ML20065J447
| ML20065J447 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/05/1994 |
| From: | Stang J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20065J448 | List: |
| References | |
| NUDOCS 9404180190 | |
| Download: ML20065J447 (2) | |
Text
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Ig UNITED STATES
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' E NUCLEAR REGULATORY COMMISSION
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WASHINGTON. D.C. 20556--0001 l
COMMONWEALTH EDIS0N COMPANY DOCKET N0. 50-237_
DRESDEN NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.126 License No. DPR-19 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated March 26, 1993, 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.
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. DPR-19 is hereby amended to read as follows:
9404180190 940405 PDR ADOCK 05000237 P
4
" (2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.126, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with tha Technical Specifications.
3.
This license amendment is effective as of the date of its issuance to be implemented within 45 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f-
'1 ames E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV i
Office of Nuclear Reactor Regulation
'l
Attachment:
.i Changes to the Technical Specifications Date of Issuance: April 5,1994
-1 s
ATTACHMENT TO' LICENSE AMENDMENT NO. 1?6 FACILITY OPERATING LICENSE NO. DPR-19 DOCKET NO. 50-237 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.
n REMOVE INSERT 1/2.1-4 1/2.1-4 3/4.2-8 3/4.2-8 3/4.2-10 3/4.2-10 B 3/4.2-29 B 3/4.2-29 8 3/4.2-31 B 3/4.2-31
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DRESDEN II DPR-19 4
Amendment No. 126 1.1 SAFETY LIMIT (Cont'd.)
2.1 LIMITING SAFETY SYSTEM SETTING (Cont'd.)
The adjustment may also be performed by increasing the APRM gain by FDLRC, which accomplishes the same degree of protection as reducing the trip setting by 1/FDLRC.
C.
Power Transient C.
Reactor low water level scram i
setting shall be greater than 1.
The neutron flux shall or equal to 144" above the top not exceed the scram of the active fuel at normal setting established in operating conditions.
Specification 2.1. A for Note:
Top of active fuel is longer than 1.5 seconds defined to be 360 inches above as indicated by the vessel zero (see Bases 3.2).
process computer.
2.
When the process computer is out of service, this safety limit shall be assumed
-l to be exceeded if the neutron flux exceeds the scram setting established by Specification 2.1. A and a control rod scram does not occur.
D.
Reactor Water Level D.
Reactor low water level ECCS (Shutdown Condition) initiation shall be greater than or equal to 84 inches Whenever the reactor is in above the top of the active-the shutdown condition with fuel at normal operating irradiated fuel in the conditions.
reactor vessel, the water Note: Top of active fuel is level shall not be less defined to be 360 inches above than that corresponding to vessel zero (see Bases 3.2).
12 inches above the top of the active fuel when it is seated in the core.
Note: Top of active fuel is defined to be 360 inches above vessel zero (see Bases 3.2).
1/2.1-4
DRESDEN II DPR 1r3 Amendment No. 126 TABLE 3.2.1 INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION FUNCTIONS MINIMUM n OF OPERABLE INST.
CHANNELS PER TRIP SYSTEM (1)
INSTRUMENTS TRIP LEVEL SETTING ACTION (3) 2 Reactor Low Greater than 144" A
Water Level above top of active fuel (9) 2 Reactor low Greater than or equal A
Low Water to 84" above to active fuel (9)p of 2
High Drywell Less than or equal A
Pressure to 2 psig (4),(5) 2 (2)
High Flow Main Less than or equal B
Steam Line to 120% of rated steam flow 2 of 4 in each High Temperature Less than or equal B
of 4 sets Main Steamline to 200*F.
Tunnel 2
High Radiation Less than or equal B
I Main Steamline to 3 times full Tunnel power background (7),(6) 2 Low Pressure Greater than or equal B
Main Steamline to 850 psig High Flow Isolation I
Condenser Line Less than or equal C
Steamline Side to 300% rated steam flow 1
Condensate Less than or equal C
i Return Side to 32" water diff on condensate return side 2
High Flow HPCI Less than or equal D
Steamline to 300% rated steam flow 4
High Temperature Less than or equal D
HPCI Steamline Area to 200*F.
