Text

Proposed Tech Specs Table 2.2-1 & Table 3.3-4 Re Sys Instrumentation Trip Setpoints,Ts 3.4.1.2 Re Reactor Coolant Sys & TS 4.4.1.2.1 Re Surveillance Requirements
	ML20134N848
Person / Time
Site:	Byron, Braidwood
Issue date:	02/18/1997
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20134N841	List: ML20134N838; ML20134N848;
References
	NUDOCS 9702250006
	Download: ML20134N848 (40)
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._ __ . . _ __ _ _ _ . _ _ _ . ._ .. _ . . . _ _ . . . _ . . _ _ _ _ _ _ __ . _ _ _ .

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINTS ALLOWA8tE VALUE mma .

gg

12. Reactor Coolant Flow-Low 290% of loop mini-m 289.3% of loop alpi-M num measured flow num measured flow gh 13. Steam Generator Water

/2

'fP';"'I"'1'If i

No Level Low-Low 7 0- / .w y NJ l ? W l h 41 'l ' a NW 3 6 6.i , (pc m. .o c pIc .1) ) 4 j,,,,,,,,,, , ,, y , _ , q ,. . s t

O$ a. Unit 1 g

m6-4

/ >_ 18.0 % (.ye(<-

't .oJ .Jh /

range h-233.0%

instrument- of-narrow-

-range. instrument a (~ I- 'E3E0 ggS b. f M .m c c.m . <

q -

-5 pan- -span-Unit 2 236.3% of narrow ;

t ,,.,,,,,_,st

. , , , n

' 234.8% of narrow n

range instrument range instrument

_ span span i

14. Undervoltage - Reactor 25268 volts -

Coolant Pumps 24920 volts - '

each bus each bus

' 15. Underfrequency - Reactor 157.0 Hz Coolant Pumps 256.08 Hz

16. Turbine Trip

a. Emergency Trip Header 21000 psig Pressure 2815 psig

b. Turbine Throttle Valve 21% open 21% open closure

17. Safety Injection Input N.A. N.A.

from ESF

18. Reactor Coolant Pump N.A. M.A.

Breaker Position Trip

Minimum measured flow = 92,850 gpe'-(97i600-gpa)**~

- **Not- Applicable to Unit 1. Applicable to Unit 2 until completion of cycle 5.-

--# Applicable to Unit 1.- Applicable to Unit 2 after cycle 5.

BYRON - UNITS 1 & 2 2 =5 ANFEMENT NO. 65

~

t a,

TABLE 3.3-4 (Continued) l

. ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRtBENTATI% TRIP ..

k FUNCTIONAL UNIT TRIP ALLOW 48LE

a SETPOINT VALUE w 4. Steam Line Isolation e.

. a. Manual Initiation !

M.A. N.A.

b. Automatic Actuation Logic ;

and Actuation Relays .;

M.A. M.A.

c. Containment Pressure-High-2 18.2 psig 19.4 psig

d. Steam Line Pressure-

. Low (Above P-11) )640 psig*

_ >614 psig* !

1 !'

w e. Steam Line Pressure 4

Negative Rate-High (Below P-11) 1100 pst** $165.3 psi **

. . _ ~

~

\ K!

5. Turbine Trip and 433.4%sf ii Feedwater Isolation f

[d yl.'?'l (pfict cigk To C.t, .bc

'f at(LT) i

/!

a. Automatic Actuation Logi:

and Actuation Relays (ogmshwenT wgecto5PA. avp

~, e :

- N.A. N.A. /

,_ r

. b. St:;am Generdar Shier -' /

Level-High-High (P-14) v/o(pKicx To t,@e p

1) Unit 1

' 4 3l .

i

-[ g,p%(c.t4cle. (7 M ah_81.45-et- 183.45-of-g p,q wngA0 'GHF e

narrow-range g Q . narrow-range instrument. instrument.

t N 5 Pan-M *Paa-

% 2) Unit 2 Q g ( ,~w~ g ~~'_ 's y-p * / s80.8x of ss2.as of g narrow range narrow range '

instrument instrument '

span span ,

a TABLE 3.2-4 (Continued)

ENGINEERED SAFETY FEATURES ACNATION SYSTEM INSTRt#ENTATION T i k FUNCTIONAL UNIT TRIP ALLOWA8LE -

a SETPOINT VALUE -

g 5. Turbine Trip and Feedwater Isolation (continued)

N '

c. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values. -

6. Auxiliary Feedwater .-

y

a. Manual Initiation S

b. Automatic Actuation M.A. M.A

/ > eq 31.67)(O M I Logic and Actuation y g(.,[ 'f;( CTh9 s

Relays M.A. N. A.

'M # y' ' -

g' (-

c. Steam Generator Water w '

..' ' T W to M T(GMC Level-Low-Low-Start l h Motor-Driven Pump and /

~'> 33.6Th (PN b 9ej' q) -

1MT#*T~ ~ ~ ~ ~~

Diesel-Driven Pump '

1) Unit 1 2 W goj' QM eje.'i o*1 ( b )/ h '~

c,Q mon'CA0 MO im%wumT ->31 E -er

/ $~>33.as-ef narrow-range- narrow-range-

2) Unit 2 x 3\M ,y- / -instrument

. span-instrument-span- ,

L 136.3E of 134.8K of narrow range narrow range instrument instrument span span

d. Undervoltage-RCP Bus-

% Start Motor Driven Pump 1526f, volts 14920 volts

= and Diesel-Driven Pump

e. Safety Injection-Start Motor-g Driven Pump and See Ites 1. above for all Safety Injectica Trip Setpoints and Diesel-Driven Pump Allowable Values.

10

REACTOR COOLANT SYSTEM HOT STANDB(

LIMITING C0pDITION FOR OPERATION _

l 3.4.1.2 At least two of the reactor coolant. loops listed below shall be '

t OPERABLE with two reactor coolant loops in operation when the Reactor Trip j System breakers are closed and one reactor coolant loop in operation when the j Reactor Trip System breakers are open:"

. a. Reactor Coolant Loop A and its associated steam generator and l reactor coolant pump, i

! b. Reactor Coolant Loop 8 and its tssociated steam generator and reactor coolant pump, 1 c. Reactor Coolant Loop C and its associated . steam generator and l reactor coolant. pump, and -

i d. Reactor Coolant Loop D and its associated steam generator and i reactor coolant pump.

4

, APPLICABILITY: MODE 3.**

! ACTION:

} !

, a. With less than the above required reactor coolant loops OPERABLE, l i

restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be i

! in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.. With only one reactor coolant loop in operation and the Reactor Trip

System breakers in the closed position, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months open the Reactor l Trip System breakers

c. With no reactor coolant loop in operation, suspend all operations l

! involving a reduction in boron concentration of the Reactor Coolant l l System and immediately initiate corrective action to return the l required reactor coolant loop to operation.

