater, Containment Vent Isolation, .

and Essential Service Water).

a. Manual Initiation 2 1 2 1,2,3,4 18

[. b. Automatic Actuation 2 1 2 1,2,3,4 14 .

Logic and Actuation U Relays

c. Containment 3 2 2 1,2,3 15^ 19 l Pressure-High-1

d. Pressurizer Pressure- 4 2 3 1, 2, 3# 1[

low (Above P-11) A

e. Steam Line Pressure- 3/ steam line 2/ steam line 2/ steam line 1, 2, 38 W l'i Low (Above P-11) any steam 1fne -

N 2. Containmint Spray g

g a. Manual Initiation 2 pair 1 pair 2 pair 1, 2, 3, 4 18 r

g

b. Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation K Relays

c. Containment Pressure- 4 2 3 1,2,3 16 High-3

TABLE 3.3-3 (Continued) en z

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 8 .

c- MINIMIRI m FUNCTIONAL UNII TOTAL HO. CHANNELS CHANNELS APPLICABLE OF CllANNELS TO TRIP OPERABLE MODES ACTION w -

> 0- 4. Steam Line Isolation

a. ' Manual Initiation

1) Individual 1/ steam line 1/ steam line 1/ operating l', 2, 3 23 steam line

2) System 2 1 2 1,2,3 22

b. Automatic Actuation 2 1 2 1,2,3 21 Logic and Actuation Relays y c. Containment Pressure- 3 2 2 1,2,3 -152- 13 7

High-2 Y

g d. Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# M*- li w Pressure-Low any steam

. (above P-11) line

e. Steam Line Pressure - 3/ steam line 2/ steam line 2/ steam line 3## 158- 11 Negative Rate'-liigh any steam (below P-11) line

5. Turbine Trip & '

Feedwater Isolation

a. Automatic Actuation 2 1 2 1, 2 24 Logic and Actuation Relays

b. Steam Generator 4/sta. gen. 2/sta. gen. 3/stm. gen. 1, 2 1 Water Level- g in any oper in each oper-High-High (P-14) ating sta.- ating stm.

gen. gen.

c. Safety Injection See Item 1. abvve for all Safety Injection initiating functions and requirem nts.

~

. TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E

c MINIMUM 5 TOTAL NO. CHANNELS CHANNELS APPLICABLE d FUNCTIONAL UNIT OF CHANNELS. TO TRIP OPERABLE MODES ACTION

p. 6.- Auxiliary Feedwater (Continued)

~

g. Auxiliary Feed- -

. water Pump Suction Pressure-Low (Transfer to I/n.w i g, , i / n..,

Essential Service Water) 2 2 -a-- 1,2,3 IS[

7. Automatic Opening of -

Containment Sump Suction

{ Isolation Valves Y a. Automatic Actuation Logic 2 1 2 1,2,3,4 la 0 and Actuation Relays

b. RWST Level - Low-Low 4 2 3 1,2,3,4 4 15 Coincident With Safety Injection See Item 1. above for Safety Injection initiating functions and requirements.

8. Loss of Power

a. ESF Bus Undervoltage 2/Eus 2/ Bus 2/ Bus 1, 2, 3, 4 25a

b. Grid Degraded Voltage 2/ Bus 2/ Bus 2/ Bus 1, 2, 3, 4 25b l e-

~ .

i. <

o .

TABLE'3.3-3 (Continued) .

- cn ,

~<

g- ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION i

x g -MINIMUM y _ .

TCTAL NO. CHANNELS CHANNELS APPLICABLE

" FUNCTIONAL UNIT OF~ CHANNELS TO TRIP , OPERABLE.

. MODES ACTION

- 9. Engineered Safety Features

"- Actuation System. Interlocks

. a. ' Pressurizer-Pressure, 3 2 '2 1,2,3 20 P-11

b. Reactor Trip, P-4 4-2/ Train 2/ Train- 2/ Train 1, 2, 3 22

~

c. Low-Low Tavg, P-12 4 2- -

3. 1, 2, 3 20 R

d. St;;; Cene. eter L'eter Level, 4/ste. 2/3t-. gs... 3/21 . . 1, 2 ZG P-li (High-"igh) g;. . in any gca. in !

Y epareting ;;;h !

o . Ste. gen. Osc.etiGg N . , e7.

s 6'

[

e h

1 l >

' i 1

- - . ,,m+r , . _ _m . ,_ --

TABLE 3.3-3 (Continued)

TABLE NOTATIONS

Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.

1. Trip function automatica11y blocked above P-11 and may be blocked below P-11 when Safety Injection on low steem line pressure is not blocked.

-- 2The-pr+v4+4ons-of4pe+4f444tton -3. 0.4 us-not-app 14cabia _ _ l r6tiu N ;upr.W d anti 4 OrcR ACLe sh4v$ wM b kootts*r ACTION STATEMENT -

ACTION 14 - With the number of OPERABLE changnel .one less than the Minimum -

- - m Channels OPERABLE requirementfbe in at least HOT STANDBY N otd "withinl6 hours 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 channel may be bypassed for up to hcurs for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE. q ACTION 15t- With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within

(,* 1 hourf l W E LT 3

ACTION 16 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable char.nel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 'for surveillance l testing per Specification 4.3.2.1. \ y _

ACTION 17 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

ACTION 18 - 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 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 . -

BYRON WITS'1 & 2 3/4 3-21

1 TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

ACTION 19 - With the number of OPERABLE channels one less than the Total 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+hourgand y l)

b. TheMinimumChannelsOPERABLErequirementis(met;however,the inoperable channel may be bypassed for up tonit hours for sur-veillance testing of other channels per Specification 4.3.2.1.

ACTION 20 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associat ! permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

testore N inervsWe channel b OPER ABLE stnivs w;Nin 6 hort, or ACTION 21 - With the number of OPERABLE (Channels one less than the Minimum Channels OPERABLE requiremen P,4 b e in at least HOT STANDBY 4henot withiM6hoursandinatleastHOTSHUTDOWNwithinthefollowing 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 + hours for surveillance testing per Specification 4.3.2.1 provide the other channel is OPERABLE. .

4 ACTION 22 - 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 SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 23 - 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.

restore he Neproie cAannel to OPEA AGLE stdes etMn 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , e ACTION 24 -Channels With theOPERABLE number of requirement, OPERABLEbe(channels in at least one HOTless than the STANDBY Minimum nen withinA6 hours; however, one channel may be bypassed for upo t' Ahours for surveillance-testingperSpecification4.3.2.1providedjtheother channel is OPERABLE. q ACTION 25 - a.- With the number of OPERABLE channels one less than the Minimum Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped condition within I hour. The inoperable 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 of tiie OPERABLE channel per Specification 4.3.2.1.

b. With the number of OPERABLE channels one less than the Mini-mum Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

BYRON - UNITS 1 & 2 3/4 3-22

r TABLE 4.3-2 h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS TRIP E ANALOG ACTUATING MODES 1 MASTER SLAVE FOR MIICH h CHANNEL CHANNEL-CHANNEL DEVICE OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

[ FUNCTIONAL UNIT m

1. Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Containment Cooling fans, Control Room Isolation, Phase "A" Isolation, Turbine Trip, Auxiliary Feedwater, Containment Vent Isolation and Essential Service Water)

Manual Initiation N.A. N.A. N.A. N.A. M.A.

$ a. N.A. R 1, 2, 3, 4 y b. Automatic Actuation N.A. N.A. M.A. N.A. M(1) M(1) Q 1, 2, 3, 4 ;

Logic and Actuation y

Relays .

c. Containment Pressure- S -M- Q N.A. N.A. N.A. N.A. 1, 2, 3 High-1

d. Pressurizer Pressure- S -M- Q N.A. N.A. N.A. N.A. 1, 2, 3 Low (Above P-11)

e. Steam Line Pressure- S R -M- Q N.A. N.A. N.A. N.A. 1, 2, 3 Low (Above P-11)

N Containment Spray g 2.

Manual Initiation N.A. M.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4 h a.

D b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) N(1) Q 1, 2, 3, 4 g Logic and Actuation Relays Y c. Containment Pressure- S R -ft- Q N.A. N.A. N.A. N.A. 1, 2, 3 High-3 A A jv -g

.. .s

TABLE 4.3-2 (Continued)

E ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

.h SUEVEILLANCE REQUIREMENTS c TRIP 2

ANALOG ACTUATING MODES

.U CHANNEL DEVICE

" MASTER SLAVE FOR WHICH CHANNEL CHANNEL' OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE' i

[.FUNCTIONALUNIT- CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED :

l

3. Containment Isolation -

a. Phase "A" Isolation

1) Manual Initiation M.A. M.A. M.A. R N.A. M.A. N.A 1, 2, 3, 4

2) Automatic Actuation M.A.. N.A. M.A. M.A. M(1) N(1) .Q 1, 2, 3, 4 Logic and Actuation Relays' -

g 3) Safety Inject!or. See Item 1. above for all Safety Injection Surveillance Requirements.

