ML20209E557
| ML20209E557 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 09/08/1986 |
| From: | Oconnor P Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20209E562 | List: |
| References | |
| NUDOCS 8609110222 | |
| Download: ML20209E557 (14) | |
Text
'
. r..;
o,,
UNITED STATES
[
g NUCLEAR REGULATORY COMMISSION 7,
E WASHINGTON, D. C. 20555
- s.,...../
UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO.
50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 17 License No. NPF-30 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Callaway Plant, Unit 1 (the facility) Facility (the licensee) dated OctoberOperating License No. NPF-30 Electric Company 16, 1985, 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 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 comon 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 Comission's regulations and all applicable requirements have been satisfied.
8609110222 860900 PDR ADOCK 00000493 P
pon
, 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-30 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 17, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license. UE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amerdment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COPNISSION
\\$\\
Paul O'Connor, Project Manager PWR Project Directorate #4 Division of PWR Licensing-A, NRR
Attachment:
Changes to the Technical Specifications OdI.e Of IbbudnCU; Sepleniver 0,1900 d
PWR#4:DPWR-A LWRpj PWR-A MC-Betp PW e / PWR-A P0'Connor:mac MDLhch Ef 64.,
BJYoungblood 08/S /86 08/ 2 ('/86 Of/J/86 d(/d /86 W
ATTACHMENT TO LICENSE AMENDMENT NO. 17 OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Corresponding overleaf page is provided to maintain document completeness.
Amended Page Overl'eaf Page 3/4 3-3 3/4 3-4 3/4 3-5 3/4 3-6 3/4 3-9 3/4 3-10 3/4 3-11 3/4 3-12 3/4 3-12a B3/4 3-1 B3/4 3-2 1
l l
TABLE 3.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION E5 MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E
11.
Pressurizer Water Level-High 3
2 2
1 6#
(
12.
Reactor Coolant Flow-Low a.
Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1
6#
l any oper-each oper-ating loop ating loop b.
Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1
6#
l below P-8) two oper-each oper-ating loops ating loop t'
13.
Steam Generator Water 4/sta. gen.
2/sta. gen.
3/sta. gen.
1, 2 6#(1)l Level-Low-Low in any oper-each oper-Y ating sta.
ating stm.
gen.
gen.
14.
Undervoltage-Reactor Coolant 6#(1)l Pumps 4-2/ bus 2-1/ bus 3
1
- 15. Underfrequency-Reactor Coolant Pumps 4-2/ bus 2-1/ bus 3
1 6#
16.
> Trip a.
Low Fluid Oil Pressure 3
2 2
1 6#
l g-b.
Turbine Stop Valve Closure 4
4 1
1 11#
E R
17.
Safety Injection Input 7
from ESF 2
1 2
1, 2 9
5
TABLE 3.3-1 (Continued) 9 h
REACTOR TRIP SYSTEN INSTRUNENTATION E
NININUN TOTAL NO.
CHANNELS CHANNELS APPLICA8LE g
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 8LE MM)ES ACTION
[
- 18. Reactor Trip System Interlocks a.
Intermediate Range Neutron Flux, P-6 2
1 2
2N 8
b.
Low Power Reactor.
Trips Block, P-7P-10 Input 4
2 3
1 8
or P-13 Input 2
1 2
1 8
c.
Power Range Neutron Flux, P-8 4
2 3
1 8
d.
Power Range Neutron Flux, P-9 4
2 3
1 8
e.
Power Range Neutron Flux, P-10 4
2 3
1, 2 8
f.
Turbine Iopulse Chamber Pressure, P-13 2
1 2
1 8
- 19. Reactor Trip Breakers 2
1 2
1, 2 9
2 1
2 3*, 4*, 58 10
- 20. Automatic Trip and Interlock Logic 2
1 2
1, 2 9
2 1
2 3*, 4*, 58 10 a
}
TABLE 3.3-1 (C*ntinu-d)
TABLE NOTATIONS i
- 0nly if the Reactor Trip System breakers happen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.
