ML18036B207

Proposed Tech Specs 3.2/4.2,3.1/4.1 & 3.7/4.7 Re Scram & Main Steamline Isolation Valve Closure Requirements

ML18036B207
Person / Time
Site:	Browns Ferry
Issue date:	03/25/1993
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18036B206	List: ML18036B205 ML18036B207
References
NUDOCS 9303290344
Download: ML18036B207 (81)

Text

PROPOSED TECHNICALSPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT I (TVA BFNP TS 322) 9303290344 930325 D'OR aOOCK Oa000aS9 P

PDR

UNIT 1 EFFECTIVE PAGE LIST REMOVE 3.1/4.1-4 3.1/4.1-6 3.1/4.1-9 3.1/4.1-11 3.1/4.1-12

-3.1/4.1-15 3.2/4.2-8 3.2/4.2-13 3.2/4.2-40 3.2/4.2-61 3.2/4.2-66 3.2/4.2-67 3.7/4.7-34 INSERT 3.1/4.1-4 3.1/4.1-6 3.1/4.1-9 3.1/4.1-11 3.1/4.1-12 3.1/4.1-15 3.2/4.2-8 3.2/4.2-13 3.2/4.2-40 3.2/4.2-61 3.2/4.2-66 3.2/4.2-67 3.7/4.7-34

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TABLE 3.1.A REACTOR PROTECTION SYSTEH (SCRAH)

INSTRUHENTATION REgUIREHENTS Hin. No.

oF Operable Instr.

Channels Per Trip 2

Tri Function High Water Level in West Scram Discharge Tank (LS-85-45A-D)

Tri Level ettin

< 50 Gallons Hodes in Which Fun tion Hust Be 0 erable Shut-Startup/

down

~Refuel 7

trio Standt Run X(2)

X(2)

A~ction 1

1.A High Water Level in East Scram Discharge Tank (LS-85-45E-H)

Hain Steam Line Isolation Valve Closure Turbine Control Valve Fast Closure or Turbine Trip Turbine Stop Valve Closure Turbine First Stage Pressure Permissive

< 50 Gallons

<10'A Valve Closure

>550 psig

<105 Valve Closure not >154 psig X(2)

X(2)

X(3)(6)

X(3)(6)

X(18)

X(18)

X 1.A X(6) 1.A or 1.C X(4) 1.A or 1.D X(4) 1.A or 1.D X(18) 1.A or 1.D (19)

OTES FOR TABLE 1 A (Cont'd) 8.

Not required to be OPERABLE when primary containment integrity is not required.

9 ~

DELETED 10.

Not required to be OPERABLE when the reactor pressure vessel head is not bolted to the vessel.

The APRM downscale trip function is only active when the reactor mode switch is in RUN.

12.

The APRM downscale trip is automatically bypassed when the IRM instrumentation is OPERABLE and not high.

13.

Less than 14 OPERABLE LPRMs will cause a trip system trip.

14.

Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Isolation Control System.

A channel failure may be a

channel failure in each system.

15.

16.

The APRM 15 percent scram is bypassed in the RUN Mode.

Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion).

A channel failure may be a channel failure in each system.

If a channel is allowed to be inoperable per J

Table'3.1.A, the corresponding function in that same channel may be inoperable in the Reactor Manual Control System (Rod Block).

17.

Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).

18.

This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first state pressure is greater than or equal to 154 psig.

19.

Action 1.A or 1.D shall be taken only if the permissive fails in such a

manner to prevent the affected RPS logic from performing its intended function., Otherwise, no action is required.

20.

DELETED 21.

The APRM High Flux and Inoperative Trips do not have to be OPERABLE in the REFUEL Mode if the Source Range Monitors are connected to give a

noncoincidence, High Flux scram, at 5 x 10 cps.

The SRMs shall be 5

OPERABLE per Specification 3.10.B.l.

The removal of eight (8) shorting links is required to provide noncoincidence high-flux scram protection from the Source Range Monitors.

BFN Unit 3.1/4.1-6

Grou 2

Electronic Level Switches (LS-85-45A, 8, G, H)

Hain Steam Line Isolation Valve Closure Turbine Control Valve Fast Closure or turbine trip Turbine First Stage Pressure Permissive (PT-1-81A and 8, PT-1-91A and 8)

Turbine Stop Valve Closure

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N tz$

High Mater Level in Scram Discharge I

Tank Float Switches (LS-85-45C-F)

A TABLE 4.1.A (Continued)

Functi nal Test Trip Channel and Alarm Trip Channel and Alarm Trip Channel and Alarm Trip Channel and Alarm Trip Channel and Alarm (7)

Trip Channel and Alarm Minimum Fre uen 3

Once/Honth Once/Honth Once/3 Months (8)

Once/Honth (1)

Every three months Once/Honth (1)

LA I

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TABLE 4.1.8 REACTOR PROTECTION SYSTEH (SCRAH)

