ML20036C190
| ML20036C190 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/07/1993 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20036C187 | List: |
| References | |
| NUDOCS 9306150188 | |
| Download: ML20036C190 (27) | |
Text
[-.
Amended Technical Soecification Paaes Page:
4 Sa 27 29 32 45 46a 47 53 60 61 68 9306150188 930607 PDR ADOCK 05000293 P
1.0 DEFINITIONS (Cont'd) 1.
At least one door in each access opening is closed.
2.
The standby gas treatment system is operable.
3.
All automatic ventilation system isolation valves are operable or secured in the isolated position.
O.
Operatina Cycle - Interval between the end of 'one refueling outage and the end of the next subsequent refueling outage.
P.
Refuelino Outaae - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant after that refueling.
For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled outage; however, where such outages occur within 11 months of completion of the previous refueling outage, l
the required surveillance testing need not be performed until the next regularly scheduled outage (Definitions U and V apply).
l Q.
Alteration of the Reactor Core - The act of moving any component in the region above the core support plate, below the upper grid and within the shroud.
Normal control rod movement with the control rod drive hydraulic system is not defined as a core alteration.
Normal movement of in-core instrumentation is not defined as a core alteration.
R.
Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.
S.
Thermal Parameters 1.
Minimum Critical Power Ratio (MCPR) - the value of critical power ratio associated with the most limiting assembly in the reactor core.
Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated to cause some point in the assembly to experience boiling transition, to the actual assembly operating power.
2.
Transition Boilina - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.
3.
Total Peakina Factor - The ratio of the fuel rod surface heat flux to the heat flux of an average rod in an identical I
geometry fuel assembly operating at the core average bundle power.
(
Amendment No. 15 4
1.0 DEFINITIONS (Continued)
U.
Surveillance Frecuency - Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.
The Surveillance Frequency establishes the limit-for which the specified time interval for Surveillance Requirements may be extended.
It permits an allowable extension of the normal surveillance interval to facilitate surveillance schedule and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities.
It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances that are not performed during refueling outages. The limitation of-Definition "U" is based on engineering judgment and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.
V.
Sure.11ance Interval - The surveillance interval is the calendar time between surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable. These tests.may be waived when the instrument, component, or system is not. required to be operable, but the instrument, component, or system shall be tested prior to being declared operable. The operating cycle interval is 24 months and l
the 25% tolerance given in Definition "U" is applicable.
W.
Fire Suppression Water System - A fire suppression water system shall consist of:
a water source (s); gravity tank (s) or pump (s);
and distribution piping with associated sectionalizing control or isolation valves.
Such valves shall include hydrant post indicator valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser.
X.
Staccered Test Basis - A staggered test basis shall consist of:
(a) a test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals; (b) the testing of one system, subsystem, train or other designated components at the beginning of each subinterval.
Y.
Source Check - A source check shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
Amendment No. 42, 89, 128, 5a
PNPS Table 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT 4
- Operable Inst.
Modes.in Which Function-Channels per Trip Function Trip Level Setting Must Be Operable Action (1}'
Trio System (1)
Refuel (7) Startup/llot Run MinimumlAvail.
Standby 1
1 Mode Switch in Shutdown X
X X
A 1
1 Manual Scram X-X X
A IRM 3
4 liigh Flux s120/125 of full scale X
X (5)
A 3
4 Inoperative X
X (5)
A APRM 2
3 liigh Flux (15)
(17)
(17)
X A or 8 2
3 Inoperative (13)
X X(9)
X A or B 2
3 liigh Flux (15%)
sl5% of Design Power X
X (16)
A or B 2
2.
