ML18033B544
| ML18033B544 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/30/1990 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18033B543 | List: |
| References | |
| NUDOCS 9011090365 | |
| Download: ML18033B544 (51) | |
Text
ENCLOSURE 1
PROPOSED TECHNICAL SPECIFICATION BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 (TVA BFN TS 288)
J 9'0li09'0365 901030 PDR
- DOCK D5000259 P
PNU
UNIT 1 EFFECTIVE PAGE LIST REMOVE INSERT 1.0-11 3.2/4.2-12 3.2/4.2-13 3.2/4.2-63a 3.2/4.2-69 3.5/4.5-30 3.5/4.5-31 3.7/4.7-2 1.0-11 3.2/4.2-12 3.2/4.2-13 3.2/4.2-63a 3.2/4.2-69 3.5/4.5-30 3.5/4.5-31 3.7/4.7-2
1.0 GG.
Site Bou da Shall be that line beyond which the land is not owned, leased, or otherwise controlled by TVA.
U s
ted rea Any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for industrial, commercial, institutional, or recreational purposes.
ose u v The DOSE EQUIVALENT I-131 shall be the concentration of 1-131 (in pdi/dm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.
The thyroid dose conversion factor used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Reactor Sites".
ou W st a
S te The charcoal adsorber vessels installed on the discharge of the steam jet air ejector to provide delay to a unit's offgas activity prior to release.
embers of t e Pub c Shall include all individuals who by virtue of their occupational status have no formal association with the plant.
This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions.
This category shall Dot include non-employees such as vending machine servicemen or postmen
- who, as part of their formal job function, occasionally enter restricted areas.
the OPERATIONAL CONDITIONS or other conditions specified for individual limiting conditions for operation unless otherwise stated in an individual Surveillance Requirements.
Each Surveillance Requirement shall be performed within the specified time interval with, (1) A maximum allowable extention not to exceed 25% of the surveillance interval, but (2) The combined time entered for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval Performance of a Surveillance Requirement within the specified time interval shall constitute compliance and OPERABILITY requirements for a limiting condition for operation and associated action statements unless otherwise required by these specifications.
Surveillance Requirements do not have to be performed on inoperable equipment.
BFN Unit 1 1.0-11
OTES FOR T B E 1.
Whenever the respective functions are required to be OPERABLE there shall be two OPERABLE or tripped trip systems for each function. If the first column cannot be met for one of the trip systems, that trip system or logic for that function shall be tripped (or the appropriate action listed below shall be taken). If the column cannot be met for all trip
- systems, the appropriate action listed below shall be taken.
A. Initiate an orderly shutdown and have the reactor in Cold Shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an orderly load reduction and have Main Steam Lines isolated within eight hours.
C.
Isolate Reactor Water Cleanup System.
D.
Isolate Shutdown Cooling.
E.
Initiate primary containment isolation within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
F.
The handling of spent fuel will be prohibited and all operations over spent fuels and open reactor wells shall be prohibited.
G.
Isolate the reactor building and start the standby gas treatment system.
H.
Immediately perform a logic system functional test on the logic in the other trip systems and daily thereafter not to exceed 7 days.
I.
Deleted J.
Withdraw TIP.
K.
Manually isolate the affected lines.
Refer to Section 4.2.E for the requirements of an inoperable system.
L. If one SGTS train is inoperable take actions H or A and F. If two SGTS trains are inoperable take actions A and F.
2.
Deleted 3.
There are four sensors per steam line of which at least one sensor per trip system must be OPERABLE.
BFN Unit 1 3.2/4.2-12
NOTES FOR ABLE 2
ont'd) 4.
Only required in RUN MODE (interlocked with Mode Switch).
5.
Deleted 6.
Channel shared by RPS and Primary Containment 6 Reactor Vessel Isolation Control System.
A channel failure may be a channel failure in each system.
7.
A train is considered a trip system.
8.
Two out of three SGTS trains required.
A failure of more than one will require actions A and F.
9.
Deleted 10.
Refer to Table 3.7.A and its notes for a listing of Isolation Valve Groups and their initiating signals.
ll.
A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
12.
A channel contains four sensors, all of which must be OPERABLE for the channel to be OPERABLE.
Power operations permitted for up to 30 days with 15 of the 16 temperature switches OPERABLE.
In the event that normal ventilation is unavailable in the main steam line tunnel, the high temperature channels may be bypassed for a period of not to exceed four hours.
During periods when normal ventilation is not available, such as during the performance of secondary containment leak rate tests, the control room indicators of the affected space temperatures shall be monitored for indications of small steam leaks.
In the event of rapid increases in temperature (indicative of steam line break),
the operator shall promptly close the main steam line isolation valves.
13.
The nominal setpoints for alarm and reactor trip (1.5 and 3.0 times background, respectively) are established based on the normal background at full power.
The allowable setpoints for alarm and reactor trip are 1.2-1.8 and 2.4-3.6 times background, respectively.
14.
Requires two independent channels from each physical location; there are two locations.
BFN Unit 1 3.2/4.2-13
Table 4.2.L Anticipated Transient Without Scram (ATWS)
Recirculation Pump Trip (RPT) Instrumentation Surveillance u ct o Functional est Channel Cal bratio Instrument C ec Reactor Vessel Water Level Low (LS-3-58Al-Dl)
Reactor Vessel Dome Pressure High (PIS-3-204A-D)
M(27)
M(27)
R(28)
R(28)
N/A N/A BFN Unit l 3.2/4.2-63a
3 '
~BUS (Cont'd)
Trip setting of 100 mr/hr for the monitors in the refueling zone are based upon initiating normal ventilation isolation and SGTS operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the SGTS.
Flow integrators and sump fillrate and pump out rate timers are used to determine leakage in the drywell.
A system whereby the time interval to filla known volume will be utilized to provide a backup.
An air sampling system is also provided to detect leakage inside the primary containment (See Table 3.2.E).
For each parameter monitored, as listed in Table 3.2.F, there are two channels of instrumentation except as noted.
By comparing readings between the two channels, a near continuous surveillance of instrument performance is available.
Any deviation in readings will initiate an early recalibration, thereby maintaining the quality of the instrument readings.
Instrumentation is provided for isolating the control room and initiating a pressurizing system that processes outside air before supplying it to the control room.
An accident signal that isolates primary containment will also automatically isolate the control room and initiate the emergency pressurization system.
In addition, there are radiation monitors in the normal ventilation system that will isolate the control room and initiate the emergency pressurization system.
Activity required to cause automatic actuation is about one mRem/hr.
Because of the constant surveillance and control exercised by TVA over the Tennessee Valley, flood levels of large magnitudes can be predicted in advance of their actual occurrence.
In all cases, full advantage will be taken of advance warning to take appropriate action whenever reservoir levels above normal pool are predicted.
Therefore, during flood conditions, the plant will be permitted to operate until water begins to run across the top of the pumping station at elevation 565.
Seismically qualified, redundant level switches each powered from a separate division of power are provided at the pumping station to give main control room indication of this condition.
At that time an orderly shutdown of the plant will be initiated, although surges even to a depth of several feet over the pumping station deck will not cause the loss of the main condenser circulating water pumps.
