ML19323D766
| ML19323D766 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 05/05/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19323D767 | List: |
| References | |
| NUDOCS 8005220309 | |
| Download: ML19323D766 (28) | |
Text
__-
O o
UNITED STATES
'g 8"
NUCLEAR REGULATORY COMMISSION o
{
.E wasumoTow, o. c. 20sss PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO.
50-277 j
i PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 68 f
License No.
DPR-44 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Philadelphia Electric Company, et al., (the licensee) dated August 27, 1979, as supplemented by letters dated November 5,1979, January 30, 1980, February 13, 1980 and March 27, 1980, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CRF 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 (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 i
have been satisfied.
2 Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-44 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 68, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
80 052 2 0M g
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
\\
M Thomas
. Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors 1
Attachment-Changes to the Technical Specifications I
Date of Issuance:
May 5, 1980
ATTACHMENT TO LICENSE AMENDMENT NO. 68 FACILITY OPERATING LICENSE NO. DPR-44 DOCKET NO. 50-277 1.
Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 41 41 42 42 44 44 45 45
- 47/48
- 47/48 65 65 66 66 81 81
- 85/86
- 85/86
- Overleaf page, provided for convenience 4
h k
i I
l
TABLE 4,1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUITS Group (2)
Functional Test Minimum Frequency (3)
Mode Switch in Shutdown A
Place Mode Switch in Each re' fueling outage.
Shutdown Manual Scram A
Trip Channel and Alarm Every 3 months.
RPS Channel Test Switch A
Trip Channel and Alarm Every refueling outage or af ter channel maintenance.
IRM High Flux C
Trip Channel and Alarm (4)
One per week during refueling or startup and before each startup.
Inoperative C
Trip channel and Alarm (4)
Once per week during refueling or startup and before each startup.
APRM High Flux B1 Trip Output Relays (4)
Once/ week.
Inoperative B1 Trip Output Relays (4)
Once/ week.
Downscale B1 Trip Output Relays (4)
Once/ week.
F20w Bias B1 Calibrate Flow Bias Signal (4)Once/ month (1).
~
tiigh Flux in Startup or Refuel C
Trip Output Relays (4)
Once per week during refueling or startup and before each startup.
High Reactor Pressure (6)
B2 Trip Channel and Alarm (4)
Every 1 month (1).
High Drywell Pressure (6)
B2 Trip Channel and Alarm (4)
Every 1 month (1).
Rtactor Low Water Level (5) (6)
B2 Trip Channel and Alarm (4)
Every 1 month (1).
Amendment No. 68
TABLE 4.1.1 (cont'd)
REACTOR PROTECTION SYSTEM (SCRAM) INS'IRUMENT FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUITS Group (2)
Functiona1 Test Minimum Frequency (3)
High Water level In Scram A
Trip Channel and Alarm Every 3 months.
Discharge Tank Turbine Condenser Low Vacuum (6)
B2 Trip Channel and Alarm (4)
Every 1 month (1).
Main Steam Line liigh Radiation B1 Trip Channel and Alarm (4)
Once/ week.
Main Steam Line Isolation A
Trip Channel and Alarm Every 1 month (1).
Valve closure Turbine control Valve A
Trip Channel and Alarm Every 1 month.
Elle Oil Pressure Turbine First Stage Pressure A
Trip Channel and Alarm Every 3 months (1).
Parmissive Turbine Stop Valve closure A
Trip Channel and Alarm Every 1 month (1).
Reactor Pressure Permissive (6) b2 Trip Channel and Alarm (4)
Every 3 months.
t Amendment No.
68
TABLE 14.1.2 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument channel Group (1)
Calibration (84)
Minimum Frequency (2)
IRM High Flux C
Comparison to APRM on Maximum frequency once Controlled Shutdown per week.
APRM High Flux Output Signal B1 Ile at Balance Twice per week.
Flow Bias Signal B1 With Standard Pressure Every refueling outage.
Source LPRM Signal B1 TIP System Traverse Every 6 weeks.
High Reactor Pressure B2 Standard Pressure Source Once per operating
- f cycle.
High Drywell Pressure B2 Standard Pressure Source Once per operating cycle.
