ML20149E583

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Proposed Tech Specs,Removing Tables 3.3-2 & 3.3-5 Re Reactor Trip Sys Instrumentation & ESF Response Times,Respectively, to Be Placed in Chapter 16 of FSAR
ML20149E583
Person / Time
Site: Byron, Braidwood, 05000000
Issue date: 01/05/1988
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20149E524 List:
References
4051K, NUDOCS 8801130391
Download: ML20149E583 (37)
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Text

. _ _ _ _ _ - _ _ _ _ _ _ - - - -

ATTACHMENT _A.

_P_ROPOSED CHANGES TO APPENDIX A.

TECHNICAL SPECIFI, CATIONS, FACILITY OPERATING LICRNSES NPP-37, NPF-66, NPV-72, and NPF-75 Revised _P_ ages :

Byron Braidwood V V l B2-5 B2-5 3/4 3-1 3/4 3-1 l

3/4 3-7 3/4 3-7 l

l 3/4 3-8 3/4 3-8 3/4 3-13 3/4 3-13 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 B3/4 3-2 B3/4 3-2 4051K 88 12.30 h 9

LIMITING CONDITIONS FOR OPE' RATION AND SURVEILLANCE REQUIREMENTS l~

t SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.................................... 3/4 2-1 FIGURE 3.2-1 AXIAL FLUX OIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER................................ 3/4 2-3 3/4.2.2 NEAT FLUX HOT CHANNEL FACT 0R............................. 3/4 2-4 FIGURE 3.2-2 K(Z)-NORMALIZED Fq (Z) AS A FUNCTION OF CORE HEIGNT... 3/4 2-5 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTNALPY RISE HOT CHANNEL FACT 0R................................................. 3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-10 3/4.2.5 D N B ^ A RAM ET E R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-13 TABLE 3.2-1 D N B P A RAM ET ER S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-14 3/4.3 INSTRUMENTATION

.r 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2

-MBi+ A k2 REA0 TOR-TRW-SYSTEM INSTRUMENTAT!0ft-RESPONSE-T!MES. . . . 3/4--3-7 #

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE R E Q U I R EM E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTJATION SYSTEM INSTRUMENTATION........................................' 3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..................................... 3/4 3-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS. . . . . . . . . . . . . . . . . . . . . . 3/4 3 23 TABbi4-34--ENGINEERED-SAFET" FEATURES-RESPONSE 4IMES m . . . . . . . . . . -3/4-3 #

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........  ?/4 3-34

)

BYRON - UNITS 1 & 2 V

LIMITING SAFETY SYSTEM SETTINGS l BASES  !

1 Power Range, Neutron Clux, High Rates (Continued) l l

The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could I cause local flux peaking which could cause an unconservative local DN8R to exist. The Power Range Negative Rate trip will prevent this from cecurring by tripping the reactor. No credit is taken for operation of the Power Range  :

Negative Rate trip for these control rod drop accidents for which DN8Rs will '

be greater than the limit value. )

Intermediate and Source Range, Neutron Flux The Ir.termediate and Source Range, Neutron Flux trips provide core protection during reactor STARTUP to mitigate the consequences of an unconl yolled rod cluster control assembly bank withdrawal from a subcritical

.b e condition.vihese trips provide redundant protection to the Low Setpoint trip The Source Range channels will initiate of thea Power Reactor trip at about Neutron Range, los counts Flux per second channelsyj[R unless manually blocked when P-6 becomes active. The IntermediateT ange channels will initiate a

- Reactor trip at a current level equivalent to approximately 25% of RATED i THERMAL POWER unless manually blocked when. P-10 becomes active. -

3 in MODE 2. whi}e. -the. Source RGnSe., Nehon F)vx ,

Overtemoerature 4T h,P prov% phy pr&% k M coMj m j MODE S 3,4 and 5 The Overtemperature .iT trip provides core protection to prevent DN8 for all combinations of pressure, power, coolant temperatur?,, anti axial power distribution, provided that the transient is slow with respecc to piping 4 transit delays from the core to the temperature detectors (atout i seconds),

and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant '.emperature to correct for temperature induced changes in density and heat capacity of water and includes oynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressuri:er pressure, and (3) axial power distribution.

With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater 1 than design, as indicated by the difference between top and bottom power range

. nuclear detectors, the Reactor trip is automatically reduced according to the <

notations in Table 2.2-1.

. \

i l

BYRON - UN O 1 & 2 B 2-5 4

_ _. , ._ , _ . _ . _ . . . _ _ _ ._ r-, _ , , , , , , .,

I 3/4.3 INSTRUMENTATTON 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERAT!0N 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE w4ch RC PONSC TI."C  :: :he-:- i ell-.

Tabla 3.3-3. Pas APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

i l

SURVEILLANCE REQUIREMENTS , 4.3.1.1 Each Reactor Trip System instrumentation enannel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one' train such that both trains ar's tested at least once per 3d months and one channol per function sucn that all channels are tested at least once every N times la months where N is the total number of redundant channels in a specific Reactor trip function as shown in the "Total Nc. of Channels" column of Table 3.3-1.