[ b. Phase "B" Isolation h 1) Manual Initiation' N.A.. N.A. M.A. R N.A. N.A. M.A 1 , 2 , 3 ,' 4

2) Automatic Actuation N.A.- N.A. M.A. N. A. M(1) M(1) Q 1, 2, 3, 4 Logic Actuation-Relays i

3) Containeent 5 R -M-Q N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-Nigh-3

^

c. Containment Vent Isolation t

1) Automatic Actuation M.A.. N.A. M.A. N.A. M(1) M(1) Q 1, 2, 3, 4 kz Logic and Actuation Relays E

G.

i

.o l

]- %

1 4

i n - - -

s1 TABLE 4 3-2 (Continued) h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

'g SURVEILLANCE REQUIREPENTS c TRIP z ANALOG ACTUATING M00ES Z

CHANNEL DEVICE MASTER SLAVE FOR idHICH CHAf00EL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE

~[FUNCTIONALUNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS PEQUIRED

" 3.c. Containment' Vent Isolation (Continued)

2) Manual Phase "A" See Item 3.a.1 above for all s.anual Phase "A" Isolation Surveillance Requirements.

i Isolation

, 3) Manual Phase #3" See Item 3.b.1 above for all manual Phase "B" Isolation Surveillance Requirements.

, Isolation

4) Safety Injection See Ites 1. above for all Safety Injection Surveillance Requirements.

$4.SteamLineIsolation f y a. Manual Initiation N.A. N.A. M.A. R N.A. N.A. M.A. 1, 2, 3

! '$ b. Automatk *.-tuation M.A. N.A N.A N.A. M(1) M(1) Q 1,2,3 Logic and' Actuation Relays-

c. Contu nnent Pressure- S it- Q M.A. N.A. M.A. M.A. 1, 2, 3 Hip't-2

d. S nam Line Pressure- S N- Cl N. A.- N.A. N.A. M.A. 1, 2, 3 Low (Above P-11)

e. Steam Line Pressure 5 4t- G - N.A. N.A. N.A. M.A. 3 kg - Negative Rate - High (Below P-11) x 3 5. Turbine Trip and Feedwater. ^

z Isolation P a. Automatic Actuation M.A. N. A. - N.A. N.A. N(1) M(1) 1, 2 Q

Logic and Actuation Relay

TABLE 4.3-2 (Continued)

?

iy.INEERED SAFEiY FEATURES ACTUATION SYSTEM INSTRUMENTATION

$ SURVEILLANCE REQUIREMENTS TRIP '

$d ANALOG ACTUATING MODES

" CHANNEL DEVICE MASTER SLAVE FOR WHICli I

" ACTUATION RELAY RELAY SURVEILLANCE CHANNEL CHANNEL OPERATIONAL OPERATIONAL TEST LOGIC TEST TEST TE5f 15 REQUIRED CHECK CALIBRATION TEST

[fUNCTIONALUNIT l

5. Turbine Trip and Feedwater (Cont'nued) g(q mg q

-NA . M-A. 1, 2

b. Steam Generator Water 5 R #Q N.A. W A-Level-High-High (F-14)

Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

c.

,6. Auxiliary Feedwater N.A. N.A. 1, 2, 3 N.A. N.A. R N.A.

A a. Manual Initiation N.A. I M(1) Q 1, 2, 3 N.A. N.A N.A. N.A. M(1) 3 b. Automatic Actuation w Logic and Actuation Relay N.A. N.A. N.A. 1, 2, 3 c.. Steam Generator Water 5 it G N.A.

Level-8nw-Low N.A. R N.A.

G j;M(3) N.A. N.A. N.A. 1, 2 g

d. Undervoltage-RCP Bus Safety Injection See ' tem 1. above for all Safety Injection Surveillance Requirements.

e. N.A 1, 2, 3, 4 1 N.A. M(2, 3) N.A. N.A.

f. Division 11 for Unit 1 N.A. R /

(Division 21 for Unit .:)

ESF Bus Undervoltage M N.A. N.A. N.A. N.A. 1, 2, 3 g g. Auxiliary Feedwater 5 R g Pump Suction P essure- --

o Low 4

67. Autom tic Opening of g Contairment Sump Suction Isolation Vaives J

P

g i

TABLE 4.'3-2 (Continued) i I i ' ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 'l

! E z

-SURVEILLANCE REQUIREMENTS

' TRIP ANALOG ACTUATING MODES !

E O CHANNEL DEVICE W IER SLAVE FOR MiICH. !

CHANNEL CHANNEL- OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE i l' TEST LOGIC TEST TEST TEST IS REQUIRED i

4-

[FUNCTIONALUNIT CHECK CALIBRATION TEST

-- l p

' 3 . Autou tic Opening (Continued) l

a. . Automatic Actuation 7 Logic and Actuation Relays N.A. N.A. N.A. N.A. M(1) N(1) Q 1, 2, 3, 4 l l.

N.A. N.A. N.A. 1,2,3,4 !

b.' RWST Level-Low-Low' S R d M.A.

i Coincident With i Safety Injection- See Item 1.-above for all Safety Injection Surveillance Requirements i 2 w

[ ~ M . Loss of Power Y

w

a. ESF Bus Undervoltage- N.A. R N.A. M(2,'3) N.A. M.A. M.A. 1, 2, 3. 4 (

' Grid Degraded Voltage N.A. R N.A. M(3) N.A. N.A. N.A. 1, 2, 3, 4 [

b.

~

]

[

f 9. Engineered Safety Feature Actuation System Interlocks

.(

4t- R N.A. N.A. N.A. M.A. 1,2,3

a. Pressurizer Pressure.- N.A.

.P-11 M.A N.A. R N.A. M.A. N.A. 1, 2, 3 i b. Reactor Trip, P-4 ..

N.A. #q N. A.- M.A. M.A. N.A. 1, 2, 3 (;

m

c. Low-Low Tavg,-P-12 o_

N(1) 1, 2

%.A. t

5 d. Sta 's.m..tv. Lie.

u 5 %"-- X- M(1) Q I

i M L .e!, 7 l' 4' 'i (;;ip.;;!p.)

z TABLE NOTATION 4 P Each train shall be tested et least every 62 days on a STAGGERED TEST BASIS.

p((1)

2) Undervoltage relay oper hility.is to be verified independently. An inoperable 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 surwell!ance testing of the OPERA 8LE channel per Specification 4.3.2.1.

j (3) Setpoint verification is not appilcable. .

g 5 Ti e meu;iseu la u.ii. ; a v ~2 , be e,amJz: ;& 32 ~.; n " & .-_ C,cie ; &^;'y.

l i . i

- . _ . _ . . ~ _

~ . - _

i 3/4.3 INSTRUMENTATION BASES ;

,3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety features Actuation System instrumentation and interlocks ensures that: (1) the associated 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 ft mrhttbed, (3) iuffkient redundancy-+s-main--

tehed-tc perdt-e-chennel te be cut of-seevke--fee-tett4ng or =ht+neneer-and-(4)-s u f44ci e n t-syr4em4u ut4 enal-capaM44 ty-4+-ava i4able-f rom-d ive rs e-paramettess Add Insui i The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements '

specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. Add Inud a, The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy. .

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4. Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation;from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of 4 the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 3.3-1, Z,+ RE + SE $ TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 3.3"4, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span,

' between the Trip Setpoint and the value used in the analysis for the actuation.

RE or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified Trip Setpoint. SE or Sensor Error is either BYRON - UNITS 1 & 2 B 3/4 3-1

10seM.1:

New Bases Paragraph #1 (Add to existing paragraph.)

...and sufficient redundancy is maintained to permit a channel to be out of service for testing or ma:ntenance consistent with maintaining an appropriate lovel of reliability '

of the Reactor Protection and Engineered Safet) Features Instrumentation and,3) sufficient system functions capability is available from diverse parameters, las.e M 2:

New Bases Paragraph #2 (Add to existing paragraph.)

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System", and supplements to that report Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

InteM3:

New Action to add to page 3/4 3-21 ACTION 15a - With the number of OPERABLE channels one less than the Total Number of Channels, declare the associated pump INOPERABLE and take the ACTION required by Specification 3.7.1.2.

l.

ZNLD/G15/121 -

l . - . - .-. _ .. . - - . - .. --

LIMITING SAFETY SYSTEM SETTINGS BASES .

Turbine Trir l A Turbine trip initiates a Reactor trip. On decreasing power the Turbine c>

trip is automatically blocked by F 7 = P-8 (a power level of approximately -

-10% (F7) or 30% (P-8) of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10%-(4'73-or 30% (P-8) of full power equivalent);

and on increasing power, reinstated automatically by-#4e P-8.

Safety injection input from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety injection.

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

React _or Coolant Pump Breaker Position Trip The Reactor Coolant Pump Breaker Position trips are anticipatory trips which provide core protection against DNB. The Open/Close Position trips assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. No credit was taken in the accident analyses for operation of these trips. Their functional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Trip System. Above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% of full power equivalent) an automatic Reactor trip will occur if more than one reactor coolant pump breaker is opened. Below P-7 the trip function is automatically blocked, becctor t+4p on hrbice-trip-4+-eneMedabwe+7-(-10)'4-unt44-the-+od444u44on-

-+r-4 mp4emeM+d-wh ieh-eneMet-B eee to e-teip-on-Twb49e--te4p-ebo ve-P-8-(-30% ) .