1 SURVEILLANCE REQUIREMENTS -

4. 4. L 2.1 At least the above required reactor coolant pumps, if no't in operation, shall be determined OPERABLE once per 7 days by verifying correct 1 i breaker alignments and indicated power availability. l 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side narrow range water level to be greater than or equal at least o ce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

to IS 5'c C 4l'7c 4e u.M t pact t 41%-Jor-Unit-14185-for-Unit-2}i c.qcle.

4.4.1.2.3 The required coolant loops shall e verified in operation and circu-lating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"All Reactor Coolant pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no operations are permitted that would cause dilution of the Reactor Coolant System baron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

- See Special Test Exceptions Specification 3.10.4.

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

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.3.2 The required steam' generator (s) shall be determined OPERABLE by verifying secondary side narrow range water level to be greater than or equal to 41%-for-Unit-1-(485-for-Unit-21 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

/71 (ytff Et LhqT~ ( pf tcr ic Cyde. 9) 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant at l o st once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l l

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

REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED 1

LIMITING CONDITION FOR OPERATION I 1

I

s. 4.1. 4.1 At least one. residual heat removal (RHR) loop shall be OPERABLE and 1 in operation *, and either:

l

a. One additional RHR loop shall be OPERA 8LE#, or I

b. The secondary side narrow range water level of at least two steam generators shall be greater than 41%-for-Unit-1d-18%-for-Unit-2), s

/5%(y/ K -Fef u'AT I prior To cple. 9)

APPLICABILITY: MODE 5 with reactor coolant loops filled ##.

~

ACTION: -

a. With one of the RHR loops inoperable and with less than the required steam generator level, immediately initiate corrective action to return the-inoperable RHR loop to OPERA 8LE status or restore the required steam generator level as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and c,irculating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ,

The RHR pump may be deenergized for up to I hour provided: (1) no operations are permitted that would cause' dilution of the Reactor Coolant System boron

concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature. ,

0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing i

provided the other RHR loop is OPERABLE and in operation.

1. A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 350*F unless the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

i BYRON - UNITS 1 & 2 3/4 4-5

m ATTACHMENT B-la MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A, IMPROVED TECHNICAL SPECFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF-66 BYRON STATION UNITS 1 & 2 REVISED PAGES 3.3-16 3.3-35 3.3-36

-3.4-10 3.4-13 3.4-16 B 3.4-29 B 3.4-34 B 3.4-37 B 3.4-40 i

1 4

l 4

L:nf a .byrbwd.sgrp:sgrievia. doc: l l

RTS Instrumentation 3.3.1 a

Table 3.3.1 1 (page 3 of 6)

Reactor Trip System Instrtmentation APPLICA8LE MODES OR OTHER SPECIFIED RIOUIRED SURVEILLANCE ALLOWA8tE CHANNELS CONDITIONS REQUIREMENTS VALUE FUNCTION CON 0!TIONS

12. Unoervoltage II 'I 4 J SR 3.3.1.9 SR 3.3.1.10 a 4920 V RCPs (per train)

SR 3.3.1.14

13. Ov,erfre:!uency l i 'l 4 J SR 3.3.1.9 SR 3.3.1.10

= 56.08 Hz RO's ;per train)

SR 3.3.1.14 gd

>ns' 14 Steam Generator (SG) [7 (5104.06(d Water Level - Low Low (per SG) u 4) _

/

a. Unit 1 1.2 4 0 SR 3.3.1.1 % = 31.01 1 SR 3.3.1.7 SR 3.3.1.10

/foA %

1 V ce +C )

SR 3.3.1.14 o 4. Qs D. Unit 2 1.2 4 D SR 3.3.1.1 ' a 34.81 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.14 l

l

',' 15 Tur cine Trio i a Emergency Trip I U3 SR 3.3.1.10 a 815 osig i 3 L Neace" Pressure SR 3.3.1.13

, toer train) o Tur oine Tneettle I III 4 L SR 3.3.1.10 = It open valve Closu re SR 3.3.1.13 j toe trains i.

l- M Sa'e:y Iaje:t'e- ($!i 1.2 2 trains N SR 3.3.1 12 NA a 1-c.,: * :e le;snee*e Sa'e:y Fea:/e a::sa:'.o-i;ra; Systee 1.2 2 trains O SR 3.3.1.a NA 8ea:*:* ;G'; t SR 3.3.1.12 E ease *s' l 3 I83 4(83 5(83 2 trains R SR 3.3.1.4 NA SR 3.3.1.12 i

i (continued)

(as atin Roc Control System cacaole of rod withdra al or all rods not fully inserted tel Acove tne P 7 (Lo= Po er Reactor Trips Block) interloct (f) ADove tre D 6 (Power Range Neutron Flux) interlock.

(g) *locluoing any reacto* trip byoats Dreaters that are racked in and closed for bypassing an RTB.

1 i

BYRON - UNITS 1 & 2 3.3-16 o.e m ier a

ESFAS fnstrumentation.

, 3.3.2 Table 3.3.2 1 (page 4 of 5)

Engineered Safety Feature Actuation System Instrtmentation APPLICA8LE MODES OR OTHER SPECIFIED RIOUIRED SURVEILLANCE ALLOWA8tE FUNCTION CON 0!TIONS CHANNELS CON 0!TIONS REQUIREMENTS VALUE

5. Turbine Trip and Feedwater Isolation

a. Automatic 1.2I9) 3(9)

. 2 trains ! SR 3.3.2.4 NA Actuation Logte SR 3.3.2.5 and Actuatton SR 3.3.2.7 Relays

b. Steam Generator (SG) Water Level - High Hign (P 14)

1) Unit i 1,2(9) 3(9) 4 per SG F SR 3.3.2.1 s 83.41

~~

SR 3.3.2.4 A_

SR 3.3.2.5 SR 3.3.2.6 " (#SL SR 3.3.2.7 *9 i SR 3.3.2.10 SR 3.3.2.12 i ya9c 't'/.)

+ t., ',

wHal -

2) unit 2 1.2(9) 3(9) 4 per SG F SR 3.3.2.1 w82 81 SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6 SR 3.3.2.7 SR 3.3.2.10 SR 3.3.2.12 c Safety lejection Refer to Function 1 (Safety injection) for all initiation functions and requirements.

l (Continued) ig. [a:e;; am?' a!' recstred fee'aate*

c Isolation Valves are closed or isolated by a close0 manual, valve 1

i f

k BYRON - UNITS 1 & 2 3.3-35 Rcvicion A

A 1 ESFAS Instrumentation i 3.3.2 .

i' Table 3.3.2 1 (page 5 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR ALLOWABLE REQUIRED SURVEILLANCE OTHER SPECIFIED REQUIREMENTS VALUE CHANNELS CONDITIONS FUNCTION CON 0!TIONS s

6. Auxiltary Feedwater l ! SR 3.3.2.4 NA

a. Automatic Actuation 1.2.3 2 trains 4

SR 3.3.2.5 i Logic and Actuation . SR 3.3.2.7 Relays j

1 $

9 e.e4 c.Nj

b. SG Water Level -Low Low 4 per SG F SR 3.3.2.1 a 31.02

$ 1) Unit 1 1.2.3 SR 3.3.2.6 6N-i SR 3.3.2.10 4 w IV '.