High Reactor Pressure s1063.5 psig X(10)
X X
- A 2
-. 2
.iligh Drywell Pressure s2.22 psig X(8)
X(8)
X A
2 2
Reactor Low Water Level
=11.7 In. Indicated Level X
'X X
A
-SDIV liigh Water Level:
s39 Gallons X(2)
X X
A 2
2 East' 2
2 West 2
2 Main Condenser Low Vacuum
=23 In. lig Vacuum X(3)
X(3)
X A or C 2
2 Main-Steam Line liigh s7X Normal Full Power
. Radiation Background (18)
X X
X(18)
A or C 4
4 Main Steam Line Isolation Valve Closure s10% Valve Closure
_X(3)(6)
X(3)(6)
X(6)
..A or C 2
2 Turbine Control Valve
=150 psig Control Oil Fast closure Pressure at Acceleration Relay X(4)
X(4)
X(4)
A or D 4
4 Turbine Stop Valve s10% Valve Closure X(4)
X(4)
'X(4)
A or 0 Closure 15 -425-86 117, 133, 147, 127 Amendment No.
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NOTES FOR TABLE 3.1.1 (Cont'd) 2.
Permissible to bypass, with control rod block, for reactor protection system reset in refuel and shutdown po:.itions of the reactor mode switch.
3.
Permissible to bypass when reactor pressure is <576 psig.
4.
Permissible to bypass when turbine first stage pressure is less than 112 psig.
5.
IRM's are bypassed when APRM's are onscale and the reactor mode switch is in the run position.
6.
The design permits closure of any two lines without a scram being initiated.
7.
When the reactor is subcritical, fuel is in the reactor vessel and the reactor water temperature is less than 212 F, only the following trip functions need to be operable:
A.
Mode switch in shutdown B.
Manual scram C.
High flux IRM D.
Scram discharge volume high level j
E.
Not required to be operable when primary containment integrity is not required.
9.
Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).
- 10. Not required to be operable when the reactor pressure vessel head is not bolted j
to the vessel, i
- 11. Deleted i
- 12. Deleted
- 13. An APRM will be considered inoperable if there are less than 2 LPRM inputs per j
level or there is less than 50% of the normal complement of LPRM's to an APRM.
- 14. Deleted i
- 15. The APRM high flux trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT, but shall in no case exceed 120% of rated thermal l
power.
startup/ hot standby modes.
- 18. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power radiation j
background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen.
The background radiation level and associated trip setpoints may be adjusted based on either calculations or' measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 20% rated power.
1
)
Amendment No. 6, 15, 27, 42, 86, 117, 118, 133, 147, 29
TABLE 4.1.2 REAC10R PROTECTION SYSTEM (SCRAM) INSTRUMENT CAllBRATION MINIMUM CAllBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Calibration Test (5)
Minimum Freauency (2)
IRM liigh Flux Comparison to APRM on Controlled Note (4)
Shutdowns Full Calibration Once per Operating Cycle APRM High Flux Output Signal lleat Balance Once every 3 Days Flow Bias Signal Calibrate Flow Comparator and At least Once Every Flow Bias Network 18 Months Calibrate Flow Bias Signal (1)
Every 3 Months LPRM Signal TIP System Traverse Every 1000 Effective Full Power llours liigh Reactor Pressure Note (7)
Note (7) liigh Drywell Pressure Note (7)
Note (7)
Reactor Low Water Level Note (7)
Note (7) liigh Water level in Scram Discharge Tanks Note (7)
Note (7)
Turbine condenser Low Vacuum Note (7)
Note (7)
Main Steam Line isolation Valve Closure Note (6)
Note (6)
Main Steam Line liigh Radiation Standard Current Source (3)
Every 3 Months Turbine First Stage Pressure Permissive flote (7)
Note (7)
Turbine Control Valve Fast Closure Standard Pressure Source Every 3 Months Turbine Stop Valve Closure Note (6)
Note (6)
Reactor Pressure Permissive Note (7)
Note (7)
'32 Amendment No. 147,
PNPS TABLE 3.2.A INSTRUMENTATION TilAT INITIATES PRIMARY CONTAINMENT ISOLATION Operable Instrument Cha1,nels Per Trip System (1)
Minimum Available Instrument Trip Level Settinn Action (2) 2(7) 2
- leactor Low Water Level
>11.7" indicated level (3)
A and D
. i 1
1 Reactor liigh Pressure
$110 psig D
l 2
2 Reactor Low-Low Water Level at or above -46.3 in.
A indicated level (4)
-2 2
Reactor liigh Water Level 545.3" indicated level (5)
B
. l-2(7) 2 liigh Drywell Pressure 52.22 psig A
4 2-2 liigh Radiation Main Steam
$7 times normal' rated B
Line Tunnel (9) full power background 2
2 Low Pressure Main Steam Line 2810 psig (8)
B 2(6) 2 liigh Flow Main Steam Line.