BRAIDWOOD - UNITS 1 & 2 B 2-8 AMEN 0 MENT NO.

l LIMITING SAFE 1Y SYSTEM SETTINGS l BASES I

l- Reactor Trip System Interlocks The Reactor Trip System Interlocks perform the following functions:

P-6 On increasing power, P-6 allows the manus) block of the Sourc- Range Reactor trip (i.e. , prevents pretnature block of Source Range trip),

provides an automatic backup block for Source Range Neutron Flux doubling, and the manual block that de-energizes the high voltage to the Source Range detectors. On decreasing power, Source Range Level trips and Neutron Flux doubling circuits are automatically reactivated and high voltage restored.

$/ C P-Y On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, more than one reactor - )

coolant pump breaker open, rea tor coolant pump bus undervoltage and underf requency, -TeMnc t-ip ressurizer low pressure and I pressurizer high level. On ecreasing power, the above listed trips are automatically blocked.

n increasing power, P-8 automatically enables Reactor trips on low N P

flow in one or more reactor coolant loops and Turbine trip. On de-creasing power, the P-8 automatically blocks the single loop low flow p

trip and Turbine trip.

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range Reactor trip and the Low Setpoint Power Range Reactor trip; and automatically blocks the Source Range Reactor trip and provides an automatic backup function to de-energize the Source Range high voltage power. On decreasing power, the Intermediate Range Reactor trip and the Low Setpoint Power Range Reactor trip are auto.matica11y reactivated and Source Range high voltage to the detectors is restorer) if power decreases below the P-6 setpoint. Provides input to P-7.

P-13 Provides input to P-7.

q b 5% %

4% b -' l g7 %T1 b# %

4 ten 1; imphmented-wMeh-enablee-Reaeur-trip-on-TurMne-tr+p-ebeve P-0 00th BRAIDWOOD - UNITS 1 & 2 B 2-9 AMEN 0 MENT NO. l

[g ee

. TA8tE 3.3 -

b REACTOR TRIP SYSTEM IniTRUE NTATION

^

MINIMIM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UND OF CHANNELS -TO TRIP OPERABLE MODES ACTION

g. .

Z

1. Manual Reactor Trip 2 1 2 1 2 1 -

2 1 2 3I . 4". 5* 10

- 2. Power Range. Neutron Flux

= 1 a.. High Setpoint 4 2 3 1. 2

b. Low Setpoint 4 2 3 188# 2

3. Power Range. Neutron Flux

, High Positive Rate 4 2 3 1, 2 2[

1

' 4. PowerRange.NeutronFIE. 4 2 3 ~ 1, 2 2 1

High Negative Rate T

j 5. Intermediate Range. Neutron Flux 2' 1 2 1###, 2 3

6. Source Range. Neutron Flux

i a. Startup 2 1 2 28C 4

b. Shutdown - g

, 2 1 2 3.4.5 5

7. Overtemperature AT i

4 2 3 1, 2 6[

8. Overpower AT 4 2 3 1. 2 f

9. Pressurizer Pressure-Low

! (Above P-7) , 4 2 3 1 l

i

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION g

b MINIMUM 8 TOTAL NO. CHANNELS . CHANNELS APPLICABLE ACTION e FUNCTIONAL UNIT' 0F CHANNELS TO TRIP OPERABLE MODES

10. ~ Pressurizer Pressure-High. 4 '2 3 1, 2 w 11. Pressurizer Water Level-High e- (Above.P-7) 3 2. '2 1 6[

m

12. Reactor Coolant Flow-Low

a. Single Loop (Above P-8) 3/locp 2/ loop in any oper-2/ loop in each oper-1 6[

ating loop - ating loop

$ b. Two Loops (Above P-7 and 3/ loc. 2/ loop in 2/ loop in 1 6[

below P-8) two oper- - each oper-T ating loops ating loop w

13. Steam Generator Water 4/sta. gen. 2/sta. gen. 3/sta. gen. 1. 2 6M- 1 Level-Low-Low in any each operating operating

,i sts. gen. stm. gen.

14. Undervoltage-Reactor Coolant Pumps (Above P-7) 4-1/ bus 2 3 1

15. Underfrequency-Reactor Coolant Pumps (Above P-7) 4-1/ bus 2 3 1 6[

E 16. Turbine Trip , ,

5 x (Above P ' cr P-8) % [ -

a. Emergency Trip Header Pressure 3/ Train 2/ Train 2/ Train 1 6[

z b. Turbine Throttle Valve Closure 4 4 1 1 P

! Y

'^^ o $ t - tri

^ "

Cr TU-bin; trip I; Cnchicd etcJE P-7 (l Z) until th; E di'icat*;n i; ig li;;nt d d i f -

k g

i en:b?r ":::ter trip :- Tu-bin trip st;sc "-S (3'").

e +r- , + r -

i L

TABLE 3.3-1 (Continued)

TABLE NOTATIONS

Vith the Reactcr Trip System breakers in the closed position and the Control Rod Drive System capable of rod withdrawal.

uThe-boron-dWtion-f4ux-doubl4ng-s4 nais-may4e44ocked-dup 9 4ng-veactor 4**.stutup-Thete-c henneh-tho-prov i de -i npu ts-to4SF A S . -The-Act4cn-Statement-fee-the-

-ehannels-4n-Tehle4.44-4-more-conservative-endr-thereforer-control 44ntr--

-#The-p rov biens-4 f-Spec 444ca Wn4r0A-a re-not-appli cabl e.

1. Below the P-6 (Intermediats Range Neutron Flux Interlock) Setpoint. ~

1. 1. Below the ' In .".ow Setpcint Power Range Neutron Flux Interlock) Setpoint.

@Whenever tht a tor Trip Bypass Breakers ate racked in and closed for by- C~

r passing a Reactor Trip Breaker. )

ACTION STATEMENTS ACTION 1 - 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 be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - With the number of OPERABLE channels.one less than the Qtal 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 ; e 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; and

c. Either, ;!ERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less,thAn or equal to BS% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT F0WER TILT RATIO is monitored at least once per 12 hnuts per Specification 4.2.4.2.

ACTION 3 - With the number of channels OPERABLE one less thaa the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6

'Setpoint; and

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable chanrel to OPERABLE status prior to

(' increasing THERMAL POWER above 10% of RATED THERMAL POWER.

BRAIDWOOD - UNITS 1 & 2 3/4 3-5 AmendmentNo.h

)

i

~

~ TABLE-3.3-1 (Continued)

ACTION STATEMENT 5 (Continued)

ACTION'4 - With the number of OPERABLE channels one less than the Minimum

-Channels OPERABLE requirement suspend all operations involving ipositive reactivity' changes.-

ACTION 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 within the next hour open the reactor trip breakers, suspend All operations involving positive reactivity changes, and verify valves CV-1118, CV-8428, CV-8439, CV-8441 and CV-8435 are closed and secured in position. With no channels OPERABLE verify compliance with the SHUTDOWN MARGIN

. requirements of Specification 3.1.1.1 or 3.1.1,2,-as applicable, and take the actions stated above within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and verify compliance at least.once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION.6 - With:the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The' inoperable channel is placed in the trippea condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and

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.

ACTION Deleted J CTION_8; With less,than the Minimum Number of Channels OPERABLE, within g g , @ a Q 1: hour' determine bv observation of the associated permissive-annunciator window'(s) that the interlock:is in its required state

[ Nd T* b 0 ,ge existing plant. condition, or apply Specification 3.0.3.

, pf, s,abegj - --

n ACTION _9'=- With the' number of OPERABLE channels one less than the Minimum l N Channels OPERABLE requirement, be in at least HOT STANDBY

@'-- 4 / hours-for w ithTnT6 hours;-howeve:. one channel may be' bypassed for up to surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 110.- 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.

.CTION 11 - With the number of OPERABLE channels less than the Total Number of. Channels, operation may continue provided.the inoperable channels are pisced in the trippeo condition within g(o hourp l ACTION 12 - a. With one of the diverse trip features (Undervoltage or C Shunt Trip Attachment) inoperable, restore it to OPERABLE )

BRAIDWOOD _ UNITS 1 & 2 3/4 3-6 Amendment No.

4

- . v ~. g TABLE 4.3-1 to I b REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

l. c2 TRIP l '5 ANALOG ACTUATING MODES FOR i

I CHANNEL DEVICE WHICH OPERATIONAL ACTUATION SURVEILLANCE E CHANNEL CHANNEL OPERATIONAL CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED l } FUNCTIONAL UNIT ~

i l [ 1. Manual Reactor Trip N.A. N.A. N.A. R(14) N.A. 1,'2, 3*, 4*, 5* ) ;

l i

2. Power Range, Neutron Flux l

a. High Setpoint S 0(2,4), Q N.A. N.A. 1, 2 q l .#

i M(3,4)

Q(4,6),gj R(4,Ja)r

b. Low Setpoint S R(4)Y/ Q N.A. N.A. 1###, 2

3. Power Range, Neutron Flux, N.A. R(4 k Q N.A. N.A. 1, 2 High Positive Rate "4

4. Power Range, Neutron Flux, H.A. R(4 k Q N.A. N. A. 1, 2 High Hegative Rate

5. Intermediate Range, S R(4, Sa Q N.A. N.A. 1###, 2 Neutron Flux

6. Source Range, Neutron Flux 5 R(4,Sb Q(9) H.A. N.A. 2##, 3, 4, 5

7. Overtemperature AT S R(13)k Q N.A. N.A. 1, 2

8. Overpower AT S R Q N.A. N.A. 1, 2

9. Pressurizer Pressure-Low 5 P N.A. N.A. 1 g (Above P-7)

W 10. Pressurizer Pressure-High 5 P Q N.A. H.A. 1, 2 K 11. Pressurizer Water Level-High S' k Q N.A. N.A. 1 (Above P-7)

L --___ _

r '. M

..n & .'