SR 3.3.2.12 e. u4 Sj) /

d d

4 per SG F SR 3.3.2.1 = 34.8%

2) Unit 2 1.2.3 SR 3.3.2.6 SR 3.3.2.10 SR 3.3.2.12 i

c. Safety Injection Refer to Function 1 (Safety injection) for all Initiation functions and recuirements.

1 F SR 3.3.2.3 a 2730 V

~

1.2.3 2 c oss

. of Offstte Power SR 3.3.2.10 i (Unce* voltage or' SR 3.3.2.11 Bas 141(241)1 l

4 0 SR 3.3.2.8 a 4920 V e unce* voltage Rea: tor 1.2 SR 3.3.2.10 Cociaet Dco (per SR 3.3.2.12 ;

tratn) I 7

1 per train M SR 3.3.2.1 a 2* Mg va: 1

A,. 'ia , Feec. ate" 1.2.3 SR 3.3.2.2 l Ae: Ss:t'or Transfee SR 3.3.2.10 !

4

1.:D oc j '*e:ss a? .c.

I !.-*: eve' :: ;ca:4 w ent j ". r: ,

1.2,3.4 2 trains C SR 3.3.2.a NA

4 8 .w r: A:t e ':a SR 3,3 2.5 j . :-: a*: A:t atter SR J.3.2.7

-+ as:

4 4 0 SR 3.3 2.1 a 44.7%

. : e 1.2.3.4 SR 3.3.2.6 l j :e* # rfa.atrMnST)

M : ase ma SR 3.3.2.10 (

1 . eve - L o. t o. SR 3.3.2.12 l i

'r:*:ea: .itn Refer to Function 1 (Safety Injection) for all intttation functions and requirements.

< Sa e:., inje:p on i

}

4 k

i BYRON - UNITS 1 & 2 3.3-36 h isicrJA 4

?

d

RCS Loops-MODE 3 3.4.5 i ,

i SURVEILLANCE RE0VIREMENTS (continued)

FREQUENCY SURVEILLANCE Verify steam generator secondary side 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.5.2 narrowrangewaterlevelis=33%doreach ___j i required Unit 1 RCS loop (= 37% for each

required Unit 2 RCS loop). :

I i

7 days !

l SR 3.4.5.3 Verify correct breaker alignment and indicated power are available to each required pump that is not in operation.

, . i e 4e % <.i %c sl 2 \f % 7'

) (b (qu~ i cal cMu -

t. /

l 1

1

(

3 4

f T

4

)

9 3.4-10 Revision A BYRON - UNITS 1 & 2

'T

RCS Loops-MODE 4 3.4.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FRE00ENCY SR 3 4.6.2 Verify SG secondary side narrow range water 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s level _isA33%sfor each required Unit 1 RCS loop (= 37% for each required Unit 2 RCS l J

1 loop).

i SR 3.4.6.3 Verify correct breaker alignment and 7 days indicated power are available to each

required pump that is not in operation.

4 a

c h ?.cc \c Hus <<

'7)"-4w%uJcdke u -

/

l J

I

~

l l

I l

l l

I 1

l i

l BYRON - UNITS 1 & 2 3.4-13 Revisjor, p 1

1

3 RCS Loops-MODE 5. Loops Filled 3.4.7 SURVEILLANCE-REQUIREMENTS i FREQUENCY !

SURVEILLANCE SR 3.4.7.1 Verify required RHR loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.'4.7.2 -

NOTE Only required when complying with LC0 3.4.7.b.

Verify SG secondary side narrow range water 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months level is = 33 Lfor each required Unit 1 RCS loop (= 37 Ffdr each required Unit 2 RCS loop).

SR 3.4.7.3 Verify correct breaker alignment and 7 days indicated power are available to each required.RHR pump that is not in operation.

Mt k.- % yek. c\ an\ z \? % ~ ,

kc 'tg ch#t ..sJ ch.- f' 4

1 i

J f

4 BYRON - UNITS 1 & 2 3.4-16 Reti:icn A

- l

RCS Loops-H0DE 3 B 3.4.5 BASES

(%Pno.-

va g'.

w e4 9 [JIi3% G enD /

SURVEILLANCE SR 3.4 5.2 ~~ >

REQUIREMENTS uired SG OPERABILITY.

(continued) SR 3.4.5.2 requires verification o that the secondary SG OPERABILITY is verified by ensuri .

side narrow range water level is = 33 'for Unit 1 (= 37% for }

Unit 2) for each required RCS loop. If the SG secondary gm M siife narrow range water leveF is : 33% for Unit 1 (< 37% for dy" become uncovered and the associated

@"N_,tMt-e), removalthe loop may not tubes be capab mafe of providing the heat sink for of the decay heat. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to a loss of SG l level.

SR 3.4.5.3 Verification that the required RCP is OPERABLE ensures that l l

an additional RCP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. ;

l Verification is performed by verifying proper breaker alignment and power availability to the required RCP. The l Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be ecceptDie by operating experience.

REFERENCES 1. UFSAR, Section 15.4.1. s I

i i

l l

P~;1:1 = ^

BYRON - UNITS 1 & 2 8 3.4-29 3 m_.

4 RCS Loops-MODE 4 I B 3.4.6 l l

BASES ACTIONS C.1 and C.2 !

(continued)

If no loop is OPERABLE, all operations involving a reduction i

of RCS baron concentration must be suspended and action ta

- restore one RCS or RHR loop to OPERABLE status must be initiated. Boron dilution requires forced circulation to provide proper mixing and preserve the margin to criticality. The inr.ediate Completion Times reflect the ,

importance of maintaining the capability for decay heat 2

4 removal.

. SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required operating RCS or RHR loop is in operation.

Verification may include flow rate, tem)erature, or pump status monitoring, which helps ensure t1at forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS and RHR loop performance. l f,e.IMc%adt,,J2.13% ./

SR 3.4.6.2 L.%- a S w wb-SR 3.4.6.2 requires verification of f required SG OPERABILITY.

i SG OPERABILITY is verified by ensurilig' that the secondary side narrow range water level is a 33Ffor Unit 1 (= 37% for i

-~- Unit 2) for each required RCS loop. If the SG secondary 4e+7 side narrow range water level"is 4-33%-for Unit 1 (: 37%-fe gM"*wUnt4),loopthe maytubes not ba may become capable uncovered of providing and the the heat sink associated necessary

4

' > for removal of decay heat. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adeque.te in view of other indications available in the control roos to alert the operator to the loss of SG level.

i 2

(continued)

J B 3.4-34 t/icien ,^

BYRON - UNITS 1 & 2

O RCS Loops-MODE 5. Loops Filled B 3.4. 7 ~ !

i

~

BASES (continued)

APPLICABLE In MODE 5. RCS circulation increases the time available for SAFETY ANALYSIS mitigation of an accidental boron dilution event. The RHR loops provide this circulation and have been identified as important contributors to risk reduction.

RCS Loops-MODE 5. Loops Filled satisfies Criterion 4 of i 10 CFR 50.36(c)(2)(ii).