5140% of rated steam flow B
2.
2 Main Steam Line Tunnel Exh aust Duct liigh Temperature
$170 F B
1 2
2 Turbine Basement Exhaust Duct liigh Temperature
$150 F B
1 1
Reactor Cleanup System liigh Flow
$300% of rated flow C
2 2
2 Reactor Cleanup System i
liigh Temperature
$150 F
.C Amendment No. 86, 147 45
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i 3.
Instrument set point corresponds to 130.96 inches above top of active fuel.
l 4.
Instrument set point corresponds to 79.96 inches above top of active fuel.
l 5.
Not required in Run Mode (bypassed by Mode Switch).
6.
Two required for each steam line.
l 7.
These signals also start SBGTS and initiate secondary containment isolation.
8.
Only required in Run Mode (interlocked with Mode Switch).
9.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of hydrogen injection with the
-l reactor power at greater than 20% rated power,- the normal full power radiation
-l background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen.
The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-i establishing normal radiation levels after completion of hydrogen injection and
. prior to withdrawing control rods at reactor power levels below 20% rated power.
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t Amendment No. 147 46a
4 PNPS TABLE 3.2.B INSTRUMENTATION TilAT INITIATES OR CONTROLS Tile CORE AND CONTAINMENT COOLING SYSTEMS' Minimum #'of 1
Operable Instrument Channels Per Trip System (1)
Trio Function Trip Level Settina Remarks 2
Reactor Low-Low Water Level at or above -46.3 in.
1.
In conjunction with Low.
l.
indicated level (4)
Reactor Pressure, initiates Core Spray and LPCI.
2.
In conjunction with liigh Drywell Pressure, 120 second time delay and LPCI or Core spray pump interlock initiates Auto Blowdown (ADS).
3.
Initiates llPCI; RCIC.
4.
Initiates starting of Diesel Generators.
2 Reactor liigh Watet Level
< +45 3" indicated Trips IIPCI and RCIC turbines'.
l-level 1
Reactor Low Level 2307" above vessel Prevents inadvertent operation (inside shroud) zero (approximately-of containment spray during 2/3 core height) accident condition.
2 Containment liigh Pressure 1 < p < 2 psig
-Prevents-inadvertent operation-of containment spray during accident condition.
Amendment No. 90' 47
. ~
2-.
.2..
. -...a. --
NOTES'FOR TABLE 3.2.B
]
1.
Whenever. any CSCS subsystem is required by Section 3.5 to be operable, there.
shall be two (Note 5) operable trip systems.
If the first column cannot be met
'l for one of the trip systems, that system shall be repaired or the reactor shall-be placed-in the Cold, Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> -after this trip ; system is made or found to be inoperable.
2.
Close isolation valves in RCIC subsystem.
3.
Close isolation valves in HPCI subsystem.
4.
Instrument set point corresponds to 79.96 inches of active fuel.
l i
5.
RCIC has only one trip system for these sensors.
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1 53 Amendment No. 105, 148 l
PNPS TABLE 4.2.A MINIMUM TEST AND CALIBRATION FREQUENCY _FOR PCIS Instrument Channel (5)
Instrument Functional Test Calibration Freauency Instrument Check 1)
Reactor High Pressure (1)
Once/3 months None 2)
Reactor low-Low Water level Once/3 Months (7)
(7)
Once/ day
-3)
Reactor liigh Water Level Once/3 Months (7)
(7)
Once/ day 4)
Main Steam liigh Temp.