\.

,A D ' '

l.

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS O

h o

TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH c- CHANNEL CHANNEL 0PERATIONAL OPERATIONAL ACTUATION SURVEILLANCE

'3 FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS PEQUIRED

12. Reactor Coolant Flow-Low 5 Q N.A. N.A. 1 m 13. Steam Generator Water Level- S H.A. N.A. 1, 2 Low-Low ,

14. Undervoltage-Reactor Coolant Pumps (Above P-7)

N.A. R N.A. Q ) N.A. 1 ('

w 15. Underfrequency-Reactor N.A. R N.A. Q(10) N.A. 1 D Coolant Pumps (Above P-7) #

16. Turbine Trip (Above ."--7 cr P-8) l

a. Emergency Trip Header N.A. R N.A. S/U(1,10') N.A. 1 Pressure

b. Turbine Throttle Valve N.A. R N.A. S/U(1,10) H.A. 1 Closure

17. Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2 ESF

18. Reactor Coolant Pump Breaker N.A. N.A. N.A. R N.A. 1 l Position Trip (Above P-7

$ 19. Reactor Trip System Interlocks z a. Intermediate Range i

% Neutron Flux, P-6 N.A. R(4) -Q-- d N.A. H.A. 2M 4 b. Low Fower Reactor

g. Trips Blcck, P-7 ,

H.A. R(4)k Q (S) R N.A. N.A. I

c. Power Range Neutron gj

X Flux, P-8 N.A. R(4)r Q (SF R N.A. N.A. 1 l

q - .Q:.

~'

.+

,. ..TA8LE 4.3-1 (Continued)

A'

o REACTOR TRI'P SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS

'k TRIP ANALOG. ACTUATING MODES FOR.-

~ CHANNEL DEVICE WHICH . . -

E ' CHANNEL- CHANNEL. OPERATIONAL OPERATIONAL ACTUATION . SURVEILLANCE ~

{UNCTIONALUNIT CHECK CALIBRATION TEST TEST- LOGIC' TEST :IS REQUIRED -

p9. Reactor Trip System Interlocks (Continued)'-

. d. Low Setpoint Power Range Neutron F1ux,'P-10 N.A.-

W R(4 W 'Q (S) k N.A. N. A. 1, 2-

e. Turbine. Impulse Chamber-Pressure, P-13

/2/

N.A. R# 4-(4) R N.A. M.A. ~1 p0. Reactor Trip' Breakers- N.A., N.A. N.A. M (11) N.A. 1, 2, 3*, 4*, Sh Y21. Automatic Trip and ' Interlock! N.A. N.' A. N.A. 'N.A. M (7) 1, 2, 3*, 4*,.5*

O Logic

22. Reactor Trip Bypass' Breakers N.A. N.A. 'N.A. (15).-R (16) N.A. 1,2,3*,4*,Sh _

i f

a a

P h

,.r,.. s c 4 - , . - - =

~1h m , hJ s w k po mt co m p sN c f incere b cue nt Axt 4 L flu t biFFCCENC d~

(c, llowi n n. re v(dte c. *do b e. fMem ed Telor h c4c. 4 n 75 'A. o E

[ TAmb Terr /fAL Tce;g.O od<c. ll" y8

.3-1 (Continued)

TABLE NOTATIONS.

Vith the Reactor Trip System breakers closed and the Control Rod Drive System capable of rod withdrawal.

'Mmcsc channels-elso-provide-inpes-to-ESFAS. The-Operc.Mwal-Test frequency-Joe-these-channels-in-Tabic 4. 3 is-more-conservat4ve-ano r -therefore,

-controll4rp

-The-+ pee 4Hed-1&-conth-4nteevd-eay-be-extended-to-32-months-foe-eycle 1-

-entyr

1. 8elow P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1. 1. Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) If not performed in previous / days. l (2) Comparison of calorimetric to excore power indication above 15% of RATED THERHAL POVER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of 5 e i-fication 4.0.4 are not applicable for entry into H00E 2 or 1. ggf (3) ,5' ingle point comparison of incore to excore AXIAL FLUX DIFFERENCE3above l 15% of RATED THERHAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are l-*

not applicable for entry into MODE 2 or 1.j (4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(Sa) Initial plateau curves shall be measured for each detector. Subsequent plateau curves shall be obtained, evaluated and compared to the initial #

curves. For the Intermediate Range and Power Range Neutron Flux channel's C the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(5b) With the high voltage setting varied as recomended by the manufacturer, an initial discriminator bias curve shall be measured for each detector. Sub-sequent discriminator bias curves shall be obtained, evaluated and compared to the initial curves.

s6) Incore - Excure Calibration, above 75% of RATED THERM /L POWER. The provi-sions of Specification 4.0.4 are not applicable for. eatry into MODE 2 or 1., l (7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

h M '(8 4!!th powcFgreeter-than-oe-equal-terthe-interlock-Setpoint-the rcquirc?

~'

ANALOG CHANNEL-CPERATIONAL TEST thal 4-consist-of-verifying-that-the-4nter-lect i: in the rcqu!r:d--+ tee-by-observing-the-permissive-annunciater-vindowe (9) Surveillance in H0 DES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Surveil-4ance ch:1' inclue veri'! cation of the Bere.a 04lution Alare Setpoint

+f-4ess-than-oe-equal--to-en-increase-of-tw4co -the ccunt-rate-withima--

-10 minute- per-14d,--

(10) Setpoint verification is not applicable.

For th e y y on G o@ %s survr$[(owc-e , menR S d m m M lpsY cmce_. 7ee M m P.b . T ( 5 .a w cgth % pee ~s o f spmGeh 4.o.3 are nc;t 9 7. i ~ ic.

{'

BRAIDWOOD - UNITS 1 & 2 3/4 3-12 Amendment No. X r" Mr~d*~ A A W M -en2mN i

i TABLE 4.3-1 (Contir.ued)

TABLE NOTATIONS

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall be performed such that each train is- tested at least every 62 days on a STAGGERED TEST BASIS and

[~

following maintenance or adjust'sent of the Reactor Trip Breakers and shall include independent verification of the OPERABILITY of the Undorvoltage

-and-Shut' Trip Attachments of tha. Reactor Trip Breakers. '

At used.

-(12)At1::: tone p:; 10 :: nth; during :h:td:un " ,y that en : :idated-Beren Diluti:n Deubliq te:t cignal CVCS-v. . A420 :nd E Ope : d

-1129- nd 0 cl::: within 30 ;;:ende.

.(13) CHANNEL CALIBRATION shall include ti.e.RTD bypass loops flow rate.

(14) Verify that the appropriate signals reach the Undervoltage and Shunt Trip relays, for both the Reactor Trip and Bypa:s Breakers froe the Manua Switches. --Initi:1 p:rferna : c' th!: :;rveill:::: r:quir := t is t: 5:

=;1 ted-prior t; th:-Stectup-feHowing-the-Unit--I Cycle 1 R:fueF-Outeger

- (15) Manual Shunt Trip prior to the Reactor Trip Bypass Breaker being racked in and closed by bypassing a-Reactor Trip Breaker. ( -

~

(16) Automatic undervoltage trip. =Initi:1 p rfemenee-of-thit-surveH4ence-r:;;ir::=t i: t b: cespl:t:d prier t; th startup-following-the-Unit Cy.1: 1 R: feel Out:;:r-u

. BRAIDWOOD - UNITS 1 & 2 3/4 3-12a (l

AmendmentNo.F

1

\

l l

l INSTRUMENTATION

'3/4.3.2- ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATION LIMlilNG CONDITION FOR OPERATION 0

3. 3. U./The Engineered Safety features Actuation System (ESFAS) instrumentation I channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints Table set consistent with the values shown in the Trip Setpoint column of 3.3-4, -

3 APPLICABILITY: As shown'in Table 3.3-3.

ACTION: .

a. With an ESFAS Instrumentation or Interlock Trip Setpoint less con-servative than the value shown in the Trip setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value,

b. With an ESFAS Instrumentation or Interlock Trip Setpoint less con-servative than the value shown in the AIL wable Values column of Table 3.3-4, either:

1, Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4 and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or C,

2. Declare the channel inoperabla and apply the applicable ACTION statement requirements of Table 3.3-3 ttntil the channel is restored to OPERABLE status with its Setpaint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z + RE + SE 5, TA Where:

Z = The value from Column Z of Iable 3.3-4 for the affected channel, RE = The "as measured" value (in percent span) of rack error for the affected channel, SE = Either the "as measured" value (in percent span) of the sensor error, or the value for Column SE (Sensor Error) of Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the af fected channel.

c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.

Centn:! her -ieleMe not required-peter-to-4nM4ebeef t4eeTky- cn Cyele-+--

- , ^ i' f aey-B*a444e-;; Vent!'atic,t-eetueMemet-cequired-prior tc init4al opers-44emat ; 20*' hted-%ema1 Pe er (MP)Myc1e 1.

!\

8RAIDWOOD - UNITS 1 & 2 3/4 3-13 AMENDMENT NO.12 .