LCO The surpose of this LCO is to require that at least one of the UiR loo)s be OPERABLE and in operation with an additional RHR loop OPERABLE or two SGs OPERABLE. One RHR loop provides sufficient forced circulation to aerform the

.' safety functions of the reactor coolant under t1ese .

conditions. An additional RHR loop is required to be OPERABLE to provide adequate redundancy for heat removal. !

However, if the standby RHR loop is not OPERABLE. an f '# ' r acceptable alternate method is two SGs with their secondary

~~

9 t *c W 1 ~ Tide Tiater leveWF33%4for Unit 1 (= 37% for Unit 2) and [

J' W I W2 L- with their associated RCS loops filled. Should the f

' had u operating RHR loop fail, the SGs via natural circulation

. % t m.. - could be used to remove the decay heat.

~' Note 1 permits all RHR pumps to be removed from operation ,

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit tests designed to validate various accident analyses values. One of the tests performed during the startup testing program is the validation of rod dro

' cold conditions. both with and without flow.p Thetimes no flowduring test may be performed in MODE 3. 4. or 5 and requires that the pumps be stopped for a short period of time. The Note permits stopping of the pumps in order to perform this test ,

and validate the assumed analysis values. If changes are i

1 made to the RCS that would cause a change to the flow

< characteristics of the RCS. the input values must be revalidated by conducting the test again. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time l period is adequate to Jerform the test and operating i

experience has shown tlat boron stratification is not likely

- during this short period with no forced flow.

l (continued) l 8 3.4-37 R~jision A BYRON - UNITS 1 & 2

RCS Loops-MODE 5. Loops Filled B 3.4.7

~

BASES ACTIONS C.1 and C.2 (continued)

If no RHR loop is OPERABLE and one or both of the required SGs are inoperable, all operations involving a reduction of RCS baron concentration mutt be suspended and action to restore one RHR loop to OPERARLE status must be initiated.

Boron dilution requires forced circulation to provide proper mixing and preserve the margin to criticality. The immediate Com31etion Times reflect the importance of maintaining t1e capability for heat removal.

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required operating RHR loop is in operation. Verification l may include flow rate. temperature. or pump status I monitoring, which helps ensure that forced flow is providing )

heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance. ~

[e.c - k w e 'i e, ,j SR 3'4'7'2 #* k 9"9 >

Ldy s Verifying that at least two SGs are OPERABLE by ensuring their secondary side narrow range water levels are = 33% Afor l Unit 1 (= 37% for Unit 2) ensures an alternate decay heat removal method via natural circulation in the event that the second RHR loop is not OPERABLE. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to the loss of SG level. This SR is modified by a Note which indicates that if both RHR loops are OPERABLE. this Surveillance does not need to be satisfied.

(continued)

BYRON - UNITS 1 & 2 8 3.4-40 Revision ,^

l ATTACHMENT B-2 MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 NPF-77 BRAIDWOOD STATION UNITS 1 & 2 REVISED PAGES; 2-5 3/4 3-25 3/4 3-26 3/44-2 3/44-4 3/44-5

) )

j

.i j

I I

T i 4

knla byrbwd.spp splevladoc:12

, . . - m.--. . ,

1ABLE 2.? (Continued)

REACTOR 1 RIP SYSTEN INSTRUNENTATION TRIP S(190lNTS

!f*

FUNCTIONAL UNIT TRIP SETP0ffLT ALLOWABLE VALUE 4

12. Reactor Coolant Flow-Low 190% of loop wini- 189.3% of loop mipi- _ '

mum measured flow num measured flow

13. Steam Generator Water level Low-Low

' ~s ~5g,/, 6d,(pucT j g (c,tpfe. 'lL Cepic.

y m4t c iMF) M 2 is.cs (cyJe a e

>(g p c,d,je, g')g g'g) }f , gtww mg. mstrmoe(wi ,

SFTp, v

a. Unit I gfg gm y 6wimumi,.233.0% of rrow 231.0% of narrow- :

9 range _ inst 2ent range-instrument '

( _. -- span- span-

b. Unit 2 217% (Cycle 3); 216.35-(Cycle 3); ,

236.3% (Cycle 234.8% (Cycle <4 and and after) of .aftert of narrow narrow range range instrument instrument span span

14. Undervoltage - Reactor 25268 volts -

Coolant Pumps 24920 volts -

each bus each bus

15. Underfrequency - Reactor 157.0 Hz Coolant Pumps 256.08 Hz

16. Turbine Trip

a. Emergency Trip Header 21000 psig Pressure 1815 psig '

b. Turbine Throttle Valve 21% open Closure 21% oPen ,

17. Safety Injection Input N.A. N.A.

from ESF

18. Reactor Coolant Pdep M.A. M.A.

Breaker Position Trip

Ninimum measured flow - 97,600 gpm (92,850 gpm)

- Applicable to -Unit-1-and- Unit-2-until completion of cycle- 5. '

Applicable-to Unit 1 and Unit-2-starting with cycle 6.

UNIT l'~- AMENOMENT NO. 56 BRAIDWOOD - UNI 15 1'& 2 2-5

~

UNIT.2 - AMENOMENT NO.. 55 .,

l

i s'

TABLE 3.3-4 ..ontinued)

. ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS i =

I D TRIP ALLOWA8LE -

f o

FUNCTIONAL UNIT

4. Steam Line Isolation SETPOINT VALUE C

a. Manual Initiation N.A. N.A.

5 a

b. Automatic Actuation H Logic and Actuation -

8* Relays M.A. N.A.

m

c. Containment Pressure-High-2 18.2 psig 19.4 psig 3

d. Steam Line Pressure-Low (Above P-11) >640 psig* >614 psig*

i e. Steam Line Pressure w Hegative Rate-High 1100 psi **, ___ $165.3. psi ** s

) (Below P-11) y 5. Turbine Trip and 'M W3.'l'1(PTick ~lo c'ylt d ,

y Feedwater Isolation 6 T ].it<'/c, ( ci g le. 5 o n t ( M E h

a. Automatic Actuation s._

("N "I ^

" W

,5 Logic and Actuation - .

Relays s y l,ti'],(gmtg 'Ib tgk. 9 N.A. N.A. /

b. Steam Generator Water $ SS.07c.(Ctyk. 3 ed (S .