INSTRUHENTATION CALIBRATION HINIHUH CALIBRATION FRE(}UENCIES FOR REACTOR PROTECTION INSTRUHENT CHANNELS Ins rum n hannel IRH High Flux APRH High Flux Output Signal Flow Bias Signal LPRH Signal High Reactor Pressure High Drywell Pressure Reactor Low Water Level High Water Level in Scram Discharge Volume Electronic Lvl Switches (LS-85-45-A, B, G, H)

Float Switches (LS-85-45C-F)

Hain Steam Line Isolation Valve Closure Turbine First Stage Pressure Permissive (PT-1-81A, B &

PT-1-91A, B)

Turbine Control Valve Fast Closure or Turbine Trip Turbine Stop Valve Closure Grou 1

Calibration Comparison to APRH on Controlled Startups (6)

Heat Balance Calibrate Flow Bias Signal (7)

TIP System Traverse (8)

Standard Pressure Source Standard Pressure Source Pressure Standard Calibrated Water Column (5)

Calibrated Water Column (5)

Note (5)

Standard Pressure Source Standard Pressure Source Note (5)

Hinimum Fre u n 2

Note (4)

Once/7 Days Once/Operating Cycle Every 1000 Effective Full Power Hours Every 3 Honths Every 3 Honths Every 3 Honths Note (5)

Note (5)

Note (5)

Once/Operating Cycle (9)

Once/Operating Cycle Note (5)

OTES FOR TABLE 4 1 B 1.

A description of three groups is included in the bases of this specification.

2.

Calibrations are not required when the systems are not required to be OPERABLE or are tripped. If calibrations are missed, they shall be performed prior to returning the system to an OPERABLE status.

3.

DELETED 4.

Required frequency is initial startup following each refueling outage.

5.

Physical inspection and actuation of these position switches will be performed once per operating cycle.

6.

On controlled startups, overlap between the IRMs and APRMs will be verified.

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8.

The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.

The instrumentation is an analog type with redundant flow signals that can be compared.

The flow comparator trip and upscale will be functionally tested according to Table 4.2.C to ensure the proper operation during the operating cycle.

Refer to 4.1 Bases for further explanation of calibration frequency.

A complete TIP system traverse calibrates the LPRM signals to the process computer.

The individual LPRM meter readings will be adjusted as a

minimum at the beginning of each operating cycle before reaching 100 percent power.

9.

Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.

BFN Unit 1 3.1/4.1-12

3.1 BAMS (Cont'd)

Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel.

This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration.

Additional IRM channels have also been provided to allow for bypassing of one such channel.

The bases for the scram setting for the IRM, APRM, high reactor pressure, reactor low water level, MSIV

closure, turbine control valve fast closure, turbine stop valve closure and loss of condenser vacuum are discussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switches) for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment.

A high drywell pressure scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality.

This instrumentation is a backup to the reactor vessel water level instrumentation.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

Reference Section 7.2.3.7 FSAR.

The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system (120/125 scram) in conjunction with the APRM system (15 percent scram) provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges.

The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping.

The discharge volume tank accommodates in excess of 50 gallons of water and is the low point in the piping.

No credit was taken for this volume in the design of the discharge piping as concerns the, amount of water which must be accommodated during a scram.

During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not BFN Unit 1 3.1/4.1-15

TABLE 3.2.A (Continued)

PRIHARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION

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Cv P 2 Hinimum No.

Instrument Channels Operable Per Tri S

1 ll Function Instrument Channel High Radiation Hain Steam Line Tunnel Tri L v 1

ttin 3 times normal rated full power background (13)

A tion 1

Remarks l.

Above trip setting initiates Hain Steam Line drain and reactor water sample line Isolation Instrument Channel-

> 825 psig (4)

Low Pressure Hain Steam Line 1.

Below trip setting initiates Hain Steam Line Isolation LA Ico 2(3) 2(12) 2(14)

Instrument Channel High Flow Hain Steam Line Instrument Channel Hain Steam Line Tunnel High Temperature Instrument Channel Reactor Mater Cleanup System Floor Drain High Temperature Instrument Channel-Reactor Water Cleanup System Space High Temperature Instrument Channel-Reactor Building Ventilation High Radiation Reactor Zone 140% of rated steam flow

< 200'F 160 - 1800F 160 1800 F

< 100 mr/hr or downscale l.

Above trip setting initiates Hain Steam Line Isolation l.

Above trip setting initiates Hain Steam Line Isolation.

l.

Above trip setting initiates Isolation of Reactor Water Cleanup Line from Reactor and Reactor Mater Return Line.

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.

NOTES FOR TABLE 3.2.A (Cont'd) 4.

Only required in RUN MODE (interlocked with Mode Switch).

5.

Deleted 6.

Channel shared by RPS and Primary Containment Ec 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 SOTS trains required.

A failure of more than one will require actions A and F.

9.

Deleted 10.

Deleted 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 isolation (1.5 and 3.0 times background, respectively) are established based on normal full power background radiation levels (setpoint may be adjusted based on a calculated value of the radiation level expected).