4 . u.

g;

~

. TABLE 3.3-3; e -

b ENGINEERED SAFETY FEATURES ACTUATIOff SYSTEM INSTRUNENTATION

- o

'k -

TOTAL'NO.

CHANNELS.

MINIMUM CHANNELS APPLICABLE

FUNCTIONAL' UNIT- 0F' CHANNELS TO TRIP- OPERABLE M00ES ACTION.

E-Z

1. . Safety Injection (Reactor Trip,-Feedwater Isolation,

. Start Diesel Generators,. -

"- Containment Cooling Fans...

Control Ro'on Isolation,

= Phase "A" Isolation, Turbine Trip, Auxiliary Feedwater,-

Containment Vent' Isolation, and Essential Service Water).

w a. Manual Initiation 2 1 2 '1, 2, 3, 4 18 4

.1 w b. Automatic Actuation .' 2 1 2 1,2,3,4 14 4 Logic and Actuation

Relays

c. Containment 3 2 2 1,2,3 --158- \% !

Pressure-High-1

d. Pressurizer Pressure- 4 2 3 1, 2, 3# . 19 Low (Above P-11) I

e. Steam Line Pressure-Low ~(Above P-11) 3/ steam line ~2/ steam line any steam 2/ steam line 1, 2, 3#

-15*-(9 b l line

2. Containment Spray

a. Manual Initiation 2 pair 1. pair 2 pair 1, 2, 3, 4 18

b. ' Automatic Actuation. -2 1 2 1,2,3,4 14 Logic and Actuation Relays

c. Containment Pressure- 4 2 3 1, 2, 3 16 High-3

~

4 har' a, .TA8tE 3.3-3_(Continued) b ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 4 al -

MINIMUM TOTAL NO. CHANNELS T CHANNELS APPLICABLE g FUNCTIONAL UNIT OF CHANNELS TO TRIP f - GPERABLE MODES ACTION .

w

4. Steas'Line Isolation '

e, - a. Manual Initiation u 1) Individual- 1/stene line 1/ steam line 1/ operating 1, 2, 3 23 '

' steme line

2) Systen 2 1 2 1,2,3 22

b. -Automatic Actuation 2 1 2 1,2,3 21 Logic and Actuation Relays 1 4:*

c. Containment Pressure- 3 2 2 1,2,3 -Ma. g b' Hig'a-2

,'a'. d. Stese Line 3/staan line gg $

" 2/ steam line 2/stene Ifne 1, 2, 3# --ga_

Pressure-tow any stese (above P-11) lipe .

e. Steae Line Pressure - 3/ steam line 2/ steam line 2/ steam line 3N --MS- \9 Negative Pate-High .

any steam (below P-11) line

5. Turbine Trip &

Feedvatar Isolation i

a. Automatic Actuation 2 -1 2 1, 2 24 Logic and Actuation !

Relays

b. Steam Generator 4/ste. gen. 2/ste gen. 3/sta. gen. 1, 2 19

~

Water Level- in any op.r in each oper- 'g High-High (P-14) ating sta ating sts.

gen. gen. t

c. Safety _ Injection See Ites 1. above for all Safesty injection initiating functions and requirements.

e TABLE 3.3-3 (Continued)

' ENGINEERED SAFETY FEATU8ES ACTUATI)N 51iSTEM INSTRUMENTATION yo .

MINIMUN TOTAL NO. CHANNELS CHANNELS APPLICA8LE E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES _. ACTION '

q

s.

6. Auxil;ary Feedwater (Co.itinued)

- g. Auxiliary feed--

- water Pump Suction .

Pressure-Low (Transfer to b* N "^ Nmb p i Essential Service Water) 2 - 2 1, 2, 3 e \9 ' I

7. Automatic Opening of' m Containment Sump Suction i Isolation Valves J. a. Automatic Actuation Logic 2 1 2 1, 2, 3, 4 14

and Actaation Relays

b. RWST Level - Low-Low 4 2 3 1,2,3,4 Coincident With ' -M-6 l Safety Injectico 'See Ites 1. above for Saf,ety Injection-initiating functions and requirements.

8. Loss of Power 25aN

a. ESF Bus Undervoltage 2/ Bus 2/8us 2/8us 1, 2, 3, 4

b. Grid Degraded Voltage 2/8us 2/ Bus 2/8us 1, 2, 3, 4 25b#

.f"

..

TABLE 3.3-3 (Co..tinued) 5 ENGINEERED SAFETY FEATURES ACTUAT ON SYSTEM INSTRUMENTATION

g. MINIMUN' TOTAL NO. CHANNELS CHANNELS -APPLICA8tE c- FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 8LE 2

4 MODES ACfION d 9. Engineered Safety features -

- Actuation System Interlocks

e.

u- a. Pressurizer Pressure. 3 26 2 1,2,3 20' P-11

b. Reactor Trip, P 2/ Train 2/ Train 2/ Train 1, 2, 3 22

c. Low-Low T,,g, P-12 4 2 ,

3 1, 2, 3 20 t'

d. St::: C:::reter' Jet:r :. . 1, t/:ts. 2/:tc. ;:- 3/:ts. 1, 2

.T

-14 'Hig. ;;'d.) ... b: :p ;::. 5:

Y ;dt t!:; :::t Es ;;;., ;;;. :;;r:t!:;

ch. ;: _

1 i

!i e

4 a

i

?

v

. . . ... . . ~ . - -. ..

s TABLE 3.3-3 (Continued)

TABLE NOTATIONS

Trip function may be blocked-in this MODE below the P-11 (Pressurizer

. . Pressure Interlock) Setpoint.

-## Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.

"he p evis f eas e' Spec"fcation 3.0.2 e-e not epp!!ceb!e.

MotMkE percMc Awwe\ .

Ohh3LE . M L6^ ION STATEMENTS ACTION 14 --With the number of OPERA 8L channels one less than the Minimum

-Channels-0PERABLE requiremen , be.in at least HOT STANDBY-.

p within96 hours and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; ;

.however, .one channel may be bypassed for .up to0 h7 ' hours for surveillance testing per Specification'4.3.2.1, provided the

- other channel is OPERA 8LE. -

4 ACTION 15 - With the nust.er of OPERABLE channels 'one less than the Total-Numbn of. Channels, ooeration may neer**n uritil performance of

- the next required ANALOG CHANNEL- OPERATIONAL TEST providad the i operable channel is placed in the tripped condition within -

ggg 3 b hourf _

I

". I ACTION 16 -- With the number of OPERABLE: channels one' less than the- Total Number of Channels, operation may proceed provided the inoperable chennel is placed in the bypassed. condition and- the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to / hours for surveillance k testing per Specification 4.3.2.1.4 ACTION 17 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

-ACTION' With the number of OPERABLE channels one less than the Minimum

_ Channels OPERABLE requirement, restore the innperable channel to OPERABLE status.within a8 hours or be in at least HOT STANDBY within the next 6Lhours and in COLD SHUTOOWN within the following-

' :' 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , i a l.

L c

w i

4 BRAIDWOOD - UNITS 1 & 2 3/4 3-21

\

TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued) ;

ACTION 19 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The ino rable channel is placed in the tripped condition within hourjanel l

b. TheMinImumChannelsOPERABLErequirementismet;however,the inoperablechannelmaybebypassedforuptoYhoursforsur- l veillance testing of other channels per Specifice. tion 4.3.2.1.

-- ,,,JC, T10N 20 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months gg 4e g g.t Q determine by observation of the associated permissive annunciator __

d% g window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

g gg, (o 64 eg

~~"'AD10f( 21. With the number of OPERABl.E Channels one less than the Minimum Channels OPERABLE requirement be in at least HOT STANDBY gA with1nt6 hours and in at least HOT SHUTOOWN within the ollowing

= -- 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 hours for surveillanca testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 22 - With the numb'er of OPERABLE channels one less than the Total Number of Ch1nnels, restare 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 NOT SHUT 00VN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 23 - 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.

- _ CTION 24 - With the number of OPERABLE channels one less than the Minimum '

g 4 Channels OPERABLE requirement,be in at least HOT STA BY 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 t hours for d*d% $ , surveillance testing per Specification 4.3.2.1 provided the other

@- G A b W 8'

, i I channel is OPERABLE.

ACTION 25 - a. With the number of OPERABLE channels one less than the Minimum Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped condition within 1 hwr. The inoperable channel say 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 of the OPERABLE channel per Specification 4.3.2.1.

b. With the number of OPERABLE channels one less than the Mini-num Ni .,er of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperabla channel is placed in the tripped condition within 1 bour.

BRAIDw000 - UNITS 1 & 2 3/4 3-22 E ___a_--_-___._______-..-___ - _ _ _ _ - . . _ . . _ . _ _ . _ . _ - . _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ - - - . - . _ - _ _ _ _ _ _ _ _ . . _ _ _ . _ _ _ _ _ _ - _ - _ _ - - _ _ _ _ _ - _ _ - _ _ - _ _ . _ _ _ _ . -

~

X ~

p; .

.A .

~

7

-TABLE 4.3-2.-

i k ENGINEERED' SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS- ~

'g TRIP.

g. . ANALOG ACTUATING. . MODES-CHANNEL. .-DEVICE . MASTER: SLAVE FOR WHICH- ,

c CHANNEL CHANNEL OPERATIONAL OPERATIONAL: ACTUATION RELAY RELAY SURVEILLANCE-3 CHECK CALIBRATION TEST TEST- -LOGIC TEST TEST- TEST IS REQUIRED g FUNCTIONAL UNIT

[ 1. Safety Injection-(Reactor Trip, y

Feedwater Isolation, Start Diesel Generators, Containment Cooling Fans, Control-Room Isolation,-

PhaseA" Isolation. Turb'ine Trip, Auxiliary Feedwater, ..