O i

l 1

SYRON - UNITS 1 & 2 3/4 3-1 l 1

. i; TAlli F 3. 3-7

!\ REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES

/

/

g FUNCTIO UNII, RESPONSE TINE b! 1. Nanual Re tor Trip- N.A.

sr -

o- 2. Power Range, Ne ron Flux 10.5 seco ru

3. Power Range, Neutron ux, liigh Positive Rate ..
4. Power Range, Neutrea Flux, '

High Negative Rate 10.5 seconda

5. Intermediate Range, Neutron Flux N.A.

R

+

6. Scurce Range, Neutron Flux N.A.

{ 7. Overtempe.sture AT 14.0 seconds *# -

. 8 ., Overpower AT N.A.

1 i

9. Pressurizer Pressure-tow 12.0 seconds (Above P-7)
10. Pressurizer Pressure- gh 12.0 s onds
11. P.essurizer Wat Level-liigh N.A. -

(Above-P-7) -

  • Neutron electors are exempt from' response time test'ing. Response time of the neutron. flux signal portion of ti channel shall be measuted from detector output or input of first electronic component in channel.

ierma_l_. lag 2a<i RIILhypats_ man'3 fold delay times are not included..

_/

-- -- - 2 - - - - . - - - -.___-- - - - - ----_ - __- -- - - ---- -- --

{'s. TABLE 3.3-2 (Continued)

REAC10R TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES f

g FUNCIIONAL UNIT RESPONSE TIME

~4 m 12. Low Reacthrs Coolant, flow - Low s- a.

K -

Single Loops (Above P-8) <1.0 second e n b. Two toops (Atieve P- 7 'and below P-8) M 0 second N .

13. Steam Generator Water Level-Low-Low <

_2.0 seconds

14. Undervoltr.ge-Reactor Coolant Pumps (Above P-7) <1.5 seconds 15.

\

Underfrequerwy-Reactor Coolant. Fumps.s (Above P-7)

$0.6 second

16. Turbine Trip (A*>ove P-7) s
a. Emergency Trip ileader Pressure N. R.

, -o

b. Turbine Thrott1e Valve Closure H.h. .

17, $atety Injection Input from ESF N N.A.

N

18. Reactor Coolant Pawnp Breake Position Trip (Above P-7)

\ N N.A.

19. Reactor Trip System I t riocks M . A,,

N

20. Reactor Trip Bre rs N.A. .
21. Automatic Tr i and I.,terlock logic H.A.

l . .

/

t INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setcoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.:nd with RESPONCE T!ME3 :: the-:- in Teele 0. 3-5.o 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 Setpo4nt column but more conservative than the value shown in the Allowable Value colume of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value,
b. With an ESFAS Instrumentation or Inter 1ccx Trip Setpoint less con-servative than the value shown in the Allowable Values column of Table r 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2 .'

was satisfied fcr the affected channel, or

2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel 'ic restored to'0PERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 I + RE + SE i TA Where: .

Z = Th's value from Column I of Taele 3.2-4 for tne.affected channel, RE = The "as measured" value (in percent span) of rack error for the affected caannel, 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 affectec channel, and TA = The value from Column TA (Total Allowance) of Ta le 3.3-4 for the affected channel,

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

i

! BYRON - UNITS 1 & 2 3/4 3-13

g , - , ag-w a -g , , ~ y ,ng -,, , ,

TABLE _1 3-4 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONO

1. Manual Initiation

\ a.

b.

Safety Injection (ECC!)

Containment Spray N.A.

N. A.

. Phase "A" Isolation N. A.

Phase "3" Isolatich N. A.

e. Containment Vent Isolation N. A. 6
f. Steam Line Isolation N. A. }
g. (eeewater Isolation N. A.

Abxiliary feedwater l h.

N.A. l

i. Essential Service Water N. A/ l j Containment Cooling Fans N A. I j k. Start'Oiesel Generator A .'
1. Centrol\Soom Isolation N. A.

Turbine i

m. N.A.
2. Containment Pressyre-High-1 1
a. SafetyInjectlen(ECCS) <27(1)/1M5)
1) Reactor Tr'ip, {2

/

FeedwaterIsobation

2) <7(3) A
3) Phase "A" Isolation
  • h0) l

\ ,

4) Containment Vent 1541 (fen 17 ,
5) Auxiliary Feedwate 160
6) Essential Servfge Water <42(1)
7) Containment ling Fans <40(1)
8) Start Diep i Genencor 112
9) Contro ' Room Isolation N.A. l
10) TurfocTrip. N.A.

I

3. Pressurize / Pressure-Low
a. SafA'ty Inje, tion (ECCS)

\ II )/12(5),

'1  !