BFN Unit 1 3.2/4.2-69
3.5
/~SOS (Cont'd) 3.5.E.
i essu e Coo a t ec ion S stem PC S
The HPCIS is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the nuclear system and loss of coolant which does not result in rapid depressurization of the reactor vessel.
The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.
The HPCIS continues to operate until reactor vessel pressure is below the pressure at which LPCI operation or Core Spray system operation maintains core cooling.
The capacity of the system is selected to provide the required core cooling.
The HPCI pump is designed to pump 5000 gpm at reactor pressures between 1120 and 150 psig.
The HPCIS is not required to be OPERABLE below 150 psig since this is well within the range of the low pressure cooling systems and below the pressure of any events for which HPCI is required to provide core cooling.
The minimum required NPSH for HPCI is 21 feet.
There is adequate elevation head between the suppression pool and the HPCI pump, such that the required NPSH is available with a suppression pool temperature up to 140'F with no containment back pressure.
The HPCIS is not designed to operate at full capacity until reactor pressure exceeds 150 psig and the steam supply to the HPCI turbine is automatically isolated before reactor pressure decreases below 100 psig.
The ADS,
Steam pressure is sufficient at 150 psig to run the HPCI turbine for OPERABILITY testing yet, still below the shutoff head of the CSS and RHRS pumps so they will inject water into the vessel if required.
The ADS provides additional backup to reduce pressure to the range where the CSS and RHRS will inject into the vessel if necessary.
Considering the low reactor pressure, the redundancy and availability of CSS,
- RHRS, and ADS during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate HPCI OPERABILITY once sufficient steam pressure becomes available.
The alternative to demonstrate HPCI OPERABILITY PRIOR TO STARTUP using auxiliary steam is provided for plant operating flexibility.
With the HPCIS inoperable, a seven-day period to return the system to service is justified based on the availability of the ADS,
The availability of these redundant and diversified systems provides adequate assurance of core cooling while HPCIS is out of service.
The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the HPCIS will be OPERABLE when required.
BFN Unit 1 3.5/4.5-30
3.5 gAS1~S (Cont'd) 3.5.F eco C
e so C o se CC The RCICS functions to provide core cooling and makeup water to the reactor vessel during shutdown and isolation from the main heat sink and for certain pipe break accidents.
The RCICS provides its design flow between 150 psig and 1120 psig reactor pressure.
Below 150 psig, RCICS is not required to be OPERABLE since this pressure is substantially below that for any events in which RCICS is required to provide core cooling.
RCICS will continue to operate below 150 psig at reduced flow until it automatically isolates at greater than or equal to 50 psig reactor steam pressure.
150 psig is also below the shutoff head of the CSS and RHRS, thus, considerable overlap exists with the cooling systems that provide core cooling at low reactor pressure.
The minimum required NPSH for RCIC is 20 feet.
There is adequate elevation head between the suppression pool and the RCIC pump, such that the required NPSH is available with a suppression pool temperature up to 140'F with no containment back pressure.
The ADS,
Steam pressure is sufficient at 150 psig to run the RCIC turbine for OPERABILITY testing, yet still below the shutoff head of the CSS and RHRS pumps so they will inject water into the vessel if required.
Considering the low reactor pressure and the availability of the low pressure coolant systems during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate RCIC OPERABILITY once sufficient steam pressure becomes available.
The alternative to demonstrate RCIC OPERABILITY PRIOR TO STARTUP using auxiliary steam is provided for plant operating flexibility.
With the RCICS inoperable, a seven-day period to return the system to service is justified based on the availability of the HPCIS to cool.the core and upon consideration that the average risk associated with failure of the RCICS to cool the core when required is not increased.
The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the RCICS will be OPERABLE when required.
3.5.G uto atic De ressu zatio S ste A
S This specification ensures the OPERABILITY of the ADS under all conditions for which the depressurization of the nuclear system is an essential response to station abnormalities.
The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the low-pressure coolant injection (LPCI) and the core spray subsystems can operate to protect the fuel barrier.
Note that this specification applies only to the automatic feature of the pressure relief system.
Specification 3.6.D specifies the requirements for the pressure relief function of the valves.
It is possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures, yet be fully capable of performing their pressure relief function.
BFN Unit 1 3.5/4.5-31
4 CO S
E S
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A.
r m Co t 3.7.A.l (Cont'd)
C ~
With the suppression pool water temperature
> 95 F initiate pool cooling and restore the temperature to g 95'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT SHUTDOWN CONDITION within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in the COLD SHUTDOWN CONDITION within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
d ~
With the suppression pool water temperature
> 105'F during testing of ECCS or relief
- valves, stop all testing, initi.ate pool cooling and follow the action in Specification 3.7.A.l.c above.
e.
With the suppression pool water temperature
> 110'F during the STARTUP CONDITION, HOT STANDBY CONDITION (with'll control rods not inserted),
or REACTOR POWER OPERATION, the reactor shall be scrammed.
With the suppression pool water temperature
> 120'F following reactor isolation, depressurize to
< 200 psig at normal cooldown rates.
BFN Unit 1 3.7/4.7-2
UNIT 2 EFFECTIVE PAGE LIST REMOVE INSERT 1.0-11 3.2/4.2-8 3.2/4.2-12 3.2/4.2-13 3.2/4.2-32 3.2/4.2-54 3.2/4.2-63a 3.2/4.2-69 3.5/4.5-28 3.5/4.5-29 3.6/4.6-32 1.0-11 3.2/4.2-8 3.2/4.2-12 3.2/4.2-13 3.2/4.2-32 3.2/4.2-54 3.2/4.2-63a 3.2/4.2-69 3.5/4.5-28 3.5/4.5-29 3.6/4.6-32
I i
owned, leased, or otherwise controlled by TVA.
U st c
e ea Any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for industrial, commercial, institutional, or recreational purposes.
Dose E
va e t
The DOSE EQUIVALENT I-131 shall be the concentration of I-131 (in pCi/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.
The thyroid dose conversion factor used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Reactor Sites".
Gaseous Waste eat e
S ste The charcoal adsorber vessels installed on the discharge of the steam jet air effector to provide delay to a unit's offgas activity prior to release.
LL.
embers o
the Pub ic Shall include all individuals who by virtue of their occupational status have no formal association with the plant.
This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions.
This category shall got include non-employees such as vending machine servicemen or postmen who, as part of their formal gob function, occasionally enter restricted areas.
OPERATIONAL CONDITIONS or other conditions specified for individual limiting conditions for operation unless otherwise stated in an individual Surveillance Requirements.
Each Surveillance Requirement shall be performed within the specified time interval with, (1) A maximum allowable extention not to exceed 25% of the surveillance interval, but (2) The combined time entered for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval Performance of a Surveillance Requirement within the specified time interval shall constitute compliance and OPERABILITY requirements for a limiting condition for operation and associated action statements unless otherwise required by these specifications.
Surveillance Requirements do not have to be performed on inoperable equipment.
BFN Unit 2 1.0-11
TABLE 3.2.A (Continued)
PRIHARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION Hinimum No.