Notes:
(See next Page) 3/4.2-8
DRESDEN II DPR-19 Amendment No. 126 TA8LE 3.2.2 INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT CDOLING SYSTEMS MINIMUM # OF OPERABLE INST.
CHANNELS PER TRIP SYSTEM (1)
TRIP FUNCTION TRIP LEVEL SETTING Remerks 4
Reactor Low Greater than or equat 1.
In conjunction with low reactor Low Water Levet to 64a above to pressure initiates core spray active fuel (5)p of and LPCI.
2.
In conjunction with high-drywell ressure, 120 sec time delay, and ow pressure core cooling interlock n' tlates auto blowdown.
3.
In' tlates HPCI and S8GTS.
4.
In tlates starting of diesel generators.
2 High Drywell Pressure Less than or equal to 1.
Initiates core spray LPCI, HPCI, and (2),(3) 2 PSIG
$8GTS.
2.
In conjunction with low low water levet 120 rec. time delay and low pressure core cooling interlock initiates auto blowdown.
3.
Initiates starting of diesel generators.
1 Reactor Low Pressure Greater than or equal 1.
Permissive for opening core spray to 300 PSIG & less and LPCI admission valves.
than or equal to 2.
In conjunction with low low reactor 350 PSIG water level initiates core spray and LPCI.
1(4)
Containment Spray Greater than or equal Prevents inadvertent operation of Interlock to 2/3 core height containment spray during accident 2/3 Core Height conditions.
2(4)
Containment High Greater than or equal Prevents inadvertent operation of 1
Pressure to 0.5 PSIG & less
. containment spray during accident than or equal to conditions.
1.5 PSIG 1
Timer Auto Blowdown Less than or equal to In conjunction with low low reactor 120 seconds water level, high dry-well pressure and low pressure core cooling Interlock initiates auto blowdown.
2 Low Pressure Core Greater than or equal Defers APR actuation pending Cooling Pump to 50 PSIG & tess confirmation of low pressure core Discharge Pressure than or equal 100 PSIG cooling system operation.
2/ Bus 4 KV Loss of Voltage Trip on 2930 volts 1.
Initiates starting of diesel i
Emergency Buses plus or minus 5%
enerators.
decreasing voltage 2.
ermissive for starting ECCS pumps.
3.
Removes nonessential loads from buses.
4.
Trips emergency bus normal feed breakers.
2 Sustained High Reactor Less than or equal to initiates isolation condenser i
Pressure 1070 PSIG for 15 i
seconds 2/ Bus Degraded Voltage on Greater than or equal to Initiates alarm and picks up time 4 KV Emergency Buses 3708 volts (equals delay relay. Diesel generator picks 3784 volts less 2%
up load if degraded voltage not tolerance) af ter less corrected after time delay.
than or equal to 5 minutes (ptus 5% tolerance) with a 7 second (plus or minus 20%) inherent time celay N0tes: (See next Page) 3/4.2-10
e DRESDEN II DPR-19 Amendment No. 126 3.2 UMITING CONDITION FOR OPERATION BASES (Cont'd.)
top of active fuel.
Retrofit 8 X 8 fuel has an active fuel length 1.24 inches longer than earlier fuel designs. However, present trip setpoints were used in the LOCA analyses.
This trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation aumps (reference SAR Section 7.7.2).
For a trip setting of 504 incies above vessel zero (144 inches above top cf active fuel) and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximum break; the setting is therefore adequate.
The low low reactor level instrumentation is set to trip when reactor water level is 444 inches above vessel zero (with top of active fuel defined as 360 inches above vessel zero, - 59 inches is 84 inches above the top of active fuel). This trip initiates closure of Group I primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems, starts the emergency diesel generator, and trips the recirculation pumps. This trip setting level was chosen to be low enough to prevent spurious operation but high enough to initiate L
ECCS operation and primary system isolation so that no melting of the fuel cladding will occur and so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded.
For the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary isolation are initiated and in time to meet the above criteria. The instrumentation also covers the full spectrum of breaks and meets the above criteria.
The high-drywell pressure instrumentation is a backup to the water level instrumentation and, in addition to initiating ECCS, it causes isolation of Group 2 isolation valves.