/)2)

Reactor Trip Feedwatar Isolation 12

<7(3) s 1

l

3) Phase "A" Isolation $2(0) <

\

4) Cantai nent Vant Isolation <7 l 1

i mor- aiim 4 --

- 374 :<3o - -=

\

-+-

1 l

i

TASLE,Nbi-4Gontinued) -

l ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECO 05

\ /

3. Pressurizer Pressure + Low (continued)

's 5) Auxiliary Feedwater 160

6) Essential Service Veter 142(1)
7) scontainment Cooling Fans

$40(1)/

8) dtart.DieselCenerator /

11?f

/

9) Control Roo's Isolation f. A,.
10) Turbine \ Trip N.A.
4. Steam Line Pressure 2.ow1 /
a. Safety Injection'(ECCS) 122O)/12(5)
1) Reactor Trip \ 12
2) FeedvsterIsoleti) 17(3) -

, 3) Phase "A" Isolation \ ,. 12 b}

4) Containment Vent Iso 'atkon 17
5) Auxiliary Feedwat r \ (60 6) 7)

Essentici Serv)c Wate'r ContainmentdoolingFans

'\ 142(1) 140(1) .

8) Start 0 el Generator <12 .
9) Contr i Roca Isolation N.A. I
10) Tur ine Trip # A.

\

b. Ste Line Isol:. tion 17 1
5. Cont.ai Ment Pressure-Hich-3  !
a. ontainnent Spray 145(1)
b. Phase "B" Isolation <22(1)/12C

~

l

6. Stam Generator Vater Level-Hich-High
a. Turbine Trip <2.5
b. Fee 6.ater Isolation g) I

--sVRON - UNFTS-1-4r 3/4 P 31 ~

JA8tE43=54Continu+d) m N

, s \ ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAt. AND FUNCTION RESPONSE TIME IN SECO S  :

7. 'Staam Generator Wa_ter Level-Low-Low
a. A Hotor-Driven Auxiliary

'Feedwater Pump

<60

b. O esel-Oriven Auxiliary F edwater Pumps $60
8. Containmerit Pressure Hich-2

\

Steam LIne Isolation 17

9. INST Level-l.ow- ow Coincident with Safety Injection ~

Automatic Open og of Containment 1100 Sump Suction Isolation Valves

10. Undervoltage_RCP Bus j
4. Motor-Oriven Awa: ilia eedwater mp 160 ,
b. Diesel-Driven Auxiliary Feedwa r Pump 160
11. Division 11 for Unit 1 (Divisions 1 for Unit 2) ESF Bus uncervoltage  ;

Motor-Driven Auxiliary eedwater ump 160 ,

12= Loss of Power

a. ESF Bus Undervolt e \

< 1. 9

b. GridDegradedVo). age 1310 m 30 delay
13. Steam Line Pressur/- Necative i Rate-Hicn (Selow -11) l Steam Lin ' Isolation 17 14 Phase "A" i, latin Con neent Vent Isolation 17 l
15. Auxill ry Fee 6*ater Purc Suction P ressiare-Lew-Low l

., Automatic Switchover to ESW N.A.

i

/

/

/ '

l SYiWM-UNITS-1-& c a/41T '

i J

N - ---7fYetM-PS--(tw4me& 1 TABLE NOTATIONS (1) Diesel generar.or starting and sequence loading delays included /

l (2) Diesel generator starting and sequence leading delay nel included.

/

Offsite power available.

(3 Hycraulic operated valves.

'(4) (eselgeneratorstartingandsequenceloadingdelayincluded. Only ;

i centrifugal charging pumps included. I (5) Dies \q generator starting and sequence lays loading not included.

l Offsita ilable. Only centrifugal cha ging pumps included.  !

(6)

Oces ti no\ lude power av,a valve closure time. -

~

l

\ -

N

's l

1 i

i'

'. \

N N-N \

4 e

SYROM - UNITS 1 a :  ;/4-3-33

'l INSTRUMENTATION I BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions.

Use of Equation 3.3-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

The methodology.to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertai,nties Sensor and ,

rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not i met its allowance. Being that there is a small statisitical chance that this +

will happen, an infrequent excessive drift is expected. Rack or sensor drif t, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

4 The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the

(

safety analyses. e credit n: t: hen 5 th: :n:1 f:r th::: chepelee.

vith 7::p:n: ti::: kdiceted ei net :ppli: M & y::: Response time may be demon-strated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response  ;

time as defined. Sensor response time verification .aay be demonstrated by '

either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. .

If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals '

to those Engineered Safety Features components whose aggregate function best '

serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate ,

the consequences of a steam line break or loss of coolant accident: (1) Safety  ;

Injection pumps start and automatic valves position, (2) Reactor trip, (3) feed-  ;

water isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position, (6) containment isolation, ,

(7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and I automatic valves position, and (11) essential service water pumps start and i automatic valves position. )

!(  !

I

)

3YRON - UNITS 1 & 2 B 3/4 3-2  ;

!u _

l.IMiiING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS _

PAGE SECTION  :

3/4.2 POWER DISTRIBUTION LIMITS s

3/4.2.1 AXIAL FLUX DIFFERENCE....... ............................ 3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMIiS~ AS A FUNCTION OF RATED THERMAL P0WER................................ 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R............................. 3/4 2-4 FIGURE 3.2-2 K(Z)-NORMALIZED F 9(Z) AS A ' FUNCTION OF CORE HEIGHT...

3/4 2-5 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNhL FACT 0R....................... ......................... 3/4 2-8 3/4.2.4 ' QUADRANT POWER TILT RATI0.........................,......

3/4 2-10 3/4.2.5 DNB PARAMETERS...........................................