Containment Vent Isolation and Essential. Service Water)

N.A. N.A. R N.A. N.A. N.A. 1, 2, 3. 4'

a. Manual Initiation N.A.

N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, . 4 -

, b. Automatic Actuation -l N. A.

g Logic and' Actuation Relays

-M-Q N.A. N.A. N.A. N.A. 1, 2, 3

c. Containment Pressure- S

-High-1

--M-- Q N.A. N.A. N.A. M.A. :1, 2, 3-

d. Pressurizer Pressure- S R Low (Above P-11)-

e. Steam Line Pressure- S Rk -M-Q N.A. N.A. N.A. N.A. 1. 2, 3 Low (Above P-11)

2. Containment Spray Manual Initiation N.A. N.A. M.A. R N.A. N.A. N.A.. 1, 2, 3,.4 t k a.

N.A. M(1) 1,2,3,4

b. Automatic Actuation- N.A. N.A. M.A. M(1) Q-h logic and~ Actuation g -

--4 Relays N.A. N.A. N.A. N.A. 1,.2, 3

% c. Containment Pressure- S R E.Q High-3 4

I v

F

.-- A I

,n ..

TABLE 4.3-2 (Continued)

ENGINEERED SAFs.TY FEATURF5 CTUATION SYSTEM' INSTRUMENTATION SURVElliANCE REQUIREMENTS U ~

6 TRIP 8 ANALOG ACTUATING MODES- !

MASTER SLAVE FOR WHICH i

CHANNEL DEVICE c ACTUATION RELAY RELAY SURVEILLANCE CHANNEL CHANNEL- OPERATIONAL OPERATIONAL i'i CALIBRAT_JN TEST TEST LOGIC TEST TEST TEST 15 REQUIRED CHECK d FUNCTIONAL UNIT w

c>3. Containment Isolation f "

a. Phase "A" Isolation N.A. R N.A. N.A. N.A 1, 2, 3, 4

1) Manual Initiation N.A. N.A.

N.A. M(1) M(1) Q 1, 2, 3, 4 N.A. N.A.

2) Automatic Actuation N.A.

Logic and Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

b. Phase "B" Isolation H.A. R N.A. N.A. N.A 1, 2, 3, 4 N.A. N.A.

$ 1) Manual Initiation M(1) 1, 2, 3, 4 N.A. N.A. N.A. M(1) Q Y 2) Automatic Actuation N.A.

O Logic Actuation Relays H.A. N.A. N.A. N.A. 1, 2, 3

3) Containment S R -M-Q Pressure-High-3

c. Containment Vent Isolation 1,2,3,4 N.A. N.A. N.A. M(1) M(1) Q

1) Automatic Actuation N.A.

Logic and Actuation Relays g

m m

E.

r .m -~

TABLE 4.3-2 (Continued) hG ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTG g- ' TRIP a

ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH c-CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE s's CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED d FUNCTIONAL UNIT

e. 3.c. Containment Vent Isolation (Cantinued)

2) Manual Phase "A" See Item 3.a.1 above for all manual Phase "A" Isolation Surveillance Requirements.

Isolation

3) Manual Phase "B" See Item 3.b.1 above for all manual Phase "B" Isolation Surveillance Requirements.

Isolation

4) Safety Injection See Item 1.:above for all Safety Injection Surveillance Requirements.

4. Steam Line Isolation

a. Manual Initiation H.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3 w

N.A. N.A N.A N.A. M(1) M(1) Q 1, 2, 3

b. Automatic Actuation-Y Logic and Actuation M Relays

c. Containaent Pressure- S --M-- Q N.A. N.A. N.A. N.A. 1, 2, 3 High-2

d. Steam Line Pressure- S Rf +Q N.A. N.A. N.A. N.A. 1, 2, 3 Low (Above P-11)

e. Steam Line Pressure 5 -M- Q N.A. N.A. N.A. N.A. 3

- Negative Rate - High g (Below P-11)

$5.TurbireTripandFeedwater 4 Isolation Automatic Actuation N.A. N.A. N. A. N.A. M(1) M(1) Q 1, 2

a. ~

5 Logic and Actuation Relay ,

^

Q O TABLE 4.3-2 (Continued)-

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E SURVEILLANCE REQUIREMENTS 6

8 TRIP i ANALOG ACTUATING MODES e CHANNEL DEVICE MASTER SLAVE 'FOR WHICH f5 CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE UFbHCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED 5.TurbineTripandFeedwater(Continued)g gg g Q u^

o. Steam Generator Water 5 RA -M-Q N.A. "8 -

"8 1, 2' l

Level-High-High (P-14)

c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

6. Auxiliary Feedwater

a. Manual Initiation N.A. N.A. H.A. R N.A. N.A. N.A. 1, 2, 3

{ 1,2,3 w b. Autonatic Actuation H.A. N.A N. A. N.A. M(1) M(1) Q

d. Logic and Actuation Relay

c. Steam Generator Water S Rk -ft-k N.A. N.A. N.A. M.A. 1,2,3 l Level-Low-Low

d. Undervoitage-RCP Bus H.A. R N.A. 3) N.A. N.A. N.A. 1, 2 i

e. Safety Injection See Item 1. above for all Safety Injection Surveillance Reqeirements.

f. Division 11 for Unit 1 N.A. R N.A. M(2,3) N.A. N.A. H.A 1,2,3,4 b (Division 21 for Unit 2)

ESF Bus Undervoltage

g. Auxiliary Feedwater S R M N.A. N.A. N.A. N.A. 1, 2, 3 Pump Suction Pressure-Low . . ,

E 9 . Automatic Opening of M Containment Sump Suction

] Isolation Valves h

..-..s w+ . . . . . ,

ENGINEERED SAFETY FEATURES ACTUA.,dM SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENIS

-3 TRIP-j ANALOG ACTUATING MASTER SLAVE MODES FOR WHICH 3 CHANNEL DEVICE

$ CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST 25 REQUIRED

7. Automatic Opening (Continued)

a. Automatic Actuation Logic and Actuation N.A. N.A. N.A. M(1) M(1) Q ] , 2, 3, 4 Relays N. A.

b. RWST Level-Low-tow 5 Rb '-M-tk N.A. N.A. N.A. N.A. 1, 2, 3, 4 l Coincident With Safety Injection See Item 1. above for all Safety Injection Surveillance Ret;uirements

8. Loss of Power o

~

a. ESF Bus Undervoltage N.A. R N.A. M(2,3) N.A. N.A. N.A. 1, 7, 3, 4 3 N.A. N.A. N.A. 1, 2, 3, 4

] b. Grid Degraded Voltage N.A. R N.A. M(3)

L 9. Engineered Safety Feature Actuation System Interlocks

a. Pressurizer Pressure, N.A. R -M- Q N.A. N.A. H.A. N.A. 1, 2, 3 P-11

b. Reactor Trip, P-4 N.A. N. . N.A. R N.A. N.A. N.A. 1, 2, 3

c. N.A. R -M- Q H.A. N.A. N.A. N.A. 1,2,3 Low-Low T,yg, P-12 d S t c c , O c .c r a t c r k'a t e r S R# M N.A. M(1) M(1) Q 1, 2

' k . '" g ,

i TABLE NGTATION

j (1) Each train shall he tested at least every 62 days on a STAGCERED TEST BASIS.

a Undervoltage relay operability is to be verified independently. An inoperable channel may be bypassed

!E(2) for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing of the.0PERABLE channel per Specificat. ion 4.3.2.1.

5(3) Setpoint verification is not applicable. g.

!h i

\g 3/4.3 INSi ><ENTATION k BASES J/4.3.1 and 3/4.3.2 REACTOR TRIP $3 TEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION The OPERABILITY of th'e Reactor Trip System and the Engineered Safety Features Actuation Systes instrumentation and interlocks ensures that: (1) the associated ACTION-and/or Reactor trip will be initiated when the p.rameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic -is-se4 eta 4*e+, (3) :;f ficf nt M=d=cy k =in-tat.md 4: ;;: =fbe-channe4-4: h ::t efn erv4<e-fee-testing-or-maintenano:,

and O) saf f4cient-sys-tem-functional-capability-is-eveilable-froediverse 1* ** * % 3'EE52T \

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall systes 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, gggg { -

The Engineered safety Features Actuation System InstrumentationT Tith Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

To accommodate the -instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated.

Allowable Values for the Setpoints have been specified in Table 3.3-4. Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel.when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 3.3-1, Z + RE + SE < TA, the -

i interactive effects of the errors in the rack and the sensor, ano the "as

! measured" values of the errors are considered. Z, as specified in Table 3.3-4, in percent span, is the statistical surnation of errors assumed in the analysis

excluding those associated with the sensor and rack drif t and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for the actuation.

RE or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified Trip Setpoint. SE or Sensor Error is either l

l -

BRAIDWOOO - UNITS 1 & 2 B 3/4 3-1

l l

LOS011#1:

New Bases Paragraph #1 (Add to existing paragraph.)

