ML20035G017

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Amends 53,53,42 & 42 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively,revising TS to Incorporate Results of Util Setpoint Reconciliation Program
ML20035G017
Person / Time
Site: Byron, Braidwood  Constellation icon.png
Issue date: 04/13/1993
From: Dyer J
Office of Nuclear Reactor Regulation
To:
Commonwealth Edison Co
Shared Package
ML20035G018 List:
References
NPF-37-A-053, NPF-66-A-053, NPF-72-A-042, NPF-77-A-042 NUDOCS 9304260029
Download: ML20035G017 (49)
v • d • e


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.o UNITED STATES

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-L NUCLEAR REGULATORY COMMISSION n

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ry WASHINGTON, D. C. 20555 i

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i COMMONWEALTH EDIS0N COMPANY i

DOCKET NO. STN 50-454 a

BYRON STATION. UNIT NO. 1-

.i AMENDMENT TO FACILITY OPERATING LICENSE

't Amendment No. 53 l

License No. NPF-37 1.

The Nuclear Regulatory Commission (the Commission) has found.that:

A.

The application for amendment by Commonwealth; Edison Company (the licensee) dated April-15, 1992, as supplemented November 23,-1992, complies with the' standards and requirements.of the' Atomic Energy.

Act of 1954, as amended.(the Act) and the Commission's rules and.

regulations set forth in 10 CFR Chapter I;-

1 The f'cility will operate in conformity with'the application, the' B.

a provisions of the Act, and the rules ~ and regulations of the 3

Commission; 1

C.

There is-reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and '(ii) that such activities will be conducted-in compliance with the Commission's ' regulations, D.

The issuance of.this amendment will not be inimical to..the common defense and security or to the health andl safety of the public-and E.

The issuance of this amendment is in accordance with 10 CFR-Part 51 of the Commission's regulations' and all applicable requirements have been satisfied.

4 2.

. Accordingly, the license is amended. by changes to the Technical Specifi-:

, cations as. indicated in the. attachment to this. license amendment, and

' paragraph 2.C.(2)'of Facility.0perating License No. NPF-37 is hereby' amended to read-as follows:

l

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l 9304260029 930413 PDR ADOCK 03000454 P-PDR; I

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_2_

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 53 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULA10RY COMMISSION Y

James E. Dyer, Director Project Directorate 111-2 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 13, 1993 i

h 1

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b, UNITED STATES NUCLEAR REGULATORY COMMISSION g

. t WASHINGT ON, D. C. 2055S k

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f COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-455 BYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 53 License No. NPF-66 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated April 15, 1992, as supplemented November 23, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the-health ar.d safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and i

E.

The issuance of this amendment is in accordance with 10.CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of facility Operating License No. NPF-66 is hereby amended to read as follows:

t

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L g

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(2)

Technical Soecifications The Technical Specifications contained in Appendix A (NUREG-1113),

l as revised through Amendment No. 53 and revised by Attachment 2 to NPF-66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37, dated February 14, 1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which is hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION h

James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor' Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 13,1993

- ~-

i e

ATTACHMENT TO LICENSE AMENDMENT NOS. 53 AND 53 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 r

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Pages indicated with an asterisk (*) are provided for convenience.

i Remove Paoes Insert Paaes 2-3 2-3 1

2-4 2-4 2-5 2-5 2-6 2-6

  • 2-7
  • 2-7 2-8 2-8
  • 2-9
  • 2-9 2-10 2-10 B 2-3 B 2-3
  • B 2-4
  • B 2-4 3/4 3-13 3/4 3-13
  • 3/4 3-14
  • 3/4 3-14 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 B 3/4 3-1 B 3/4 3-1 B 3/4 3-2 B 3/4 3-2 i

b

l v

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

2.2 LIMITING SAFETY SYSTEM SETTINGS t

REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent within the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY:

As shown for each channel in Table 3.3-1.

ACTION:

a.

With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint

value, i

b.

With the Reactor Trip System Instrumentation or Interlock-Setpoint

[

1ess conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the I

applicable ACTION statement requirement of Specification 3.3-1 until i

the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

i I

s i

BYRON - UNITS 1 & 2 2-3 Amendment No. 53

TABLE 2.2-1

.E,'

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS b

d FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE H

1.

Manual Reactor Trip N.A.

N.A.

e.

N 2.

Power Range, Neutron Flux a.

High Setpoint

$109% of RTP*

1111.36% of RTP*

b.

Low Setpoint

$25% of RTP*

127.36% of RTP*

3.

Power Range, Neutron Flux,

<5% of RTP* with

<6.3% of RTP* with High Positive Rate i time constant i time constant 22 seconds 22 seconds 4.

Power Range, Neutron Flux,

<5% of RTP* with

<6.3% of RTP* with High Negative Rate i time constant i time constant y

i 12 seconds 12 seconds 5.

Intermediate Range, 125% of RTP*

131.5% of RTP*

Neutron Flux 6.

Source Range, Neutron Flux 110 cps

$1.42 x 105 cps 5

7.

Overtemperature AT See Note 1 See Note 2 8.

Overpower AT See Note 3 See Note 4 9.

Pressurizer Pressure-Low 11885 psig 11869 psig N 10.

Pressurizer Pressure-High 12385 psig 12393 psig

$ 11 Pressuri7.er Water Level-High

$92% of instrument

$93.5% of instrument g

span span g

h ^ RIP = RAIED THERMAL POWER l.

CD

g TABLE 2.2-1 (Continued)

E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE M

_ 12.