3/4 2-13 TABLE 3 2-1 DNB PARAMETERS........................................ 3/4 2-14 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION. . . . . . . . . . . . . . . . . . . . . . 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION. . . . . . . . . . . . . . . . . . .

3/4 3-2

-TABtE-3-3PRf ACTOR-TRIP-SYSTEMiNSTRUMENTAT404-RESPONSE 41MES. . . .

-3/4-3-K l l

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................................ 3/4 3-9 f 3/4.3.2 ENGINEERED SAFETY FEATURES AC1UATION SYSTEM l INSTRUMENTATION........................................ 3/4 3-13 l TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM l INSTRUMENTATION..................................... 3/4 3-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS...................... 3/4 3-23

-T ABtf-3-3-5--ENGINEERED-S AFETY-FEATURES-RESPONSE-TIMES. . . . . . . . . . . . .

-3/4-3-30" TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM IN9TRUMENTATION SURVEILLANCE REQUIREMENTS........... 3/4 3-34 V

> BRAIDWOUD - UNITS 1 & 2

1 I

1 LIMITING SAFETY SYSTEM SETTINGS j 1

BASES 1

1 Power Range, Nei! tron Flux, High Rates (Continued)

The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the limit value.

Intermediato and Sourge Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor STARTUP to mitigate the consequences of an uncontrolled rod cluster control assembly bank withd.awal fram a subcritical 6 A of iondition? /hese tript provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels / The Source Range channels will l

, initiate a Reactor trip at about 105 counts per second unless manually blocked when Pd6 becomes active. The Intermediate Range channels will initiate a Reactor trip et a current level equivalent to approximately 25% of RATED THERMAL POWER unle w minu111y blocked when P-10 becomes active.

% hade 2. A;l<. Ae. % ec e. Ru m e. , NsJcon Flu tr@ powdes p % c for fla. ecre. ;n Overtemperature AT Mod e s 3, 4 and C.7 pof e.c hn The Overtemperature aT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping i transit delays from the core to the temperature detectors (about 4 secords),

and pressure is within the range between the Pressurizer High and Low Pressure I tript. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays froai the core to the loop tenperature detectors, (2) pressurizer pressure, and (3) axial power distribution.

With normel axial power distribution, this Reactor trip limit is always below ,

the core Safety Limit as shown in Figure 2.1'1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically redexed according to tiie notations in Table 2.2-1. ,

i r

l 1

i BRAIDWOCD - UNITS 1 & 2 B 2-5

, .- - . . w , n-- , - - - - . - - . . , - - . a., - , - - -

- - - , a , -w. -

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE,eth-RESPONSE--T4HES-as-shown-in- "

-Tabl e-3+2+

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1 1 Each Reat +n- Trip System instrumentation channel and interlock and

  • ne automatic trip <]ic shall be demonstrated OPcRABLE by the performance of the Reactor Trip Sysiam Instrumentation Surveillance Requirements specified in Table 4.3-1.

4.3.1.2 The RFACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channeis are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.

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

V TABLE 3.3-2 N b

o REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES l5 o

FUNCTIO A1._ UNIT RESPONSE TIME

1. Manual tor Trip N.A.

,E 4

2. Power Range, Ne ron Flux 50.5 second*

]

  • 3. Power Range. Neutron ux,

" High Positive Rate N.A.

4. Power Range, Neutron Flux, High Negative Rate 10.5 second*
5. Intermediate Range, Neutron Flux N.A.

Source Range, Neutron Flux N.A.

g 6.

Y 7. Overtemperature AT 14.0 seconds *# f

~ .

8. Overpower AT N.A.

l

9. Pressurizer Pressure-Low -<2.0 seconds (Above P-7)
10. Pressurizer Pressure-H's 12. seconds
11. Pressurizer Water evel-High N.A. ~

(Above P-7) i

\

  • Neutron tectors are exempt from response time testing. Response time of the neutron fl signal portion of th channel shall be measured from detector output or input of first electronic component ' channel.

hermal lag and RTD bypass manifold delay times are not included.

~

1

r-- ,

TABLE 3.3-2 (Ccntinued)

E R REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES 2

8 o

FUNCTIONAL UNIT RESPONSE TIME E 42. Low React olant Flow - Low

~

a a.

x Single Loop ( bove P-8)

_ .0 second

b. Two Loops (Abo P-7 and below P-8)
e. 11.0 second N
13. Steam Generator Water Le a -Low-Low , 12.0 secor.ds
14. Undervoltage-Reactor Coolant ps (Above P-7)

, 11.5 sr.conds

15. Underfrequency-Reactor Coolant Pum (Ab P-7) 10.6 second
16. *~-^ Trip (Above P-7)

R

'.xrgency Trip Header) essure N.A.

y -

iurbine Throttle V hie Closure N.A.

co

17. Safety Inject i ut from ESF N.A.
18. Reactor CooJa t Pump Breaker Position Trip (Above P-7) N. A.
19. React / Trip System Interlocks b
20. ctor Trip Breakers N.A .

21 Automatic Trip and Interlock Logic N.A.

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INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2* The Engineered Safety Features Actua+. ion System (ESFAS) instrumentation

, channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.,and-wftHESPONSE-T4ME4-es-shown-k-Tabk-3v3-5 e - .