For the breaks discussed above, this instrumentation will initiate ECCS operation at about the same time as the low low water level instrumentation; thus the results given above are applicable here, also Group 2 isolation valves include the drywell vent, purge and sump isolation valves. _High-drywell pressure activates only these valves because high drywell pressure could occur as the result of non-safety-related causes such as not purging the drywell air during startup.
Total system isolation is not desirable for these conditions, and only the valves in Group 2 are required to close. The low low water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes a trip of Group 1 4
primary system isolation valves.
B 3/4.2-29
DRESDEN II DPR-19 Amendment No.126 3.2 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
and/or bypass valves _to open. With the trip set at 850 psig, inventory loss is limited so that fuel is not uncovered and peak clad temperatures are much less than 1500 degrees F; thus, there are no fission products
)
available for release other than those in the reactor water. (Ref.
l Section 11.2.3 SAR) 1 Two sensors on the isolation condenser supply and return lines are provided to detect the failure of isolation condenser line and actuate isolation action. The sensors on the supply and return sides are arranged in a 1 out of 2 logic and, to meet the single failure criteria, all sensors and instrumentation are required to be operable.
The trip settings of s 300% rated steam flow and 32 inches of water differential and valve closure time are such as to prevent uncovering the core or exceeding site limits.
The sensors will actuate due to high flow in either direction.
The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI piping.
Tripping of this instrumentation results in actuation of HPCI isolation valves, i.e., Group 4 valves.
Tripping logic for this function is the same as that for the isolation condenser and thus all sensors are required to be operable to meet the single failure of design flow and valve closure time are such that core uncovery is prevented and fission product release is within limits.
The instrumentation which initiates ECCS action is arranged in a dual bus system.
As for other vital instrumentation arranged in this fashion the Specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.
The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not go below the MCPR fuel cladding integrity safety limit. The trip logic for this function is 1 out of n, e.g., any trip on one of the six APRM's, 8 IRM's, or 4 SRM's will result in a rod block.
The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria are met.
The minimum instrument channel requirements for the REM may be reduced by one for a short period of time to allow for maintenance, testing or calibration.
This time period is only approximately 3% of the operating time in a month and does not significantly increase the risk of preventing and inadvertent control rod withdrawal.
During-Single Loop Operation, the flow biased RBM is reduced by 4 percent to compensate for reverse flow in the idle loop jet pumps.
The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at B 3/4.2-31
t$( **o uq UNITED STATES
[
LLfi I NUCLEAR REGULATORY COMMISSION
- fg ij 6
WASHINGTON, D.C. 20555 0001 gw j
COMMONWEALTH EDISON COMPANY DOCKET NO. 50-249
_l DRESDEN NUCLEAR POWER STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 120 License No. DPR-25 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated March 26, 1993, complies with the standards l
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; 8.
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 3.B. of facility Operating License No. DPR-25 is hereby
.l amended to read as follows,
- B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. _120, are hereby incorporated in the
' license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its-issuance to be implemented within 45 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Ja James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation i
Attachment:
Changes to the Technical Specifications l
Date of Issuance:
April 5, 1994
L ATTACHMENT TO LICENSE AMENDMENT NO.120 FACILITY OPERATING LICENSE NO. DPR-25 1.
DOCKET NO 50-249
/
~
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.
REMOVE INSERT 1/2.1-4 1/2.1-4 3/4.2-8 3/4.2-8.
3/4.2-10 3/4.2-10 B 3/4.2-29 83/4.2-29 B 3/4.2-31 B3/4.2-31 I
r n
ww e-r
DRESDEN III DPR-25 Amendment No. 120 1.1 SAFETY LIMIT (Cont'd.)
2.1 LIMITING SAFETY SYSTEM SETTING (Cont'd.)
The adjustment may also be performed by increasing the APRM gain by FDLRC, which accomplishes the same degree of protection as reducing the trip setting by 1/FDLRC.
C.
Power Transient C.
Reactor low water level scram setting shall be greater than 1.
The neutron flux shall or equal to 144" above the top not exceed the scram of the active fuel at normal setting established in operating conditions.
Specification 2.1.A for Note: Top of active fuel is longer than 1.5 seconds defined to be 360 inches above as indicated by the vessel zero (see Bases 3.2).
process computer.
2.