The allowable setpoints for alarm and isolation 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

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Function Instrument Channel Reactor Low Mater Level (LIS-3-203A-D, SM 2-3)

Instrument Channel Reactor High Pressure None once/3 months TABLE 4.2.A SURVEILLANCE REgUIREHENTS FOR PRIHARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Functional Test Calibration Fre uenc Instrument Ch ck (1)

(5) once/day Instrument Channel-Reactor Low Mater Level (LIS-3-56A-D, SM 01)

Instrument Channel High Drywell Pressure (PS-64-56A-0)

Instrument Channel High Radiation Hain Steam Line Tunnel once/3 months (32) once/3 month (5) once/18 months (31) once/day N/A once/day Instrument Channel-Low Pressure Hain Steam Line (PT-1-72, -76, -82, -86)

Instrument Channel High Flow Hain Steam Line (dPT-1-13A-D, -25A-D, -36A-D, -50A-D) once/3 months (27) (29) once/operating cycle (28) once/3 months (27) (29) once/operating cycle (28)

None once/day

NOTES FOR TABLES 4 2.A THROUGH 4.2.L exce t 4.2 D AND 4.2.K (Cont'd) 26.

This instrument check consists of comparing the background signal levels for all valves for consistency and for nominal expected values (not required during refueling outages).

27.

Functional test consists of the injection of a simulated signal into the electronic trip circuitry in place of the sensor signal to verify OPERABILITY of the trip and alarm functions.

28.

Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of-the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.

29.

The functional test frequency decreased to once/3 months to reduce challenges to relief valves per NUREG-0737, Item II.K.3.16.

30.

Calibration shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a

one-point source check of the detector below 10 R/hr with an installed or portable gamma source.

31.

Calibration consists of using a current source to provide an instrument channel alignment of the monitor electronics and the radiation source provides a calibration of the primary sensor.

32.

This instrumentation is exempted from the instrument channel test definition.

This instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels.

BFN Unit 1 3.2/4.2-61

3.2 BASES (Cont'd)

The low reactor water level instrumentation that is set to trip when reactor water level is 378 inches above vessel zero (Table 3.2.B) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators.

These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated.

For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves.

For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation;

thus, the results given above are applicable here also.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.

The primary function of the instrumentation is to detect a break in the main steam line.

For the worst case accident, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000'F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines.

Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas.

Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.

The setting of 200'F for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks.

For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.

In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200'F.

The temperature increases can cause an unnecessary main steam line isolation and reactor scram.

Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure.

A trip setting of three times normal full-power background is established to close the main steam line drain isolation valves, reactor water sample line isolation valves, and trip BFN Unit 1 3.2/4.2-66

3.2 BASES (Cont'd) the mechanical vacuum pump.

For changes in the background radiation level, the setpoint may be adjusted based on a calculated value of the radiation level expected.

An alarm with a nominal setpoint of 1.5 x normal full-power background is provided also.

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 psig.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping.

Tripping of this instrumentation results in actuation of HPCI isolation valves.

Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE.

High temperature in the vicinity of the HPCI equipment is sensed by four sets of four bimetallic temperature switches.

The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system.

The HPCI trip settings of 90 psi for high flow and 200'F for high temperature are such that core uncovery is prevented and fission product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI.

The trip setting of 450" H20 for high flow and 200'F for temperature are based on the same criteria as the HPCI.

High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system.

When high temperature

occurs, the cleanup system is isolated.

The instrumentation which initiates CSCS action is arranged in a dual bus system.

As for.other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.

An exception to this is when logic functional testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 1.07.

The trip logic for this function is 1-out-of-n:

e.g.,

any trip on one of six APRMs, eight IRMs, or four SRMs will result in a rod block.

The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met:

The minimum instrument channel requirements for the RBM may be reduced by one for maintenance,

testing, or calibration.

This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.

BFN Unit 1 3.2/4.2-67

'J I

3.7/4.7 BASES (Cont'd)

Demonstration of the automatic initiation capability and OPERABILITY of filter cooling is necessary to assure system performance capability. If one standby gas treatment system is inoperable, the other systems must be tested daily.

This substantiates the availability of the OPERABLE systems and thus reactor operation and refueling operation can continue for a limited period of time.

3.7.D/4.7.D Prima Containment Isolation Valves The Browns Ferry Containment Leak Rate Program and Procedures contains the list of all the Primary Containment Isolation Valves for which the Technical Specification requiremen'ts apply.

The procedures are subject to the change control provisions for plant procedures in the administrative controls section of the Technical Specifications.

The opening of locked or sealed closed" containment isolation valves on an intermittent basis under administrative control includes the following considerations:

(l) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment.

Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.

Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA.

~Gros 1 Process lines are isolated by reactor vessel low water level (378")

in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems.

The main steam line drain isolation valves in Group 1 are also closed when process instrumentation detects excessive main steam line flow, high radiation, low

pressure, or main steam space high temperature.