Reactor Coolant Flow-Low

>90% of loop mini-

>89.3% of loop' mini-9, e,

mum measured flow

  • mum measured flow
  • ro 13.

Steam Generator Water Level Low-low a.

Unit 1

>33.0% of narrow

>31.0% of narrow Fange instrument Fange instrument span span b.

Unit 2

>36.3% of narrow

>34.8% of narrow range instrument range instrument span span

?

14.

Undervoltage - Reactor

>5268 volts -

>4920 volts -

Coolant Pumps iach bus iach bus on 15.

Underfrequency - Reactor

->57.0 Hz

>56.8 Hz Coolant Pumps 16.

Turbine Trip a.

Emergency Trip Header

>1000 psig

>315 psig Pressure b.

Turbine Throttle Valve

>1% open

>1% open Closure ll17.

Safety Injection Input N.A.

N.A.

g from ESF x

5j 18.

Reactor Coolant Pump N.A.

N.A.

Breaker Position Trip

,e P

g

  • Minimum measured flow = 97,600 gpm

h TABLE 2.2-1 (Continued) o 7

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E

h FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE 19.

Reactor Trip System Interlocks y

a.

Intermediate Range

->l x 10 10 amp

~>6 x 10 21 amp Neutron Flux, P-6 b.

Low Power Reactor Trips Block, P-7

1) P-10 input

$10% of RTP*

>7.9% to $12.1% of RTP*

2) P-13 input

<10% RTP* Turbine

<12.1% RTP* Turbine Impulse Pressure Impulse Pressure m

cn Equivalent Equivalent c.

Power Range Neutron 530% of RTP*

$32.1% of RTP*

Flux, P-8 d.

Power Range Neutron

-<10% of RTP*

->7.9% to <12.1% of RTP*

Flux, P-10 e.

Turbine Impulse Chamber

<10% RTP* Turbine

<12.1% RTP* Turbine Pressure, P Impulse Pressure Impulse Pressure Equivalent Equivalent

$ 20.

Reactor Trip Breakers N.A.

N.A.

z E 21.

Automatic Trip and Interlock N. A.

N.A.

3 Logic

% ~22.

Reactor Trip Bypass Breakers N.A.

N.A.

d ^ RIP = RATED 1HERMAL POWER

~

q t

TABLE 2.2-1 (Continued) to

'E' TABLE NOTATIONS a:.

1

~$

NOTE.1:

OVERTEMPERATURE AT AT Il + TaS} i

{K

-K

[- (1 tsS 2

8

)~

1(O

+

~

~

g t

0 e

ro

'Where:

AT

=

Measured aT by RTD. Manifold Instrumentation, 1

lead-lag compensator on measured AT,

=

7 3

t-11, ' T2 Time constants utilized in lead-lag compensator for AT, t2=8s,

=

-T2 = 3 s,.

]

yf 3

Lag compensator en measured AT,.

=

ilme constants utilized in the lag compensator for AT, t3=0s, T3

-=

.AT,.

Indicated AT at RATED' THERMAL POWER,

=

K

=

i 1.1Es, Kr.

=' O.0265/ F, 1

t,5 The function generated by the. lead-lag compensator for T,yg

=

3 dynamic compensation, Time constants. utilized in the lead-lag compensator for T,yg, 14 = 33 s, t4, Is

=

ts.= 4 s, T-

=

Average' temperature,F, 1

1y 15 Lag compensator on measured T,yg,.

=

8

..a.-_.:..._...--,-..-.:..

.......-. :- - - :- -.. -...... a -.. -

--.2..,.

..:....-.--.-...~...u-....--

g TABLE 2.2-1 (Continued).

o E

TABLE NOTATIONS (Continued)

I g

NOTE 1:

(Continued)

M Time constant utilized in the measured T,, lag compensator, 76 - O s, 7

6 T'

$ 588.4*F (Nominal T,y at RATED THERMAL POWER),

K3 0.00134, P

Pressurizer pressure, psig, P'

2235 psig (Nominal RCS operating pressure),

o S

Laplace transform operator, s and f,(al) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:

~

lo (i) for q - a between -32% and +13% f (AI) - 0, where q and q3 are percent RATED THERMAL l

POWER,in the top and bottom halves of the core respec,tively, and q, + q, is total g

THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of q -q exceeds +13%, the4T Trip Setpoint shall be automatically reduced by 1.74% of its,valu,e at RATED THERMAL POWER.

(iii) for each percent that the magnitude of q - 93 exceeds -32%, the AT Trip Setpoint shall be automatically reduced by 1.67% of its,value at RATED THERMAL POWER NOTE 2:

The channel's maximum Trip setpoint shall not exceed its computed Trip Setpoint by more than y

3.71% of dT span.

l E

e 5

b e

m W

m i

n On

=

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4 i

L e-4 v

w u

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s 80 t*

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h r,

t j

m D)

  1. 0 O

0 W

80 b

85

>=

n U

>=

e L

e V

b o

ce O

L e4 +

C b

L O

M O

+d C

+J f5 v

80 m

W V

in C

C C

G aa fp C

M W

D L

0 U

3 u

a J

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  1. 0 O

fU 80 m

A

^

w w

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  • J m

o e

t 3

W c

.J s0 C

4J 80 b

3 e4 +

L r

C

+J W

W CD W

4 C

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+J O

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y L

+J i

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W C"

W

^

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80 J2 Z

f*

t O

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em o

s.,o e

s C

o N

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6 M

M M

e4 e4 e4 e4 c4

  • J e

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m 80 m

M W

W W

W W

W 80 L

  • e-W W

N k

e4

  • J eJ sJ eJ sJ
  • J G

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+b

  • J
  • J O

v O

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O L

C 3

O O

W 2.