Instrument Channels Operable Pr Tri Ssl 11 2(3) 2(12) 2(14)
Fn in Instrument Channel High Radiation Hain Steam Line Tunnel (6)
Instrument Channel-Low Pressure Hain Steam Line (PIS-1-72, 76, 82, 86)
Instrument Channel-High Flow Hain Steam Line (PdIS-1-13A-D, 25A-O, 36A-O, 50A-D)
Instrument Channel-Hain Steam Line Tunnel High Temperature Instrument Channel-Reactor Water Cleanup System Floor Drain High Temperature Tri L v 1
in 3 times normal rated full power background (13)
> 825 psig (4) 140% of rated steam flow
< 2000F 160 - 180'F A
i n 1
Rmrk 1.
Above trip setting initiates Hain Steam Line Isolation 1.
Below trip setting initiates Hain Steam Line Isolation l.
Above trip setting initiates Hain Steam Line Isolation l.
Above trip setting initiates Hain Steam Line Isolation.
l.
Above trip setting initiates Isolati on of Reactor Water Cleanup Line from Reactor and Reactor Water Return Line.
Instrument Channel-Reactor Mater Cleanup System Space High Temperature Instrument Channel Reactor Mater Cleanup System Pipe Trench Instrument Channel Reactor Building Ventilation High Radiation Reactor Zone 160 - 180'F
< 1500F
< 100 mr/hr or downscale 1.
Same as above 1.
Same as above l.
1 upscale or 2 downscale will a.
Initiate SGTS b.
Isolate reactor zone and refueling floor.
c.
Close atmosphere control system.
0 S
0 Whenever the respective functions are required to be OPERABLE there shall be two OPERABLE or tripped trip systems for each function. If the first column cannot be met for one of the trip systems, that trip system or logic for that function shall be tripped (or the appropriate action listed below shall be taken). If the column cannot be met for all trip
- systems, the appropriate action listed below shall be taken.
A. Initiate an orderly shutdown and have the reactor in Cold Shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an orderly load reduction and have Main Steam Lines isolated within eight hours.
C.
Isolate Reactor Water Cleanup System.
D.
Isolate Shutdown Cooling.
E.
Initiate primary containment isolation within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
F.
The handling of spent fuel will be prohibited and all operations over spent fuels and open reactor wells shall be prohibited.
G.
Isolate the reactor building and start the standby gas treatment system.
H.
Immediately perform a logic system functional test on the logic in the other trip systems and daily thereafter not to exceed 7 days.
I.
Deleted J.
Withdraw TIP.
K.
Manually isolate the affected lines.
Refer to Section 4.2.E for the requirements of an inoperable system.
L. If one SGTS train is inoperable take actions H or A and F. If two SGTS trains are inoperable take actions A and F.
2 ~
Deleted 3 ~
There are four sensors per steam line of which at least one sensor per trip system must be OPERABLE.
BFN Unit 2 3.2/4.2-12
g
0 S
0 TABL 4.
Only required in RUN MODE (interlocked with Mode Switch).
5.
Deleted 6.
Channel shared by RPS and Primary Containment
& Reactor Vessel Isolation Control System.
A channel failure may be a channel failure in each system.
7.
A train is considered a trip system.
8 ~
Two out of three SGTS trains required.
A failure of more than one will require actions A and F.
9.
Deleted 10.
Refer to Table 3.7.A and its notes for a listing of Isolation Valve Groups and their initiating signals.
A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
12.
A channel contains four sensors, all of which must be OPERABLE for the channel to be OPERABLE.
Power operations permitted for up to 30 days with 15 of the 16 temperature switches OPERABLE.
In the event that normal ventilation is unavailable in the main steam line tunnel, the high temperature channels may be bypassed for a period of not to exceed four hours.
During periods when normal ventilation is not available, such as during the performance of secondary containment leak rate tests, the control room indicators of the affected space temperatures shall be monitored for indications of small steam leaks.
In the event of rapid increases in temperature (indicative of steam line break),
the operator shall promptly close the main steam line isolation valves.
13.
The nominal setpoints for alarm and reactor trip (1.5 and 3.0 times background, respectively) are established based on the normal background at full power.
The allowable setpoints for alarm and reactor trip are 1.2-1.8 and 2.4-3.6 times background, respectively.
14.
Requires two independent channels from each physical location; there are two locations.
BFN Unit 2 3.2/4.2-13
Hinimum ¹ of Operable Instrument hnn1 1/Valve
~In ryan H2H 76 94 H2H 76 104 PdI-64-137 PdI-64-138 RR-90-272CO RR-90-273CO LI-64-159A XR-64-159 PI-64-160A XR-64-1 59 TI-64-161 TR-64-1 61 TI-64-1 62 TR-64-162 RH-90-306 RR-90-360 TABLE 3.2.F (cont'd)
Surveillance Instrumentation In rum n Orywell and Torus Hydrogen Concentration Orywell to Suppression Chamber Differential Pressure Relief Valve Tailpipe Thermocouple Temperature or Acoustic Honitor on Relief Valve Tailpipe High Range Primary Containment Radiation Recorders Suppression Chamber Mater Level-Wide Range Orywell Pressure Mide Range Suppression Pool Bulk Temperature Hide Range Gaseous Effluent Radiation Honitor and recorder Type Indication and Ran 0.1 205 Indicator 0 to 2 psid Recorder 1-10'/Hr Indicator, Recorder 0-240" Indicator, Recorder) 0-300 psig
)
Indicator, Recorder)
)
300 2300 F
)
Honi ter and recorder (Ho)le Gas 10 10+5 yCi/cc)
(1) (2) (3)
(5)
(7)(8)
(1) (2) (3)
(1) (2) (3)
(1) (2) (3) (4) (6)
(7)(8)(9)
TABLE 4.2.F HINIHUH TEST AND CALIBRATION FREQUENCY FOR SURVEILlANCE INSTRUHENTATION In rum n harm 1
- 1) Reactor Water Level (LI-3-58A&B)
- 2) Reactor Pressure (PI-3-74A&B)
- 3) Drywell Pressure (PI-64-678) and XR-64-50
- 4) Drywell Temperature (TI-64-52AB) and XR-64-50
- 5) Suppression Chamber Air Temperature (XR-64-52)
- 8) Control Rod Position
- 9) Neutron Honitoring
- 10) Drywell.Pressure (PS-64-67B) ll) Drywell Pressure (PIS-64-58A)
- 12) Drywell Temperature (TS-64-52A)
- 13) Timer (IS-64-67A) 14)
CAD Tank Level
- 15) Containment Atmosphere Honitors Calibr i n Fr n
Once/6 months Once/6 months Once/6 months Once/6 months Once/6 months N/A (2)
Once/6 months Once/6 months Once/6 months Once/6 months Once/6 months Once/6 months In rum n h
Each Shift Each Shift Each Shi ft Each Shift Each Shift Each Shift Each Shift N/A N/A N/A N/A Once/day Once/day
Table 4.2.L Anticipated Transient Without Scram (ATWS)
Recirculation Pump Trip (RPT) Instrumentation Surveillance Functional est Channel C
r o
Instrument Reactor Vessel Water Level Low (LS-3-58Al-Dl)
Reactor Vessel Dome Pressure High (PIS-3-204A-D)
M(27)
M(27)
R(28)
R(28)
N/A N/A BFN Unit 2 3.2/4.2-63a
3 '
BASES (Cont'd)
Both instruments are required for trip but the instruments are set so that the instantaneous stack release rate limit given in Specification 3.8 is not exceeded.