The main steam line isolation valves close on the same condition except for main steam line high radiation.

The reactor water sample line valves isolate only on reactor low water level at 378" or main steam line high radiation.

~Gros 3 isolation valves are closed by reactor vessel low water level (538")

or high drywell pressure.

The Group 2 isolation signal also "isolates" the reactor building and starts the standby gas treatment system.

It is not desirable to actuate the Group 2 isolation signal by a transient or spurious signal.

~Grou 3 Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from nonsafety related causes.

To protect the reactor from a possible pipe break BFN Unit l 3.7/4.7-34

1

PROPOSED TECHNICALSPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 2 (TVA BFNP TS 322)

UNIT 2 EFFECTIVE PAGE LIST REMOVE INSERT 3.1/4.1-4 3.1/4.1-6 3.1/4.1-9 3.1/4.1-11 3.1/4.1-12 3.1/4.1-15 3.2/4.2-8 3.2/4.2-13 3.2/4.2-40 3.2/4.2-61 3.2/4.2-67 3.2/4.2-68 3.2/4.2-69 3.2/4.2-70 3.2/4.2-71 3.2/4.2-72 3.2/4.2-73 3.2/4.2-73a 3.7/4.7-34 3.1/4.1-4 3.1/4.1-6 3.1/4.1-9 3.1/4.1-11 3.1/4.1-12 3.1/4.1-15 3.2/4.2-8 3.2/4.2-13 3.2/4.2-40 3.2/4.2-61 3.2/4.2-67 3.2/4.2-68*

3.2/4.2-69*

3.2/4.2-70*

3.2/4.2-71*

3.2/4.2-72*

3.2/4.2-73*

3.2/4.2-73a*

3.7/4.7-34

• Denotes Spill-Over Pages

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TABLE 3.1.A REACTOR PROTECTION SYSTEH (SCRAH)

INSTRUHENTATION RE(UIREHENTS Hin. No. of Operable Instr.

Channels Per Trip Tri Functi n

Tri Level Settin H d s in whi h F n tion Hu Be

~OeraOI e Shut-Startup/

down

~Rfoe'I 7

~Hot Standb Ron A~ction I

2 I

4 High Water Level in West Scram Discharge Tank (LS-85-45A-D)

High Water Level in East Scram Discharge Tank (LS-85-45E-H)

Hain Steam Line Isolation Valve Closure Turbine Control Valve Fast Closure or Turbine Trip Turbine Stop Valve Closure Turbine First Stage Pressure Permissive (PIS-1-81A&B, PIS-1-91A&B)

Low Scram Pilot Air Header Pressure

< 50 Gallons

< 50 Gallons

<105 Valve Closure

>550 psig

<lOL Valve Closure not >154 psig

>50 psig X(2)

X(2)

X(2)

X(2)

X(18)

X(18)

X(2)

X(2) 1.A X

1.A X(6) 1.A or 1.C X(4) 1.A or 1.0 X(4) 1.A or 1.D X(18) 1.A or 1.D (19) 1.A

NOTES FOR TABLE 1 A (Cont'd) 8 ~

Not required to be OPERABLE when primary containment integrity is not required.

9.

DELETED 10.

11.

Not required to be OPERABLE when the reactor pressure vessel head is not bolted to the vessel.

The APRM downscale trip function is only active when the reactor mode switch is in RUN.

12.

The APRM downscale trip is automatically-bypassed when the IRM instrumentation is OPERABLE and not high.

13.

14.

Less than 14 OPERABLE LPRMs will cause a trip system trip.

Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Isolation Control System.

A channel failure may be a

channel failure in each system.

15.

The APRM 15 percent scram is bypassed in the RUN Mode.

16.

Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion).

A channel failure may be a channel failure in each system.

If a channel is allowed to be inoperable per Table 3.1.A, the corresponding function in that same channel may be inoperable in the Reactor Manual Control System (Rod Block).

17.

Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).

18.

This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first stage pressure is greater than or equal to 154 psig.

19.

Action 1.A or 1.D shall be taken only if the permissive fails in such a

manner to prevent the affected RPS logic from performing its intended function.

Otherwise, no action is required.

20.

DELETED 21.

The APRM.High Flux and Inoperative Trips do not have to be OPERABLE in the REFUEL Mode if the Source Range Monitors are connected to give a noncoincidence, High Flux scram, at 5 x 10 cps.

The SRMs shall be 5

OPERABLE per Specification 3.10.B.l.

The removal of eight (8) shorting links is required to provide noncoincidence high-flux scram protection from the Source Range Monitors.