Z Z

Z Z

Z Z

U G8 Z

Z J

an C

CD m

C3 W

M C

C C

C C

C ar-4J C

C

-J e

  • c-ac-e CC C

ar-e W

CD e

LW OO fD "O

V V

V V

T CL e ae= -

+3 T

t H

v W

W c

W c

%3

  • J 4J in W

D M

C

.C

.C

.C-C C

  • J U EU C

C C

e

  • w
  • c-
u. 80 Cm O
  • e-e i.

=

0 %

3C U

O C

W C

W W

N NW W

W t

t t

t

'O V

N Nh CL W

D t

C.

CE

&E E

in en m

m sa m

.W

.c O sa m

v1 e4 o *J rv t-mW II Il li il 11 li d

Il ll ll.

Il 11 -

r1 W

e4 +

m m m

m e

N N

81 f*

e v'M W

H H

H W

H f*

e-4

+ +'

  • e4

+

O w +

+

^^

w

  1. 1 H

w so e..

e H M M

4 P4 M

N

.H W

4 M

M e4 he c4 w

4 N

W H

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+

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.-4 W

Q.

vv b

CC W

Lad C

3 o

h

&&J W

C-Z F

BYRON - UNITS 1 & 2 2-9

=

ag TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued) e

-j!

NOTE 3:

(Continued) a

((

Ko 0.00170/ F for T > T" and Ks = 0 for T $ T",

=

[*

T As defined in Note 1,

=

T" Indicated T,yg at. RATED THERMAL POWER (Calibration temperature for AT

=

instrumentation, 5 588.4*F),

S

=

As defined in Note 1, and f (AI) 0 for all al.

=

2 os NOTE 4:

The channel's maximum Trip'Setpoint shall not exceed its computed Trip Setpoint by more than g';

2.31% of AT span.

I 8

9 8

- o t

1

=

.--=c.--i.,--wwerr ww-

- -,. -..wcvw-.-

e

-e

=w w

w-v w.wr,=-v*-

w

-we*-

e w

-e-w=v-.,w a

v.c--

-i-=,e

,y---

1-~--

wv.

yv -

v

--r-ww -+-, - -

e-----

i 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES t

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS

+

The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit.

The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engi -

i neered Safety Features Actuation System in mitigating the consequences of-accidents.

The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the'"as measured" Setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1.

Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.

The methodology to derive the Trip Setpoints is based upon combing all of-the uncertainties in the channels.

Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.

Sensors and other 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 statistical chance that this will happen, an infrequent excessive drift is expected.

Rack or sensor drift, i

in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

j

-f i

l l

BYRON - UNITS 1 & 2 8 2-3 Amendment No. 53-I

\\

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level.

In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore, providing Trip System functional diversity.

The fuac-tional capability at the specified trip setting is required for those anticf-patory or diverse Reactor trips for which no direct credit was assumed in the accident analysis to enhance the overall reliability of the Reactor Trip System.

The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated.

This prevents the reactivity addition that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability.

Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting.

The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.

The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERHAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

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

i BYRON - UNITS 1 & 2 B 2-4

~. -..

.~._

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 3.5-3 shall be OPERABLE with their Trip l

Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a.

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

b.

With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the l

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.

l_

c.

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

t s

-i BYRON'- UNITS 1 & 2 3/4 3-13 AMENDMENT NO. 53.

l

l INSTRUMENTATRON-j SURVEILLANCE REQUIREMENTS i

4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance l

of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the 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 channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.

e I

9 d

4 t

i BYRON - UNITS 1 & 2 3/4 3-14 e

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS i

. TRIP ALLOWABLE e3 FUNCTIONAL' UNIT SETPOINT VALUE

[

1.

Safety Injection

-(Reactor Trip, Feedwater

- Isolation,. Start Diesel g

~ Generators, Containment

' Cooling. Fans,.. Control Room Isolation,~ Phase "A"-

Isolation, Turbine Trip, Auxiliary Feedwater,-

Containment Vent

- Isolation and Essential

-Service Water) y a.

Manual Initiation N.A.

N.A.

'T

b. _

Automatic. Actuation N. A.

N.A.

O Logic:and Actuation-

' Relays c.

' Containment Pressure-High-1 5 3.4 psig 1 4.6 psig d.

Pressurizer Pressure-Low (Above P-11) 1 1829 psig 1 1813 psig Steam Line Pressure-e.

Low (Above P-11) 1 640 psig*

1 614 psig*

. g.

g. 2.

Containment Spray-a

a.. ' Manual-Initiation N.A.

N.' A.

w z-

- b..

Automatic' Actuation P

Logic and' Actuation

- g-

' Relays

' N.A.

N.A.

c.

. Containment Pressure-High 1 20.0 psig 5.21.2 psig w w o,-m: ws

+.

+.=wiav-iw.3 w e -n es.ww rme w

.=.:.r

~,- w 2 m. m + w,

  • e w n

,.++w.e m n wrwes-4

  • =r, m v r % ew.. * --

.-,.+-=%-

wv---aw,-.

we...m% v-n.-

a r.e,,,,-.cr-,w

, w-c e s e, ww,-..

.-are-%,,,.s rem

. c e w - e s.-m -o,

. -+ -v,

TABLE 3.3-4 (Continued)

+

oo ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS i

_e TRIP ALLOWABLE

25. FUNCTIONAL UNIT SETPOINT VALUE

.-4

((

3.

Containment Isolation a.