'(1)
Once/3 months None 5)
Main Steam liigh Flow (1) (7)
(7)
Once/ day 6)
Main Steam Low Pressure Once/3 Months (7)
(7)
_Once/ day
.l.
7)
Reactor Water Cleanup liigh Flow (1)
Once/3 months once/ day.
8)
Reactor Water Cleanup High Temp.
(1)
Once/3 months None logic System functional Test (4) (6)
Frecuency 1)
Main Steam Line Isolation Vvs.
Once/18 months Main Steam Line Drain Vvs.
Reactor-Water Sample Vvs.
2)
RilR - Isolation Vv. Control Once/18 months Shutdown. Cooling Vvs.
Ilead Spray Discharge-to Radwaste 3)
Reactor Water Cleanup Isolation once/18 months 4)
Drywell isolation Vvs.-
Once/18 months TIP-Withdrawal
-Atmospheric Control Vvs.
Sump Drain Valves" 5)
Standby Gas Treatment System Once/18 months Reactor-Building Isolation Amendment No. 107, 130 60
-TABLE'4.2.B MINIMUM TEST AND CALIBRATION FREQUENCY FOR CSCS Instrument Channel (5)-
Instrument Functional Test Calibration Freauency Instrument Check 1)
Reactor Water Level Once/3 Months (7)
(7)
Once/ day 2)
Drywell Pressure (1) (7)
(7)
Once/ day 3)
Reactor Pressure (1) (7)
(7)
Once/ day 4)
Auto Sequencing Timers NA Once/ operating cycle None 5)
Pressure Interlock (1)
Once/3 months None 6)
Start-up Transf. (4160V) a.
Loss of Voltage Relays Monthly-Once/ operating cycle None b.
Degraded Voltage Relays Monthly' Once/ operating cycle None 7)
Trip System Bus Power Monitors Once/ operating cycle NA Once/ day 8)
Recirculation System d/p (1)
Once/3 months Once/ day 9)
Core Spray Sparger d/p NA Once/ operating cycle Once/ day
- 10) Steam Line liigh Flow (llPCI & RCIC)
(1)
Once/3 months None 11)
Steam Line liigh Temp. (llPCI & RCIC)
(1)
Once/3 months None
- 12) Safeguards Area liigh Temp.
(1)
Once/3 months None 13)
RCIC Steam Line Low Pressure (1)
Once/3 months None
- 14) 11PCI Suction Tank Levels (1)
Once/3 months None 15)
Emergency 4160V Buses A5 & A6 Monthly Once/ operating Cycle None Loss of Voltage Relays 61 Amendment No. 42,-61;-99, 148
BASES:
3.2 In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block, and standby gas treatment systems. The objectives of the Specifications are, (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.
When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.
The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required.
Such instrumentation must be available whenever primary con binment integrity is required.
i The instrumentation which initiates primary system isolation is connected in a i
dual bus arrangement.
l The low water level instrumentation set to trip at 130.96 inches above-the l
l top of the active fuel closes all isolation valves except those in Groups 1, 4 and 5.
This trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second valve closing time.
Required closing times are less than this.
The low low reactor water level instrumentation is set to trip when reactor water level is 79.96 inches above the top of the active fuel
(-46.3" on the instrument). This trip closes Main Steam Line Isolation.
l l
l Amendment No. 105, 113 68
e 5
6 Attachment C to BEco letter 93-072.
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1.0.DEFI'NITIONS (Cont'd)
At least one door in cach access opening is c1csed.
1.
e The standby gas treat =ent system is operable.,
2.
All automatic venti 1'ation system isolation valves are operable 1
3.
or secured in the isolated position.
Operatino Cycle - Interval between the end of one refueling outage 0.
~and tne enc of the next subsequent refueling outage.