Level-High-High (P-14) i g wow scugc. twfisc

1) Unit 1 SNM . _ (, 181.4%-of- 183.45-of-

' - - - - 7 narrow range narrow range-instrument instrument span- span-g

! E 2) Unit 2 <78.1% (Cycle <79M(Cycle-3};

3); <80.8% 782.85 (Cycle-4-and k (CycTe and lif ter) of narrow 3 after) of range instrument narrow range span instrument A span 3 , . , . , , , , , , -,.,e.. - , . , - , my . ~ - ,- ~ - , , ,,.,-.._-~,,-,..w . , , ,,=,m .~,,..--,y - ~ . ..-w . , , --

TABLE 3.3-4$bentinued) l

, ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS l

I C

~ TRIP ALLOWABLE SETPOINT VALUE FUNCTIONAL UNIT

, 5. Turbine Trip and Feedwater Isolation (continued)

E Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and y c.

e Allowable Values. ~

6. Auxiliary Feedwater I

)

7 '

a. Manual Initiation N.A. N.A.I{}3lfd(ptvrle-

. cy, g)

b. Automatic Actuation g 1 (sI ,} ( (' e cle Y logic and Actuation Relays N.A. N.A. 5 M dgikt)')d-,

c. Steam Generator Water - lu:Ounemf Spm ) -

i Level-Low-Low-Start / - s ~

/

R Motor-Driven Pump and / 2 3 OT LIN'T .I - hPbN) --

g g<g,g y',( Cty.le 3 oxd oklar) \

Diesel-Driven Pump p, E 1) Unit 1 \ O Y 'br W U W T M .\C 1

M " W-M) >33.0%-of >31.0% of.

narrow range narrow range '

3~ \ % instrument instrument span span.

2) Unit 2 >17% (Cycle >16.3%-(Cycle 3);

3); >36.3% 534.8% (Cycle 4 (CycTe 4 and and after) of after) of narrow range narrow range instrument instrument span g span z

E d. Undervoltage-dCP Bus- ->5268 volts >4920 volts y Start Motor Driven Pump and Diesel-Driven Pump B

e. Safety Injection-e start Motor-Driven Pump and See Item 1. above for all Safety Injection Trip Setpoints and Diesel-Driven Pump Allowable Values.

REACTOR COOLANT SYSTEM HOT STANDBY LINITING CONDITION FOR OPERATION f

3.4.1.2 At least two of the reactor coolant loops listed below shall be ;

OPERA 8LE with two reactor coolant loops in operation when the Reactor Trip 1 System breakers are closed and one reactor coolant loop in operation when the Reactor Trip System breakers are open:*

a. Reactor Coolant Loop A and its associated steam generator and l reactor coolant pump, l

b. Reactor Coolant Loop 8 and it: associated steam generator and reactor coolant pump,

c. Raactor Coolant Loop C and its associated steam generator and reactor coolant pump, and j d. Reactor Coolant Loop D and its associated steam generator and

! reactor coolant pump.

I j APPLICABILITY: MODE 3.**

ACTION

i

a. With less than the above required reactor coolant loops OPERABLE, i restore the required loops to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be j in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With only one reactor coolant loop in operation and the Reactor Trip

System breakers in the closed position, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months open the Reactor i Trip System breakers. l

[ c. With no reactor coolant loop in operation, suspend all operations i involving a reduction in boron concentration of the Reactor Coolant l System and immediately initiate corrective action to return the 1- required reactor coolant loop to operation.

I SURVEILLANCE' REQUIREMENTS i 4.4.1.2.1 At least the above required reactor coolant pumps, if not in j operation, shall-be determined OPERABLE once per 7 day: by verifying correct j breaker alignments and indicated power availability.

t' 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side narrow range water level to be greater than or equal to 41%-fee-Unit-1--(185-feNinit-2-)- at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

15%(,Mi % fcN U.vhT I pucx To C.upic O 4.4.1.2.3 The required coolant loops shall be verified in operation and circu-

, lating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l "All Reactor Coolant pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet teinperature is maintained at least 10*F below saturation temperature.

, **See Special Test Exceptions Specification 3.10.4.

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

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.3.2 The required steam generator (s) shall be detereined OPERABLE by verifying secondary side narrow range water level to be grester than or equal to 41%-for- Unit-1-(18%-for-Unit-2) at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , l % ( 41% 90s DeniT I ptet To t,y cle. S )

4.4.1.3.3 At least one reactor coolan't or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 1

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

_ _ - _ _O

REACTOR COOLANT SYSTEM l

COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, .and either:

a. One additional RHR loop shall be OPERA 8LE#, or

b. The secondary side narrow range water level of at least two steam generators shall be greater than 41%-for-Unit-1-(-18%-for-Unit-2). . s IS%(4f % oY LWT ! prict Ic C.(CM S /

4 APPLICA8ILITY: MODE 5 with reactor coolant loops filledN.

ACTION:

a.- With one of the RHR loops inoperable and with less than the required steam generator level, immediately initiate corrective action to return the inoperable RHR loop to OPERA 8LE status or restore the required steam generator level as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months .

4.4.1.4.1.2~ At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1

The RHR pump may be deenergized for up to I hour provided: (1) no operations i

are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

0ne RNR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.

MA reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 350*F unless the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

1 i BRAIDWOOD - UNITS 1 & 2- 3/4 4-5

ATTACIIMENT B-2a MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX AIMPROVED TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72, NPF-77 BRAIDWOOD STATION UNITS 1 & 2 REVISED PAGES; 3.3-16 3.3-35 3.3-36 3.4-10 3.4-13 3.4-16 B 3.4-29 B 3.4-34 B 3.4-37 B 3.4-40 1

1 J

k:nttbybwd:sgrp.splevladoc:13

I RTS Instrumentation i 3.3.1 l i

Taele 3.3.1 1 (page 3 of 6)

Reactor Trip System Instrumentation APPLICA8tE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWA8LE FUNCTION CON 0!TIONS CHANNELS COM)lTIONS REQUIREMENTS VALUE i

12. Undervoltage 1(') 4 J SR 3.3.1.9 a 4920 V l RCPs (per train) SR 3.3.1.10 4 SR 3.3.1.14 l

'13. Underfrequency 1(') 4 J SR 3.3:1.9 a 56.08 Hz RCPs (per train) SR 3.3.1.10 SR 3.3.1.14 ,

, # ~N, !

14, Stean Generator (SG) #N Water Level - Low Low (per SG) f hidL I(M J l

{ cs4py j

a. Unit 1 1.2 4 0 SR 3.3.1.1 = 1.01 SR 3.3.1.7 6 1 SR 3.3.1.10 W u +,o SR 3.3.1.14 4d 4. "$),/

b. Unit 2 1.2 4 0 SR 3.3.1.1 in 34.81 SR 3.3.1.7 SR 3.3.1.10 i SR 3.3.1.14 l l

I

15. Turbine Trio

a. Emergency trip 1(f) 3 L SR 3.3 1.10 a 815 psig Header Pressure SR 3.3.1.13 (per train)

e. Turoine Throttle 1(f) 4 L SR 3.3.1.10 a 11 open Valve Closure SR 3.3.1.13 (per trains i

i 16 Safety injection (SI) 1.2 2 tra?ns N SR 3.3.1.12 NA InDut from Engineereo Safety Feature

.Act ation System

j. (E5' AS i 1'

! 17 Rea:to ' lo 1.2 2 trains O SR 3.3.1.4 NA i Brease's '

SR 3.3.1.12 3I8) a(a) S I8) 2 trains A SR 3.3.1.4 NA SR 3.3.1.12

(continued) i (a) Witn Roc Control System capable,of rod withdrawal or all rods not fully inserted.

(e) Acove tne P 7 (Low Power Reactor Trips Block) interlock.

j (f) Acove tne P+8 (Power Range Neutron Flux) interlock.