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 Vait -alumn of Table 3.3-4 adjust the Setpoint consistent with the Trip Sstpoint value.
b. With an ESFAS Instrumentation or Interlock Trip Setpoint less con-servative than the value shown in the Allowable 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or
2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

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

Z = The value from Column Z of Table 3.3-4 for the affected channel, RE " ihe "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 affected channel,

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

"Control Room isolation not required prior to initial criticality on Cycle 1.

Auxiliary Building Ventilation actuation not required prior to initial opera-tion at > 20% Rated Thermal Power (RTP) on Cycle 1.

BRAIDWOOD - UNITS 1 & 2 3/4 3-13

T

~

TABLE 3.3-5

~

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITLATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual Initiation
a. afety Injection (ECCS) N.A.
b. tainment Spray N.A.

~ c. P e "A" Isolation

d. N.A.

Pha "B" Isolation N.A.

e. Conta qment Vent Isolation N.A.
f. Steam Li e Isolation N.A.
g. Feedwate solation N.A.
h. Auxiliary edwater N.A.
i. Essential S vice Water j Containment ooling Fans N.A.

N.A.

k.

1.

StartDieselGherator N.A.

Control Room Isolation N.A.

m. Turbine Trip N.A.
2. Containment Pressuro-High 1
a. SafetyInjection(ECCS) 27(1)/12(5)
1) Reactor Trip 12
2) Feedwater Isolation 17(3)
3) Phase "A" Isolation 12(6)
4) Containment Vent Isolation /

17

5) Auxiliary Feedwater 160
6) Essential Service L.'ater 142(1)
7) Containment Cooling ns (40(1)
8) Start Diesel Gener tor 1-
9) Control Room Is lation N.A.
10) Turbine Trip / N.A.
3. Pressurizer Pressur4-Low
a. Safety Inje tion (ECCS) 5'/(1)/12(

2

1) Reactor Trip <2
2) edwater Isolation 17 (3) 3 Phase "A" Isolation 12(6)
4) Containment Vent Isolation 17

' AIDWOOD - UNITS 1 & 2 3/4 3-30 k

w w

f TABLE 3.3-5 (Continued)  :

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIA NG SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

3. Pre izer Pressure-Low (continued)

I

. 5) uxiliary Feedwater 160 ~

6) Es ntial Service Water 142(1)
7) Con ment Cooling Fans $40(1)
8) Start Die el Generator 112
9) Control Room Isolation N.A.

T4.cbine Trip N. .

10)

4. Steam Line )ressure-Low
a. Safety Iniact 4n (ECCS) 122(4)/12(5)
1) Reactor Trip 12 l
2) Feedwater Isolation 17(3) ,
3) Phase "A" Isolation $2(6) i
4) Containment Vent Isolatio - 17
5) Auxiliary Feedwater 106  !
6) Essential Service W ser 142(1)
7) Containment Cooli g Fans <40(1)
8) Start Diesel nerator 12
9) Control Roo Isolation H.A.
10) Turbine rip N.A.
b. Steam Line Isolation <7
/

Containment fressure-High-3 5.

j a. Co ninent Spray 1 45(1)

b. P ase "B" Isolation <22(1)/12(2)
6. StegmGeneratorWaterLevel-High-Hig

. Turbine Trip <2.5

b. Feedwater Isolation 17(3) l

, _ BRAIDWOOD - UNITS 1 & 2 3/4 3-31

~

, , _ . _ - _ _ _ . ~

h - --

TABLE 3.3-5 (Continued)

^

ENGINEERED SAFETY FEATURES RESPONSE TIMES IN TIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS

7. MteamGeneratorWaterLevel-Low-Low
a. Motor-Driven Auxiliary Feedwater Pump b, 160

, esel-Driven Auxiliary Fe dwater Pumps 160

8. Containmen Pressure-High-2 Steam Li e Isolation 17
9. RWST Level-Low-L_w Coincident with Safety Injection Automatic Openihg of Containment Sump Suction Isol tion Valves 1100
10. Undervoltage RCP Bus
a. Motor-Driven Auxiliar Feedwater Pump
b. 160 Diesel-Driven Auxiliary Feedwater Pump 560
11. Division 11 for Unit 1 (Divisi n 21 for Unit 2) ESF Bus Undervoltage Motor-Driven Auxiliary Feedwat r ump 160
12. Loss of Power
a. ESF Bus Undervoltage < 1. 9
b. Grid Degraded Voltage {310130 delay
13. Steam Line Pressure - Negative Rate-High (Below P-11)

Steam Line Iselati 1

14. Phase "A" Isolation -

Containment Vent Isolation $7 l

15. Auxiliary Feedw'ater Pump Suction Pressure-Low-Low

/

Automatic Switchover to ESW N.A. l I

BRpIDWOOD-UNITS 1&2 3/4 3-32

l TABLE 3.3-5 (Continued)

TABLE NOTATIONS (1 Diesel generator starting and sequence loading delays included.

(2) iesel generator starting and sequence loading delay not includ d.

Of ite power available.

(3) Hydr lic operated valves.