Reactor Low Water Imvel B2 Pressure Standard once per operating cycle.
High Water Level in Scram A
Water Column Discharge Volume Every refueling outage.
Turbine Condenser Low Vacuum B2 Standard Vacuum Source Once per operating cycle.
Main Steam Line Isolation A
Note (S)
Note (5).
Valve Closure Main Steam Line High Radiation B1 Standard Current Source (3)
Every 3 months.
i Turbine First Stage Pressure A
Standard Pressure Source Every 6 months.
Permissive Amendment No.
68
~.
l 1
TABLE 4.1.2 (cont ' d)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Group (1)
Calibration (4)
Minimum Frequency (2)
Turbine Control Valve Fast A
Standard Pressure Source Once per operating Closure Oil Pressure Trip cycle.
Turbine Stop valve Closure A
Note (5)
Note (5).
Reactor Pressure Premissive B2 Standard Pressure Source Once per operating Cycle.
1 I
Amendment No.
68
.:n.
~"?
3.]
BASJj_S The reactor protection system automatically initiates a reactor scram to:
1.
Preserve the integrity of the fuel cladding.
2.
Preserve the-integrity of the reactor coolant system.
3.
Minimize the energy which must be absorbed following a loss of coolant accident, and prevent inadvertant crit-icality.
This specification provides th'e limiting conditions for op-eration necessary to preserve the ability of.the system to perform its intended function even during periods when in-strument channels may be out of service because of mainten-ance.
When necessary, one channel may be made. inoperable f6r brief intervals to conduct required functional tests and calibrations.
The reactor protection system is of the dual channel type (Ref erence subsection 7.2 FSAR).
The system'is made up of two independent trip systems, each having two subchannels
'of tripping devices.
Each subchannel has an input from at least one instrument channel which monitors a criticcl para-meter.
The outputs of the s'ubchannels arc combined in a 1 out of 2 j
logic; i.ei an input signal on either one or both of the subchannels will cause a trip system trip.
The outputs of the trip systems are arranged so that a trip on both sys-tems is required to produce a reactor scram.
This system meets the intent df IEEE - 279 for Nuclear Power Plant Protection Systems.
The system has a reliability greator than that of a 2 out of 3 system and somewhat less than that of a 1 out of 2 system.
With the exception of the Average Power Range Monitor (APRM) channels, the Intermediate Range Monitor (IRM) channels, the Main Steam Isolation Valvo closure and the Turbine Stop Valvo closure, each subchannel has one instrument channel.
When the minimum condition for operation on'the number of operable instrument channels per untripped protection trip system is met or if it cannot be met and the affected pro-tection trip system is placed in a tripped condition, the effectiveness of the protection system is preserved.
The APRM instrument channels are provided for each protec-tion trip system.
APRM's A and E operate contacts in one subchannel and APRM's C and E operato contacts in the other subchannel.
APRM's B, D and F are arranged similarly in -
October 1973
PSAPS 3 0 3ASES (Cont'd) the other protection trip system.
Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel.
This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration.
Additional IRM channels have also been provided to allow for bypassing of one such channel.
The bases for the scram setting for the IRM, APRM, high reactor pressure, reactor low water level, MSIV closure, generator load rejection, turbine stop valve closure and loss of condenser vacuum are discussed in Specification 2.l and 2.2.
Instrumentation s ensing drywell pressure is provided to detect a loss of coolant accident and initiate the core standby cooling equipment.
A high drywell p ressu re scram is provided at the same setting as the core standby cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality.
This instrumentation is a backup to the reactor vessel water level instrumentation.
High radiation levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel.
A scram is initiated whenever such radiation level exceeds three times normal background.
The purpose of this scram is to limit fission product release so that 10 CFM Part 100 guidelines are not exceeded.
Discharge of excessive amounts of radioactivity to the site environs is prevented by the air ejector off-gas monitors which cause an isolation of the main condenser off-gas line.
A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.
Ref. paragraph 7.2.3.7 FSAR.
The manual scram function is active in all modes, thus providing f or a manual means of rapidly inserting control rods during all modes of reactor operation.