(g) Inclucing any rea: tor trip bypass breakers that are racked in and closed for bypassing an RTB.

l f

BRAIDWOOD , UNITS 1 & 2 3.3-16 Rsisica A

j. ,

ESFAS Instrumentation- )

3.3.2 '

(

Table 3.3.21 (page 4 of 5)

Engineered Safety Feature Actuation System Instementation l i

! APPLICABLE N00ES OR t OTHER SPECIFIED REQUIRED SURVE!LLANCE ALLOWA8LE j FUMTION CON 0!TIONS CHANNELS CONDITIONS REQUIREMENTS VALUE I 1 l $. Turoine Tetp and l Feeonater Isolation

! a. Automatic 1,2(93.3(9) 2 trains ! SR 3.3.2.4 NA 4 Actuation Logic SR 3.3.2.5

! and Actuation SR 3.3.2.7 i Relays 1

a

b. Steam Generator i (SG) Water j Level - High Hign

(P 14)

! !) Unit 1 1.2(9) 3(9)

. 4 per SG F SR 3.3.2.1 s 83.4%

! SR 3.3.2.4

, SR 3.3.2.5 Q ,v,(c, % 5

} SR l3: Ef CR N

i SR 3.3.2.10 A T4 97c j SR 3.3.2.12 bysl4. *f ;W j 2) Unit 2 1.2(93.3I9I 4 per SG F SR 3.3.2.1 s'edi SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6 ,

SR 3.3.2.7 I SR 3.3.2.10 l SR 3.3.2.12 l l

i

Safety Injection Refer to function 1 (Safety Injection) for all initiation functions and requirements.

l (Continued) 19' ['te t ="e* al' re0 aired Feed = ate

15clation Valves are closed of Isolated by a closed manual valve.

)

l I

c0 BRAIDWOOD - UNITS 1 & 2 3.3-35 Re.i:icn A

a ESFAS Instrumentation 3.3.2 '

Table 3.3.2 1 (page 5 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICA8LE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CON 0!TIONS REQUIREMENTS VALUE

6. Auxtliary Feedwater

a. Automatte Actuation 1.2.3 2 trains ! SR 3.3.2.4 NA Loote and Actuation SR 3.3.2.5 Relays SR 3.3.2.7 IM W 7

~

b. SP Water Level - Low

1) Unit i 1.2.3 4 per SG F SR 3.3.2.1 a 31.01 SR 3.3.2.6 fa .

SR 3.3.2.10 v- -' to SR 3.3.2.12 titl<. T), ?

2) Unit 2 1.2.3 4 per SG F SR 3.3.2.1 ' a 34.8%

SR 3.3.2.6 SR 3.3.2.10 SR 3.3.2.12

c. Safety injection Refer to Function 1 (Safety Injection) for all Initiation functions and requirements,

d. Loss of Offstte Power 1.2.3 2 F SR 3.3.2.1 = 2730 V (Uncervoltage on SR 3.3.2.10 Bus 141(241)J SR 3.3.2.1;

e. Oncervoltage Reactor 1.2 4 0 SR 3.3.2.8 a 4920 V Coolant Pumo (per SR 3.3.2.10 train) SR 3.3.2.12

f. Auxiliary Feeewater 1.2.3 1 per train M SR 3.3.2.1 a 2* Hg Vac Puno Suction Transfer SR 3.3.2.2 on L.c. ion SR 3.3.2.10 hcS5ure - Low

7. Seitenover to Containtnent Sumo

4. automatic Actuation 1.2.3.4 2 trains C SR 3.3.2.4 NA Logic and Actuation SR 3.3,2.5 Relays SR 3.3.2.7 D. iicfueling Water 1.2.3.4 4 0 SR 3.3.2.1 a 44.73 Storage Tant (RWST) SR 3.3.2.6 Level - Low Low SR 3.3.2.10 SR 3.3.2.12 Coinctoent with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety injection BRAIDWOOD - UNITS 1 & 2 3.3-36 R ;ici= ^

RCS Locps-MODE 3 3.4.5 SURVEILLANCE RE0UIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.5.2 Verify steam generator secondary side 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months narrow range water level is = 33%.for each required Unit 1 RCS loop (= 37% for each required Unit 2 RCS loop).

SR 3.4.5.3 Verify correct breaker alignment and 7 days indicated power are available to each required pump that is not in operation.

.. /- . n - -

/ - ,=

u.ac

- ,c, a.w C .3'.- k. : .-a x.a n ru i

4 BRAIDWOOD - UNITS 1 & 2 3.4-10 Revision A

j RCS Loops-MODE 4

3.4.6 ~

l SURVEILLANCE RE0VIREMENTS (continued)

SURVEILLANCE FREQUENCY 4

i SR 3.4.6.2 Verify SG secondary side narrow range water 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months level is = 33%+for each required Unit 1 RCS j loop (= 37% for each required Unit 2 RCS i loop).

] .

)

! SR 3.4.6.3 Verify correct breaker alignment and 7 days

indicated power are available to each required pump that is not in operation.

+-s i

%rbe.fusSeu l 2 Irb )

b ejdt Y a,,J ch..- /

('W _

2 ,

l

[..

J 4

i i

i l

1 i

1 1

i.

i

! l I

'BRIIDWOOO-UNITS 1&2 3.4-13 Prficien A l-i

.RCS Loops-H00E 5. Loops Filled .

3.4.7 SURVEILLANCE RE0VIREMENTS

i. SURVEILLANCE FREQUENCY l SR -3.4.7.1 Verify required RHR loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months a

SR 3.4.7.2 NOTE Only required when complying with

LCO 3.4.7.b.

Verify SG secondary side narrow range water 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

, level is = 33%+for each required Unit 1 RCS

, loo (= 37% for each required Unit 2 RCS loo ).

SR 3.4.7.3 Verify correct breaker alignment and 7 days indicated xwer are available to each required R.iR pump that is not in operation.

j- y - %

f.,c.,-rchpu 3:41\fh l k.- i

_k ~ Mc.k7 -

wiqSe -

4 i

i l

l l

BRAIDWOOD - UNITS 1 & 2 3.4-16 ,tei icn ^ ,

. RCS Loops-MODE 3 L

8 3.4.5 ~

~

BASES ,.~-

ic que3 w 215% la era /

[ 8.. M SURVEILLANCE SR 3.4 5.2 d W "b'

, REQUIREMENTS (continued) SR 3.4.5.2 requires verification oft required SG OPERABILITY.

SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is a 33bfor Unit 1 (= 37% for

-* If the SG secondary W, M ', Unit 2) for each required RCS ioop. side narrow range water leve111s +3 2 % cQ -+ Unit 210 the tubes may become uncovered and the associated L7 loop may not be capable of providing the heat sink for removal of the decay heat. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to a loss of SG level.

SR 3.4.5.3 Verification that the required RCP is OPERABLE ensures that an additional RCP can be placed in operation. if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power availability to the required RCP. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

REFERENCES 1. UFSAR. Section 15.4.1.