...and sufficient redundancy is maintained to aermit a channel *.o be out of service for testing or ma nienance consistent with maintairl..'g an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation and,3) rufficient system functions capability is available from diverse parameters.

Ins.ett#2:

New Bases Paragraph #2 (Add to existing paragraph.)

Specified surveillance intervals and surveillance and b maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System", and supplements to that report. Surveillance intervals and out of service times were c etermined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

InseILR3: f l ~

New Action to add to page 3/4 3-21 j

ACTION 15a - With the number of OPERABLE channels .

one less than the Total Number of Channels, declare the associated pump INOPERABLE and take the ACTION required by Specification 3.7.1.2.

ZNLD/615/121

- ___--_ -__-________________________--_-_____--____________________O

ATTACHMENT 3A EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS WCAP 10271- AND EDITORIAL CHANGES e Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards considerations.

According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

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

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

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

Because these changes are generically applicable to the Chapter

-15 rnalyses, the above attributes are being summarily addressed.

1.Does the change irsolve a significant increase the consequences of an accident previously evaluated? probability or The determination that the results of the proposed change are within all acceptable criteria have been estab'ished in the SERs prepared for WCAP-10271, WCAP-10271 Sup31ement 1, WCAP-10271 Su salement 2 and WCAP-1027' Supplement ~2, Revision 1 issuec. by_ References 1,2 and 5. Implementation of the proposed changes is expected to result in an acceptable increase in total Reactor Protection System yearly unavailability. This increase, which is pdmarily due to less frequent surveillance, results in an increase of similar magnitude in the probability of an Anticipated Transient Without Scram (ATWS) and in the probability of core melt resulting from an ATWS and also results in a smallincrease in core 3

damage fre uency (CDF) due to Engineered Safety Features Actuation S stem unavailability.

Implementation of the pro aosed changes is expected to result in a 1 significant reduction in the pro aability of core melt from inadvertent reactor trips. This !s a result of a reduction in the number of inadvertent reactor trips (0.5 fewer inadvertent reactor trips per unit 0 per year) occurring during testing of RPS instrumentation. This reduction is primarily attributable to less frequent surveillance.

L ZNLD/615/12?

l

The reduction in inadvertent core melt frequency is sufficiently large to counter the increase in ATWS core melt probability resulting in an overall reduction in total core melt probability.

The values determined by the WOG and presented in the WCAP for the increase in CP' were verified by Brookhaven National BF.J as . art of an audit and sensitivity analyses for the Laboratory NRC Staff. (Basso on the small value of the increase compared to the t range of uncertainty in the CDF, the increase is considered acco 3 table. The one plant-specific function evaluated on a plant spec fic basis for the Byron and Braldwood Nuclear Stations falls within the same crite-ia and is also considered to be acceptable.

The changes of an editorial nature have no impact on the severity or consequences of an accident previously evaluated.

Changes to Surveillance Test Frequencies for the Reactor Trip System Interlocks do not represent a significant reduction in testing.

The currently specified test interval for interlock channels allows the surveillance requirement to be satisfied by verifying that the aermissive logic is in its required state using the annunciator status ight. The surveillance as currently required only verifies the star _., ut the permissive logic and does not address verification of channel setpoint or operability. The setpoint verification and channel oaerability are verified after a refueling shutdown. The definition of the channel check includes comparison of the channel status with other channels for the sama parameter. The requirement to rcutinely verify permissive status is a different consideration than the availability of trip or actuation channels which are required to change state on the occurrence of an event and for which the function availability is more dependent on the surveillance interval. The change in curveillance requirement to at least once every 18 months does not therefore

. represent a significant change in channel surveillance and does not -

involve a significant increase in unavailability of the Reactor Protection System.

The proposed changes do not result in an increase in the severity or conseauences of an accident previously evaluated. Implementation of the proposed changes affects the probability of failure of the RPS but does not alter the manner in which protection is afforded nor the manner in which limiting criteria are established.

ZNLD/615/123

2.Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes do not involve hardware changes and do not result in a change in the manner in which the Reactor Protection System provides plant protection. No change is being made which alters the functioning o" the Reactor Protection System. Rather the likelihood or probability of the Reactor Protection System functioning properly is affected as described above. Therefore the proposed changes do not create'the possibility of a new or different kind of accident from any accident previously evaluated.

3.Does the change involve a significant reduction in a margin of safety.

The proposed changes do not alter the manner in which safety limits, limiting safety system setpoints or limiting conditions for operation are determined. The impact of reduced testing other than as addressed above is to allow a longer time ir'terval over which instrument uncertainties (e.g., drift may act. F Tm ace has shown that the initial uncertainty assumpt) ions are vallu for reduced testing.

Implementation of the proposed changes is expected to result in an overallimprovement in eafety by:

a. Less frequent testing will result in less inadvertent reactor trips and actuation of Engineered Safety Features Actuation System components.

b. -

Higher quality repairs leading to improved equipment reliability due to longer repair times,

c. improvements in the effectiveness of the operat!ng staff in monitoring and controlling plant operation. This is due to less frequent distraction of the operator and shift supervisor to attend to instrumentation testing.

The foregoing analysis demonstrates that the proposed amendment to Byron and Braidwood Nuclear Station Technical Specifications does not involve a significant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident from any accident previously evaluated, and does not involve a significant reduction in a margin of safety.

l ZNLD/615/124

REFERENCES:

1. Letter from C. O. Thomas (NRC) to J. J. Sheppard (WOG) dated February 21,1985 " Safety Evaluation by the Office of Nuclear Reactor Regulation WCAP-10271, Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System".

2. Letter from Charles E. Rossi (NRC) to Rugor A. Newton (WOG) dated February 22,1989 " Safety Evaluation by the Office of Nuclear Reactor Regulation Review of Westinghouse Report WCAP-10271 Supplement 2 and WCAP-10271 Su aplement 2, Revision 1 on Eva uation of Surveillance Frequenc es and Out of Service Times for the Engineered Safety Features Actuation System".

3. WCAP-10271 Supplement 1-P-A, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrument:tlon System", May 1986.

4. WCAP-10271-P A Supplement 2, Revision 1 " Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuttison System", May 1989.

5. Letter Charles E. Rossi NRC) to Gerard T. Goering (WOG) dated April 30,1990 (NRC Suo lementa! Safety Evaluation for WCAP-10271 Supplement 2, R vision 1).

6. Technical Specification Optimization Program, RWST Switchover Justification for Byron and Braidwood Nuclear Stations.

ZNLD/615/125

hpy o

s f ATI'ACHMENT 3B EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS AUXILIARY FEEDWATER PUMP SUCTION PRESSURE-LOW (TRANSFER TO ESSENTIAL SERVICE WATER)

Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards considerations.

According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

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

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

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

Specification Table 3.3-3, Functional Unit 6 9, Auxiliary Feedwater Pum a Suction Pressure-Low (Transfer to Essential Service Water)

Tota' Number of Channels, Channels to Trip, and Minimum Channels

- OPERABLE would be changed from two to one per train. The ACTION would also be changed such that if a channel were to become inoperable the associated AF pump would be declared inoperable and Specification 3.7.1.2 would be applied rather than placing the inoperable channel in the tripped condition and continuing operation until the performance of the next required ANALOG CHANNEL OPERATIONAL TEST.

The as-built plant configuration has only one suction pressure transmitter installed at the suction of each AF pump. A low suction l3ressure condition sensed by that transmitter in conjunction with an ESFAS actuation signal for its associated AF pump will initiate a transfer of the associated AF pump suction from the CST to the SX water sup aly. This actuation is train dependent and has a one-out-o -one actuation logic, i

ZNLD/615/126

The current ACTION allows for continued operation until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the triaped condition within one hour. Rather than reducing the actuation logic to one-out-o'-one, placing the channel in the tripped condition arms the transfer of the associated AF pump suction to the SX water supply. If the associated AF pump were to subsequently receive an ESFAS actuation signal, then the associated AF pump would start, its suction would be transferred to the SX water supply, and SX water would be injected into the steam generators. Injection of untreated SX water into the steam generators would have a devastating effect on secondary water chemistry and potentially shorten steam generator life. At a minimum, an extended outage would be ret utred for secondary water chemistry cleanup and evaluation of long term effects. In orc er to preclude this potential event from occurring, current operating practice is to place the control switch for the associated AF pump in the pull out position rendering that pump inoperable prior to placing the inoperable AF pump suction pressure channel in the tripped condition. Specification 3.7.1.2 then becomes limiting requiring the associated AF pump to be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY in the next six hours and in HOT SHUTDOWN within the following six hours.

The proposed ACTION would effectively impose a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed outage time (AOT) by invoking Specification 3.7.1.2. By not requiring the inoperable AF pump suction pressure channel to be placed in the tripped condition the control switch for the associated AF pump need not be placed in the pull out position leaving that pump available to manually or automatically respond in the event of an ESFAS actuation signal during the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AOT.

The effect on plant o aeration will be to increase the availability of the affected AF pump to respond manual y or automatically to an ESFAS actuation signal by not requiring the inoperable AF pump suction pressure channel to be placed in the tripped condition. The remainder of the changes only serve to reflect the as-built plant configuration and mimic current plant operating practice.

This change will have no effect on reactivity management.