The APRM (High flux in Start-up or Refuel) system provides protection a3ainst excessive power levels and short reactor periods in the start-up and intermediate power ranges.
The IRM~ system provides protection against short reactor periods in these ranges.
The control rod drive scram system is designed so that all of j
the water which is discharged from the reactor by a scram can l
be accommodated in the discharge piping.
The scram discharge volume accommodates in excess of 50 gallons of water and is
(
the low point in the piping.
No credit was taken for this volume in the design of the discharge piping as concerns l Amendment No.
68
TABLE 3. 2.B. (Cont ed)
INSTRUMENTATION T!!AT INITIATES OR CONTROLS Tile CORE AND CONTAINMENT COOLING SYS'IEMS Minimum No.
of Operable Number of Instru-In2trument Trip Function Trip Level Setting ment Channels Pro Remarks Channels Per vided by Design
> Trip System (1) 2 Reactor liigh Water f*45 in. in,dicated 2 Inst. Channels Trips llPCI and RCIC Level level turbines.
1 Reactor Iow Level 2+312 in. above 2 Inst. Channels Prevents inadvertent (inside shroud) vessel zero (2/3 operation of contain-i core height) ment spray during accident condition.
2 Containment liigh 1 < p < 2 psig 4 Inst. Channels Prevents inadvertent Pressure operation of contain-a ment spray during i
accident condition.
1 Confirmatory Low 2t6 in. indicated 2 Inst. Channels ADS Permissive Level level 0
1 2
111gh Drywell 52 psig 4 Inst. Channels 1.
Initiates Core Spray;
{
Pressure LPCI; ilPCI 2.
Initiates starting of Diesel Generators 3.
Initiates Auto Blow-down (ADS) in conjunction with Low-Low Reactor i
Water Level, 120 I
second time delay, and LPCI or Core Spray pump running.
I Amendment No.
68
TABLE 3.2.B (Cont'd)
INSTRUMENTATION TilAT INITI ATES Oh CONTROLS T!!E CORE AND CONTAINMENT COOLING SYST EMS Minimum No.
Cf Operable Number of Instru-In trument Trip Function Trip Letal Setting ment Channels Pro Remarks Channels Per vided by Design Trip System (1) 2 Reactor Low 400-500 psig 4 Inst. Channels Permissive for opening Pressure Core Spray and LPCI Admission valves.
Coincident with high dry well pressure, starts LPCI and Core Spray pumps.
2 Reactor Low 200-250 psig 4 Inst. Channels Permissive for closing Pressure Recirculating Pump Discharge Valve.
h 1
Reactor Low 505P575 psig 2 Inst. Channels In conjunction with PCI e
Pressure signal permits clouure of RilR (LPCI) injection valves.
Amendment No.
68
9 I
TABLE al.2.B MINIMUM TEST AND CALIBRATION FREQUEtK:Y FOR CSCS Instrument Channel Instrument Ftuictional Test Calibration Frequency Instrument check 1)
Feactor Water Level (7)
(1) (3)
Once/ operating cycle once/ day 2)
Drywell Pressure (7)
(1) (3)
Once/ operating cycle once/ day l
3)
Reactor Pressure (7)
(1) (3)
Once/ operating cycle Once/ day 4)
Auto Sequencing Timers NA Once/ operating cycle None 5)
(1)
Once/3 months None Pressure Interlock 6)
Trip System Bus Power Monitors (1)
NA None 7)
Core Spray Sparger d/p (1)
Once/6 months Once/ day 8)
Steam Line Ifigh Flow (IIPCI S RCIC)
(1)
Once/3 months None
,9)
Steam Line fligh Temp. (IIPCI L RCIC)
(1) (3)
Once/oferating cycle once/ day
- 10) Safeguards Area Iligh Temp.
( 1)
Once/3 months None
- 11) IIPCI and RCIC Steam Line (1)
Once/3 months None Low Pressure
- 12) IlPCI Suction Source Levels (l)
Once/3 months None
- 13) 4KV Emergency Ibwer System Once/ operating cycle Once/5 year None Voltage Relays Ill) ADS Relief Valves Bellows Once/ operating cycle Once/ operating cycle None Pressure Switches
- 15) LPCI/ Cross Connect Valve Position Once/ refueling cycle N/A N/A Amendment No.