Refueli'no Outace - Refueli69 outage is the period of time between P.
tne snutcown of the unit prior to a refueling and the startup of For the purpose of designating the plant after that refueling.
frequency of testing and surveillance, a refueling outage shall mean a regulagiy scheduled outage; however, where such outages occur witnin(&,1 months of the conpletion of the previous refueling y
outage, the required surveillance. testing need not be performed until the next regularly scheduled outagg (p u a VQ)
Alteration of the Reactor Core - The act of mo'ving any ce=ponent Q.
in tne region acove tne core support plate, below the upper grid Normal control rod movement with the control and within the shroud.
rod drive, hydraulic system is not defined as a core alteration.
Hom.a1 movement of in-core instrumentation is not defined as a core alteration.
Reactor Vessel Pressure - Unless otherwise indicated', reactor R.
vessel pressures 11:;teo in the Technical Specifications are those measured by the reactor vessel steam space detectors.
i S.
Thermal Parameters y
7 Mini:-u. C-itical ?cver Ratio (ML"HO - the value of critical 1.
power ratio asse:1sted with the most limiting asse=bly in the
~
Critical Power Ratio (CPR) is the ratio of that reacter core.
power in a fuel assechly, which is eticulated to cause so=e point in the assembly to experience boiling transitien, to the j
actual asse bly operating power.
':tensition Boiling - Ott.nsition boiling neans the boiling Transition boiling 2.
regime between nucleate and film boiling.
is the rerf.ee in which both nucleate and film boiling occur inter =ittetly with neither type being completely stable.
Total Peakirc Tacter - The ratio of the fuel rod surface best 3
~ flux to the beat flux of r.n average.-od in an identical geometry fuel assembly cperating at the cere averge bt:sile pcver.
~J (namen. E er
i 1.0 DEFINfT10M (Continued)
U.
Surveillance Frecuency - Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum i
allowable extension not to exceed 25 percent of the specified
??
surveillance interval.
The Surveillance Frequency establishes the limit for which the
< ~'
specified time interval for Surveillance Requirements may be extended.
It permits an allowable extension of the normal surveillance interval to facilitate surveillance schedule and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other 1(,,
ongoing surveillance or maintenance activities. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specifiEd for surveillantes that are not performed during refueling outages.
The limitation of k
Definition "U" is based on engineering judgment and the recognition J
that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements.
This provision is sufficien,t to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified r
surveillance interval.
4 V.
Surveillance Interval - The surveillance interval is the calendar time between surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable.
These tests may be waived when the instrument, component, or system is not required to be operable, but the instrument, component, or system shall be teste ior to being I
declared operable. The operating cycle interval i months and the 257. tolerance given in Definition "U" is applica N.4
/
2 W.
Fire Sucoression Water System - A fire suppression water system shall consist of: a water source (s); gravity tank (s) or pump (s);
and distribution piping with associated sectionalizing control or isolation valves.
Such valves shall include hydrant post indicator valves and the first valve ahead of the water flow alara device on each sprinkler, hose standpipe or spray system riser.
X.
Staacered Test Basis - A staggered test basis shall consist of:
(a) a test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals; (b) the testing of one system, subsystem, train or other designated components at the beginning of each subinterval.
Y.
Source Check - A source check shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
Ame E 89 Sa
PNPS Table 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Operabie Inst.
Modes in Which Function Channels per Trip Function
-Trip Level Setting Must Be Operable Action (I)
Trip System (1)
Refuel (7) Startup/ Hot Run MinimumlAvail.