Four radiation monitors are provided for each unit which initiate Primary
. Containment Isolation (Group 6 isolation valves) Reactor Building Isolation and operation of the Standby Gas Treatment System.
These instrument channels monitor the radiation in the reactor zone ventilation exhaust ducts and in the refueling zone.
Trip setting of 100 mr/hr for the monitors in the refueling zone are based upon initiating normal ventilation isolation and SGTS operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the SGTS.
Flow integrators and sump fillrate and pump out rate timers are used to determine leakage in the drywell.
A system whereby the time interval to filla known volume will be utilized to provide a backup.
An air sampling system is also provided to detect leakage inside the primary containment (See Table 3.2.E).
For each parameter monitored, as listed in Table 3.2.F, there are two channels of instrumentation except as noted.
By comparing readings between the two channels, a near continuous surveillance of instrument performance is available.
Any deviation in readings will initiate an early recalibration, thereby maintaining the quality of the instrument readings.
Instrumentation is provided for isolating the control room and initiating a pressurizing system that processes outside air before supplying it to the control room.
An accident signal that isolates primary containment will also automatically isolate the control room and initiate the emergency pressurization system.
In addition, there are radiation monitors in the normal ventilation system that will isolate the control room and initiate the emergency pressurization system.
Activity required to cause automatic actuation is about one mRem/hr.
Because of the constant surveillance and control exercised by TVA over the Tennessee Valley, flood levels of large magnitudes can be predicted in advance of their actual occurrence.
In all cases, full advantage will be taken of advance warning to take appropriate action whenever reservoir levels above normal pool are predicted.
Therefore, during flood conditions, the plant will be permitted to operate until water begins to run across the top of the pumping station at elevation 565.
Seismically qualified, redundant level switches each powered from a separate division of power are provided at the pumping station to give main control room indication of this condition.
At that time an orderly shutdown of the plant will be initiated, although surges even to a depth of several feet over the pumping station deck will not cause the loss of the main condenser circulating water pumps.
BFN Unit 2 3.2/4.2-69
~
3.5 Q~SS (Cont
)
3.5.E.
essu e
Coo a
n ect o
S ste C
S The HPCIS is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the nuclear system and loss of coolant which does not result in rapid depressurization of the reactor vessel.
The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.
The HPCIS continues to operate until reactor vessel pressure is below the pressure at which LPCI operation or Core Spray system operation maintains core cooling.
The capacity of the system is selected to provide the required core cooling.
The HPCI pump is designed to pump 5000 gpm at reactor pressures between 1120 and 150 psig.
The HPCIS is not required to be OPERABLE,below 150 psig since this is well within the range of the low pressure cooling systems and below the pressure of any events for which HPCI is required to provide core cooling.
The minimum required NPSH for HPCI is 21 feet.
There is adequate elevation head between the suppression pool and the HPCI pump, such that the required NPSH is available with a suppression pool temperature up to 140 F with no containment back pressure.
The HPCIS is not designed to operate at full capacity until reactor pressure exceeds 150 psig and the steam supply to the HPCI turbine is automatically isolated before reactor pressure decreases below 100 psig.
The ADS,
Steam pressure is sufficient at 150 psig to run the HPCI turbine for OPERABILITY testing yet still below the shutoff head of the CSS and RHRS pumps so they will inject water into the vessel if required.
The ADS provides additional backup to reduce pressure to the range where the CSS and RHRS will inject into the vessel if necessary.
Considering the low reactor pressure, the redundancy and availability of CSS,
- RHRS, and 99S during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate HPCI OPERABILITY once sufficient steam pressure becomes available.
The alternative to demonstrate HPCI OPERABILITY PRIOR TO STARTUP using auxiliary steam is provided for plant operating flexibility.
With the HPCIS inoperable, a seven-day period to return the system to service is justified based on the availability of the ADS,
The availability of these redundant and diversified systems provides adequate assurance of core cooling while HPCIS is out of service.
The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the HPCIS will be OPERABLE when required.
BFN Unit 2 3.5/4.5-28
3.5
~BS!S (Cont'd) 3.5.F Reactor Co e
so t o Cool S
te C
CS The RCICS functions to provide core cooling and makeup water to the reactor vessel during shutdown and isolation from the main heat sink and for certain pipe break accidents.
The RCICS provides its design flow between 150 psig and 1120 psig reactor pressure.
Below 150 psig, RCICS is not required to be OPERABLE since this pressure is substantially below that for any events in which RCICS is required to provide core cooling.
RCICS will continue to operate below 150 psig at reduced flow until it automatically isolates at greater than or equal to 50 psig reactor steam pressure.
150 psig is also below the shutoff head of the CSS and RHRS, thus, considerable overlap exists with the cooling systems that provide core cooling at low reactor pressure.
The minimum required NPSH for RCIC is 20 feet.
There is adequate elevation head between the suppression pool and the RCIC pump, such that the 'required NPSH is available with a suppression pool temperature up to 140'F with no containment back pressure.
The ADS,
Steam pressure is sufficient at 150 psig to run the RCIC turbine for OPERABILITY testing, yet still below the shutoff head of the CSS and RHRS pumps so they will inject water into the vessel if required.
Considering the low reactor pressure and the availability of the low pressure coolant systems during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate RCIC OPERABILITY once sufficient steam pressure becomes available.
The alternative to demonstrate RCIC OPERABILITY PRIOR TO STARTUP using auxiliary steam is provided for plant operating flexibility.
With the RCICS inoperable, a seven-day period to return the system to service is justified based on the availability of the HPCIS to cool -the core and upon consideration that the average risk associated with failure of the RCICS to cool the core when required is not increased.
The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the RCICS will be OPERABLE when required.
3.5.G utomatic De essu zat o
S ste DS This specification ensures the OPERABILITY of the ADS under all conditions for which the depressurization of the nuclear system is an essential response to station abnormalities.
The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the low-pressure coolant injection (LPCI) and the core spray subsystems can operate to protect the fuel barrier.
Note that this specification applies only to the automatic feature of the pressure relief system.
Specification 3.6.D specifies the requirements for the pressure relief function of the valves.
It is possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures, yet be fully capable of performing their pressure relief function.
BFN Unit 2 3.5/4.5-29
3.6/4.6 QLSSS 3.6.E/4.6.E (Cont'd) resistance to the recirculation pump is also reduced;
- hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure).
If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed.
Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.
In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate.
The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight, decrease (3 percent to 6 percent) in the total core flow measured.
This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.
Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.
A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true.
The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.
3.6.F/4.6.F ec cul t o u
0 erat o
Operation without forced recirculation is permitted for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is not in the RUN MODE.
And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F.
This reduces the positive reactivity insertion to an acceptably low value.