- The boron dilution flux doubling signals may be blocked during reactor startup in accordance with approved procedures.
- The provisions of Specification 3.0.4 are not applicable.
- Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
- 8elow the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
(1) The applicable MODES and ACTION statement for these channels noted in Table 3.3-3 are more restrictive and, therefore, applicable.
ACf!ON STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
a.
The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, l
b.
The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l
for surveillance testing of other channels per Specification 4.3.1.1, and c.
Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron flux Trip 5etpoint is reauced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.
ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
a.
Below the P-6 (Intermediate Range Neutron Flux interlock)
Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint; or b.
Above the P-6 (Intermediate Range Neutron Flux interlock)
Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.
l CALLAWAY - UNIT 1 3/4 3-5 Amendment No. 17
=__
O TABLE 3.3-1 (Continued) l ACTION STATEMENTS (Continued)
ACTION 4 - With the number of OPERABLE channels one less than the Minimum l
Channels 0PERABLE requirement suspend all operations involving positive reactivity changes.
l ACTION 5 - a.
With the number of OPERABLE channels one less than the Minimum Channels 0PERABLE requirement, restore the in-operable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open 1
the Reactor trip breakers, suspend all operations involving i
positive reactivity changes and verify Valves 8G-V178 and BG-V601 are closed and secured in position within the next l
hour.
i b.
With no channels OPERA 8LE, open the Reactor Trip Breakers, suspend all operations involving positive reactivity changes i
and verify compliance with the SHUTDOWN MARGIN requirements i
j of Specification 3.1.1.1 or 3.1.1.2, as applicable, within I hour and ever i
BG-V178 and SG y 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, and verify valves 1
V601 are closed and secured in position within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and verified to be closed and secured in position every 14 days.
i ACTION 6 - With the number of OPERA 8LE channels one less than the Total i
Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
l a.
The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and l
j b.
The Minimum Channels OPERABLE requirement is met; however, j
the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l
for surveillance testing of other channels per Specification 4.3.1.1.
j ACTION 7 - Deleted.
l ACTION 8 - With less than the Minimum Number of Channels 0PERABLE within 1hourdeterminebyobservationoftheassociatedpermIssive i
annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.
ACTION 9 - With the number of OPERA 8LE channels one less than the Minimum i
Channels 0PERA8LE requirement, be in at least NOT STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1, provided the other channel is 0PERA8LE.
i ACTION 10 - With the number of OPERA 8LE channels one less than the Minimum i
Channels OPERA 8LE requirement, restors the inoperable channel f
to OPERA 8LE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the Reactor trip l
breakers within the next hour.
(
ACTION 11 - With the number of OPERA 8LE channels less than the Total Number of Channels, operation may continue provided the inoperable l
channels are placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
i I
l L
CALLAWAY - UNIT 1 3/4 3-6 Amendment No. 17 i
l l
t
~
TABLE 4.3-1 h
REACTOR TRIP SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS E
TRIP ANALOG ACTUATING N00ES FOR CHANNEL DEVICE WHICH E
CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE O
FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRE 0 w
1.
Manual Reactor Trip N.A.
M.A.
M.A.
R N.A.
1, 2, 3*,
4*, 5*
2.
Power Range, Neutron Flux a.
High Setpoint S
O(2, 4),
Q(14)
N.A.
N.A.
1, 2 N(3, 4),
Q(4, 6),
R(4, 5) b.
Low Setpoint S
R(4)
S/U(1)
N.A.
N.A.
1W,2 3..
Power Range, Neutron Flux, M.A.
R(4)
Q(14)
N.A.
M.A.
1, 2
{
High Positive Rate Y
4.
Power Range, Neutron Flux, M.A.
R(4)
Q(14)
N.A.
N.A.
1, 2 High Negative Rate 5.
Intermediate Range, S
R(4, 5)
S/U(1)
M.A.
N.A.
IM,2 Neutron Flux 6.
Source Range, Neutron Flux S
R(4, 5, 12)
S/U(1),Q(9,14)
N.A.
