ML20092A073
| ML20092A073 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 09/05/1995 |
| From: | Assa R, Dick G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20092A076 | List: |
| References | |
| NPF-37-A-074, NPF-66-A-074, NPF-72-A-066, NPF-77-A-066 NUDOCS 9509080196 | |
| Download: ML20092A073 (24) | |
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UNITED STATES p
.g NUCLEAR REGULATORY COMMISSION
'f WASHINGTON, D.C. 20555-0001 49.....,o COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 74 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 February 21, 1995, 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; 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 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 and 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.
Accordingly, the license is amended by changes to the Technical Specifi-2.~
cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows:
9509000196 950905 PDR ADOCK 05000454 P
, (2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 74 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 accordante 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 C0ftllSSION George F. Dick, Senior Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 5, 1995
I pMEC y
4 UNITED STATES p
g j
NUCLEAR REGULATORY COMMISSION
't WASHINGTON, D.C. 2055f4001 o
%...../
COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-455 BYRON STATION. UNIT NO. 2 j
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 74 License No. NPF-66 l
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated February 21, 1995, 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 i
provisions of the Act, and the rules and regulations of the Commission; 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 and 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.
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:
P (2)
Technical Specifications The Technical Specifications contained in Appendix A (NUREG-lll3),
as revised through Amendment No. 74 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 George F. Dick, Senior Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 5, 1995
~
C i
ATTACHMENT TO LICENSE AMENDMENT NOS. 74 AND 74 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 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 only.
Remove Paaes Insert Paaes 2-7 2-7 2-8 2-8 2-10 2-10 B 2-5 B 2-5
- B 2-6
- B 2-6 3/4 3-12a 3/4 3-12a 3/4 3-28 3/4 3-28 i
i
-o oc TABLE 2.2-1 (Continued)
TABLE NOTATIONS NOTE 1: OVERTEMPERATURE 6T A T (1+t N) ( 1+T S) s A T [K -K (1+T S) [T( 1+T S) -T'] +K (
1 1
(1+T S) 1 4
o 1
2 3
1 (1+T S) 2 3
s 6
Measured AT by RTD Instrumentation, Where:
AT
=
1+T S 1
1+t g Lead-lag compensator on measured AT, a
3 2
Time constants utilized in lead-lag compensator for AT, r, = 8 s, 7, r 72-3s, 1
1+r,s Lag compensator on measured AT, Time constants utilized in the lag compensator for AT, r3 - O s,' s 2s,"
r 3
AT, Indicated AT at RATED THERMAL POWER, 1.325*, (1.164)**
K, 0.0297/* F*, (0.0265/*F)"
K 2
1 + r,s 1+t s The function generated by the lead-lag compensator for T,y 3
dynamic compensation, Time constants utilized in the lead-lag compensator for T,,, r, - 33 s, 7,
i 4
s r3 - 4 s, T
Average temperature,
- F,
=
- ,, Applicable to Unit 1.
Applicable to Unit 2 after cycle 5.
Not applicable to Unit 1.
Applicable to Unit 2 until completion of cycle 5.
j
" Applicable to Unit I until the completion of cycle 7.
Applicable to Unit 2 until the completion of cycle 6.
" Applicable to Unit I starting with cycle 8.
Applicable to Unit 2 starting with cycle 7.
RVRON - IINTTS 1 A 7 77 AMFNDMFNT NO 74
o %
i TABLE 2.2-1 (Continued)
TABLE NOTATIONS (Continued)
NOTE 1:
(Continued) 1 Lag compensator on measured T,,,,
1+gs Time constant utilized in the measured T,,, lag compensator, 76 = 0 s,' s 2s,"
7 6
T' s 588.4*F (Nominal T,,, at RATED THERMAL POWER),
0.00181*, (0.00134)**
K
=
3 P
Pressurizer pressure, psig,
=
F' 2235 psig (Nominal RCS operating pressure),
=
laplace transform operator, s,
S
=
and f (AI) is a function of the indicated difference between top and bottom detectors of the 3
power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:
for q
-q between -24%* (-32%)** and +10%* (+13%)** f AI 0 where q (i)
RATED,THEf0mL POWER in the top and bottom halves of th(e c)or=e r,espective*1y, and q, + q, isa total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnjtude of,q, f Ets value at RATED THERMAL POWER.* (+13%)**, the A q exceeds +10%
automatically reduced by 4.11% (1.74%)
o (iii) for each percent that the magnitude of,q, f Its value at RATED THERMAL POWER. exceeds -24%*
q automatically reduced by 3.35% (1.67%)
o NOTE 2:
The ch*annel's g ximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.16%
(3.71%),1.33%" of AT span.