Requiring at least one recirculation pump to be OPERABLE while in the RUN MODE (i.e., requiring a manual scram if both recirculation pumps are tripped) provides protection against the potential occurrence of core thermal'-hydraulic instabilities at low flow conditions.
Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.
3.6.G/4.6.G St u tu a te t
The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.
The program reflects the built-in limitations of access to the reactor coolant systems.
It is intended that the required examinations and inspection be completed during each 10-year interval.
The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
BFN Unit 3.6/4.6-32
UNIT 3 EFFECTIVE PAGE LIST REMOVE 1.0-11 3.2/4.2-12 3.2/4.2-13 3.2/4.2-62a 3.2/4.2-68 3.5/4.5-31 3.5/4.5-32 INSERT 1.0-11 3.2/4.2-12 3.2/4.2-13 3.2/4.2-62a 3.2/4.2-68 3.5/4.5-31 3.5/4.5-32
GG.
owned, leased, or otherwise controlled by TVA.
U res cted e
Any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for industrial, commercial, institutional, or recreational purposes.
ose uiva e t
The DOSE EQUIVALENT I-131 shall be the concentration of Z-131 (in pdi/tm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.
The thyroid dose conversion factor used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Reactor Sites".
Gaseous W ste e t t
S s e
The charcoal adsorber vessels installed on the discharge of the steam jet air ejector to provide delay to a unit's offgas activity prior to release.
embe s of t e ub c Shall include all individuals who by virtue of their occupational status have no formal association with the plant.
This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions.
This category shall got include non-employees such as vending machine servicemen or postmen
- who, as part of their formal job function, occasionally enter restricted areas.
LL.
the OPERATIONAL CONDITIONS or other conditions specified for individual limiting conditions for operation unless otherwise stated in an individual Surveillance Requirements.
Each Surveillance Requirement shall be performed within the specified time interval with, (1) A maximum allowable extention not to exceed 25% of the surveillance interval, but (2) The combined time entered for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval Performance of a Surveillance Requirement within the specified time interval shall constitute compliance and OPERABILITY requirements for a limiting condition for operation and associated action statements unless otherwise required by these specifications.
Surveillance Requirements do not have to be performed on inoperable equipment.
BFN Unit 3 1.0-11
OTES FO 1.
Whenever the respective functions are required to be OPERABLE, there shall be two OPERABLE or tripped trip systems for each function. If the first column cannot be met for one of the trip systems, that trip system or logic for that function shall be tripped (or the appropriate action listed below shall be taken). If the column cannot be met for all trip
- systems, the appropriate action listed below shall be taken.
A. Initiate an orderly shutdown and have the reactor in COLD SHUTDOWN CONDITION in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an orderly load reduction and have main steam lines isolated within eight hours.
C.
Isolate Reactor Water Cleanup System.
D.
Isolate Shutdown Cooling.
E.
Initiate primary containment isolation within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
F.
The handling of spent fuel will be prohibited and all operations over spent fuels and open reactor wells shall be prohibited.
G.
Isolate the reactor building and start the standby gas treatment system.
H.
Immediately perform a logic system functional test on the logic in the other trip systems and daily thereafter not to exceed 7 days.
I.
DELETED J.
Withdraw TIP.
K.
Manually isolate the affected lines.
Refer to Section 4.2.E for the requirements of an inoperable system.
L. If one SGTS train is inoperable take action H or actions A and F. If two SGTS trains are inoperable take actions A and F.
2.
Deleted 3.
There are four sensors per steam line of which at least one sensor per trip system must be OPERABLE.
BFN Unit 3 3.2/4.2-12
0 S
0 4.
Only required in RUN MODE (interlocked with Mode Switch).
5.
Deleted 6.
Channel shared by RPS and Primary Containment 6 Reactor Vessel Isolation Control System.
A channel failure may be a channel failure in each system.
7.
A train is considered a trip system.
8.
Two out of three SGTS trains required.
A failure of more than one will require actions A and F.
9.
DELETED 10.
Refer to Table 3.7.A and its notes for a listing of Isolation Valve Groups and their initiating signals.
11.
A channel may be placed in an inoperable status for up to four hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
12.
A channel contains four sensors, all of which must be OPERABLE for the channel to be OPERABLE.
Power operations permitted for up to 30 days with 15 of the 16 temperature switches OPERABLE.
In the event that normal ventilation is unavailable in the main steam line tunnel, the high temperature channels may be bypassed for a period of not to exceed four hours.
During periods when normal ventilation is not available, such as during the performance of secondary containment leak rate tests, the control room indicators of the affected space temperatures shall be monitored for indications of small steam leaks.
In the event of rapid increases in temperature (indicative of steam line break),
the operator shall promptly close the main steam line isolation valves.
13.
The nominal setpoints for alarm and reactor trip (1.5 and 3.0 times background, respectively) are established based on the normal background at full power.
The allowable setpoints for alarm and reactor trip are 1.2-1.8 and 2.4-3.6 times background, respectively.
14.
Requires two independent channels from each physical location; there are two locations.
BFN Unit 3 3.2/4.2-13
Table 4.2.L Anticipated Transient Without Scram (ATWS)
Recirculation Pump Trip (RPT) Instrumentation Surveillance Fu etio Functional Test Channel Cal bratio Instrument C eck Reactor Vessel Water Level Lov (LS-3-58Al-Dl)
Reactor Vessel Dome Pressure High (PIS-3-204A-D)
M(28)
M(28)
R(29)
R(29)
N/A N/A BFN Unit 3 3.2/4.2-62a
Trip setting of 100 mr/hr for the monitors in the refueling zone are based upon initiating normal ventilation isolation and SGTS operation so that none of the activity released during'he refueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the SGTS.
Flow integrators and sump fillrate and pump out rate timers are used to determine leakage in the drywell.
A system whereby the time interval to filla known volume will be utilized to provide a backup.
An air sampling system is also provided to detect leakage inside the primary containment (See Table 3.2.E).
For each parameter monitored, as listed in Table 3.2.F, there are two channels of instrumentation except as noted.
By comparing readings between the two channels, a near continuous surveillance of instrument performance is available.
Any deviation in readings will initiate an early recalibration, thereby maintaining the quality of the instrument readings.
Instrumentation is provided for isolating the control room and initiating a pressurizing system that processes outside air before supplying it to the control room.
An accident signal that isolates primary containment will also automatically isolate the control room and initiate the emergency pressurization system.
In addition, there are radiation monitors in the normal ventilation system that will isolate the control room and initiate the emergency pressurization system.
Activity required to cause automatic actuation is about one mRem/hr.
Because of the constant surveillance and control exercised by TVA over the Tennessee Valley, flood levels of large magnitudes can be predicted in advance of their actual occurrence.
In all cases, full advantage will be taken of advance warning to take appropriate action whenever reservoir levels above normal pool are predicted.
Therefore, during flood conditions, the plant will be permitted to operate until water begins to run across the top of the pumping station at elevation 565.
Seismically qualified, redundant level switches each powered from a separate division of power are provided at the pumping station to give main control room indication of this condition.
At that time an orderly shutdown of the plant will be initiated, although surges even to a depth of several feet over the pumping station deck will not cause the loss of the main condenser circulating water pumps.