Standby 1
1 Mode Switch in Shutdown X
X X
A 1
1 Manual Scram X
X X
A 4-IRM 3
4 liigh Flux s120/125 of full scale X
X (5)
A 3
4 Inoperative X
X (5)
A APRM 2
3 liigh Flux (15)
(17)
(17)
X A or B 2
3 Inoperative (13)
X X(9)
X A or B 2
3 liigh Flux (15%)
sl5% of Design Power X
X (16)
A or 8 2
2 High Reactor Pressure X(10)
X X
A
.m 2
2 High Drywell Pressure s2.5 /psig X(8)
X(8)
X A
2 2
Reactor low Water Level te vel X
X X
A SDIV liigh Water level:
s39 Gallons X(2)
X X
A 2
2 East 2
2 West 2
2 Main Condenser low Vacuum
=23 In. Hg Vacuum X(3)
X(3)
X A or C 2
2 Main Steam Line liigh s7X Normal Full Power Radiation Background (18)
X X
X(18)
A or C 4
4 Main Steam Line Isolation Valve Closure s10% Valve Closure X(3)(6)
X(3)(6)
X(6)
A or C 2
2 Turbine Control Valve
=150 psig Control Oil Fast Closure Pressure at Acceleration Relay X(4)
X(4)
X(4)
A or D 4
4 Turbine Stop Valve s10% Valve Closure X(4)
X(4)
X(4)
A.or D Closure ichh na-u_
kS b e M C 15 -421-86 117 -133, 27 1
1 7
1
i NOTES FOR TABLE 3.1.1 (Cont'd) 2.
Permissible to bypass, with control rod block, for reactor protection system reset in refuel and shutdown positions of the reaA or mode switch.
y pf 3.
Permissible to bypass when reactor pressure is gpsig.
4.
Permissible to bypass when turbine first stage pressure is less tha psig.jg 5.
IRM's are bypassed when APRM's are onscale and the reactor mode swit is in the run position.
6.
The design permits closure of any two lines without a scram being initiated.
7.
When the reactor is subcritical, fuel is in the reactor vessel and the reactor water temperature is less than 212 F, only the-following trip functions need to be operable:
A.
Mode switch in shutdown B.
Manual scram C.
High flux IRM D.
Scram discharge volume high level E.
Not required to be operable when primary containment integrity is not required.
9.
Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).
- 10. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
- 11. Deleted i
- 12. Deleted
- 13. An APRM will be considered inoperable if there are less than 2 LPRM inputs per -
i level or there is less than 50% of the normal complement of LPRM's.to an APRM.
- 14. Deleted
- 15. The APRM high flux trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT, but shall in no case exceed 120% of rated thermal power.
- 17. The APRM flow biased high flux scram is bypassed when in the refuel or startup/ hot standby modes.
- 18. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power radiation i
background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen.
The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 20% rated power,
&s lAE. MLA -
6,15,27,42,86,117,118,133,h 29 Amenoment No.
TABLE 4.1 2 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CAllBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CilANNELS Instrument Channel Calibration Test (5)
Minimum Freauency (2)
IRM liigh Flux Comparison to APRM on Controlled Note (4)
Shutdowns Full Calibration Once/ Operating Cycle APRM liigh Flux Output Signal lleat Balance Once every 3 Day q f(Gaeh-RTfueHtig-outyge Flow Bias Signal Calibrate Flow Comparator and Flow Bias Network Calibrate Flow Bias Signal (1)
Every 3 Months LPRM Signal TIP System Traverse Every 1000 Effective Full Power flours liigh Reactor Pressure Note (7)
AT AEAST ogcg Note (7) liigh Drywell Pressure Note (7)
Y N M#M Note (7)
Reactor low Water Level Note (7)
Note (7) liigh Water Level in Scram Discharge Tanks Note (7)
Note (7)
Turbine Condenser low Vacuum Note (7)
Note (7)
Main Steam Line Isolation Valve Closure Note (6)
Note (6)
Main Steam Line liigh Radiation Standard Current Source (3)
Every 3 Months Turbine First Stage Pressure Permissive Note (7)
Note (7)
Turbine Control Valve Fast Closure Standard Pressure Source Every 3 Months Turbine Stop Valve Closure Note (6)
Note (6)
Reactor Pressure Permissive Note (7)
Note (7)