68
,9 a
ga i
TABLE 4.2.E MINIMUM TEST AND CALIBRATION FREQUENCY FOR DRYWELL LEAK DETECTION Instrument Channel Instrument Functional Calibration Instrument Test Frequency Check 1)
Equipment Drain Sump Flow Integrator (1)
Once/3 months Once/ day 2)
Floor Drain Sump Flow Integrator (1)
Once/3 months once/ day 3)
Air Sampling System (1)
Once/3 months Once/ day m
I i
N 9
3
- I TABLE 4. 2. P i
j MINIMUM TEST AND CALIBRATION FREQUENCY FOR SURVEILANCE INSTRUMENTATION Instrument channel Calibration Frequency Instrument Check l
l 1)
Reactor Level Once/ operating cycle Once Each Shift i
2)
Reactor Pressure Once/6 months Once Each shift 3)
Drywell Pressure Once/6 months Once Each Shif t 1
1 4)
Drywell Temperature Once/6 months Once Each Shift I
l e
-l E,
5)
Suppression Chamber Temperature Once/6 months Once Each Shift
-l 6)
Suppression Chamber Water Level Once/6 months once Each Shift 1
I 7)
Control Rod Position NA Once Each Shift l
8)
Neutron Monitoring (APRM)
Twice Per Week Once Each Shitt
)
t I
- I 1
',l Amendment No.
68
s 8
k UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
WASHlblGTON, D. C. 20666 y
\\*****/
PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY i
DOCKET NO. 50-278 PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 3
_ AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 67 License No. DPR-56 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Philadelphia Electric Company, et al.,
(the licensee) dated August 27, 1979, as supplemented by letters dated November 5, 1979, January 30, 1980, February 13,1980 and March 27, 1980, complies with the standaras 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 Comission; C.
There is reasonable assurance (i) that the activities authorized by J
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 Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-56 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 67, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
. C 3.
This license amendment is effective as of the date of its issuance.
l FOR THE NUCLEAR REGULATORY COMMISSION
/?
n Thomas Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 5, 1980
=
)
,,4.,
-g
--4
,,7.
ATTACHMENT TO LICENSE AMENDMENT NO. 67 FACILITY OPERATING LICENSE NO. DPR-56 D0CKET NO. 50-278 1.
Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Insert 41 41 42 42 44 44 45 45
- 47/48
- 47/48 65 65 66 66 81 81
- 85/86
- 85/86
- Overleaf page, provided for convenience l
daleted when modification authorized by Amendment No.
are completed
- Effective when modifications authorized by Amendment No.
are completed TABLE 4,1.1 REACTOR PROTECTION SYSTEM (SCRAM) INS *1RUMENT FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUITS Group (2)
Functional Test Minimum Frequency (3)
Mode switch in Shutdown A
Place Mode Switch in Each re' fueling outage.
Shutdown Manual Scram A
Trip Channel and Alarm Every 3 months.
RPS Channel Test Switch A
Trip channel and Alarm Every refueling outage or af ter channel maintenance.
IRM S
liigh Flux C
Trip Channel and Alarm (4)
One per week during refueling or startup ar)d before each startup.
Inoperative C
Trip Channel and Alarm (4)
Once per week during refueling or startup and before each startup.
APRM High Flux B1 Trip Output Relays (4)
Once/ week.
Inoperative B1 Trip Output Relays (4)
Once/ week.
Downscale B1 Trip Output Relays (4)
Once/ week.
Flow Dias B1 Calibrate Flow Dias Signal (4)Once/ month ( 1).
liigh Flux in Startup or Refuel C
Trip Output Relays (4)
Once per week during refueling or startup and before each startup.
liigh Reactor Pressure (6)
B2 Trip Channel and Alarm (4)
Every 1 month
- High Drywell Pressure
'A Trip Channel and Alarm Every 1 month (1)(1).
e;,11igh Drywell Pressure (6)
D2 Trip Channel and Alarm (4)
Every 1 month (1).
Reactor Low Water Level (5) (6)
B2 Trip Channel and Alarm (4)
Every 1 month (1).