BFN Unit 3.1/4.1-6

Grou 2

High Water Level in Scram Discharge Tank Float Switches (LS-85-45C-F)

A Electronic Level Switches (LS-85-45A, B, G, H)

Hain Steam Line Isolation Valve Closure Turbine Control Valve Fast Closure or turbine trip Turbine First Stage Pressure Permissive (PIS-1-81A and B, PIS-1-91A and 8)

Turbine Stop Valve Closure Low Scram Pilot Air Header Pressure (PS 85-35 Al, A2, Bl, 8 B2)

TABLE 4.1.A (Continued)

Functional Test Trip Channel and Alarm Trip Channel and Alarm (7)

Trip Channel and Alarm Trip Channel and Alarm Trip Channel and Alarm (7)

Trip Channel and Alarm Trip Channel and Alarm Hinimum Fre u

n Once/Month Once/Honth Once/3 Honths (8)

Once/Honth (1)

Every three months Once/Honth (1)

Once/6 Honths

N td tt Instrument Charm 1

Grou 1

~Calibra i n

TABLE 4.1.B REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENT CALIBRATION HINIHUH CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Hinimum Fre u nc 2

IRH High Flux APRH High Flux Output Signal Flow Bias Signal LPRH Signal High Reactor Pressure (PIS-3-22 AA, BB, C, D)

High Drywell Pressure (PIS-64-56 A-D)

Reactor Low Water Level (LIS-3-203 A-D)

High Water Level in Sc I

Discharge Volume ram Float Switches (LS-85-45-C-F)

Electronic Level Switches (LS-85-45 A, B, G, M)

Hain Steam Line Isolation Valve Closure Turbine First Stage Pressure Permissive (PIS-1-81 A&B, PIS-1-91 A&B)

Turbine Stop Valve Closure Low Scram Pilot Air Header Pressure (PS 85-35 Al, A2, Bl, & 82)

Turbine Control Valve Fast Closure on Turbine Trip Comparison to APRH on Controlled Startups (6)

Meat Balance Calibrate Flow Bias Signal (7)

TIP System Traverse (8)

Standard Pressure Source Standard Pressure Source Pressure Standard Calibrated Water Column Calibrated Water Column Note (5)

Standard Pressure Source Note (5)

Standard Pressure Source Standard Pressure Source Note (4)

Once/7 Days Once/Operating Cycle Every 1000 Effective Full Power Hours Once/6 Months (9)

Once/18 Honths (9)

Once/18 Months (9)

Once/18 Months Once/18 Honths (9)

Note (5)

Once/18 Honths (9)

Note (5)

Once/Operating Cycle Once/18 Honths

OTES FOR TABLE 4 1.B l.

A description of three groups is included in the bases of this specification.

2.

Calibrations are not required when the systems are not required to be OPERABLE or are tripped. If calibrations are missed, they shall be performed prior to returning the system to an OPERABLE status.

3.

DELETED 4.

Required frequency is initial startup following each refueling outage.

5.

Physical inspection and actuation of -these -position-switches will be performed once per operating cycle.

6.

On controlled startups, overlap between the IRMs and APRMs will be verified.

7.

The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.

The instrumentation is an analog type with redundant flow signals that can be compared.

The flow comparator trip and upscale will be functionally tested according to Table 4.2.C to ensure the proper operation during the operating cycle.

Refer to 4.1 Bases for further explanation of calibration frequency.

8.

A complete TIP system traverse calibrates the LPRM signals to the process computer.

The individual LPRM meter readings will be adjusted as a

minimum at the beginning of each operating cycle before reaching 100 percent power.

9.

Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.

BFN Unit 2 3.1/4.1-12

~,

3.1 BASES (Cont'd)

Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel.

This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration.

Additional IRM channels have also been provided to allow for bypassing of one such channel.

The bases for the scram setting for the IRM, APRM, high reactor pressure, reactor low water level, MSIV

closure, turbine control valve fast closure and turbine stop valve closure are discussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switches) for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment.

A high drywell pressure-scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality.

This instrumentation is a backup to the reactor vessel water level instrumentation.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

Reference Section 7.2.3.7 FSAR.

The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system (120/125 scram) in conjunction with the APRM system (15 percent scram) provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges.

The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping.

The discharge volume tank accommodates in excess of 50 gallons of water and is the low point in the piping.

No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram.

During normal" operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not BFN Unit 2 3.1/4.1-15

4 0

~

~

~

~

TABLE 3.2.A (Continued)

PRIMARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION Hinimum No.

Instrument Channels Operable Per Tri 1 ll 2(3) 2(12)

Fun ion Instrument Channel High Radiation Hain Steam Line Tunnel 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-D, 36A-D, 50A-D)

Instrument Channel-Main Steam Line Tunnel High Temperature Instrument Channel Reactor Building Ventilation High Radiation - Reactor Zone Tri L v 1

ettin 3 times normal rated full power background (13)

> 825 psig (4)

< 140$ of rated steam flow

< 2000F

< 100 mr/hr or downscale Action 1

Remarks l.

Above trip setting initiates Hain Steam Line drain and reactor water sample line Isolation l.

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.

1 upscale or 2 downscale will a.

Initiate SGTS b.

Isolate reactor zone and refueling floor.

c.

Close atmosphere control system.

~

~

~

~

NOTES FOR TABLE 3.2.A (Cont'd) 4.

Only required in RUN MODE (interlocked with Mode Switch).