When the process computer is out of service, this safety limit shall be assumed to be exceeded if the j
neutron flux exceeds j
the scram setting established by Specification 2.1.A and 1
a control rod scram d
does not occur.
D.
Reactor Water level D.
Reactor low water level ECCS (Shutdown Condition) initiation shall be greater than or equal to 84 inches j
Whenever the reactor is in above the top of the active the shutdown condition with fuel at normal operating irradiated fuel in the conditions.
reactor vessel, the water Note:
Top of active fuel is level shall not be less defined to be 360 inches above than that corresponding to vessel zero (see Bases 3.2).
12 inches above the top of the active fuel when it is seated-in the core.
Note: Top of active fuel is defined to be 360 inches above vessel zero (see Bases 3.2).
1/2.1-4
i 1
DRESDEN III-DPR-25 l
Amendment No. 120 TABLE 3.2.1 j
INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION FUNCTIONS' RINIMUM # OF OPERABLE INST.
CHANNELS PER
' TRIP SYSTEM (11 JNSTRUMENTS TRIP LEVEL SETTING '
ACTION (3) 2 Reactor Low Greater than 144" A
Water Level above top of active fuel (8) 2 Reactor Low Greater than or equal A
Low Water to 84" above to active fuel (8)p of 2
High Drywell Less than or equal A
Pressure to 2 psig (4),(5) 2 (2)
High Flow Main Less than or equal B
Steam Line to 120% of rated steam flow 2 of 4 in each High Temperature Less than or equal B
of 4 sets Main Steamline to 200*F.
Tunnel 2
High Radiation Less than or egual B
Main Steamline to 3 times full Tunnel power background (6) 2 Low Pressure Greater than or equal B
Main Steamline to 850 psig High flow Isolation 1
Condenser Line Less than or equal C
Steamline Side to 300% rated steam flow 1
Condensate Less than or equal C
Return Side to 14.8" water diff on condensate return side 2
High Flow HPCI Less than or equal D
Steamline to 300% rated steam flow 4
High Temperature Less than or equal D
HPCI Steamline Area to 200*F.
Notes:
(See next Page) 3/4.2-8
. =.
DRESDEN III DPR-25 Amendment No. 120 TABLE 3.2.2 INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS RTRIMvM # or OPERABLE INST.
CHANNELS PER TRIP SYSTEM (1)
TRIP FUNCTION TRIP LEVEL SETTING Remarks 4
Heactor Low Greater than or equal 1.
In conjunction with low reactor l
Low Water Level active fuel (5)p of pressure initiates core spray l
to 84" above to and LPCI.
2.
In conjunction with high-drywell pressure, 120 sec. time delay, and low pressure core cooling interlock in tlates auto blowdown.
3.
In tietes HPCI and SaGTS.
4.
In tlates starting of diesel generatora.
2 High Dr m LL Pressure Less than or equal to 1.
Initiates core spray LPCI, HPCI, and (2),(3) 2 PSIG SBGTS.
2.
In conjunction with low low water level 120 sec. time delay and low pressure core cooling interlock initiates auto blowdown.
3.
Initiates starting of diesel generators.
1 Reactor Low Pressure Greater than or equal 1.
Permissfve for openi M core spray to 300 PSIG & less and LPCI admission valves, than or equal to 2.
In conjunction with low low reactor 350 PSIG water level initiates core spray and LPCI.
1(4)
Containment Spray Greater than or equal Prevents inadvertent operation of Interlock to 2/3 core height containment spray during accident 2/3 Core Height conditions.
2(4)
Containment High Greater than or equal Prevents inadvertent operation of Pressure to 0.5 PSIG & less containment spray du*ing accident than or equal to conditions.
1.5 PSIG 1
1 Timer Auto Blowdown Less than or equal to inconjunctionwithlowlowreactor 120 seconds water tevel, high dry well pressure and low pressure core cooling interlock initiates auto blowdown.
2 Low Pressure Core Greater than or equal Defers APR actuation pending Cooling Pump to 50 PSIG & less confirmation of low pressure core Discharge Pressure than or equal 100 PSIG cooling system operation.
2/ Bus 4 KV Loss of Voltage Trip on 2930 volts 1.
Initiates starting of diesel Emergency Buses plus or minus 5%
generators.
decreasing voltage 2.