BFN Unit 3 3.2/4.2-68
3. 5
~B S !S (Cont'd) 3. 5. E.
i P ess e Coo a t I ect o
S ste CIS The HPCIS is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the nuclear system and loss of coolant which does not result in rapid depressuri zation of the reactor vessel.
The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized The HPCIS continues to operate until reactor vessel pressure is below the pressure at which LPCI operation or Core Spray system operation maintains core cooling.
The capacity of the system is selected to provide the required core cooling.
The HPCI pump is des igned to pump 5000 gpm at reactor pressures between 1120 and 15 0 ps ig.
The HPCI S is not required to be OPERABLE below 15 0 psig since this is well within the range of the low pressure cooling systems and below the pressure of any events for which HPCI is required to provide core cooling.
The minimum required NPSH for HPCI is 21 feet.
There is adequat e elevation head between the suppression pool and the HPCI pump, such that the required NPSH is available with a suppress ion pool temperature up to 140' with no containment back pressure.
The HPCIS is not designed to operate at full capacity until reactor pressure exceeds 150 ps ig and the steam supply to the HPCI turbine is automatically isolated before reactor pressure decreases below 100 ps ig.
The ADS,
CSS, and RHRS (LPCI) must be OPERABLE when starting up from a COLD CONDITION.
Steam pressure is sufficient at 150 ps ig to run the HPCI turbine for OPERABILITY testing, yet still below the shutoff head of the CSS and RHRS pumps so they will inject water into the vessel if requi red.
The ADS provides additional backup to reduce pressure to the range where the CSS and RHRS will inject into the vessel if necessary.
Cons idering the low reactor pressure, the redundancy and availability of CSS,
RHRS, and ADS during startup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate HPCI OPERABILITY once sufficient steam pressure becomes available.
The alternative to demonstrate HPCI OPERABILITY PRIOR TO STARTUP using auxiliary steam is provided for plant operat ing flexibility.
With the HPCIS inoperable, a seven-day period to return the system to service is justified based on the availability of the ADS,
CS S,
RHRS (LPCI) and the RCICS.
The availability of these redundant and diversified systems provides adequate assurance of core cooling while HPCIS is out of service.
The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the HPCIS will be OPERABLE when required.
BFN Unit 3 3.5/4.5-31
3.5
~!~S (Cont'd) 3.5.F e c Co e
o t o Coo in S ste C
C The RCICS functions to provide core cooling and makeup water to the reactor vessel during shutdown and isolation from the main heat sink and for certain pipe break accidents.
The RCICS provides its design flow between 150 psig and 1120 psig reactor pressure.
Below 150 psig, RCICS is not required to be OPERABLE since this pressure is substantially below that for any events in which RCICS is required to provide core cooling.
RCICS will continue to operate below 150 psig at reduced flow
.until it automatically isolates at greater than or equal to 50 psig reactor steam pressure.
150 psig is also below the shutoff head of the CSS and RHRS, thus, considerable overlap, exists with the cooling systems that provide core cooling at low reactor pressure.
The minimum required NPSH for RCIC is 20 feet.
There is adequate elevation head between the suppression pool and the RCIC pump, such that the required NPSH is available with a suppression pool temperature up to 140 F with no containment back pressure.
The ADS,
Steam pressure is sufficient at 150 psig to run the RCIC turbine for OPERABILITY testing, yet still below the shutoff head of the CSS and RHRS pumps so they will inject water into the vessel if required.
Considering the low reactor pressure and the availability of the low pressure coolant systems during sta'rtup from a COLD CONDITION, twelve hours is allowed as a reasonable time to demonstrate RCIC OPERABILITY once suffi.cient steam pressure becomes available.
The alternative to demonstrate RCIC OPERABILITY PRIOR TO STARTUP using auxiliary steam is provided for plant operating flexibility.
With the RCICS inoperable, a seven-day period to return the system to service is justified based on the availability of the HPCIS to cool-the core and upon consideration that the average risk associated with failure of the RCICS to cool the core when required is not increased.
The surveillance requirements, which are based on industry codes and standards, provide adequate assurance that the RCICS will be OPERABLE when required.
3.5.G utom tic D essu at o
S s e
DS This specification ensures the OPERABILITY of the ADS under all conditions for which the depressurization of the nuclear system is an essential response to station abnormalities.
The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the low-pressure coolant injection (LPCI) and the core spray subsystems can operate to protect the fuel barrier.
Note that this specification applies only to the automatic feature of the pressure relief system.
Specification 3.6.D specifies the requirements for the pressure relief function of the valves. It is possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures, yet be fully capable of performing their pressure relief function.
BFN Unit 3 3.5/4.5-32
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f 4
I)
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ENCLOSURE 2
SUMMARY
OF CHANGES i.
Revise definition 1.0.II (Dose Equivalent I-131) for units 1, 2, and 3.
Existing definition 1.0.II reads in part:
concentration of I-131 (in mCi/gm) which alone Revised definition 1.0.II would read in part:
concentration of I-131 (in pCi/gm) which alone II
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II
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2.
Delete the following notes from Notes for Table 3.2.A (Primary Containment and Reactor Building Isolation Instrumentation) for units 1, 2,
and 3.
"2. When it is determined that a channel is failed in the unsafe condition, the other channels that monitor the same variable shall be functionally tested immediately before the trip system or logic for that function is tripped.
The trip system or the logic for that function may remain untripped for short periods of time to allow functional testing of the other trip system or logic for that function."
"5. Not required in RUN Mode (bypassed by Mode Switch)."
3.
Revise Table 3.2.A (Primary Containment and Reactor Building Isolation Instrumentation) on page 3.2/4.2-8 for unit 2 only.
Existing trip level setting for "Instrument Channel High Radiation Main Steam Line Tunnel" reads:
"g3 times normal rated full power background" Revised trip level setting for "Instrument Channel High Radiation Main Steam Line Tunnel" would read:
"3 times normal rated full power background (13)"
4.
Revise Table 3.2.F (Surveillance Instrumentation) on page 3.2/4.2-32 for unit 2 only.
Existing "Type Indication and Range" for instruments "RM-90-306 and RM-90-360" reads:
"Monitor and recorder (Noble gas 10 7 10+5 Ci/cc)"
Revised "Type Indication and Range" for instruments "RM-90-306 and RM-90-360" would read:
"Monitor and recorder (Noble gas 10 10+
pCi/cc)"
5.
Revision to Table 4.2.F (Minimum Test and Calibration Frequency for Surveillance Instrumentation) for unit 2 only.
C
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'h lb
Enclosure 2
Page 2 of 3 Existing Table 4.2.F reads in part:
- 2) Reactor Pressure (PI-3-74 A&B)
Once/12 months Revised Table 4.2.F would read in part:
- 2) Reactor Pressure (PI-3-74 ARB)
Once/6 months Revise Table 4.2.L (Anticipated Transient Without Scram
[ATWS]
Recirculation Pump Trip [RPT] Instrumentation Surveillance) as follows for units 1, 2, and 3.
a.
Existing title of table reads in part:
. Recirculation Pump Test Revised title of table would read in part:
Recirculation Pump Trip b.