(4) Diesel gAnerator starting and sequence loading delay in uded. Only centrifugh charging pumps included.

(5) Diesel generk or starting and sequence loading dela s not included.

Offsite power ailable. Only centrifugal chargi pumps included.

(6) Does not include alve closure time.

/

/

BR DWOOD - UNITS 1 & 2 3/4 3-33 I

INSTRUMENTATION

_4 BASES REACTOR TRIP SYSTEM and ENGICEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

L the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions.

Use of Equation 3.3-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudet, Pack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statisitical chance tiiai, this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses. -No credit was-taken in l.he-analyses-for-those-channet W

-Wth-response-times-indicated-as-not-appl 4 cable & Response time may be demon-strated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to conhinations indicative of various accidents, events, and transients. Once the required logic combination is complated, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident: (1) Safety t Injection pumps start and automatic valves position, (2) Rcactor trip, (3) feed-water isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pump start and automatic valves position, (10) containment cooling fans start and automatic valvas position, and (11) essential service water pumps start and j automatic valves position. .

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BRAIDWOOD - UNITS 1 & 2 B 3/4 3-2 l

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ATTACHMENT B PROPOSED CHANGES _TO BYRON /BRAIDWOOD.

PGAR C1(APTER 16 Revised Pages:

16.0-1 16.1-1 16.2-1 16.3-1 16.3-2 16.3-3 16.3-4 16.3-5 16.3-6 16.3 *1 16.3-8 4051K

B/B-FSAR AMENDMDJT C11 APTER 16.0 - TECIUJICAL SPECIFICATIONS TALLE OF CONTENTS PAGE 16.0 TECIDJICAL SPECIFICATIONS 16.1-1 16.1 PRELIMINARY TECIDJICAL SPECIPICATIONS 16.1-1 16.2 PROPOSED FINAL TEC10JICAL SPECIFICATIONS 16.2-1 16.3 TECIDJICAL SPECIFICATION IMPROVEMENT PROGRAM 16.3-1 16.3.1 Reactor Trip System Instrumentation 16.3-1 Response Times 16.3.2 Engineered Safety Features Response Times 16.3-1 16.0-i P

(1399M/0165M)

B/B-FSAR AMENDMDIT CllAPTER 16.0 - TECIDJICAL SPECIFICATIONS 16.1 PRELIMINARY TECIDJICAL SPECIFICATIONS Refer to Section 16.2 4

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B/B-FSAR AMENDMDJT 16.2 PROPOSED FINAL TECIDJICAL SPECIFICATIONS The Technical Specifications for Byron /Braidwood Stations are issued by the NRC as Appendix A to the Operating License. Because of their evolving nature during original definition, and their subsequent revision during plant operations, the Technical Specifications are placed in a separate document.

This deviation f rom Regulatory Guide 1.70 (Revision 2) on FSAR format and content is justified on the basis of NRC's proof and Review Draft of the Byron Technical Specifications, dated December 16, 1983.

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B/B-PSAR AMENDMDIT 16.3 TECIDJICAL SPECIFICATION IMPROVEMENT PROGRAM Regulatory Guide 1.70 does not address a FSAR Section 16.3 but this section has been created to address the Technical Specification Improvement Program.

The Technical Specification Improvement Program has resulted in the inclusion of certain technical requirements into the FSAR. These items are provided as follows.

16.3.1 Reactor Trip System Instrumentation Response Times The response time of each Reactor Trip function shown in Technical Specification Table 3.3-1 shall be as shown in FSAR Table 16.3-1. No credit was taken in the analyses for those channels with response times indicated as not applicable.

16.3.2 Engineered Safety Features Response Times The response time of each Engineered Safety Feature Actuation System function shown in Technical Specification Table 3.3-3 shall be as shown in FSAR Table 16.3-2. ESF response times specified in Table 16.3-2 which include sequential operation of the RWST and VCT valves (Notes 4 and 5) are based on values assumed in the non-LOCA safety analyses. These analyses take credit for injection of borated water from the RWST. Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening of the RWST charging pump suction valves. When the sequential operation of the RWST and VCT valves is not included in the response times (Note 7), the values specified are based on the LOCA analyses.

The LOCA analyses take credit for injection tiow regardless of the source.

Verification of the response times specified in Table 16.3-2 will assure that the assumptions used for the LOCA and non-LOCA analyses with respect to operation of-tne VCT and RWST valves are valid. No credit was taken in the analyses for those channels with response times indicated as not applicable.

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B/B-FSAR AMENDMENT TABLE 16.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

1. Manual Reactor Trip N.A.
2. Power Rango, Neutron Flux $ 0.5 second*
3. Power Range, Neutron Flux, N.A.

High Positive Rate

4. Power Range, Neutron Flux, i 0.5 second*

High Negative Rate

5. Intermediate Range, Neutron Flux N.A.
6. Source Range, Neutron Flux 1 0.5 second*
7. Overtemperature OT 1 4.0 seconds *#
8. Overpower OT N.A.
9. Pressuriser Pressure-Low i 2.0 seconds (Above P-7)
t. .
10. Pressurizer Pressure-High 1 2.0 seconds
11. Pressurizer Water Level-High N.A.