Phase "A" Isolation

1) Manual Initiation N.A.

N.A.

n,

2) Automatic Actuation Logic and Actuation Relays N.A.

N.A.

.3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

b.

Phase "B" Isolation s

((

1) Manual Initiation N.A.

N.A.

SI

2) Automatic Actuation N.A.

N.A.

Logic and Actuation Relays

3) Containment Pressure-High-3 5 20.0 psig 5 21.2 psig c.

Containment Vent Isolation 1). Automatic Actuation Ei Logic and Actuation j[

Relays N.A.

N.A.

a

2) Manual Phase "A" N.A.

N.A.

ll Isolation O

3) Manual Phase "B"-

N.A.

N.A.

83 Isolation 4)- Safety Injection See Item 1 above for all Safety Injection Trip Setpoints and Allowable Values.

-.-~

TABLE 3.3-4 (Continued) mM ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS TRIP ALLOWABLE 5

FUNCTIONAL UNIT SETPOINT VALUE d

4.

Steam Line Isolation g

a m

a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic and Actuation Relays N.A.

N.A.

c.

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

Steam Line Pressure-

>640 psig*

>614 psig*

Low (Above P-11) w1 e.

Steam Line Pressure w4 Negative Rate-High 1100 psi **

1165.3 psi **

(Below P-11)

S.

Turbine Trip and Feedwater Isolation a.

Automatic Actuation Logic and Actuation Re1ays N.A.

H.A.

b.

-Steam Generator Water Level-High-High (P-14)

1) Unit 1 181.4% of 183.4% of narrow range narrow range k

instrument instrument g

span span h

2) Unit 2 180.8% of-182.8% of

+

narrow range narrow range g

instrument instrument span span g.

oo

-TABLE 3.3-4-(Continued)

Y

. 5!-

_ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS TRIP ALLOWABLE

- c:

25 FUNCTIONAL' UNIT SETPOINT VALUE

-4, v

5.

Turbine Trip and g, -

Feedwater Isolation (continued)

' ^2 c.

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

6.~

. Auxiliary Feedwater a.

Manual Initiation-N.A.

N. A.

b.

Automatic Actuation Logic and Actuation

. m -

. Relays.

N.A.

N.A.

30 c.-

~ Steam Generator Water 32-Level-Low-Low-Start

- ;g.

Motor-Driven Pump and Diesel-Driven. Pump

1) Unit.1

>33.0% of-

>31.0% of Harrow range Harrow range instrument-instrument-span span i

2)' Unit 2

>36.3% of'

>34.8% of l

narrow range narrow range i

instrument instrument span span d.

Undervoltage-RCP Bus-

>5268 volts i

j L Start Motor Driven Pump

->4920 volts

~

gg Land Diesel-Driven Pump ER e..

Safety Injection-

"i Start Motor-

.jg '

' Driven Pump and-See Item 1. above for all Safety Injection Trip Setpoints and Diesel-Driven Ptagi Allowable Values.

- u, W'

i -

-r,w.,.

v.-,-,

+re*.,--...eh

-,--.m-,-[1,,

4,.

,,,,.r-~.~-.

..-*-+,.%

  1. -.w.

+ + -,-,-w,..-r--w,-.-v.

+.,., -,

.....---*v-,,+d-..,~,

-,I w.

,-.,-y

,+,,.-

...-.~.-. <.... - --

,*m~

r..-m-..-..

TABLE 3.3-4 (Continued) 5-

.g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS e

TRIP ALLOWABLE-3 FUNCTIONAL' UNIT-SETPOINT VALUE 6.

Auxiliary Feedwater (Continued)_

[

f.

Division 11'for Unit 1 (Division.21 for Unit 2)

ESF Bus Undervoltage-Start Motor-Driven Pump 2870 volts 2730 volts g.

Auxiliary Feedwater Pump Suction Pressure-

-Low (Transfer to Essential Service Water) 1.22" Hg vac 2" Hg vac

-y E

7.

Automatic' Opening of Containment Sump Suction

. Isolation Valves a.

Automatic Actuation N.A.

N. A.

Logic and Actuation.

Relays b.

.RWST Level-Low-Low

46. M 44.M Coincident with-Safety Injection.

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

E 8

n' ar.

O..

TABLE 3.3-4 (Continuad) m5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS h

TRIP ALLOWABLE 5

FUNCTIONAL UNIT SETPOINT VALUE d

8.

Loss of Power g

e a.

ESF Bus Undervoltage 2870 volts

>2730 volts m

w/1.8s delay 9/<1.9s delay b.

Grid Degraded Voltage 3804 volts

>3728 volts w/310s delay U/310 1 30s delay

-9.

Engineered Safety Feature Actuation System Interlocks a.

Pressurizer Pressure, P-11 11930 psig 11936 psig m

b.

Reactor Trip, P-4 N.A.

N.A.

Low-Low T,yg, P-12

>550 F

>547.2. F c.

d.

Steam Generator Water See. Item 5.b. above for all Steam Generator Water Level Trip Level, P-14 Setpoints and Allowable Values.

(High-Nigh) si ait 8-o O.

3/4.3 INSTRUMENTATION BASES i

s 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION i

The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the

(

associated ACTION and/or Reactor trip will be initiated when the parameter j

monitored by each channel or combination thereof reaches its Setpoint, (2) the i

specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance, and (4) sufficier,t system functional capability is available from diverse i

parameters.

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

l The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability R

is maintained comparable to the original design standards.

The' periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables' i

are set for each functional unit.

A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4.

Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable i

Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.

v i

-l i

i

)

1 BYRON - UNITS 1 AND 2 B 3/4 3-1 Amendment No. 53' j

a

- =

a h

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

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties ;n 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 magnitudes.

Rack 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 that 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.

Response time may be demonstrated 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 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, i

(2) Reactor trip, (3) feedwater 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 automatic valves position, and-(11) essential service water pumps start and automatic' valves position.

i BYRON - UNITS 1 & 2 B 3/4 3-2 AMENDMENT NO. 53

d#

UNITED STATES

.8 i

~ NUCLEAR REGULATORY COMMISSION

{

,E wAsmuovou. o. c.20sss

\\*..../

COMMONWEALTH EDISON COMPANY DOCKET N0; STN 50-456 BRAIDWOOD STATION. UNIT NO. I AMENDMENT T0' FACILITY OPERATING LICENSE

' Amendment No. 42

~

License No. NPF-72 1.

The Nuclear Regulatory Commission (the Commission) has found that-A.

The application for amendment by Commonwealth Edison Company (the licensee) dated April 15, 1992, as. supplemented November 23,:1992, j

complies with the standards and requirements of the Atomic Energy' Act of 1954, as amended (the Act) and the Commission's ' rules and t

regulations set forth in 10 CFR Chapter I;--

l B.

The facility will operate'in' conformity with the application, the-i l

provisions of the Act, and the rules and regulations of the.

Commission;

)

C.

There is reasonable assurance (i) that the activities authorized.

l by this amendment can be conducted without endangering the health d

and safety of the public, and (ii) that such activities will:be conducted in compliance with the Commission's regulations;.

j D.

The issuance of this amendment will not be inimical to the common 1

defense and security or to the health and safety.of.the..public; j

~

and 1

E.

The issuance of this amendment is'in accordance with 10 CFR' i

Part 51 of the Commission's regulations"and all' applicable

]

requirements have been satisfied.

2.

Accordingly, the license is amended by changes. to. the -Technic'al Specifi -

j cations as indicated in the attachment to this license amendment,;and J

paraoraph 2.C.'(2) of facility Operating License No.-.NPF-72 is:hereby' 1

amended to read as followsi i

f

.l 1

?

l, r -

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 42 and the Environmental Protection Plan contained in Appendix B, both of which are attacned hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

t 3.

This license amendment is effective as of the date of its issuance.

1 FOR THE NUCLEAR REGULATORY COMMISSION 4 V TM h James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation i

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 13,1993 t

I l

i l

4

'o

- UNITED STATES NUCLEAR REGULATORY COMMISSION n

gp t.

E WASHINGTON, D. C. 20555 g$

COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FAClllTY OPERATING LICENS1 Amendment No. 42 License No. NPF-77 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison _ Company (the licensee) dated April 15, 1992, as supplemented November 23, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; i

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; i

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR i

Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-i cations as' indicated in the attachment to this license amendment, and' paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby r

amended to read as follows:

i i

i i

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 42 and the Environmental Protection Plan l

contained in Appendix B, both of which were attached to License-No. NPF-72, dated July 2,1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection i

Plan.

3.

This license amendment is effective as of the date if its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION l

i QvT Q

James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V l

Office of Nuclear Reactor Regulation l

Attachment-Changes to the Technical Specifications j

Date of Issuance:

April 13, 1993 h

5 b

f I

[

l f

F l

t t

ATTACHMENT TO LICENSE AMENDMENT NOS, 42 AND 42 FACILITY OPERATING LICENSE NOS, NPF-72 AND NPF-77

{

DOCKET NOS. STN 50-456 AND STN 50-457 Y

Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.

The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Pages indicated with an asterisk (*) are provided for convenience.

Remove Pages Insert Paaes 2-3 2-3 2-4 2-4 2-5 2-5 2-6 2-6

  • 2-7
  • 2-7 2-8 2-8
  • 2-9
  • 2-9 2-10 2-10 B 2-3 B 2-3
  • B 2-4
  • B 2-4 3/4 3-13 3/4 3-13
  • 3/4 3-14
  • 3/4 3-14 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-26 3/4 3-26 3/4 3-26a (Unit 2 only) l 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 B 3/4 3-1 B 3/4 3-1

(

B 3/4 3-2 9 3/4 3-2 i

t i

i 9

C l

t SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent within the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY:

As shown for each channel in Table 3.3-1.

ACTION:

i With a Reactor Trip System Instrumentation or Interlock Setpoint less a.

conservative than the value shown in.the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.

b.

With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the

-l applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

l t

l 1

1

-)

l i

3 BRAIDWOOD - UNITS _1 &'2 2-3 Amendment No. 42 1

4 T,

2:

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE y

1.

Manual Re&ctor Trip N.A.

N.A.

2.

Power Range, Neutron Flux a.

High Setpoint

$109% of RTP*

<111.36% of RTP*

m b.

Low Setpoint

$25% of RTP*

$27.36% of RTP*

3.

Power Range, Neutron Flux,

<5% of RTP* with

<6.3% of RTP* with High Positive Rate i time constant i time constant 12 seconds 12 seconds 4.

Power Range, Neutron Flux,

<5% of RTP* with

<6.3% of RTP* with High Negative Rate i time constant i time constant m

A 12 seconds 12 seconds 5.

Intermediate Range, 125% of RTP*

$31.5% of RTP*

Neutron Flux 6.

Source Range, Neutron Flux

$10 cps

$1.42 x 105 cps 5

7.

Overtemperature AT See Note 1 See Note 2 8.

Overpower AT See Note 3 See Note 4 9.

