ML20155F430

Amends 23 to Licenses NPF-37,NPF-66 & Amends 12 to Licenses NPF-72 & NPF-77,respectively,removing Two Tables from Tech Specs Which List Reactor Trip Sys Instrumentation & ESF Response Times

ML20155F430
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	09/27/1988
From:	Muller D Office of Nuclear Reactor Regulation
To:	Commonwealth Edison Co
Shared Package
ML20155F434	List: ML20155F430
References
NPF-37-A-023, NPF-66-A-023, NPF-72-A-012, NPF-77-A-012 NUDOCS 8810130264
Download: ML20155F430 (32)

Text

,

J' i

/

'o UNITEO STATES

~ ',,

NUCLEAR REGULATORY COMMISSION e

g,.....):

WASHINGTON D. C 20666 COMMONWEALTH EDISON COMPANY t

DOCKET NO. STN 50-454 BYRON STATION. UNIT 1

&'NDMENTTOFACILITYOPERATINGLICENSE Amendment No. 23

~

License No. NPF-37 1.

The Nuclear Regulatory Camission (the Comission) has found that:

j A.

The application for amendment by Comonwealth Edison Company (the licensee) dateo January 5,1988, complies with the standards and i

requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in conformity with the application, the i

i provisions of the Act, a.$d the rules and regulations of the j

Comission; i

C.

There is reasonable assurance (1) that the activities authorized by this amendent can be conducted without endangering the health and i

safety of the public, and (11) that such activities will be conducted in compliance with the Comission's regulations; i

1 0.

The issuance of this amendment will not be inimical to the comon i

defense and security or to the health and safety of the public; and l

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have 1

been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tion as indicated in the attachment to this license amendment, and i

paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby arended to read as follows:

I l

8810130264 sso927 DR p

ApoCM 05000454

]

PEC

2 T

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 23 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 amendrent is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMi!SSION k

/

Daniel R. Muller, Director Project Directorate !!!-2 Division of Reactor Projects - !!!,

!Y, Y and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: September 27, 1988 l

/*

'~g UNITED 8 TATE 8 NUCLEAR REGULATO3Y COMMISSION

{

W ASHING TON, O. C. 20664

• \\,,,,,/

4 L

r COMMONWEALTH EDISON COMPANY 1

DOCKET NO. STN 50-455 BYRON STATION UNIT 2 MENDMENTTOFACILITYOPERATINGLICENSE t

Amendment No. 23 4,

j License No. NPF-66 i

1.

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

l A.

The application for amendment by Comonwealth Edison Company (the licensee) deted January 5,1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

Comission; j

C.

There is reascnable 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 Comission's regulations; j

i D.

The issuance of this amendment will not b2 inimical to the comon l

l 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 l

of the Comission's regulations and all applicable requirements have j

been satisfied, i

2.

Accordingly, the license is amended by changes to the Technical Specifica-tiens as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby I

amended to read as follows i

i

)

i

}

2 l

(2) Technical Specifications The Technical Specifications contained in Appendix A (NUREG-11!3), as revised through Amendment No. 23 and revised by Attachment 2 to i

hPF-60, and the Environmental Protection Plan contained in Appendix B, both of which are attached to License No. NPF-37, dated February 14, l

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 f

with the Technical Specifications and the Environmental Protection f

Plan.

3.

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

i FOR THE NUCLEAR REGULATORY COMMIS$10N Daniel R. Huller. Directcr Project Directorate !!!-2 Division of Reactor Projects - 111 t

IV, Y and Special Projects 1

Attachment:

i Changes to tiie Technical i

Specifications l

J f

Date of Issuance: September 27, 1988 i

I 1

a l

[

I L

4

i 1

i i

ATTACHMENT TO LICENSE AMENDMENT N05.23 AND 23 I

j FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 t

DOCKET NOS.,$TN-50 454 AND,,$7N 50-455 i

i Revist Appendix A as follows:

I t

Remove Pages Insert Pages l

l V

V i

1 i

j 8 2-5 B 2-5 4

i 3

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

1 3/4 3-7 3/4 3-7 i

3/4 3-8 3/4 3-8 f

3/4 3 13 3/4 3-13

)

3/4 3-30 3/4 3-30 I

3/4 3-31 3/4 3-31 j

1 h

l 3/4 3-32 3/4 3-32 l

1 i

3/4 3 33 3/4 3-33 i

j B 3/4 3-2 B 3/4 3 2 I

i I

f t

l 1

i i

1 i

l i

l

]

J

I-s LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3,'4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE....................................

3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER................................

3/4 2-3 1

3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R.............................

3/4 2-4 FIGURE 3.2-2 K(Z)-NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGH 1...

3/4 2-5 q

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R..........................................

4 3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATI0................................

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

3/4 2-13 TABLE 3.2-1 DNB PARAMETERS........................................

3/4 2-14 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

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

3/4 3-2 TABLE 3.3-2 (THIS TABLE IS NOT USED)..............................

3/4 3-7 l

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE a

REQUIREMENTS........................................