This change will not affect the failure of AF pump suction pressure channels. "

However, should a channel fail the associated AF pump will be declared inoperable without placing the inoperable channel in the tripped condition. If the channel was inoperable for reasons other than f ailing low, then the AF pump, ahhough inoperable, would be available to respond manually or automatically in the event of an ESFAS actuation signal during the period of time the suction pressure channel is inoperable without injecting SX water into the steam generators.

1 l

ZNLD/615/127

The accidents that require AF system initiation are:

Inadvertent Opening of a Steam Generator Relief or Safety Valve, Steam System Piping Failure, Loss of External Load, Loss of Non-emergency AC Power to the Plant Auxiliaries, Loss of Normal Feedwater Flow, Feedwater System Pipe Break, Steam Generator Tube Rupture, Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary, and Anticipated Transients Without Scram.

The function of the AF system is the same for all accidents. The following is a generic discussion which applies to all accidents listed above.

The probability of an accident will not be increased by this change. This change is being made to accuratoly reflect the as-built plant configuration and to prevent an inadvertent injection of SX water into the steam generators should an AF pump suction pressure channel become inoperable.

The offsite dose consequences of previously analyzed accidents will not be

' increased if an AF pump suction aressure channel was to become inoperable and its associated AF pump was also dec ared inoperable, then the other 100% ca aacity train of AF would be ava table to automatically respond to an ESFAS actuation s gnal to mitigate the consequences of these accidents. This is consistent with the initial assumptions of the accident analyses.

- The probability of a malfunction of equipment important to safety is not affected by this changes. This change is being made to accurately reflect the as-built plant configuration. The probability of a failure of an AF pump suction pressure channel will not increase as result of this change. As a result ol this change if an AF pump suction pressure channel was to become Inoperable, then the associated AF pump will also be declared inoperable. This is consistent with the current operating practice that is required to ensure that SX water is not inadvertently injected into the steam generators when the inoperable AF pump suction pressure channelis placed in the tripped condition.

The consequences of a malfunction of equipment important to safety will not be increased. In the event that an AF pump suction pressure channel becomes inoperable and its associated AF pump :s subsequently declared inoperable, the other 100% capacity tralr of AF would be available to automatically respond to an ESFAS actuation signal to mitigate the consequences of these accidents. Additionally, the revised ACTION limits the window of vulnerability during which an accident could occur while in this degraded condition to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . And, by not placing the inoperable channelin the tripped condition, the inoperable AF pump may, in some circumstances, remain available to help mitigate the consequences of these accic%nts.

ZNLD/615/128

The possibility of a new or di'ferent kind of accident or malfunction is not introduced. This change does not introduce any new plant equipment or require any installed plant equipment to be operated in a different manner. Declaring the affected AF pump inoperable has no impact on the initial assumptions of these accident analyses.

This chan00 will not reduce the margin of safety. This chanDe is be!ng made to accurately reflect the as built plant configuration and to prevent an inadvertent injection of SX water into the steam generators s1ould an AF pump suction pressure channel become Inoperable consistent with current operating practice of declaring the affected AF pump inoperable, imposition of *e 72 hout AOT for the AF pump willlimit the window of vulnerability during whici, an accident could occur while in this degraded condition and will not reduce the margin af safety. The margin of safety will also not be reduced by not placing the Inoperable AF pump sucilon pressure channelin the tripped condition since the other 100% capacity train of AF would be available to automatically respond to an ESFAS actuation signal to mitigate the consequences of these accidents and under certain circumstances the inoperable AF pump would also be available to rospond manually or autonmtically to an ESFAS octuation signal.

ZNLD/615/129

ATTACHMENT 30 EVALUAT80N OF SIGNIFICANT HAZARDS CONSIDERATIONS RWST LEVEL CHANNEL CHANGE l Commonwealth Edison has evaluated this proposed amendment and determined that it inv0lves no significant hazards considerations. AccorJing to 10 CFR 50.92(c), a propostd amendment to an operating license involves no significant hazards considerations if operation of the fac lity in accordance vilth the proposed amendment would not:

1. Involve a significant incr9ase in the probability or consequences of an accident previously evaluateo; or !

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

3. Involve a sl0nificant reduction in a margin of esfoty.

The proposed change will clarify the Action Statement for an Inuperable RWST level channel. The current Action Statement requires an inoperable channel to be bypassed. Byron and Braldwood are configured such that an inoperable channel is

. removed from service by placing that channel in the irlaped conclltion. In order to

- bypass the channel, temporary jumpers must be instal ed, or a circuit card removed.

Testing In this configurat on is contrary to IEEE 279 and the SER associated with WCAP 10271. -

The new action statement will require RWST channels that are inoperable to be placed in a trioped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Operation may continue until the next required ANALOG CHANNEL Or)ERATIONAL TEST.

Each RWST is equipped with 4 level channels. TMs is one more than required by lEEE 2791971, because there are no control functions associated with the subject

' channels.-- Other than alarm functions, these level channels provide an input to the .

SSPJ. These inputs are associated with the semi automatic switchover of the ECCS to the containment recirculation sumps. Upon reaching a level of 46.7% on two of the four level chen*, concurrent with an SI signal, the containment sump suction valves . '

e ill open. Manuai action is then required to complete the realignment, which would isolate the RWST from the RH system. Prior to the colupletion of this realignment, the RH pum as take suction jointly from the containment sump and the RWST, i.e. the sump and the 3WST are crosstled.

While a channel is being surveilled or is otherwise inoperable, it is alaced in a tripped condition, consistent with the installed configuration. This resulLs in a 1/3 coincidence concurrent with an Si si

. _ suction valves. In this configuration,gnal full complianceto effect with IEEE the opening 2791971 of the containm is maintained.

This change has no effect on reactivity management.

For the period during which an RWST channel is ino 3erable, the f ailure of an

~ additional RWST channel (channel falls lowl will not resu t in the undesirable 03ening of the sump suction valves because an SI signalis also required. The affectec modes are Modes 1 inrough 4, which coincides with the modes of applicability for the affected l ECCS systems. Although an SI pump la required to be available in some Mode 5 and l 6 configurations ( during periods of reduced inventory operation), no credit for the semiautomatic switchover to the sump is assumed. ;

- ZNLD/615/130

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The accidents which result in an SI signal are:

Increased heat removal by the secondary, Feedwater line break, S:)urious SI, Tne range of LOCAs, and Certain ATWS scenarios.

The limiting accident is the Large Break LOCA, which results in the greatest demand for RWST inventory, and thus results in the need to switch the RH pump suction to the containment recirculation sump at the earliest time. This accident bounds all other transiento for the purposes of this change.

The prc,bability of an accident will not be increased by this change. The large Break LOCA is analyzed assuming a full complement of ECCS equipment, wl;h the subsequent failure of an entire train of this equipment. No allowance is made for the initiation of this transient with inoperable equipment, and it is recognized that the single failure criterion may not be met while operating under an Action Stalomont. The configuration used to remove inoperable RWST channels from service is unrelated to the probability that a catastrophic failure of the RCS piping will occur.

The offsite dose consequences cf an accident are not increased by this change. As analyzed the Large Break LOCA does not result in unacceptable offsite dose consec uences. In the event that a LOCA occurred with an RWST channel noperable the actuation logic to automatically open the sump suction valvos would be one out of three for the remaining operable RWST channels. This action would occur at the proper time assuming no failure of an additional RWST channel. It is rocognized that s!ngle failure criterion cannot always be met when in an Action Statement due to already identified inoperable oquipment.

The proposed revision will not increase the probability of a malfunction of eclulament important to safety. No change is being made to installec p ant equipment. The change is limited to method of removin; Tn inoperable RWST Level channel from service.

1' ZNLD/615/131

1 l

The consequences of a malfunction of equipment important to safety is unchanged. This change doea not render affected ec ulpment ,

vulnerable to a loss of suction, which could result in equipment failure.

l Multiple failures are required to resu;; ;n the undesirable transfer of RWST Inventory to the containment sump. The consaquences of a loss of suction to the ECCS pumps due to multiple failures in the switchover circuitry are no worse than the consequences of a loss of suction to the ECCS pumps due to other causes, such as RWST catastrophic failure or personnel error. It must be reemphasized that the scenario leading to a loss of suction event requires multiple failures, which is beyond the design basis for the plant.

This change does not create the possibility for a new or different kind of accident or malfunction from those previously evaluated No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. Placing an inoperable RWST level channel in the tripped configuration does not render the plant vulnerable to a loss of suction which would result in equipment unavailability.

The margin of safety is not adversely impacted by the proposed change. The proposed change deals with the configuration of an inoperable RWS1 level channel. There is no change in the point at which a switchaver of the ECCS pump suctions to the containment sumps it required. This change, as proposed, does not impact any analysis assum ations, and therefore, does not impact the analysis resu ts. As suc t the design margin of safety is unaffected.

ZNLD/615/132 -

ATTACHMENT 4 ENVIRONMENTAL ASSESSMENT Commonwealth Edison Company has evaluated the proposed changes and determined that:

1.The char.ges do not Involve a significant hazards consideration, 2.The changes do not involve a significant change in the types or sl0nificant increase in the amounts of any offluents inat may be released offsite, or 3.The changet, do not involve a significant increase in Individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the ollgibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, ptacuant to 10 CFR 51.22(b), an environmental ascessment of the proposed changes is not required.

ZNLD/615/133

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ML20099E070

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