Permissive for starting ECCS pumps.
3.
Removes nonessential loads from l
buses.
i 4.
Trips emergency bus normal feed breakers.
2 Sustained High Reactor Less than or equal to Inittstes isolation condenser Pressure 1070 PSIG for 15-seconds 2/ Bun Degraded Voltage on Greater than or equal to Initiates alarm and picks up time j
4 EV Emergency Buses 3708 volts (equals delay relay. Diesel generator picks 3784 volts less 2%
up load if degraded voltage not tolerance) after less corrected af ter time delay.
than or equat to 5 m nutes (plus 5% tolerance) w' th a 7 second (plus or m'nus 20%) inherent time delay Notes: (see next Page) 3/4.2-10
DRESDEN III DPR-25 Amendment No. 120 3.2 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
top of active fuel.
Retrofit 8 X 8 fuel has an active fuel length 1.24 inches longer than earlier fuel designs. However, present trip setpoints were used in the LOCA analyses. This trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation pumps (reference SAR Section 7.7.2).
For a trip setting of 504 inches above vessel zero (144 inches above top of active fuel) and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximum break; the setting is therefore adequate.
The low low reactor level instrumentation is set to trip when reactor water level is 444 inches above vessel zero (with top of active fuel defined as 360 inches above vessel zero, - 59 inches is 84 inches above the top of active fuel). This trip initiates closure of Group I primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems, starts the emergency diesel generator, and trips the recirculation pumps.
This trip setting level was chosen to be low enough to prevent spurious operation but high enough to initiate ECCS operation and primary system isolation so that no melting of the fuel cladding will occur and so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded.
For the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary isolation are initiated and in time to meet the above criteria.
The instrumentation also covers the full spectrum of breaks and meets the above criteria.
The high-drywell pressure instrumentation is a backup to the water level instrumentation and, in addition to initiating ECCS, it causes isolation of Group 2 isolation valves. For the breaks discussed above, this instrumentation will initiate ECCS operation at about the same time as the low low water level instrumentation; thus the resulte given above are applicable here, also Group 2 isolation valves include tne drywell vent, purge and sump isolation valves. High-drywell pressure activates only these valves because high drywell pressure could occur as the result of non-safety-related causes such as not purging the drywell air during startup. Total system isolation is not desirable for these conditions, and only the valves in Group 2 are required to close. The low low water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes a trip of Group 1 primary system isolation valves.
B 3/4.2-29
DRESDEN III DPR-25 Amendment No. 120 3.2 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
and/or bypass valves to open. With the trip set at 850 psig, inventory loss is limited so that fuel is not uncovered and peak clad temperatures are much less than 1500 degrees F; thus, there are no fission products i
available for release other than those in the reactor water. (Ref.
Section 11.2.3 SAR)
Two sensors on the isolation condenser supply line and two sensors on the return line are provided to detect the failure of isolation condenser line and actuate isolation action. The sensors on the supply and return sides are arranged such that any one of the four sensors can cause isolation and, to meet the single failure criteria, all sensors and instrumentation are required to be operable. The trip settings of f 300%
rated steam flow and 14.8 inches of water differential and valve closure time are such as to prevent uncovering the core or exceeding site limits.
The sensors will actuate due to high flow in either direction.
The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI piping.
Tripping of this instrumentation results in actuation of HPCI isolation valves, i.e., Group 4 valves. Tripping logic for this function is the same as that for the isolation condenser and thus all sensors are required to be operable to meet the single failure of design flow and valve closure time are such that core uncovery is prevented and fission product release is within limits.
The instrumentation which initiates ECCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion the Specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.
The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not go below the MCPR fuel cladding integrity safety limit.
The trip logic for this function is I out of n, e.g., any trip on one of the six APRM's, 8 IRM's, or 4 SRM's will result in a rod block.
The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria are met.
The minimum instrument channel requirements for the RBM may be reduced by one for a short period of time to allow for maintenance, testing or calibration.
This time period is only approximately 3% of the operating time in a month and does not significantly increase the risk of preventing and inadvertent control rod withdrawal.
During Single Loop Operation, the flow biased RBM is reduced by 4 percent to compensate for reverse flow in the idle loop jet pumps.
The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at B 3/4.2-31
-