Amendment No. 67
TABLE 4.1.1 (cont'd)
REACTOR PROTECTION SYSTEM (SCRAM) INS *IRUMENT FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUlTS Group (2)
Functional Test Minimum Frequency (3)
High Water Imvel In Scram A
Trip channel and Alarm Every 3 months.
Discharge Tank Turbine Condenser Low Vacuum (6)
B2 Trip Channel and Alarm (4)
Every 1 month (1).
Main Steam Line High Radiation B1 Trip Channel and Alarm (4)
Once/ week.
Main Steam Line Isolation A
Trip Channel and Alarm Every 1 month (1).
p Valve Closure Turbine Control Valve A
Trip Channel and Alarm Every 1 month.
EHC Oil Pressure Turbine First Stage Pressure A
Trip Channel and Alarm Every 3 months (1).
Permissive Turbine Stop Valve Closure A
Trip Channel and Alarm Every 1 month (1).
oReactor Pressure Permissive (6)
D2 Trip Channel and Alarm (4)
Every 3 months.
N Reactor Pressure Permissive A
Trip Channel and Alarm Every 3 months, E
o Deleted when modification authorized by Amendment No.
are completed.
CO Effective when modifications authorized by Amendment No.
are completed.
Amendment No. 67
s TABLE 4.1.2 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CliANNELS l
Instrument Channel Group (1)
Calibration (4)
Minimum Frequency (2)
IRM High Flux C
Comparison to APRM on Maximum frequency once Controlled Shutdown per week.
APRM High Flux Output Signal B1 Ile a t Balance Twice per week.
Flow Bias Signal B1 With Standard Pressure Every refueling outage.
Source LPRM Signal B1 TIP System Traverse Every 6 weeks.
g High Reactor Pressure B2 Standard Pressure Source Once per operating OHigh Drywell Pressure A
Standard Pressure Source cycle.
i Once per operating cycle.
Ogligh Drywell Pressure B2 Standard Pressure Source Once per operating cycle.
Reactor Low Water Level B2 Pressure Standard Once per operating cycle.
e R
High Water Level in Scram A
Water Column 3
Discharge volume Every refueling outage.
5 Turbine condenser Low vacuum B2 Standard vacuum Source 2
,o Once per operating cycle.
cn Main Steam Line Isolation A
Note (5)
Note (5).
N valve Closure Main Steam Line liigh Radiation B1 Standard current Source (3)
Every 3 months.
Turbine First Stage Pressure A
Standard Pressure Source Every 6 months.
Permissive O Deleted when modification autho M zed by Amendment So.
are completed.
CO Effective when modifications authorized by Amendment No.
are completed.
TABLE 4.1.2 (cont ' d)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument channel Group (1)
Calitration (4)
MinkmumFrequency (2)
Turbine Control Valve Fast A
Standard Pressure Source Once per ooerating Closure Oil Pressure Trip cycle.
Turbine Stop Valve closure A
Note (5)
Note (5).
fdReactor Pressure Permissive B2 Standard Pressure Source Once per operating Cycle.
CR: actor Pressure Permissive A
Standard Pressure Source Every 6 months
- Deleted when modifications authorized by Amendment No.
are completed
- Effective when modifications authorized by Amendment No.
are completed Amendment No. 67
'(7.e 4
3.1 BASpS The reactor protection system automatically initiates a reactor scram to:
>1.
Preserve the integrity of the fuel cladding.
~
2.
Preserve the integrity of the reactor coolant system.
3.
Minimize the energy which must be absorbed following a loss of coolant accident, and prevent inadvertant crit-icality.
This specification provides th'e limiting conditions for op-eration necessary to preserve the ability of the system to perform its intended function even during periods when in-strument channels may be out of service because of mainten-ance.
When necessary, one channel may be made inoperable f6r brief intervals to conduct required functional tests and calibrations.
The reactor protection system is of the dual channel type (Reference subsection 7.2 FSAR).
The system is made up of two independent trip systems, each having two subchannels of tripping devices.
Each subchannel has an input from at least one instrument channel which monitors a critical para-meter.
The outputs of the subchannels are combined in a 1 out of 2 logic; i.eg an input signal on either one or both of the subchannels will cause a trip system trip.