.BRAIDWOOD - UNITS 1 & 2 B 3.4-29 Revisicn A-

RCS L oops -MODE 4

, ., B 3.4.6 '

BASES

} ACTIONS C.1 and C.2 (continued)

If no loop is OPERABLE. all operations involving a reduction 1 of RCS boron concentration must be suspended and action to I restore one RCS or RHR locp to OPERABLE status must be initiated. Boron dilution requires forced circulation to L provide proper mixing and preserve the margin to criticality. The imediate Completion Times reflect the

importance of maintaining the capability for decay heat F removal.

I SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the j required operating RCS or RHR loop is in operation.

Verification may include flow rate, temperature, or pump

status monitoring, which helps ensure t1at forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS l and RHR loop performance. - - - -

m

! /ne / w tp S cV I IW L )

SR 3.4.6.2 s' ~~

i SR 3.4.6.2 requires verification of,_ceguired SG OPERABILITY.

l SG OPERABILITY is verified by ensuring that the secondary

._ side narrow range water level is a 33%A for Unit 1 (= 37% for l

!" lg _ g Unit 2) for each reauired RCS loop. If the SG secondary

-- a side narrow range water levellis : 33% for Unit 1 (' 37! for

'66"*Urit 2b the tubes may become uncovered and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available

in the control room to alert the operator to the loss of SG

level.

k b

i i

i (continued) j 'BRAIDWOOD -UNITS 1 & 2 B 3.4-34 t/i icn A 4

-, ~,m_ _

e- -e -

% . _m- . - , , , . . _ _ . _ _ __

RCS Loops-MODE 5. Loops Filled B 3.4.7' l

BASES (continued)

APPLICABLE In MODE 5. RCS circulation increases the time available for SAFETY ANALYSIS mitigation of an accidental boron dilution event. The RHR loops provide this circulatica and have been identified as important contributors to risk reduction.

RCS Loops-MODE 5. Loops Filled satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO The surpose of this LC0 is to require that at least one of the IHR locas be OPERABLE and in operation with an additional RHR loop OPERABLE or two SGs OPERABLE. One RHR loop provides sufficient forced circulation to ]erform the safety functions of the reactor coolant under t1ese conditions. An additional RHR loop is required to be OPERABLE to provide adequate redundancy for heat removal.

However. if the standby RHR loop is not OPERABLE, an

/, 7 acceptable alternate method is two SGs with their secondary

> he w W : side water levels a 33%Afor Unit 1 (= 37% for Unit 2) and s with their associated RCS loops filled. Should the

%(L f}C"dU fWhcould d'3 W /

D be used to remove the decay heat. operating RHR loop fail, the Note 1 permits all RHR pumps to be removed from operation 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit tests designed to validate various accident analyses values. One of the tests performed during the startup testing program is the validation of rod dro 1 cold conditions both with and without flow.p times The noduring flow l test may be performed in MODE 3. 4. or 5 and requires that .

the pumps be stopped for a short period of time. The Note l permit; stopping of toe pumps in order to perform this test i and validate the assumed analysis values. If changes are made to the RCS that would cause a change to the flow l characteristics of the RCS. the input values must be i revalidated by conducting the test again. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time '

period is adequate to aerform the test, and operating i experience has shown t1at boron stratification is not likely ;

during this short period with no forced flow.

l l

I l

l (continued) l BRAIDWOOD - UNITS 1 & 2 B 3.4-37 Revicier, A .

1 l

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! RCS Loops-HODE 5. Loops Filled i l'-

B 3.4.7 l j

~ BASES j ACTIONS C.1 and C.2 (continued)

If no RHR loop is OPERABLE and one or both of the required .

SGs are inoperable. all operations involving a reduction of l RCS boron concentration must be suspended and action to restore one RHR loop to OPERABLE s htus must be initiated.

Boron dilution requires forced circulation to provide proper mixing and preserve the margin to criticality. The imediate Comaletion Times reflect the importance of maintaining t1e capability for heat removal.

I 1

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required operating RHR loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance. c/"%

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SR 3.4.7.2 W% '

CS Verifying that at least two SGs are OPERABLE by ensuring i their secondary side narrow range water levels are = 33% 4for l l Unit 1 (= 37% for Unit 2) ensures an alternate decay heat l removal method via natural circulation in the event that the l second RHR loop is not OPERABLE. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is I considered adequate in view of other indications available in the control room to alert the operator to the loss of SG level. This SR is modified by a Note which indicates that if both RHR loops are OPERABLE. this Surveillance does not need to be satisfied. l 1

1 i

(continued)

BRAIDWOOD - UNITS 1 & 2 8 3.4-40 -Revision a

=

3 l*.

ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF j FACILITY OPERATING LICENSES i NPF-37, NPF-66, NPF-72, AND NPF-77 j

i

_ Comed has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to Title 10 Code of Federal Regulations

. Section 50 Subsection 92 Paragraph c (10 CFR 50.92 (c)), a proposed amendment to an

operating license involves no significant hazards considerations if operation of the facility j 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. Involve a significant reduction in a margin of safety.

i A. INTRODUCTION 4

Commonwealth Edison (Comed) proposes to revise Byron and Braidwood Technical Specification (TS) Table 2.2-1 (functional unit 13.a), " Reactor Trip System l- Instrumentation Trip Setpoint: Steam Generator Water Level - Low-Low"; TS Table 3.3-j 4 (functional unit 5.b.1)," Engineered Safety Features Actuation System Instrumentation 1 Trip Setpoint: Steam Generator Water Level - High-High"; TS Table 3.3-4 (6.c.1),

! " Engineered Safety Features Actuation System Instrumentation Trip Setpoint: Steam Generator Water Le',el - Low-Low Motor-Driven Pump and Diesel Driven Pump Start",

TS Surveillance Requirement (TSSR) 4.4.1.2.2, required steam generator inventory during hot standby, TSSR 4.4.1.3.2, required steam generator inventory during hot ,

shutdown, and TS Section 3.4.1.4.1.b, limiting condition for operation during cold l shutdown with loops filled.

1 The installation of Babcock and Wilcox,- International (BWI), replacement steam I generators (RSGs) at the Byron Unit I and Braidwood Unit 1 Nuclear Power Stations i necessitates an increase to the operating range of the steam generators due to the decrease I l

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l

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l in narrow range span from 233 inches for the original Westinghouse Model D4 steam

! generators (OSGs) to 180 inches for the BWI RSGs. The increase in operating range will minimize the possibility ofinadvertent plant trips following load changes and feedwater transients.-

l Commonwealth Edison also proposes to eliminate notations from page 2-5 for both l

Braidwood and Byron and pages 3/4 3-25 and 3/4 3-26 (for Braidwood only) since they are related to cycles already completed and, therefore, are no longer valid.

B. NO SIGNIFICANT HAZARDS ANALYSIS

1. - The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

This proposed change includes changing the low-low and high-high SG level setpoints.

The setpoints are being changed to increase the SG level operating range. The change in acceptable operating range will decrease the possibility ofinadvertent plant trips following load changes and feedwater transients. Therefore, the probability ofinadvertent plant trips i will decrease with this change.