5.

Deleted 6.

Channel shared by RPS and Primary Containment Sc 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.

Deleted 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.

V 13.

The nominal setpoints for alarm and isolation (1.5 and 3.O times background, respectively) are established based on normal full power background radiation levels (setpoint may be adjusted based on a calculated value of the radiation level expected).

The allowable setpoints for alarm and isolation are 1.2-1.8 and 2.4-3.6 times background, respectively.

BFN Unit 2 3.2/4.2-13

0

~

TABLE 4.2.A SURVEILLANCE RE()UIREHENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Function Functional Test Calibration Fre u nc Instrumen Ch ck Instrument Channel Reactor Low Water Level (LIS-3-203A-D)

(1) (27)

Once/18 Honths (28)

Once/day Instrument Channel Reactor High Pressure (PS-68-93

& 94)

Instrument Channel Reactor Low Water Level (LIS-3-56A-D)

Instrument Channel High Drywell Pressure (PIS-64-56A-D)

Instrument Channel High Radiation Hain Steam Line Tunnel 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-D, 36A-D, 50A-D)

(31)

(1) (27)

(1) (27)

Once/3 months (33)

(29) (27)

(29) (27)

Once/18 months Once/18 months Once/18 Honths Once/18 months (32)

Once/18 Honths (28)

Once/18 Honths (28)

None (28)

Once/day (28)

N/A Once/day None Once/day

NOTES FOR TABLES 4 2.A THROUGH 4 2.L exce t 4 2 D AND 4.2.K (Cont'd) 26.

This instrument check consists of comparing the background signal levels for all valves for consistency and for nominal expected values (not required during refueling outages).

27.

Functional test consists of the injection of a simulated signal into the electronic trip circuitry in place of the sensor signal to verify OPERABILITY of the trip and alarm functions.

28.

Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of the=electronic-'trip -circuitry,. so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.

29.

The functional test frequency decreased to once/3 months to reduce challenges to relief valves per NUREG-0737, Item II.K.3.16.

30.

Calibration shall consist of an electronic calibration of.,the channel, not including the detector, for range decades above 10 R/hr and a

one-point source check of the detector below 10 R/hr with an installed or portable gamma source.

31.

Functional Tests shall be performed once/3 months.

32.

Calibration consists of using a current source to provide an instrument channel alignment of the monitor electronics and the radiation source provides a calibration of the primary sensor.

33.

This instrumentation is exempted from the instrument channel test definition.

This instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels.

BFN Unit 2 3.2/4.2-61

~

~

3.2 BASES (Cont'd) flow instrumentation is a backup to the temperature instrumentation.

In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200'F.

The temperature increases can cause an unnecessary main steam line isolation and reactor scram.

Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure.

A trip setting of three times normal full-power background is established to close the main steam line drain isolation valves, reactor water sample line isolation valves, and trip the mechanical vacuum pump.

For changes in the background radiation level, the setpoint may be adjusted based on a calculated value of the radiation level expected.

An alarm with a nominal setpoint of 1.5 x normal full-power background is provided also.

Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 825 psig.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping.

Tripping of this instrumentation results in actuation of HPCI isolation valves.

Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE.

High temperature in the vicinity of the HPCI equipment is sensed by four sets of four bimetallic temperature switches.

The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system.

Each trip system consists of two elements.

Each channel contains one temperature switch located in the pump room and three temperature switches located in the torus area.

The RCIC high flow and high area temperature sensing instrument channels are arranged in the same manner as the HPCI system.

The HPCI high steam flow trip setting of 90 psid and the RCIC high steam flow trip setting of 450" H20 have been selected such that the trip setting is high enough to prevent spurious tripping during pump startup but low enough to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.

The HPCI and RCIC steam line space temperature switch trip settings are high enough to prevent spurious isolation due to normal temperature excursions in the vicinity of the st'earn supply piping.

Additionally, these trip settings ensure that the primary containment isolation steam supply valves isolate a break within an acceptable time period to prevent core uncovery and maintain fission product releases within 10 CFR 100 limits.

BFN Unit 2 3.2/4.2-67

3.2 BASES (Cont'd)

High temperature at the Reactor Water Cleanup (RWCU) System in the main steam valve vault, RWCU pump room 2A, RWCU pump room 2B, RWCU heat exchanger room or in the space near the pipe trench containing RWCU piping could indicate a break in the cleanup system.

When high temperature

occurs, the cleanup system is isolated.

The instrumentation which initiates CSCS action is arranged in a dual bus system.

As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.

An exception to this is when logic functional testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 1.07.

The trip logic for this function is 1-out-of-n:

e.g.,

any trip on one of six APRMs, eight IRMs, or four SRMs will result in a rod block.

A General Electric study, GE-NE-770-06-0392 shows for the unit 2 cycle 6

core that if the initial MCPR is as specified in item 7e or 7f of Table 3.2.C, then no single rod withdrawal error can cause the MCPR to decrease below the MCPR safety limit.