(Above P-7)

12. Low Reactor Coolant Flow - Low w

~

a. Single Loop (Above P-8) i 1.0 second
b. Two Loops (Above P-7 and below P-8) i 1.0 secon1
13. Steam Generator Water Level-Low-Low i 2.0 seconds
14. Undervoltage-Reactor Coolant Pumps (Above P-7) i 1.5 seconds
15. Underfrequency-Reactor Coolant Pumps (Above P-7) $ 0.6 second
  • Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
  1. Thermal lag and RTD bypass manifold delay times are not included.

16.3-2 (1399M/0165M)

l B/B-PSAR AMD1DMDIT TABLE 16.3-1 (Continued) l I

REACTOR TRIP SYSTEM INSTRUMDITATION RESPONSE TIMES j l

l PUNCTIONAL UNIT RESPONSE _TIbE l 1

16. Turbine Trip (Above P-7)
a. Emergency Trip Header Pressure N.A.
b. Turbine Throttle Valve Closure N.A.
17. Safety Injection Input from ESP N.A.
18. Reactor Coolant Pump Breaker N.A.

Position Trip (Above P-7)

19. Reactor Trip System Interlocks N.A.
20. Reactor Trip Breakers N.A.
21. Automatic Trip and Interlock Logic N.A.

N 16.3-3 (1399M/0165M)

B/B-PSAR AMENDMENT TABLE 16.3-2 ENGINEERED SAFETY FEATURES RESPONSE TIMIS INITIATING SIGNAL AND FUNCTIC'N RESPCNSE TIME IN SECONDS

1. Manual Initiation
a. Safety Injection (ECCS) N.A.
b. Containment Spray N.A.
c. Phase "A" Isolation N.A.
d. Phase "B" Isolation N.A.
e. Containment Vent Isolation N.A.
f. Steam Line Isolation N.A.
g. Feedwater Isolation N.A.
h. Auxiliary Feedwater N.A.
i. Essential Service Water N.A.
j. Containment Cooling Fans N.A.
k. Start Diesel Generator N.A.
1. Control Room Isolation N.A.
m. Turbine Trip N.A.
2. Containment Pressure-liigh-1
a. Safety Injection (ECCS) i 27(7)/27(5)
1) Reactor Trip 'a,. 52
2) Feedwater Isolation i 7(3) .
3) Phase "A" Isolation i 2(6)
4) Containment Vent Isolation i7 -

s

5) Auxiliary Feedwater i 60
6) Essential Service Water i 42(1)
7) Containment Cooling Fans i 40(1)
8) Start Diesel Generator i 12
9) Control Room Isolation N.A.
10) Turbine Trip N.A.

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1 16.3-4 (1399M/0165M)

J

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. i B/B-PSAR AMENDMENT TABLE 16.3-2 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES l INITIATING SIGNAL AND FUNCTICN RESPONSE TIME IN SECONDS

3. P_ressurizer Pressure-Low
a. Safety Injection (ECCS) i 27(7)/27(5)
1) Reactor Trip i2
2) Feedwater Isolation 1 7(3)
3) Phase "A" Isolation i 2(6)
4) Containment Vent Isolation i7
5) Auxiliary Feedwater i 60
6) Essential Service Water i 42(1)
7) Containment Cooling Pans i 40(1)
8) Start Diesel Generator i 12
9) Control Room Isolation b. N.A.
10) Turbine Trip N.A.
4. Steam Line Pressure-Low
a. Safety Injection (ECCS) 1 37(4)/27(5
1) Reactor Trip i2
2) Feedwater Isolation 1 7(3)
3) Phase "A" Isolation i 2(6) i
4) Containment Vent Isolation i7
5) Auxiliary Feedwater 1 60 l l

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6) Essential Service Water i 42(1)
7) Containment Cooling Fans i 40(1) 16.3-5 l

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B/B-FSAR AMDJDMDIT TABLE 16.3-2 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TDIE IN SECONDS

4. Steam Line Pressure-Low (Continued}
8) Start Diesel Generator i 12
9) Control Room Isolation N.A.
10) Turbine Trip N.A.
b. Steam Line Isolation i7
5. Containment Pressure-lligh_-3
a. Containment Spray i 45(1)
b. Phase "B" Isolation i 22(1)/12(2)
6. Steam Generator Water Level-liigh-liigh
a. Turbine Trip i 2.5
b. Feadwater Isolation '.

$ 7(3)

7. Steam Generator Water Level-Low-Low
a. Motor-Driven Auxiliary'Feedwater Pump i 60
b. Diesel-Driven Auxiliary Feedwater Pumps i 60 s
8. Containment Pressure-High-2 Steam Line Isolation i7
9. RWST Level-Low-Low Coincident with Safety Injection Automatic Opening of Containnent i 100 l Sump Suction Isolation Valves I I'
10. Undervoltage RCP Bus
a. Motor-Driven Auxiliary Feedwater Pump i 60 I
b. Diesel-Driven Auxiliary Feedwater Pump i 60

\

16.3-6 l i

(1399M/0165M)

B/B-FSAR AMENDMENT TABLE 16.3-2 (Continued)

ENGINEERED SAFETY FEATURES RESPCNSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

11. Division 11 for Unit 1 (pivision 21 for Unit 2) ESP Bus Undervoltage Motor-Driven Auxiliary Feedwater Pump i 60
12. Loss of Power
a. ESF Bus Undervoltage i 1.9
b. Grid Degraded Voltage 1 310 1 30 delay
13. Steam Line Pressure - Negative Rate-High (Below P-ll)

Stea.n Line Isolation i7

14. Phase "A" Isolation Containment Vent Isolation i7
15. Auxiliary Feedwater Pump Suction Pressure-Low-Low Automatic Switchover to ESW N.A.