M.A.
2N, 3, 4, 5 7.
Overtemperature AT S
R(13)
Q(14)
N.A.
N.A.
1, 2 S.
Overpower AT S
R Q(14)
N.A.
M.A.
1, 2 k
9.
Pressurizer Pressure-Low S
R Q(14)
N.A.
N.A.
1 2
10.
Pressurizer Pressure-High 5
R Q(14)
N.A.
N.A.
1, 2 5
11.
Pressurizer Water Level-High S
R Q(14)
N.A.
N.A.
I 12.
Reactor Coolant flow-low 5
R Q(14)
N.A.
M.A.
1 0
TA8LE 4.3-1 (Continued) h REACTOR TRIP SYSTEN INSTRUENTATION SURVEILLANCE REQUIREENTS 5$
TRIP ANALOG ACTUATING N00ES FOR CHANNEL DEVICE WilCH E
CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE Q
FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED w
- 13. Steam Generator Water Level-S R
Q(14,15)
N.A.
N.A.
1, 2 Low-Low 14.
Undervoltage - Reactor Coolant N. A.
R N.A.
Q(14,15)
N.A.
1 Pays
- 15. Underfrequency - Reactor N.A.
R N.A.
Q(14)
N.A.
1 Coolant Pumps
- 16. Turbine Trip w
a.
Low Fluid Oil Pressure N. A.
R N.A.
S/U(1,10)
N.A.
1 b.
Turbine Stop Valve N.A.
R N.A.
S/U(1,10)
N.A.
1 Closure 17.
Safety Injection Input from M.A.
N.A.
M.A.
Rf N.A.
1, 2 ESF 18.
Reactor Trip System Interlocks a.
Intermediate Range i
Neutron Flux, P-6 N.A.
R(4)
R N.A.
N.A.
2##
b.
Power Range Neutron g.
Flux, P-8 N.A.
R(4)
R N.A.
N.A.
1 c.
Power Range Neutron y
Flux, P-9 N.A.
R(4)
R N.A.
M.A.
1
.U L
TABLE 4.3-1 (Continued)
REACTOR TRIP SYSTEN INSTRtBENTATION SURVEILLANCE REQUIRENENTS E
E TRIP ANALOG ACTUATING N00ES FOR CHANNEL DEVICE WHICH E
CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE U
FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED s
18.
Reactor Trip System l
Interlocks (Continued) i d.
PowerRah Neutron Flux, P-10 N.A.
R(4)
R N.A.
N.A.
1, 2 l
e.
Turbine Impulse Chamber l
Pressure, P-13 N.A.
R R
N.A.
M. A.
1 l
l w
19.
Reactor Trip Breaker M.A.
M.A.
M.A.
N (7, 11)
N.A.
1, 2, 3*, 4*, 5*
l N
w 20.
Automatic Trip and l
g Interlock Logic M.A.
N.A.
N.A.
N.A.
M(7) 1, 2, 3*, 4*, 5*
1
~
t+
1 2
l 0
l l
l
{
i
?
.=
TA8LE 4.3-1 (Continued) l 3
I TABLE NOTATIONS
- 0nly if the Reactor Trip System breakers happen to be closed and the Con-i trol Rod Drive System is capable of rod withdrawal.
- The specified 18 month frequency may be waived for Cycle I provided the i
surveillance is perfomed prior to restart following the first refueling i
outage or June 1 1986, whichever occurs first. The provisions of j
Spec <fication4.6.2areresetfromperformanceofthissurveillance, j
l NBelow P-6 (Intermediate Range Neutron Flux interlock) Setpoint.
l N#BelowP-10(LowSetpointPowerRangeNeutronFluxinterlock)Setpoint.
j (1) If not performed in previous 31 days..
(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 25.
The provisions of Specification 4.0.4 are not applicable for entry into M00E 2 or 1.
(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 155 4
of RATED THERMAL POWER.
Recalibrate if the absolute difference is greater l
than or equal to 35. The provisions of Specification 4.0.4 are not appli-
)
cable for entry into MODE 2 or 1.
l l
(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.