The minimum setpoint change proposed in this request establishes controls to ensure that an adequate heat sink is maintained by providing an adequate secondary liquid mass to 3

remove primary system sensible heat and core decay heat shortly after reactor trip and

- initiating auxiliary feedwater flow for long-term cooling. The accidents evaluated for this requirement are the Loss of Normal Feedwater and Feedwater Line Break transients.

The maximum setpoint ensures the steam lines and turbine remain undamaged from the

introduction oflow quality, two-phase flow from the steam generators into the steam

lines. The accident evaluated for this requirement is the Feedwater System Malfunction j that results in an increase in feedwater to one or more steam generators.

j The steam generator water level setpoints are not considered a precursor to any of the

analayzed accidents, and, therefore, these proposed changes do not result in an increase in the probability of occurrence of any accident previously analyzed.

The accidents evaluated for the low-low setpoint are the Loss of Normal Feedwater and

Feedwater Line Break transients. These accidents were both analyzed using approved i methodologies. All acceptance criteria were shown to be met for both these events. In addition, it was demonstrated that the Feedwater System Pipe Break response with the RSGs and the proposed low-low setpoint were bounded by the response with the original Model D4 steam generators. Therefore, the proposed low-low level setpoint change is

demonstrated not to result in an increase in the consequences for these accidents.

4 j The accident evaluated for the high-high setpoint is the Feedwater System Malfunction 4

that results in an increase in feedwater to one or more Steam Generators. All acceptance e a -

i e- -w-m yeme v - --c-, -----,- -1 y --s--r- = ,-3 e--- % v - .- . ,- -

e criteria were shown to be met. In addition, it was shown that the RSGs do not completely fill with liquid. This assures that the steam lines and turbine remain undamaged with no introduction oflow quality, two-phase flow from the steam generators into the steam lines during the transient. With all acceptance criteria met, the proposed high-high level setpoint change is demonstrated not to result in an increase in the consequences for these accidents.

TSSR 4.4.1.2.2, TSSR 4.4.1.3.2, and TS 3.4.1.4.1.b assure a minimum inventory (i.e.,

' level) to provide decay heat removal. The requirement for a minimum inventory to remove decay heat is met with assurance that the tube bundle is completely covered. The 1

steam generator operating water level during shutdown conditions are not considered a precursor to any accident, and, therefore, these proposed changes do not result in an increase in the probability of occurrence of any accident previously analyzed.

The elimination of outdated cycle speci'.ic notations from page 2-5 for both Braidwood and Byron and pages 3/4 3-25 and 3/4 3-26 (Braidwood only) are only administrative and does not impact the probability or consequences of any accidents previously analyzed.

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

-l The proposed setpoint changes do not create any new operating conditiens or modes. The proposed change only revises the setpoints for the Reactor Trip System and Engineered Safety Features Actuation System. The actions of these systems will continue to be performed in accordance with existing requirements which are sufIlcient to ensure plant safety is maintained.

Shutdown conditions steam generator water level is necessary to assure adequate decay heat removal capacity. Assurance that the tube bundle is completely covered along with existing technical specification controls on the Auxiliary Feedwater System and on the Condensate Storage Tank ensure adequate heat removal capacity is maintained and that plant safety is maintained.

Thus, this proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The elimination of outdated cycle specific notations from page 2-5 for both Braidwood and Byron and pages 3/4 3-25 and 3/4 3-26 (Braidwood only) are only administrative and does not create the possibility of a new or different accident.

3. The proposed change does not involve a significant reduction in a margin of safety.

k nla bytud:sgrp sgrlehdoc:16

e A safety evaluation was performed to determine the effect of the RSGs with the revised setpoints.

The accidents potentially affected by the change in the Reactor Trip Steam Generator Water Level low-low setpoint (TS 2.2.1, Table 2.2-1, functional unit 13.a) and Engineered Safety Features Actuation System low-low AFW start setpoint (TS 3.3.2, Table 3.3-4, functional unit 6.c.1) are the Loss of Normal Feedwater and Feedwater Line Break transients. These accidents were both analyzed using approved methodologies. All acceptance criteria were shown to be met for both these events. In addition, it was demonstrated that the Feedwater System Pipe Break response with the RSGs with the proposed low-low setpoint were bounded by the response with the OSGs. Therefore, the proposed low-low level setpoint change is demonstrated net to result in an reduction in the margin of safety for these accidents.

The accident potentially affected by the change in the Engineered Safety Features Actuation System high-high SG level trip (TS 3.3.2, Table 3.3-4, functional unit 5.b.1) is a Feedwater System Malfunction that results in an increase in feedwater to one or more steam generators. This accident was analyzed using an approved methodology. In the evaluation of the Feedwater System Malfunction, all acceptance criteria were shown to be met. In addition, it was shown that the RSGs do not completely fill with liquid. This assures that the steam lines and turbine remain undamaged with no introduction oflow quality, two-phase flow from the steam generators into the steam lines during the transient. With all acceptance criteria met, the proposed high-high level setpoint change is demonstrated not to result in a reduction in the margin of safety l There are no design basis accidents involving shutdown condition steam generator water level. Existing TS controls on the Auxiliary Feedwater System and on the Condensate Storage Tank ensure adequate heat removal capacity is maintained and that plant safety is maintained during shutdown conditions. Therefore, a change to the shutdown condition steam generator water level does not result in a reduction in the margin of safety.

The elimination of oatdated cycle specific notations from page 2-5 for both Braidwood and Byron and pages 3/4 3-25 and 3/4 3-26 (for Braidwood only) are only administrative and does not result in a reduction in the margin of safety for any analyzed event.

Therefore, this amendment request does not result in a significant decrease in a margin of safety.

Based on the above evaluation, Comed has concluded that these changes involve no l

significant hazards considerations.

knisby, ..ssgrp agrievladoc 17

/o 1

e ATTACHMENT D ENVIRONMENTAL ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 Commonwealth Edison Company (Comed) has evaluated this proposed License Amendment Request against the criteria for identification oflicensing and regulatory actions requiring environmental assessment in accordance with Title 10, Code of Federal Regulations, Part 51, Section 21 (10 CFR 51.21). Comed has determined that this proposed License Amendment Request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based upon the following.

1. The proposed licensing action involves the issuance of an amendment to a license for a reactor pursuant to 10 CFR 50 which changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or which changes an inspection or a surveillance requirement. This proposed License Amendment Request changes the level setpoints for Byron and Braidwood Unit I steam generators after refueling outages BIR08 and AIR 07, respectivily, due to the decrease in the narrow range span resulting from replacement of the original Westinghouse D4 steam generators with BWI steam generators. i

2. This proposed License Amendment Request involves no significant hazards considerations as demonstrated in Attachment C;

3. There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite; and

4. There is no significant increase in individual or cumulative occupational l

l radiation exposure.

Therefore, pursuant to 10 CFR 5L22(b), neither an environmental impact statement nor i an environmental assessment is necessary for this proposed License Amendment Request.

L l

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ML20134N848

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