Existing table reads in part:
. Reactor Vessel Water Level Low LS-3-58A-D Reactor Vessel Dome Pressure High PS-3-204-D Revised table would read in part:
. Reactor Vessel Water Level Low (LS-3-58Al-Dl)
Reactor Vessel Dome Pressure High (PIS-3-204A-D)
Delete the following phrase from bases section 3.2 (Protective Instrumentation) for units 1, 2, and 3.
. however, the plant flood protection is always in place and does not depend in any way on advanced warning Add the following paragraph to bases section 3.5.E (High Pressure Coolant Injection System) for units 1, 2, and 3.
"The minimum required NPSH for HPCI is 21 feet.
There is adequate elevation head between the suppression pool and the HPCI pump, such that the required NPSH is available with a suppression pool temperature up to 140'F with no containment back pressure."
Enclosure 2
Page 3 of 3 Add the following paragraph to bases section 3.5.F (Reactor Core Isolation Cooling System) for units 1, 2, and 3.
"The minimum required NPSH for RCIC is 20 feet.
There is adequate elevation head between the suppression pool and the RCIC pump, such that the required NPSH is available with a suppression pool temperature up to 140 F with no containment back pressure."
Revise bases section 3.6.F/4.6.F (Recirculation Pump Operation) for unit 2 only.
Existing bases reads in part:
. Requiring at least one recirculation pump to be operable while in the RUN mode provides protection against the potential occurrence of core thermal-hydraulic instabilities at low flow conditions.
Revised bases would read in part:
. Requiring at least one recirculation pump to be operable while in the RUN mode (i.e., requiring a manual scram if both recirculation pumps are tripped) provides protection against the potential occurrence of core thermal-hydraulic instabilities at low flow conditions.
Revise paragraph 3.7.A.1.F (Primary Containment) for unit 1 only.
Existing paragraph 3.7.A.1.F reads in part:
. depressurize to >200 psig at NORMAL Revised paragraph 3.7.A.1.F would read in part:
. depressurize to <200 psig at NORMAL.
1 H
4 W
ENCLOSURE 3
REASON AND JUSTIFICATION FOR THE PROPOSED CHANGES Reaso o
Chan es These proposed changes to the BFN technical specifications are administrative in nature or revise the bases section for flood protection to be consistent with the FSAR.
They are being made to resolve open issues from NRC inspection
- reports, to resolve an open item in an NRC safety evaluation, and to correct errors in previous technical, specification. submittals and implementation.
A detailed summary of the changes is provided by Enclosure 2 (Summary of Changes).
Justification for the C an es The justification for each change is provided below in the order in which it appears in Enclosure 2 (Summary of changes).
The change number used in Enclosure 2 is used here.
Change 1 Def. 1.0.II (Dose Equivalent I-131) was added to the technical specifications by amendments
- 132, 128, and 103 to units 1, 2, and 3 respectively.
The approved definition included the units "pCi/gm" for the concentration of I-131.
The current technical specifications for all three units erroneously have "mCi/gm" for the concentration of I-131.
This change will correct the technical specifications to be as they were approved by NRC.
Change 2 Note 2 is being deleted from the Notes for Table 3.2.A for all three units.
Note 2 has been in the notes since the original issuance of the unit 1 technical specifications (TS).
However, this note has never been attached to any item in the table.
TVA has researched the similar sections of TSs for six other Boiling Water Reactors (BWRs) with custom TSs and has not found a similar note which requires testing of other channels.
Also, the General Electric BWR Standard TSs (NUREG 0123) do not have a similar note or requirement for the isolation actuation instrumentation in Table 3.3.2-1.
Operations takes the actions in note 1 whenever instrument channels are tripped so note 2 is unnecessary.
Functional testing of the instrument channels within the required surveillance intervals provides reasonable assurance that the instrument channels are OPERABLE.
Additional testing prior to tripping a channel is unnecessary.
The note is therefore being deleted from the table.
Note 5 is also being deleted from the Notes for Table 3.2.A.
This note appeared in the original technical specifications for unit 1 and was associated with "Instrument Channel Reactor High Water Level".
When the unit 2 technical specifications were issued (amendment 3 for unit 1), this reactor high water level setpoint was deleted because it was not required for BWRs similar to BFN.
However, the note was never deleted from the Notes for Table 3.2.A.
This change removes this unnecessary note for all three units.
Change 3 Amendment numbers
- 108, 102, and 75 for units 1, 2, and 3, respectively, were issued by NRC on August 13, 1984 and added note 13 to Table 3.2.A (Instrument Channel High Radiation Main Steam Line Tunnel).
TVA did not receive the amendment until about August 22, 1984.
On August 23,
- 1984, TVA sent a letter (TS-199) to NRC requesting
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Enclosure 3
Page 2 of 3 another change to the same page in Table 3.2.A.
Because the amendments approved August 13, 1984 had not been received when the letter TS-199 was in the approval cycle it went to NRC without the reference to note 13.
NRC approved the change requested in TS-199 on August 19, 1986 (Amendment 125 for unit 2) utilizing the "old" version of the page from Table 3.2.A.
Therefore, the reference to note 13 was inadver'tently deleted.
That same August 23, 1984 submittal included the "g" symbol in front of the "3" for unit 2.
Amendment 102 for unit 2 did not include this symbol so it should be deleted.
Units 1 and 3 are correct as is.
Unit 2 must be revised to match units 1 and 3 and to meet the intent of amendment 102.
Change 4 A typographical error exists in the range of the noble gas monitors as currently listed.
The units currently listed (Ci/cc) would be too high and would not provide the required monitoring range information.
This change would reflect the correct range (pCi/cc) of the instruments.
The error exists because TVA's submittal to NRC dated June 20,
- 1989, TS 266 Supplement 1 Correction to Tables 3.2.F and 4.2.F, erroneously omitted the "p" symbol.
Change 5 The calibration frequency for instruments PI-3-74ARB for unit 2 is being corrected to be once per six months.
Amendment 167 to the BFN unit 2 technical specifications was approved by NRC on July 7, 1989.
That amendment revised the calibration frequency for the reactor pressure instruments (page 3.2/4.2-54) to a more conservative interval of 6 months.
This was based on the recommendation of Tobar, Inc., the manufacturer of these instruments.
Amendment 171 to the BFN unit 2 technical specifications was approved by NRC on August 22, 1989.
An overleaf page for this amendment was page 3.2/4.2-54, the page which includes the reactor pressure instruments.
The clean pages for Amendment 171 were sent to NRC on July 25, 1989.
The BFN technical specification clerk had apparently not yet received Amendment 167 and therefore sent a clean page 3.2/4.2-54 with the old calibration frequency of 12 months on it.
When Amendment 171 was issued, it had the incorrect calibration frequency (12 months) which had been transmitted on the clean page.
This proposed change revises page 3.2/4.2-54 to correctly indicate that the calibration frequency for the Reactor Pressure instruments is 6
months as approved by Amendment 167.
Change 6 Table 4.2.L (Anticipated Transient Without Scram
[ATWS]
Recirculation Pump Trip [RPT] Instrumentation Surveillance) is being revised to correct the title and to incorporate the correct instrument numbers into the table.