,,' Applicable to Unit 1.
Applicable to Unit 2 after cycle 5.
pot applicable to Unit 1.
Applicable to Unit 2 until completion of cycle 5.
Applicable to Unit I until the completion of cycle 7.
Applicable to Unit 2 until the completion of cycle 6.
" Applicable to Unit I starting with cycle 8.
Applicable to Unit 2 starting with cycle 7.
DVDO*f itFt T TC 1 9 9 9 O AMEMDMrHT HM 7A
. ~.....
m o et i
TABLE 2.2-1 (Continued)
TABLE NOTATIONS (Continued)
NOTE 3:
(Continued) 0.00245/*F* (0.00170/*F)**, for T > T" and K6 - 0 for T s T",
K 6
T As defined in Note 1, T"
Indicated T, at RATED THERMAL POWER (Calibration temperature for AT
=
instrumental. ion,s588.4*F),
S As defined in Note 1, and f,(AI) 0 for all AI.
=
NOTE 4:
Thechannel'smaximumTripSetpointshallnotexceeditscomputedTripSetpointbymorethan 3.08%
(2.31%), 3.65%** of AT span.
I l
,,' Applicable to Unit 1.
Applicable to Unit 2 after cycle 5.
Not applicable to Unit 1.
Applicable to Unit 2 until completion of cycle 5.
" Applicable to Unit I until the completion of cycle 7.
Applicable to Unit 2 until the completion of cycle 6.
- Applicable to Unit I starting with cycle 8.
Applicable to Unit 2 starting with cycle 7.
BYRON - UNITS 1 & 2 2-10 AMENDMENT NO. 74
7 9.
LIMITING SAFETY SYSTEM SETTINGS I
i i
BASES Power Ranae. Neutron Flux. Hiah Rates (Continued)
The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist.
The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.
No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the limit value.
]
Intermediate and Source Ranae. 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 3,54 and 5.
The Source Range channels will initiate a Reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active.
t The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually l
blocked when P-10 becomes active.
t Overtemperature oT 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 and pressure is l
within the range between the Pressurizer High and Low Pressure trips. The l
Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capecity of water and includes dynamic compensation for piping delays from the core to the loop temperature i
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 6
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.
l l
BYRON - UNITS 1 & 2 B 2-5 AMENDMENT N0 74
-l
t e
LIMITING SAFETY SYSTEM SETTINGS BASES Overpower AT The Overpower AT Reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip.
The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat generation rate (kW/f t) is not exceeded.
The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."
Pressurizer Pressure In each of the pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted.
The Low Setpoint trip protects against lcw pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.
On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and
~;
on increasing power, automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.
Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves.
On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
BYRON - UNITS 1 & 2 8 2-6
I TABLE 4.3-1 (Continued)
TABLE NOTATIONS (10) Setpoint verification is not applicable.
(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall be performed such that each train is tested at least every 62 days on a STAGGERED TEST BASIS and following maintenance or adjustment of the Reactor Trip Breakers and shall include independent verification of the OPERABILITY of the Undervoltage and Shunt Trip Attachments of the Reactor Trip Breakers.
(12) Not used.
(13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.*
(14) Verify that the appropriate signals reach the Undervoltage and Shunt Trip Relays, for both the Reactor Trip and Bypass Breakers from the Manual Trip Switches.
(15) Manual Shunt Trip prior to the Reactor Trip Bypass Breaker being racked in and closed for bypassing a Reactor Trip Breaker.
(16) Automatic Undervoltage trip.
l
- This note is applicable to Unit I until completion of cycle 7 and Unit 2 until completion of cycle 6.
BYRON - UNITS 1 & 2 3/4 3-12a AMENDMENT N0. 74
- c TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS EUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE 8.
Loss of Power a.
ESF Bus Undervoltage 2870 volts 22730 volts w/1.8s delay w/sl.9s delay b.
Grid Degraded 3804 volts 23728 volts Voltage w/310s delay w/310 i 30s delay 9.
Engineered Safety Feature Actuation System interlocks a.
Pressurizer Pressure,
$1930 psig s1936 psig P-ll b.
Reactor Trip, P-4 N.A.
N.A.