Pressurizer Pressure-Low 11885 psig 11869 psig il 8

10.

Pressurizer Pressure-High

$2385 psig

$2393 psig E

g 11.

Pressurizer Water level-High

<92% of instrument

<93.5% of instrument ipan ipan 5

4 "RTP = RATED THERMAL POWER

TABLE 2.2-1 (Continued) m REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS s

a

$FUNCTIONALUNIT TRIP SETPOINT ALLOWABLE VALUE

$12 Reactor Coolant Flow-Low

>90% of loop mini- >89.3% of loop g

mum measured flow

  • minimum measured g

flow

  • e 13.

Steam Generator Water N

Level Low-Low a.

Unit 1

>33.0% of narrow

>31.0% of narrow range instrument range instrument span span b.

Unit 2

>17% (Cycle 3);

>16.3% (Cycle 3);

T36.3% (Cycle 4 534.8% (Cycle 4 and ind after) of after) of narrow y

narrow range range instrument instrument span span u,

14.

Undervoltage - Reactor

>5268 volts -

>4920 volts -

Coolant Pumps iach bus iach bus 15.

Underfrequency - Reactor

->57.0 Hz

->56.08 Hz Coolant Pumps 16.

Turbine Trip e.

Emergency Trip Header

>1000 psig

>815 psig Pressure b.

Turbine Throttle Valve

>1% open Closure

->1% open 5 17.

Safety Injection Input N.A.

N.A.

M from ESF 18.

Reactor Coolant Pump N.A.

N. A.

5 Breaker Position Trip

% ^ Minimum measured flow s 97,600 gpm

O TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 8

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE h

19.

Reactor Trip System vi Interlocks s

N Neutron Flux, P-6

->l x 10 10 amp

~>6 x 10 11 amp R-a.

Intermediate Range b.

Low Power Reactor Trips Block, P-7

1) P-10 input

<10% of RTP*

>7.9% to $12.1% of RTP*

2) P-13 input

<10% RTP* Turbine

<12.1% RTP* Turbine Impulse Pressure Impulse Pressure ry Equivalent Equivalent c.

Power Range Neutron

$30% of RTP*

$32.1% of RTP*

Flux, P-8 d.

Power Range Neutron

-<10% of RTP*

->7.9% to <12.1% of RTP*

Flux, P-10 e.

Turbine Impulse Chamber

<10% RTP* Turbine

<12.1% RTP* Turbine Pressure, P-13 Tmpulse Pressure Impulse Pressure Equivalent Equivalent 20.

Reactor Trip Breakers N.A.

N.A.

g 21.

Automatic Trip and Interlock N.A.

N.A.

g Logic E 22.

Reactor Trip Bypass Breakers N.A.

N.A.

5 5 *RTP = RATED THERMAL-POWER iO

TABLE 2.2-1 (Continued) m b

TABLE NOTATIONS o5 O

NOTE 1:

OVERTEMPERATURE AT ATfl (1 f g) $ AT, {K 1 5) - T'] + K (P -'P') - f (al)}

[T (1

-K2 3

y ta t

8 O

Measured AT by RTO Manifold Instrumentation, Where:

AT

=

e.

f Lead-lag compensator on measured AT,

=

i.

Time constants utilized in lead-lag compensator for AT, t

=8s, I

=

It, t2 T2=3s, l

t Lag compensator on measured AT,

=

1 tS 3

r

-s Time constants utilized in the lag compensator for AT,13=0s,

=

is Indicated AT at RATED THERMAL POWER, AT,

=

1.164, K

=

i 0.0265/'F, K

=

2 f

The function generated by the lead-lag compensator for T,yg

=

dynamic compensation,

. Time constants utilized in the lead-lag compensator for T,yg, 14 = 33 s,

=

14, ts is = 4 s, Average temperature, 'F, T

=

I 1 + ts5 Lag c spensator on measured T,yg,

tog TABLE 2.2-1 (Continued)

G$

TABLE NOTATIONS (Continued) 8 NOTE 1:

(Continued)

C Time constant utilized in the measured T,yg lag compensator, To = 0 s, Ts

=

v T'

5 588.4 F (Nominal T at RATED THERMAL POWER),

~

avg n.

K3

= 0.00134, m

Pressurizer pressure, psig, P

=

P' 2235 psig (Nominal RCS operating pressure),

=

Laplace transform operator, s 1, S

=

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

'T 3

00 power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:

(i) for qt ~9b between -32% and +13% f (AI) = 0, where qt and q are percent RATED THERMAL POWER 3

b in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of qt ~9b exceeds 13%'the AT Trip Setpoint shall be automatically-reduced by 1.74% of its value at RATED THERMAL POWER.

h (iii) for each percent that the magnitude of_q g gb exceeds -32%, the AT trip setpoint shall be-E automatically reduced by 1.67% of its value at RATED THERMAL POWER.

x NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than

-g 3.71% of AT span.

si

4 TABLE 2.2-1 (Continued)

TABLE NOTATIONS _(Continued) 8-7 NOTE 3:

OVERPOWER AT (1 + 1 5)

(t5

)

( 1

)

e

.3 7

aT.(1

,,3) (1 ),,3) 1 AT, {K,'- Ks

  • (

}

~

~

^

8 1 + tyS 1

teS 1 + tsS H

e.