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Enclosure 3
Page 3 of 3 Change 7 The statement in the bases for technical specification 3.2 that ".
. however, the plant flood protection is always in place and does not depend in any way on advanced warning
. " is being deleted to agree with the plant Final Safety Analysis Report (FSAR).
The flood doors to the reactor and radwaste buildings are the flood protection referred to as always being in place in the technical specification 3.2 bases.
The FSAR was revised in 1987 by amendment 5 to reflect the practice of leaving the flood doors open under normal circumstances.
The FSAR had previously indicated that these doors are normally closed.
The Unreviewed Safety Question Determination in support of the FSAR revision concluded that because of the constant surveillance and control exercised by TVA over the Tennessee River and the relatively short amount of time required to close the flood doors, leaving them normally open would not degrade plant flood protection.
Administrative controls are in place to ensure the flood doors are closed when they are needed to provide flood protection.
Changes 8 and 9 The paragraphs in bases sections 3.5.E (HPCI) and 3.5.F (RCIC) regarding net positive suction head for the HPCI and RCIC systems were inadvertantly deleted by TVA when TS 274 was submitted to NRC.
Amendments
- 173, 176, and 144 were therefore approved for units 1, 2,
and 3 respectively with these paragraphs deleted.
They are still applicable and are therefore being re-inserted.
Change 10 A statement is being added to bases section 3.6.F/4.6.F (Recirculation Pump Operation) to note the intention to scram the reactor if the operating recirculation pump trips when in single loop operation.
TVA's commitment to make this change was documented in NRC's safety evaluation supporting unit 2 amendment 174 (TAC 73435).
Change 11 This paragraph is being revised to correct the "greater than" symbol (>) to a "less than" symbol (<) to reflect the requirement to depressurize the reactor to less than 200 psig.
This change corrects a typographical error and makes this paragraph consistent with the equivalent unit 2 and unit 3 requirements.
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ENCLOSURE 4 PROPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Descri tion of Pro osed Technical S ecification Amendment The BFN technical specifications are being revised as follows:
Revise definition 1.0.II for units 1, 2, and 3 to make the units for I-131 concentration be "pCi/gm".
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Delete notes 2 and 5 from the Notes for Table 3.2.A for units 1, 2, and 3 ~
3.
Revise the trip level setting for "Instrument Channel High Radiation Main Steamline Tunnel" in Table 3.2.A for unit 2 to add the reference to note 13 and to delete the "g" symbol.
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Revise the units for the noble gas monitors in Table 3.2.F for unit 2 to be "pCi/cc".
5.
Revise calibration frequency for reactor pressure instruments PI-3-74ASB in Table 4.2.F for unit 2.
6.
Revise Table 4.2.L for units 1, 2, and 3 to correct the title and instrument numbers.
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Revise bases section 3.2 for units 1, 2, and 3 to delete the statement that plant flood protection is always in place.
8.
Add the reference to Net Positive Suction Head (NPSH) for the High Pressure Coolant Injection system back into bases section 3.5.E for units 1, 2, 3.
9.
Add the reference to NPSH for the Reactor Core Isolation Cooling system back into bases section 3.5.F for units 1, 2, and 3.
10.
Revise bases section 3.6.F/4.6.F for unit 2 to note the intention to scram the reactor if the operating recirculation pump trips when in single loop operation.
ll.
Correct typographical error for unit 1 to specify the requirements for reactor depressurization when suppression pool temperature exceeds its limit.
Basis for Pro osed o Si nificant Hazards Consideration Determinatio NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c).
A proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety.
1.
The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
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Enclosure 4
Page 2 of 3 Definition 1.0.II (Dose Equivalent I-131) is being revised to show the correct units for I-131 concentration (pCi/gm) for all these units.
Notes 2 and 5 are being deleted from Table 3.2.A (Primary Containment and Reactor Building Isolation Instrumentation) for all three units.
Note 2 involves functional testing of redundant instrument channels when it is determined that a channel has failed in the unsafe condition.
This note was in the original technical specifications for unit 1 but was never referenced in Table 3.2.A.
Operations takes the actions in note 1 whenever instrument channels are tripped so note 2 is unnecessary.
Note 5 applies to a reactor high water level instrument which is not required for Boiling Water reactors similar to BFN and is not installed in the plant.
The note is therefore being deleted.
The trip level setting in Table 3.2.A for Instrument Channel High Radiation Main Steam Line Tunnel is being corrected for unit 2 to agree with the NRC approved technical specification and to reference note 13.
Table 3.2.F (Surveillance Instrumentation) is being corrected for the instruments RM-90-306 and RM-90-360 on unit 2.
The range for these instruments is being corrected to show units of pCi/cc.
Table 4.2.F (Minimum Test and Calibration Frequency for Surveillance Instrumentation) for reactor pressure instruments PI-3-74 A S
B on unit 2 is being revised to six months.
The current value of twelve months was changed to six months by amendment 167 based on the manufacturer's recommendation.
This change corrects the technical specification.
Table 4.2.L (Anticipated Transient Without Scram
[ATWS] Recirculation Pump Trip [RPT] Instrumentation Surveillance) is being revised to correct the title and instrument numbers in the table for all three units.
Bases section 3.2 for all three units is being revised to delete the statement that flood protection is always in place.
The FSAR was revised in 1987 to delete a similar statement.
Because of the constant surveillance and control exercised by TVA over the Tennessee River and the relatively short amount of time to close the flood doors, leaving them normally open will not degrade plant flood protection.
Paragraphs in bases sections 3.5.E (High Pressure Coolant Injection) and 3.5.F (Residual Core Isolation Cooling) regarding the net positive suction head for each system were inadvertantly deleted for units 1, 2, and 3
when TVA submitted TS 274 to NRC.
These descriptive paragraphs are still applicable and are therefore being re-inserted.
A statement is being added to bases section 3.6.F/4.6.F (Recirculation Pump Operation) for unit 2 to note the intention to scram the reactor if the operating recirculation pump trips when in single loop operation.
TVA's commitment to make this change was documented in NRC's safety evaluation supporting unit 2 amendment 174.
These changes do not affect any of the design basis accidents.
They are administrative in nature and are being made to delete non-applicable notes from tables, to correct administrative errors in previous technical specification submittals and implementation, and to make the technical specifications agree with the Final Safety Analysis Report (FSAR).
They do not involve an increase in the probability or consequences of an accident previously evaluated.
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Enclosure 4
Page 3 of 3 The proposed changes do not create the possibility of a new or different kind of accident from an accident previously evaluated.
The proposed changes are administrative in nature.
They are being made to delete non-applicable notes from tables, to correct administrative errors in previous technical specification submittals and implementation, and to make the technical specifications agree with the FSAR.
No modifications to any plant equipment are involved.
There are no effects on system interactions made by these changes.
The changes will correct the technical specifications so that they are more accurate and more closely reflect actual plant conditions.
The proposed changes do not involve a significant reduction in a margin of safety.
The proposed changes are administrative in nature.
They delete non-applicable notes from tables, correct administrative errors in previous technical specification submittals and implementation, and make the technical specifications agree with the FSAR.
No safety margins are affected by these changes.
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