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16.3-7 (1399M/0165M)

B/C-FSAR AMENDMElfr TABLE 16.3-2 (Cor.tinued)

TABLE NOTATIONS

1. Diesel generator starting and sequence loading delays included.
2. Diesel generator starting and sequence loading delay ngt included.

Offsite power available.

3. liydraulic operated valves.
4. Diesel generator starting and sequence loading delay included. Only centrifugal charging pumps included. Sequential transfer of centrifugal charging pump suction from the VCT to the RWST (CV112D and E open, then CV112B and C close) is included.
5. Diesel generator starting and sequence loading delays not included.

Offsite power available. Only centrifugal charging pumps included.

Sequential transfer of centrifugal cl ging pump suction from the VCT to the RWST (CV112D and E open, then CV112B and C close) is included.

6. Does not include valve closure time.
7. Diesel generator starting and sequence loading delaya included.

Sequential transfer of centrifugal charging pump suction from the VCT to the RWST (CVil2D and E open, then CV112B and C close) is not included.

Only the opening of CV112Diand E is included, s

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1 16.3-8 (1399M/0165M)

_ _ . - ._ __ _~. _ . _ . . _ _ _ _ _ . . . . _

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ATTACHMENT _C BACKGROUND INFORP!AT.I.ON 4

i As part of the technical specification improvement effort. '

Commonwealth Edison proposes to remove Table 3.3-2, "Reactor Trip System Instrumentation Response Times", and Table 3.3-5, "Engineered Safety Features Response Times", from the Byron and Braidwood Technical Specifications. These tables will be placed in a new section of the Byron /Braidwood FSAR, Chapter 16.3. The Technical Specifications will still contain a surveillance requirement to verify that the reactor trip system and engineered safety features response times are within their limits at least once per 18 months. l

(

If a response time limit is not met, the affected channel is declared inoperable. The Technical Specifications will still contain the action ,

requirements to be implemented when a channel is declared inoperable. [

Therefore, sufficient control will remain in the Technical Specifications to s

! ensure response times are periodically checked and instrument operability is maintained. (

A change to a value listed in the new tables in Chapter 16.3 of the FSAR would require a 10 CPR 50.59 evaluation to be performed. The 10 CFR

  • 50.59 evaluation involves a review of the three items in 10 CPR 50.59(a)(2) to j determine if the proposed change involves an unreviewed safety question. l i Beyond the 10 CFR 50.59 review, Byron and Braldwood Station administrative  :

! procedures also require a reviisw to evaluate the impact of the change on j Technical Specifications, license conditions and other potential interfacing areas such as mechanical, electrical, and instrumentation and control  ;

1 systems. These administrative procedures also require the Stations to ,

consider the acceptability of the proposed change relative to other 10 CFN, as J well as NRC Regulatory Guides, s

j' Subsequently, an on-site review of a change to these tables would be  !

performed in accordance with Technical Specification 6.5.2. The on-site '

4 review will involve a minimum of two station personnel, one of whom holds a senior reactor operator license. If the on-sito review finds the change acceptable, it will be submitted to the Station Manager or his designated I alternate for final approval. Once approved by the Station Manager, assuming no unreviewed safety question exists from the 10 CFR 50.59 review, the change to a value listed in these FSAR tables will be available for implementation in l the plant. The changed value will also be incorporated into the next update of the FSAR.

1 The process outlined above will ensure a thorough review is performed

on proposed changes to items relocated from the Technical Specifications to i the FSAR.

I 4051K

ATTACHMENT D EVALUATION OF STGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards consideration. According to 10 CFR 50.92(c), a proposed aniendment to an operating license involves no significant hazards consideration if operation of the facility !n accordance with the 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) Involve a significant reduction in a margin of safety.

This proposed amendment removes tables 3.3-2 and 3.3-5 from the technical specifications. These tables list the reactor trip system instrumentation response timeL and engineered safety features response times.

The tables will be placed in a new section of the Byron /Braidwood PSAR, Chapter 16.3. The response times listed in the tables are not being changed by this proposed amendment. The requirements to periodically measure the response times will rumain in the technical specifications.

Relocating the reactor trip and engineered safety features response time tables from the technical specifications to the FSAR is an administrative change. The response times assumed in previously evaluated accidents are not being changed by this proposed amendment. As a result, the probability or consequences of previously evaluated accidents are not affected by this amandment, similarly, this administrative change does not create the possibility of a new or different kind of accident, nor does it affect a margin of safety.

For the reasons stated above, Commonwealth Edison believes this proposed amendment involves no significant hazards consideration.

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4051K

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