3/4 3-9 I

3/4.3.2 ENGINEERED SAFLTY FEATURES ACTUATION SYSTEM INSTRUMENTATION.................................

3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.....................................

3/4 3-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS......................

3/4 3-23 TABLE 3.3-5 (THIS TABLE IS NOT USED)..............................

3/4 3-30 l

i TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........

3/4 3-34 BYRON - UNITS 1 & 2 V

AMENDMENT NO. 23

LIMITING SAFETY SYSTEM SETTINGS BASES Power Range, Neutron Flux, High Rates (Continued)

The Power Range Negative Rete trip provides protection for control rod drop accidents.

At high power a single or multiple rod drop accident could cause local flux peaking hich could cause an unconservative local DNBR to exist.

The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the limit value.

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

Both of these trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels in MODE 2 while the Source Range, Neutron Flux trip provides primary protection for the core in MODES 3 4 and 5.

The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active.

The Intermediate Rango channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the Pressurizer High and Low Pressure

trips, The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution.

With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If exial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

BYRON - UNITS 1 & 2 B 2-5 AMENDMENT NO. 23

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERACLE.

l APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

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

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

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

I BYRON - UNITS 1 & 2 3/4 3-1 AMENDMENT NO. 23

T S

S co TABLE 3.3-2 "g

(THIS TABLE IS NOT USED)

C Z

v.

W Q*

ro Las N4 u

e

-9 o

7na Z-4 5

Lab

~ _ _. _ _

o" j

k TABLE 3.3-2 (Continued)

=

5 (THIS TABLE IS NOT USED)

Z 1

e l'

C2 1

--4 4

t/h H

i N

l 1

1 1

I w

N 1

w s

U N

i E

j ZO3m ZH E

N 4

e 1

A.

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

l APPLICABILITY:

As shown in Table 3.3-3.

ACTION:

a.

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

b.

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

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or 2.

Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the chonnel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

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

2 = The value from Column Z of Table 3.3-4 for the affected

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

With an ESFAS instrumentation channel or interlock inoperable, take the c.

ACTION shown in Table 3.3-3.

i BYRON - UNITS 1 & 2 3/4 3-13 AMENDMENT NO. 23 J

-.,-_m_

• A TABLE 3.3-5 (THIS TABLE IS NOT USED) e' BYRON - UNITS 1 & 2 3/4 3-30 AMENDHENT NO. 23

i TABLE 3.3-5 (Continued) j (THIS TABLE IS NOT USED)'

i f

f p

l f

BYRON - UNITS 1 & 2 3/4 3 31 AMENDMENT NO.

23

. _ =

.s TABLE 3.3-5 (Continued)

(THIS TABLE IS NOT USED)

?

)

6 i

l a.

J

)

i i

i I

i l

L J

t i

i 1

4 4

l 4

1 1

)

g I

a 1

j i

1 i

1 i

BYRON - UNITS 1 & 2 3/4 3-32 AMENOMENT NO. 23 i,

I 2

Q TABLE 3.3-5 (Continued)

(THIS TABLE IS NOT USED)

L I.

l l

l 1

{

a s

l L

t

?

i e

g l

l l

{

1 1

1 i

i I

I s

I BYRON - UNITS 1 & 2 3/4 3-33 AMENDMENT NO. 23 i

I J

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

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

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

Inherent to the determination of the T.'ip 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 ant 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 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.

BYRON - UNITS 1 & 2 B 3/4 3-2 AMENDMENT NO, 23

• pa ns-;

+

UNITED STATES

[

o,i NUCLEAR REGULATORY COMMISSION j

5

'p WASHINGTON, D. C. 20555 f

%,.....,o COMMONWEALTH E0! SON COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 12 License No. NPF-72 1.

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

A.

The application for amendment by Comonwealth Edison Company (the licensee) dated January 5, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifica-tion as indicated in the attachment to this license auendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:

~

i I

f 2

i (2) Technical Specification 1 The Technical Specifications contained in Appendix A as revised through Amendment No.12 and the Environmental Protection Plan l

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 Environt. ental Protection Plan.

3.

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

FOR THE NUCLEAR REGU!.ATORY COMMISSION Daniel R. Muller, Director Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects Atte onent:

Changes to the Technical Specifications Date of Issuance: September 27, 1988 l

i*

UNITED STATES o,,

[

g NUCLEAR REGULATORY COMMISSION g

.t WASHINGTON, D. C. 20555 I

t, COMMONWEALTH EDIS0N COMPANY DOCKET d0. STN 50 457 8_RAIDW000 STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.12 License No. NPF-77 1.

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

A.

The application for amendment by Comonwealth Edison Company (the licensee) dated January 5, 1988, complies with the standards and requirementsoftheAtomicEnergyActof1954,asamended(theAct) and the Comission's rules and regulations set forth in 10 CFR Chapter I; R

The #4cility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.

The issuance of this amendment will not be inimical to the corrrron 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 Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tion as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:

\\

2 I

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No.12 and the Envircamental Protection Plan 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 Plan.

3.