Low-Low T,,, P-12 2550*F 2547.2*F*, 2546.9'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)
- Applicable to Unit I until completion of cycle 7.
Applicable to Unit 2 until completion of cycle 6.
,, Applicable to Unit I starting with cycle 8.
Applicable to Unit 2 starting with cycle 7.
BYRON - UNITS 1 & 2 3/4 3-28 AMENDMENT NO. 74
m t
3 CER h
k UNITED STATE 3 NUCLEAR REGULATORY COMMISSION If WASHINGTON. D.C. 20566-0001
\\****/
COMMONWEALTH EDIS0N COMPANY DOCKET N0. STN 50-456 BRAIDWOOD STATION. UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 66 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 February 21, 1995, 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; 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 and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulaticns; 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 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-72 is hereby amended to read as follows:
.g.
- (2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 66 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 Ramin R. Assa, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 5, 1995
(s9 Cth
,f
't UNITED STATES g
,j NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 20666 4001
%...../
COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 5 License No. -
-77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
i A.
The application for amendment by Commonwealth Edison Company (the licensee) dated February 21, 1995, 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 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 and 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.
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-77 is hereby amended to read as follows:
-- (2)
Technical Specifications i
The Technical Specifications contained in Appendix A as revised l
through Amendment No. 66 and the Environmental Protection Plan contained in Apper. dix B, both of which were attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this licenso.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
l 3.
This license amendment is effective as of the date if its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A_
Ramin R. Assa, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV l
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 5, 1995 i
i
ATTACHMENT TO LICENSE AMENDMENT NOS. 66 AND 66 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET N05. STN 50-456 AND STN 50-457 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 only.
[
Remove Paaes Insert Paaes 2-7 2-7 l
2-8 2-8 2-10 2-10 i
B 2-5 B 2-5 i
- B 2-6
- B 2-6 i
3/4 3-12a 3/4 3-12a 3/4 3-28 3/4 3-28 1
I j
TABLE 2.2-1 (Continued)
TABLE NOTATIONS NOTE 1: OVERTEMPERATURE AT A T (1+t S) ( 1+t S) s A T [K -K (1+t S) [T( 1+t S) -T'] +K (P-P') -f ( AI)]
1 1
(1+T S) 1 4
o 1
2 3
1 (1+t S) 2 3
3 s
Measured AT by RTD Instrumentation, l
Where:
AT 1+T S 1
Lead-lag compensator on measured AT, 1+t s a
Time constants utilized in ' a d-lag compensator for AT, ri - 8 s, 7, 7 3
2 r2 - 3 s, 1
Lag compensator on measured AT, j
1+T s 3
Time constants utilized in the lag compensator for AT, 73 - O s*, s2s,**
T 3
Indicated AT at RATED THERMAL POWER, AT, 1.164,* 1.325**
K, 0.0265/~F,* 0.0297/* F**
K 2
1+T S 4
1+r S The function generated by the lead-lag compensator for T,,
3 dynamic compensation, Time constants utilized in the lead-lag compensator for T,,, r4 - 33 s, 7, 7 4
3 73 - 4 s, I
T Average temperature,
- F, 4
i
- ,, Applicable to Unit I and Unit 2 until completion of cycle 5.
Applicable to Unit I and Unit 2 starting with cycle 6.
BRAIDWOOD - UNITS 1 & 2 7-7 AMENDNENT NO. 66
TABLE 2.2-1 (Continued)
I TABLE NOTATIONS (Continued)
NOTE 1:
(Continued) 1 Lag compensator on measured T,,,
1+r s Time constant utilized in the measured T,, lag compensator, 76 - O s,* s 2s,**
7 6
T' s 588.4*F (Nominal T,, at RATED THERMAL POWER),
0.00134*, 0.00181**
K
=
3 i
P Pressurizer pressure, psig,
=
P' 2235 psig (Nominal RCS operating pressure),
=
Laplace transform operator, s",
S
=
and f (AI) is a function of the indicated difference between top and bottom detectors of the i
power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:
l (i) for q - q between -32%', -24%
and +13%, +10%** f (AI) = 0, where q1 and q3 are percent RATED,THERNAL POWER in the top and bottom halves of the core respectively, and q, + q, is 3
total THERMAL POWER in percent of RATED THERMAL POWER; (ii)
- q3 exceeds 13%', +10%
the AT Trip Setpoint shall be for each percent that the magnjtude of,,q, f its value at RATED THERMAL POWER.
automatically reduced by 1.74%, 4.11%
o l
(iii)
- q, exceeds -32%*, -24%
the AT trip setpoint shall be for each percent that the magnjtude of,,q, f its value at RATED THERMAL POWER automatically reduced by 1.67%, 3.35%
o L
NOTE 2:
The channel's.