As defined in Note 1, Where:

AT

=

1+tS t

As defined in Note 1,

=

i 1+T52 As defined in Note 1,

'=

ti,12 As defined in Note 1,

=

1+T53

= As defined in Note 1, t3 As defined in Note 1, AT,

=

1.072,

.K

=

4 0.02/*F for increasing average temperature and 0 for decreasing average Ks temperature,

.7 tyS The function generated by the rate-lag compensator for T,y9 dynamic

=

1 + tyS compensation,

' Time constants utilized in the rate-lag compensator for T,yg, tr.= 10 s,

=

t7-

.=.As defined in Note 1,

!l 1

y, 3

As defined in Note 1,

=

ts

....-.-...---.-.-,.-.--..,.-.-.-,-.-a...-..-...--

i 4-TABLE 2.2-1 (Continued)

>G TABLE NOTATIONS (Continued) 6

.8 NOTE 3:

(Continued) i Ks 0.00170/ F for T > T" and Ko = 0 for T $ T",

=

T

=

As defined in Note 1, s

w T"

Indicated T,yg at RATED THERMAL POWR (Calibration temperature for AT

=

instrumentatior, 5 588.4 F),

}

t S

As defined in Note 1, and

=

f (aI) 0 for all AI.

=

2 T

O NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than i

2.?t% of AT span.

a t

.m l

5

c l

2.2 LIMITING SAFETY SYSTEM SETTINGS l

BASES l

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit. -The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents.

The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1.

Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.

j 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.

Sensors and other 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 met its allowance.

Being that there is a small statistical chance that 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.

I r

BRAIDWOOD - UNITS 1 & 2 B 2-3 Amendment No. 42

-. ~ -

I 9

LIMITING SAFETY SYSTEM SETTINGS

[

BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level.

In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore, providing Trip System functional diversity.

The func-tional capability at the specified trip setting is required for those antici-patory or diverse Reactor trips for which no direct credit was assumed in the accident analysis to enhance the overall reliability of the Reactor Trip System.

The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated.

This prevents the reactivity addition that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

t Manual Reactor Trip i

The Reactor Trip System includes manual Reactor trip capability.

Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range' trip setting.

The Low Setpoint trip provides protection during subtritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power-operations to mitigate the consequences of a reactivity excursion from all power levels.

The Low Setpoint trip may be manually blocked above P-30 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

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

1 i

i BRAIDWD0D - UNITS 1 & 2 B 2-4

=

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 i

Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

APPLICABILITY: As shown in Table 3.3-3.

r ACTION:

a.

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

7 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, declare the channel inoperable and apply the applicable l

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.

c.

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

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BRAIDWOOD - UNITS 1 & 2 3/4 3-13 AMENDMENT NO. 42 i

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j INSTRUMENTATION SURVEILLANCE REQUIREMENTS I

4.3.2.1 Each ESFAS instrumentation channel and-interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2..

4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function i

shall be demonstrated to be within the 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 channels.

are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of i

Channels"' Column of Table 3.3-3.

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-BRAIDWOOD - UNITS 1 & 2-3/4 3-14 i

TABLE 3.3-4 a3

. gg.

g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 6g TRIP ALLOWABLE FUNCTIONAL UNIT SETPOINT VALUE c5 1.

Safety Injection.

d (Reactor Trip, Feedwater Isolation, Start Diesel g

Generators, Containment g,

Cooling Fans, Control

.m Room Isolation, Phase "A" Isolation, Turbine Trip, Auxiliary Feedwater, Containment Vent Isolation and Essential Service Water) a.

Manual Initiation N.A.

N.A.

[

b.

Automatic Actuation N.A.

N. A.

Logic and Actuation w

Relays c.

Containment Pressure-High-1

$ 3.4 psig 5 4.6 psig d.

Pressurizer Pressure-Low (Above P-11) 2 1829 psig 1 1813 psig

.e.

Steam Line Pressure-Low (Above P;-11) 1 640 psig*

1 614 psig*

. ' 2.

-Containment' Spray

  • g a.

Manual Initiation N.A.

N.A.

f.

b.

Automatic Actuation.

. Logic and' Actuation-Relays N.A.

N.A.

N c.

Containment Pressure-High-3

< 20.0~psig

$ 21.2 psig

~

TABLE 3.3-4 (Continued) m 5g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 58 TRIP ALLOWABLE FUNCTIONAL UNIT SETPOINT VALUE 3.

Containment Isolation

l a.

Phase "A" Isolation

1) Manual Initiation N.A.

N.A.

.g-

[

2) Automatic-Actuation Logic and Actuation Relays N.A.

N. A.

3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

b.

Phase "B" Isolation ws

[

1) Manual Initiation N.A.

N.A.

2) Automatic Actuation N.A.

N.A.

Logic and Actuation Relays

3) Containment Pressure-High-3

$ 20.0 psig

< 21.2 psig c.

Containment Vent Isolation

1) Automatic Actuation E

logic and Acttation

[-

Relays N.A.

N.A.

a 3

2) Manual Phase "A" N.A.

N.A.

{

Isolation O

3) Manual Phase "B" N.A.

N.A.

O Isolation

4) Safety Injection See Item l'above for all Safety Injection Trip Setpoints and Allowable Values.

TABLE 3.3-4 (Continued) 1 ENGINEERED SAFETY FEATURES ACTUATION' SYSTEM INSTRUMENTATION TRIP SETPOINTS

' ~-

..5 TRIP ALLOWABLE 6

FUNCTIONAL UNIT SETPOINT VALUE 8

-..4.

~ Steam Line Isolation e

5

'a Manual Initiation N.A.

N.A.

u

' Automatic Actuation L

b.

.. Logic and Actuation' vo

. Relays; N.A.

N.A.

Y 1

c.

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

Steam Line' Pressure-Low-(Above P-11)

>640 psig*

>614 psig*

e.