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

FOR THE NUCLEAR REGULATORY C069tISSION Daniel R. Muller, Director Project Directorate III-2 Division of Reactor Projects - III, IV, Y and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 27, 1988 1

1

\\

ATTACHMENT TO LICENSE ANENDMENT N05.12 AND 12 AND FACILITY OPERATING LICENSE N05. NPF-72 AND NPF-77 DOCKET NOS. STN-50-456 AND STN 50-457 e Appendix A as follows:

Remove Pages Insert Pages V

V B 2-5 B 2-5 3/4 3-1 3/4 3-1 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 3/4 3-13 3/4 3-13 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 B 3/4 3-2 B 3/4 3-2

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS f

SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX 01FFERENCE....................................

2/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER................................

3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R.............................

3/4 2-4 FIGURE 3.2-2 K(Z)-NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT...

3/4 2-5 q

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT0R.................................................

3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATI0................................

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

3/4 2-13 TABLE 3.2-1 DNB PARAMETERS........................................

3/4 2-14 3/4.3 INSTRUMENTATION

• 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

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

3/4 3-2 TABLE 3.3-2 (THIS TABLE IS NOT USED)..............................

3/4 3-7 l

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................................

3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................................

3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.....................................

3/4 3-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0!NTS......................

3/4 3-23 TABLE 3.3-5 (THIS TABLE IS NOT USED)..............................

3/4 3-30 l

TABLE 4.3-2 ENGIf>EERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREtiENTS...........

3/4 3-34 BRAIDWOOD - UNITS 1 & 2 V

AMENDMENT NO. 12

LIMITING SAFETY SYSTEM SETTINGS BASES Power Range. Neutron Flux, High Rates (Continued)

The Power Range Negative Rate trip provides protection for control rod drop accidents.

At high power a single or multiple rod drop accident could i

cause local flux peaking which could cause an unconservative local DNBR to exist.

The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the limit value.

Inte mediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core pro-tection during reactor STARTUP to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition.

Both of these trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels in Mode 2 while the Source Range, Neutron Flux trip provides primary protection for the core in Modes 3, 4 and 5.

The Source Range channels will initiate a Reactor trip at about 10s counts per second unless manually blocked when P-6 becomes active.

The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the Pressurizer High and Low Pressure trips.

The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution.

With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

i r

BRAIDWOOD - UNITS 1 & 2 B 2-5 AMENDMENT NO. 12 I

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

l APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

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

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to b6 within its limit at least once per 18 months.

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

1 BRAIDWOOD - UNITS 1 & 2 3/4 3-1 AMENDMENT NO. l?

9 9

e e

O M=

0 m

=

m m

LAJ CD E

e-N i5 m

-i i

i BRAIDWOOD - UNITS 1 & 2 3/4 3-7 AMENDMENT NO,12

TABLE 3.3-2 (Contin: led)

,x3 (THIS TABLE IS NOT U51D) o o

O e

3, t

1 v

N l

1 l

m 9

E 2

5 4

N

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

l APPLICABILITY:

As shown in Table 3.3-3.

ACTION:

a.

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

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

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or 2.

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

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

= The value from Column 2 of Table 3.3-4 for the affected

channel, RE = The "as measured" value (in percent span) of rack error for the affected channel, SE = Either the "as measured" value (in percent n.:.a) of the sensor error, or the value for Column SE (Sensor Error) of Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel, With an ESFAS instrumentation channel or interlock inoperable, take the c.

ACTION shown in Table 3.3-3.

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

Auxiliary Building Ventilation actuation not required prior to initial opera-tion at > 27% Rated Thermal Power (RTP) on Cycle 1.

BRAIDWOOD - UNITS 1 & 2 3/4 3-13 AMENDMENT NO 12

TABLE 3.3-5 (THIS TABLE IS NOT USED) i BRAIDWOOD - UNIT 5 1 & 2 3/4 3-30 AMENDHENT NO.12 l

U

TABLE 3.3-5 (Continued)

(THIS TABLE IS NOT USED) l l

BP.AIDWOOD - UNITS 1 & 2 3/4 3-31 AMENDHENT NO. 12

I TABLE 3.3-5 (Continued)

(THIS TABLE IS NOT USED) 4 BRAIDWOOD - UNITS 1 & 2 3/4 3-32 AMENOMENT NO. 12

9 TABLE 3.3-5 (Continued)

(THIS TABLE IS NOT USED)

I l

L i

a b

h I

a J

O h

i L

t 4

)

I

't I

i l

{

4 l

i t

1 e

I 1

i 1

I l

{

i i

p t

1 4

4 1

i i

SRAIDWOOD - UNITS 1 & 2 3/4 3 33 AMENDHENT NO. 12

.w-

.-v.

,,w,.m.--y

.*y v

,v,,,-nw-,-m---..-,v,---

,m.m e-,---m.

-e---

~

--#,,,r-+-,,,

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

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

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

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

Sensor anu 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 allewance.

Being that there is a small statisitical chance that this will happen, an infrequent excessive drif t 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, l

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 t

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.

l BRAIDWOOD - UNITS 1 & 2 B 3/4 3-2 AMENDMENT Nd.12

. -