@disie _ _
~
32 Amendment No.
e O
O PNPS TABLE 3.2.A INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT IS01ATION Operable Instrument Channels Per Trip System (1)
Minimum l Available Instrument y
. Trip Level Settinn Action (2) 2(7) 2 Reactor Low Water Level indicated level (3)
A and D j
/
1 1
Reactor liigh Pressure
$110 psig D
2 2
Reactor Low-Low Water Level at or above A
/
indicated level (4) f 4'_l23"
/,
2 2
Recetor liigh Water Level Q
B" indicated level (5)
B
/,
2(7) 2 liigh Drywell Pressure f g, g - [ psig A
/
j, 2
2 Iligh Radiation Main Steam
<7 times normal rated B
Line Tunnel (9) full power background
[l Uly 2
2 Low Pressure Main Steam Line psig (8)
B
/
/![j 2(6) 2 liigh Flow Main Steam Line
$140% of rated steam flow B
2 2
Main Steam Line Tunnel
/
Exhaust Duct liigh Temperature 5170 F B
f.
2 2
Turbine Basement Exhaust y -
Duct liigh Temperature
$150 F B
!(
l 1
Reactor Cleanup System e
liigh Flow
$300% of rated flow C
/
/ '
2 2
Reactor Cleanup System j
liigh Temperature
$150 F C
.s n.
, c cX, w_....
Amendment No. 86, 147 45
13o. %
t 3.
Instrument set point corresponds to 120.26 inches above top of active fuel
- 99,96 4.
Instrument set point corresponds to 77.26 inches above top of active fuel.
5.
Not required in Run Mode (bypassed by Mode Switch).
6.
Two required for each steam line.
7.
These signals also start SBGTS and initiate secondary containment isolation.
8.
Only required in Run Mode (interlocked with Mode Switch).
9.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen.
The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection.
The background i
radiation level shall be determined and associated trip setpoints shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 20% rated power.
1 h/ici 165b Amendmentlo.g 46a
'f PNPS TABLE 3.2.B INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Hinimum # of Operable Instrument Channels Per Trip _ System (1)
Trip Function Trip level Setting
'D3 Remarig 2'
Reactor low-Low Hater at or above -
in.
1.
In conjunction with low Level Indicated level (4)
Reactor Pressure, initiates Core ';pi ay and LPCI.
2.
In conjunction with lingh Drywell Pressure, 120 second time delay and LPCI or Core Spray pump interlock initiates Auto Blowdown (ADS).
3.
4 45,3'#
4.
Initiates starting of
/
Diesel Generators.
2 Reactor High Hater Level
_ g indicated Trips HPCI and RCIC turbines.
level 1
~~seactor low Level
>307" above vessel Prevents inadvertent operation (Inside shroud) iero (approximately of containment spray during f
]g 2/3 core height) accident condition.
i of containment spray during accident condition.
k' lk ! ' k h.
NOTES FOR TABLE 3.2.B 1.
Whenever any CSCS subsystem is required by Section 3.5 to be operable, there shall be two (Note 5) operable trip systems.
If the first column cannot be met for one of the trip systems, that system shall be repaired or the reactor shall be placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after this trip system is made or found to be inoperable.
2.
Close isolation valves in RCIC subsystem.
3.
Close isolation valves in HPCI subsystem. p gg,4 M
4.
Instrument set point corresponds tc Q inches of active fuel.
5.
RCIC has only one trip system for these sensors.
i Q : d 108, h n?%
mendmir 53
f(< 7h s....
PNPS TABLE 4.2.A HINIMUM TEST AND CALIBRATION FREOUENCY FOR PCIS Instrument Channel (5)
Instrument Functional Test Calibration Freqttency Instrument Check 1)
Reactor High Pressure (1
Once/3 months None 2)
Reactor Low-Low Hater Level 4 (7)
(7)
Once/ day 3
3)
Reactor High Hater Level (7)
(7)
Once/ day 4)
Hain Steam High Temp.
(1)
Once/3 months None 5)
Hain Steam High Flow (1) (7)
(7)
Once/ day 6)
Main Steam Low Pressure (7)
(7)
Once/ day 7)
Reactor Water Cleanup High Flow (1)
Once/3 months Once/ day 8)
Reactor Water Cleanup High Temp.