When core operating conditions have been verified to be within the limits of items 7e or 7f of Table 3.2.C, the RBM is not required.

When the RBM is required, the minimum instrument channel requirements apply.

These requirements assure sufficient instrumentation to assure the single failure criteria is met.

The minimum instrument channel requirements for the RBM may be reduced by one for maintenance,

testing, or calibration.

This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.

The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at reduced flow.

The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence.

The trips are set so that MCPR is maintained greater than 1.07.

The RBM rod block function provides local protection of the core; i.e.,

the prevention of critical power in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.

If the IRM channels are in the worst condition of allowed bypass, the sealing arrangement is such that for unbypassed IRM channels, a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10.

A downscale indication is an indication the instrument has failed or the instrument is not sensitive enough.

In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.

The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position.

BFN Unit 2

3.2 BASES (Cont'd)

For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time.

The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate.

The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation.

The trip settings given in the specification are adequate to assure the above criteria are met.

The specification preserves the effectiveness of the system during periods of maintenance,

testing, or calibration, and also minimizes the risk of inadvertent operation; i.e.,

only one instrument channel out of service.

Two post treatment off-gas radiation monitors are provided and, when their trip point is reached, cause an isolation of the off-gas line.

Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip or both have a downscale trip.

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.

BFN Unit 2 3.2/4.2-69

3.2 BASES (Cont'd)

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.

The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation dose to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.

The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the seismic response of those features important to safety.

This capability is required to permit comparison of the measured response to that used in the design basis for Browns Ferry Nuclear Plant and to determine whether the plant can continue to be operated safely.

The instrumentation provided is consistent with specific portions of the recommendations of Regulatory Guide 1.12 "Instrumentation for Earthquakes."

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm/trip setpoints for these instruments will be calculated in accordance with guidance provided in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.

This instrumentation also includes provisions for monitoring the concentration of potentially explosive gas mixtures in the off-gas holdup system.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

BFN Unit 2 3.2/4.2-70 i

2.2 BASES (Cont'd)

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.

The alarm/trip setpoints for these instruments shall be calculated in accordance with guidance provided in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20 Appendix B, Table II, Column 2.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

ATWS/RPT, Anticipated Transients without Scram/Recirculation Pump Trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an ATWS event.

The response of the plant to this postulated event (ATWS/RPT) follows the BWR Owners Group Report by General Electric NEDE-31096-P-A and the accompanying NRC Staff Safety Evaluation Report.

ATWS/RPT utilizes the engineered safety feature (ESF) master/slave analog trip units (ATU) which consists of four level and four pressure channels total.

The initiating logic consists of two independent trip systems each consisting of two reactor dome high pressure channels and two reactor vessel low level channels.

A coincident trip of either two low levels or two high pressures in the same trip system causes initiation of ATWS/RPT.

This signal from either trip system opens one of two EOC (end-of-cycle) breakers in series (the other system opens the other breaker) between the pump motor and the Motor Generator set driving each recirculation pump.

Both systems are completely redundant such that only one trip system is necessary to perform the ATWS/RPT function.

Power comes from the 250 VDC shutdown boards.

Setpoints for reactor dome high pressure and reactor vessel low level are such that a normal Reactor Protection System scram and accompanying recirculation pump trip would occur before or coincident with the trip by ATWS/RPT.

4.2 BASES The instrumentation listed in Tables 4.2.A through 4.2.F will be functionally tested and calibrated at regularly scheduled intervals.

The same design reliability goal as the Reactor Protection System of 0.99999 generally applies for all applications of (1-out-of-2) X (2) logic.

Therefore, on-off sensors are tested once/3 months, and bistable trips associated with analog sensors and amplifiers are tested once/week.

Those instruments which, when tripped, result in a rod block have their contacts arranged in a 1-out-of-n logic, and all are capable of being bypassed.

For such a tripping arrangement with bypass capability provided, there is an optimum test interval that should be maintained in order to. maximize the reliability of a given channel (7).

This takes account of the fact that testing degrades reliability and the optimum interval between tests is approximately given by:

BFN Unit 2 S.ud.2-22

]

4.2 BASES (Cont'd) 2t Where:

i =

the optimum interval between tests.

t =

the time the trip contacts are disabled from performing their function while the test, is in progress.

r =

the expected failure rate of the relays.

To test the trip relays requires that the channel be bypassed, the test

made, and the system returned to its initial state.

It is assumed this task requires an estimated 30 minutes to complete in a thorough and workmanlike manner and that the relays have a failure rate of 10 failures per hour.

Using this data and the above operation, the optimum test interval is:

i=

=lx10 10

= 40 days For additional mar in a test interval of once er month will be used initiall The sensors and electronic apparatus have not been included here as these are analog devices with readouts in the control room and the sensors and electronic apparatus can be checked by comparison with other like instruments.

The checks which are made on a daily basis are adequate to assure OPERABILITY of the sensors and electronic apparatus, and the test interval given above provides for optimum testing of the relay circuits.