The outputs of the trip systems are arranged so that a trip on both sys-tems is required to produce a reactor scram.
This system meets the intent of IEEE - 279 for Nuclear Power Plant Protection Systems.
The system has a reliability greater than that of a 2 out of 3 system and somewhat less than that of a 1 out of 2 system.
With the exception of the Average Power Range Monitor (APRM) channels, the Intermediate Range Monitor (IRM) channels, the Main Steam Isolation Valve closure and the Turbine Stop Valve closure, each subchannel has one instrument channel.
When the minimum condition for operation on'the number of operable instrument channels per untripped protection trip system is met or if it cannot be met and the affected pro-tection trip system is placed in a tripped condition, the effectiveness of the protection system is preserved.
The APRM instrument channels are provided for each protec-tion trip system.
APRM's A and E operate contacts in one subchannel and APRM's C and E operate contacts in the other subchannel.
APRM's B, D and P are arranged similarly in October 1973
FBAFS 3 0 BASES (Cont'd) the other protection trip system.
Each protection trip 3
system has one more APRM than is necessary to meet the i
minimum number required per channel.
This allows the bypassing of one APRM per protection trip system for l
maintenance, testing or calibration.
Additional IRM channels have als o b een p rovided to allow for bypassing of one such channel.
The bases for the scram setting f or the IRM, APRM, high reactor pressure, reactor low water level, MSIV closure, Senerator load rejection, turbine stop valve closure and loss of condenser vacuum are discussed in Specification 2.1 and 2.2.
Instrumentation s ensing dryvell p ressure is provided to detect a loss of coolant accident and initiate the core standby cooling equipment.
A high dryvell pressure scram is provided at the same setting as the core standby cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality.
This instrume,ntation is a backup to the reactor vessel water level instrumentation.
High radiation levels in the main steam line tunnel above i
that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel.
A scram is initiated whenever such radiation level exceeds three times normal background..
The purpose of this scram is to limit fission product release so that 10 CTM Part 100 guidelines are not exceeded.
Discharge of excessive amounts of radioactivity to the site environs is prevented by the air ejector off-gas monitors which cause an isolation of the main condenser off-gas line.
A reactor mode switch is provided which actuates or bypasses the various s cram'.f unc ti ons appropriate to the particular plant operating status.
Ref. paragraph 7.2 3.7 TSAR.
The manual scram function is active in all modes, thus providing f or a manual means of rapidly inserting control rods during all modes of reactor operation.
The APRM (High flux in Start-up or Refuel) system provides protection against excessive p ow e r levels and short reactor periods in the start-up and intermediate power ranges.
The IRM' system provides protection against short reactor periods in these ranges.
The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping.
The scram discharge volume accommodates in excess of 50 gallons of water and is the low point in the p ip in g.
No credit was taken for this volume in the design of the d?echarge oiping as concerns l
Amendment No. 67 r
l
TABLE 3. 2.B. (Cont'd)
INSTI(UMENTATION TilAT INITIATES Olt CONTI<OLS Tile CORE AND CONTAINHENT COOLING SYSTEMS Minimum No.
of Operable Number of Instru-In trument Trip Function Trip Level Setting ment Channels Pro Remarks ChEnnels Per vided by Design Trip System (1) 2 Reactor liigh Water 56 45 in. in,dicated 2 Inst. Channels Trips IIPCI and RCIC Level level turbines.
1 Reactor Iow Level 2+312 in. a bove 2 Inst. Channels Prevents inadvertent (inside shroud) vessel zero (2/3 operation of contaisi-core height) ment spray during accident condition.
2 Containment Iligh 1 < p < 2 psig 4 Inst,. Channels Prevents inadvertent Pressure operation of contain-ment (pray during accident condition.
1 Confirmatory Low 2+6 in. indicated 2 Inst, Channels ADS Permissive Level level 2
liigh Orywell 52 psig 4 Inst,. Channels
- 1. Initiates core Spray;
)
Pressure LPCI; ilPCI i
I 2.