3.71%, l'.33%, maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than of AT span.
l
- ,, Applicable to Unit I and Unit 2 until completion of cycle 5.
l Applicable to Unit 1 and Unit 2 starting with cycle 6.
l BRAIDWOOD - UNITS 1 & 2 2-8 AMENDMENT NO. 66 i
y l
TABLE 2.2-1 (Continued)
TABLE NOTATIONS (Continued)
NOTE 3:
(Continued) 0.00170/*F', 0.00245/*F** for T > T" and K - 0 for T s T",
K.
As defined in Note 1, T
=
i Indicated T at RATED THERMAL POWER (Calibration temperature for AT T"
=
instrumentaI., ion,s588.4*F),
As defineu in Note 1, and l
S O for all AI.
f (AI) 2 NOTE 4:
The channel's, maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than j
2.31%, 3.65%
of AT span.
l i
t
- i i
i t
',, Applicable to Unit I and Unit 2 until completion of cycle 5.
Applicable to Unit I and Unit 2 starting with cycle 6.
1 t
l
[
1 4
BRAIDWOOD - UNITS 1 & 2 2-10 AMENOMENT-NO. 66
LIMITING SAFETY SYSTEM SETTINGS 1
BASES Power Ranas. Neutron Flux. Hiah Rates (Continued)
The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the limit value.
i Intermediale and Source Ranoe. 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 Modeg 3, 4 and 5.
The Source Range channels will initiate a Reactor trip at about 10 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 i
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 and pressure is l
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 l
BRAIDWOOD - UNITS 1 & 2 B 2-5 AMENDMENT NO. 66
LIMITING SAFETY SYSTEM SETTINGS BASES Overpower AT The Overpower AT Reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip.
The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded.
The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."
Pressurizer Pressure In each of the pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted.
The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.
On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and i
on increasing power, automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.
i Pressurizer Water Level 1
i The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves.
On decreasing power the Pressurizer i
High Water Level trip is automatically blocked by P-7 (a power level of approximately 100 of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
i BRAIDWOOD - UNITS 1 & 2 B 2-6
- -. = _ _.
TABLE 4.3-1 (Continued)
TABLE NOTATIONS (10) Setpoint verification is not applicable.
(11) The TRIP ACT'JATING DEVICE OPERATIONAL TEST shall be performed such that each train is tested at least every 62 days on a STAGGERED TEST BASIS and following maintenance or adjustment of the Reactor Trip Breakers and shall include independent verification of the OPERABILITY of the Undervoltage and Shut Trip Attachments of the Reactor Trip Breakers.
(12) Not Used.
(13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.*
1 (14) Verify that the appropriate signals reach the Undervoltage and Shunt Trip relays, for both the Reactor Trip and Bypass Breakers from the Manual Trip Swi tches.
i (15) Manual Shunt Trip prior to the Reactor Trip Bypass Breaker being racked in and closed by bypassing a Reactor Trip Breaker.
(16) Automatic undervoltage trip.
1 l
j
- This note is applicable to Unit I and Unit 2 until completion of cycle 5.
BRAIDWOOD - UNITS 1 & 2 3/4 3-12 a AMENDMENT NO. 66
v TABLE 3.3-4 (Continued) l ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS TRIP ALLOWABLE FUNCTIONAL UNIT SETPOINT VALUE i
8.
Loss of Power a.
ESF Bus Undervoltage 2870 volts 22730 volts w/1.8s delay w/s1.9s delay b.
Grid Degraded 3804 volts 23728 volts Voltage w/310s delay w/310 i 30s delay I
9.
Engineered Safety Feature Actuation System Interlocks a.
Pressurizer Pressure, P-ll
$1930 psig s1936 psig b.
Reactor Trip, P-4 N.A.
N.A.
Low-low T,,,
P-12 2550*F 2 547.2*F*, 2 546.9'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.
i (High-High) l l
i i
i t
- Applicable to Unit 1 and Unit 2 until completion of cycle 5.
,, Applicable to Unit 1 and Unit 2 starting with cycle 6.
L BRAIDWOOD - UNITS 1 & 2 3/4 3-28 AMEN 0 MENT NO. 66