I (5) Detector plateau curves shall be obtained, evaluated and compared to manu-facturer's data.
For the Intemediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not app icable for entry f
l into MODE 2 or 1.
(6) Incore - Excore Calibration above 75% of RATED THERMAL POWER.
The provi-sionsofspecification4.0.4arenotapplicableforentryintoMODE2or1.
l (7) Each train shall be tested at least every 62 days on a $TAGGERED TEST BASIS.
j j
(8) Deleted i
i (9)
QuarterlysurveillanceinMODES3*Intheirrequiredstateforexistingplant4*, and that permissives P-6 and P-10 are i
I conditions by observation of the permissive annunciator window.
Quarterly l
l surveillance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to an increase of twice the count rate within a t
j 10 minute period.
j (10)Setpointverificationisnotrequired.
[
(11) At least once per 18 months and following maintenance or adjustment of the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall i
include independent verification of the Undervoltage and Shunt trips.
(12)Atleastonceper18monthsduringshutdown,verifythatonasimulated t
Boron Dilution Doubling test signal the normal CVC5 discharge valves will close and the centrifugal charging pumps suction valves from the RWST will open within 30 seconds.
L CALLAWAY - UNIT 1 3/4 3-12 Amendment No. 17 I
L
TABLE 4.3-1 (Continued)
)
TABLE NOTATIONS (13) CHANNEL CALIBRATION shall include the RTO bypass loops flow rate.
(14) Each channel shall be tested at least every 92 days on a STAGGERED TEST BASIS.
(15) The surveillance frequency and/or MODES specified for these channels in Table 4.3-2 are more restrictive and, therefore, applicable.
CALLAWAY - UNIT 1 3/4 3 12a Amendment No. 17
\\.
l 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-i tained to permit a channel to be out of service for testing or maintenance, l
and (4) sufficient system functional capabil,ity is available from diverse l
parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and tilversity 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 l
assumptions used in the safety analyses.
The Surveillance Requirements specified l
for tnese systems ensure that the overall system functional capability is main-l tained comparable to the original design standards.
The periodic surveillance i
tests performed at the minimum frequencies are sufficient to demonstrate this l
capability.
Specified surveillance intervals and surveillance and maintenance i
outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service times for the Reactor Protection l
Instrumentation System," supplements to that report, and the NRC's Safety Evaluation dated February 21, 1985.
Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.
The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit.
A betpoint 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.
l 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, An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value.
The methodology of this option I
utilizes the "as measured" deviation from the specified calibration point i
for rack and sensor components in conjunction with a statistical combination of I
the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.
In Equation 3.3-1, Z + R + 5 < TA, the interactive effects of the errors in the rack and the l
sensor, and the "as measured" values of the errors are considered.
Z, as specified in Table 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and CALLAWAY - UNIT 1 0 3/4 3-1 Amendment No.17
l l
INSTRUMENTATION 8ASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM l
INSTRUMENTATION (Continued) rack drift and the accuracy of their measurement.
TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for the actuation.
R or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified Trip Setpoint.
S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions.'
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 t
l 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 drif t, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
The measurement of response time at the specified frequencies provides assurance that the Reactor Trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
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 l
parameters and determines whether or not predetermined limits are being exceeded.
l If they are, the signals are combined into logic matrices sensitive to combina-tions 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 a0gregate function best l
serves the requirements of the condition.
As an example, the following actions l
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 trips.
(3) Feedwater System isolates, (4) the emergency diesel generators start, (5) containment spray pumps start and automatic valves position, (6) contain-ment isolates, (7) steam lines isolate, (8) Turbine trips, (9) auxiliary feedwater pumps start and automatic valves position. (10) containment cooling fans start and automatic valves position, (11) essential service water pumps start and automatic valves position, and (12) isolate normal control room ventilation and start Emergency Ventilation System.
CALLAWAY - UNIT 1 B 3/4 3 2 Amendment No. 17
<