(1)
Once/3 months None nbN 3 M
FreauencJ f
Loaic System Functional Test (4) (6) 1)
Hain Steam Line Isolation Vvs.
Once/IB months
/
Main Steam Line Drain Vvs.
/
Reactor Hater Sample Vvs.
/
2)
RHR - Isolation Vv. Control Once/iB months
/
Shutdown Cooling Vvs.
Head Spray
/
Discharge to Radwaste
/
3)
Reactor Hater Cleanup Isolation Once/18 months
/
Once/18 months 4)
Drywell Isolation Vvs.
TIP Hithdrawal Atmospheric Control Vvs.
Sump Drain Valves y
Once/18 months
~j 5)
Standby Gas Treatment System Reactor Building Isolation h-lan30 60 Amendment No. 107,
PHPS TABLE 4.2.B MINIMUM TEST AND CAllBRATION FREQUENCY FOR CSCS Instrument Channel Instrument Functional Test Calibration Frequency Instrument Check 1)
Reactor Water Level
~
7)
(7)
Once/ day 2)
Drywell Pressure (1) (7)
(7)
Once/ day 3)
Reactor Pressure (1) (7)
(7)
Once/ day 1
4)
Auto Sequencing Timers NA Once/ Operating Cycle None 5)
Pressure Interlock (1)
Once/3 months None 6)
Start-up Transf. (4160V) a.
Loss of Voltage Relays Monthly Once/ Operating Cycle None b.
Degraded Voltage Relays Monthly Once/ Operating Cycle None 7)
Trip System Bus Power Monitors once/ operating cycle NA Once/ day 8)
Recirculation System d/p (1)
Once/3 months once/ day 9)
Core Spray Sparger d/p NA Once/ Operating Cycle Once/ day 10)
Steam Line liigh Flow (llPCI & RCIC)
(1)
Once/3 months None 11)
Steam Line liigh Temp. (llPCI & RCIC)
(1)
Once/3 months None 12)
Safeguards Area liigh Temp.
(1)
Once/3 months None 13)
RCIC Steam Line low Pressure (1)
Once/3 months None 14) llPCI Suction Tank levels (1)
Once/3 months None 15)
Emergency _4160V Buses A5 & A6 Monthly Once/ Operating Cycle None loss of Voltage Relays
@ bie- ! M 61 Amendinent No. 42;-61,-99,
i i
BASES 1 3.2 In' addition 'to reactor protection instrumentatio' n which initiates a reactor scram, protective instrumentation has been provided which
- i L
initiates action to mitigate the' consequences of accidents which are beyond the operator's ability to control, or terminates operator errors-before they result in serious consequences.: This set of specifications-provides the limiting conditions of operation for ~ the primary system isolation function, initiation of the core cooling systems', control rod block and standby gas treatment systems. The objectives of the; Specifications are (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate.a-
' 3 single failure of any componert of such systems even during periods when t
portions of such systems are out of service' for maintenance, and (ii) to.
prescribe the trip settings required to assure adequate performance.
When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.
The set points of other instrumentation, where only the high or low end of the setting has a direct. bearing on safety, are chosen at-a level away from the normal operating range to prevent inadvertent
- r actuation of the safety system involved and exposure to abnormal situations.
Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for p}
which isolation is required.
Such instrumentation must be available whenever primary containment integrity is required.
The instrumentation which initiates primary system isolation is-connected in a dual bus arrangement.
The low water level instrumentation set. to trip 'at fBr inches above the top of the active fuel closes all isolation valve xcept those in Groups 1, 4 and 5.
This trip setting is adequate to prevent core i
uncovery in the case of a break in the largest line assuming a 60 second valve closing time.
Pequired closing times are less than this.
j The low low reactor wat A e1 instrumentation is set to trip when re +or water level is 7.U inches above the top of the active fuel
' on the instrument).
This trip closes Main Steam Line Isolation I
4.3 7Mf I
r 1
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Amenoment
- 705, 68
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