. Steam Line Pressure-Negative Rate-High

,w.)-

(Below P-11) 5100 psi **

1165.3 psi **

r w

~5.

Turbine Trip and

[ft Feedwater Isolation-a.

' Automatic Actuation Logic and Actuation

- Relays-N.A.

N.A.

b.

Steam Generator Water Level-High-High (P-14)-
1) Unit.1

'<81.4% of

<83.4% of a

Harrow range Harrow range instrument-instrument-g-'

span-span j

2) Unit 2

.. <78.1% (Cycle <79.7% (Cycle 3);

3); Te 4 and 182.8% (Cycle 4 and

<80 8% -

L :n -

~

h

- (Cyc after)'of narrow 4

. after) off range instrument' p

narrow range span instrument.

,m

- span 1

un-e rs e' Ww w-6 4e etw===-evrw Ir w ee-wr gst"w

'm erriv 6 a -tu1119

-*-49Wn'D-f Dew w-WMvvvev.' mew

  • W--

w-'9't'wed

Vr-'*

1 4

W yery y"e-g3

+4 wwm wMwmwyt ug e vwswe W yytm v awvyw e g@v

  • -y_7WW W's wrWw*-tv*--w-DMy wir-9* 1t ye go t Me www g.

wpwar agre 1-4Wmtfeue-V ry-7 w *t y w

  • v-w'yyw rr W

TABLE 3.3-4 (Continued)

ENGINEERED ~ SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS m5g TRIP ALLOWABLE g

FUNCTIONAL UNIT SETPOINT VALUE O

7 5.

Turbine Trip and Feedwater Isolation (continued) c c.

Safety Injection See Iten 1. above for all Safety Injection Trip Setpoints and m

Allowable Values.

[

6.

Auxiliary Feedwater a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic and Actuation Relays N.A.

N.A.

c.

Steam Generator Water Level-Low-Low-Start R

Motor-Driven Pump and Diesel-Driven Pump 5

1) Unit 1

>33.0% of

>31.0% of narrow range narrow range instrument instrument span span

2) Unit 2

>17% (Cycle

>16.3% (Cycle 3);

3); Te 4 and F34.8% (Cycle 4

>36.3%

(Cyc and after) of after) of narrow range narrow range instrument instrument span

,g-span z

E d.

Undervoltage-RCP Bus-

>5268 volts >4920 volts y

Start Motor Driven Pump

~

and Diesel-Driven Pump O

e.

Safety Injection-2 Start Motor-

. Driven Pump and See Item 1. above for all Safety Injection Trip Setpoints and

. Diesel-Driven ~ Pump' Allowable Values.

e.

TABLE 3.3-4 (Continued)

- $l ~

' ;g -

- ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

-8[

TRIP ALLOWABLE FUNCTIONAL UNIT SETPOINT VALUE C,i5 6.

Auxiliary Feedwater (Continued) wu, f.

~ Division 11 for Unit 1-(Division-21 for. Unit 2)

' ESF Bus Undervoltage-

n,

. Start Motor-Driven Pump-2870 volts 2730 volts Auxiliary 1Feedwater-g.

' Pump Suction Pressure-Low-(Transfer to Essential Service u,

g;;

Water) 1.22" Hg vac 2" Hg vac y

' E$

7.

Automatic Opening of Containment--Sump Suction Isolation 1 Valves-a.

. Automatic Actuation N.A.

N.A.

Logic and Actuation Relays b.

RWST Level-Low-Low 46.7%

44.7%

Coincident with-

. g7 Safety Injection See Item 1. above for: Safety Injection Trip Setpoints and Allowable Values.

8 9

8 O

p**w y

,py..c q,m%g yes g

---g-Mtr *aiir e -

9 ' ' W 1sPr e-

=4 ht*549 7MhW-ide

  • Mr-S'

+--~b" WCMST'

-v M '%

7-~9*pe - I

--' W af 's

    • M*W4***W8w-T5-+e+--

a*

93-'4'1"-e i -- PtW i

TABLE 3.3-4 (Continued) 5g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS h:

TRIP ALLOWABLE' FUNCTIONAL UNIT SETPOINT VALUE 8.

Loss of Power w

[

a.

ESF Bus Undervoltage 2870 volts

>2730 volts w/1.8s delay w/ l.9s delay i

b.

Grid Degraded m

Voltace 3804 volts

>3728 y 'ts w/310s delay w/310 1 30s delay f

9.

Engineered Safety Feature Actuation i

System Interlocks a.

Pressurizer Pressure, P-11 11930 psig 11936 psig b.

Reactor Trip, P-4 N. A.

N.A.

Low-Low'T,yg, P-12

>550 F

>547.2 F c.

d.

Steam Generator Water See Item 5.b. above for all Steam Generator Water Level Trip Level, P-14 Setpoints and Allowable Values.

(High-High)

E Ba 8

O.

t 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

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

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

The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.

The periodic surveillance tests performed at the minimum frequencies are suf'ficient to demonstrate this capability.

[

l The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables i

are set for each functional unit.

A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4.

Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.

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BRAIDWOOD - UNITS 1 & 2 B 3/4 3-1 Amendment No. 42

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

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 magnitudes.

Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not 4

met its allowance.

Being that there is a small statisitical chance that 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 measarement 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 r

safety analyses.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demon-strate the total channel response time as defined.

Sensor response time veri-fication 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 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 automatic valves position, and (11) essential service water pumps start and automatic valves position.

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l ORAIDWOOD - UNITS 1 & 2 B 3/4 3-2 AMENDMENT NO. 42

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