Initiates starting of Diesel Generators j
- 3. Initiates Auto Blow-down (ADS) in conjunction with Low-Low Reactor Water Level, 120 uecond time delay, and j
LPCI or Core Spray
[
pump running.
I Amendment No.
67
TABLE 3.2.B (Cont'd)
INSTit0 MENTATION TilAT INITI ATES OR CONTROLS T!!E CORE AND CONTAINHENT COOLING SYSTEMS Minimum No.
cf Operable Number of Instru-Instrument Trip Function Trip Level Setting ment Channels Pro Remarks Ch:nnels Per vided by Design Trip System (1) 2 Reactor Low 400-500 psig 4 Inst. Channels Permissive for opening Pressure Core Spray and LPCI Admission valves.
Coincident with high dry well pressure, starts LPCI and Core Spray pumps.
2 Reactor Low 200-250 psig 4 Inst. Channels Permissive for closing Pressure Uccirculating Pump Discharge Valve.
h 1
Reactor Low 505PS75 psig 2 Inst. Channels In conjunction with PCI Pressure signal permits closure of EllR (LPCI) injection valves.
Amen &ent No.
67
.i t
i j
TABLE 4. 2. B MINIMUM TEST AND CALIBRATION FREQUENCY FOR CSCS
_ Instrument Channel I nst rumen t Functional Test Calibration Frequency Instrument Check 1)
Feactor Water Level (7)
(1) (3)
Once/ operating cycle once/ day
- 2) **Drywell Pressure (7)
(1) (3)
Once/ operating cycle Once/ day l
- Drywell Pressure (1)
Once/3 months None
- 3) ** Reactor Pressure (7)
(1) (3)
Once/ operating cycle Once/ day
- Reactor Pressure (1)
Once/3 months None 4)
Auto Sequencing Timers NA Once/operat ing cycle None 5)
(1)
Once/3 mont.hs None Pressure Interlock 6)
Trip System Bus Poser Monitors (1)
NA None 7)
Core Spray Sparger d/p (1)
Once/6 months Once/ day Y 8)
Steam Line liigh Flow (HPCI G RCIC)
(1)
Once/3 months None
,9)
Steam Line High Temp. (HPCI & RCIC)
(1) (3)
Once/ operating cycle Once/ day
- 10) Safeguards Area High Temp.
( 1)
Once/3 months None
Once/3 months None Low Pressure
- 12) HPCI Suction A urce Levels (1)
Once/3 months None
- 13) 4KV Emergency Power oystem Once/ operating cycle Once/5 year None Voltage Relays
- 15) LPCI/ Cross Connect valve Position Once/ refueling cycle N/A N/A
- Deleted when modification authorized by Amendment No.
are completed.
J ** Effective when modifications authorized by Amendment No.
are completed
^1 Amendment No. 67 1
i !
TABLE'4.2.E MINIMUM TEST AND CALIBRATION FREQUENCY FOR DRYWELL LEAK DETECTION Instrument Channel Instrument Functional Calibration-Instrument
~
Test Frequency Check 1)
Equipment Drain Sump Flow Integrator (1)
Once/3 months Once/ day 2)
Floor Drain Sump Flow Integrator (1)
Once/3 months Once/ day 3)- Air Sampling System (1)
Once/3 months Once/ day 5
L E
P
- Jw i
i
TABLE 4. 2. F MINIMUM TEST AND CALIBRATION FREQUENCY FOR SURVEILANCE INSTRUMENTATION Instrument Channel Calibration Frequency Instrument Check
- 1) * Reactor Level Once/ operating cycle Once Each Shift
- Reactor Level Once/6 months Once each shift 2)
Reactor Pressure Once/6 months Once Each S'hift 3)
Drywell Pressure Once/6 months Once Each Shift 4)
Drywell Temperature Once/6 months Once Each Shift a
5)
Suppression Chamber Temperature once/6 months once Each Shift B-1 6)
Suppression Chamber Water Level Once/6 months Once Each Shift 7)
Control Rod Position NA Once Eacli Shift 8)
Neutron Monitoring (APRM)
Twice Per Week Once Each Shitt O Deleted when modifications authorized by Amendment No, are completed.
Co Effective when modifications authorized by Amendment No, are completed.
Amendment No. 67
-