The above calculated test interval optimizes each individual channel, considering it to be independent of all others.

As an example, assume that there are two channels with an individual technician assigned to each.

Each technician tests his channel at the optimum frequency, but the two technicians are not allowed to communicate so that one can advise the other that his channel is under test.

Under these conditions, it is possible for both channels to be under test simultaneously.

Now, assume that the technicians are required to communicate and that two channels are never tested at the same time.

(7) UCRL-50451, Improving Availability and Readiness of Field Equipment Through Periodic Inspection, Benjamin Epstein, Albert Shiff, July 16,

1968, page 10, Equation (24), Lawrence Radiation Laboratory.

Forbidding simultaneous testing improves the availability of the system over that which would be achieved by testing each channel independently.

These one-out-of-n trip systems will be tested one at a time in order to take advantage of this inherent improvement in availability.

Optimizing each channel independently may not truly optimize the system considering the overall rules of system operation.

However, true system BFN Unit 2 3.2/4.2-72 j

4.2 BASES (Cont'd) optimization is a complex problem.

The optimums are broad, not sharp, and optimizing the individual channels is generally adequate for the system.

The formula given above minimizes the unavailability of a single channel which must be bypassed during testing.

The minimization of the unavailability is illustrated by Curve No.

1 of Figure 4.2-1 which assumes that a channel has a failure rate of O.l x 10

/hour and 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is required to test it.

The unavailability is a minimum at a test interval i, of 3.16 x 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

If two similar channels are used in a 1-out-of-2 configuration, the test interval for minimum unavailability changes as a function of the rules for testing.

The simplest case is to test each one independent of the other.

In this case, there is assumed to be a finite probability that both may be bypassed at one time.

This case is shown by Curve No. 2.

Note that the unavailability is lower as expected for a redundant system and the minimum occurs at the same test, interval.

Thus, if the two channels are tested independently, the equation above yields the test interval for minimum unavailability.

A more usual case is that the testing is not done independently.

If both channels are bypassed and tested at the same time, the result is shown in Curve No. 3.

Note that the minimum occurs at about 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, much longer than for cases 1 and 2.

Also, the minimum is not nearly as low as Case 2 which indicates that this method of testing does not take full advantage of the redundant channel.

Bypassing both channels for simultaneous testing should be avoided.

The most likely case would be to stipulate that one channel be bypassed,

tested, and restored, and then immediately following, the second channel be bypassed,

tested, and restored.

This is shown by Curve No. 4.

Note that there is no true minimum.

The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel.

The best test procedure of all those examined is to perfectly stagger the tests.

That is, if the test interval is four months, test one or the other channel every two months.

This is shown in Curve No. 5.

The difference between Cases 4 and 5 is negligible.

There may be other arguments, however, that more strongly support the perfectly staggered

tests, including reductions in human error.

The conclusions to be drawn are these:

1.

A 1-out-of-n system may be treated the same as a single channel in terms of choosing a test interval; and 2.

more than one channel should not be bypassed for testing at any one time.

BFN Unit 2 3.2/4.2-73

4.2 BASES (Cont'd)

The radiation monitors in the refueling area ventilation duct which initiate building isolation and standby gas treatment operation are arranged in two l-out-of-2 logic systems.

The bases given for the rod blocks apply here also and were used to arrive at the functional testing frequency.

The off-gas post treatment monitors are connected in a 2-out-of-2 logic arrangement.

Based on experience with instruments of similar design, a testing interval of once every three months has been found adequate.

The automatic pressure relief instrumentation can be considered to be a l-out-of-2 logic system and the discussion above applies also.

The criteria for ensuring the reliability and accuracy of the radioactive gaseous effluent instrumentation is listed in Table 4.2.K.

The criteria for ensuring the reliability and accuracy of the radioactive liquid effluent instrumentation is listed in Table 4.2.D.

The RCIC and HPCI system logic tests required by Table 4.2.B contain provisions to demonstrate that these systems will automatically restart on a RPV low water level signal received subsequent to a RPV high water level trip.

BFN Unit 2 3.2/4.2-73a

3.7/4.7 BASES (Cont'd)

Demonstration of the automatic initiation capability and OPERABILITY of filter cooling is necessary to assure system performance capability. If one standby gas treatment system is inoperable, the other systems must be tested daily.

This substantiates the availability of the OPERABLE systems and thus reactor operation and refueling operation can continue for a limited period of time.

3.7.D/4.7.D Prima Containment Isolation Valves The Browns Ferry Containment Leak Rate Program and Procedures contains the list of all the Primary Containment Isolation Valves for which the Technical Specification requirements apply.

The procedures are subject to the change control provisions for plant procedures in the -administrative controls section of the Technical Specifications.

The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations:

(1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment.

Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.

Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA.

~Grou I Process lines are isolated by reactor vessel low water level (g 398") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems.

The main steam line drain isolation valves in Group I are also closed when process instrumentation detects excessive main steam line flow, high radiation, low pressure, or main steam space high temperature.

The main steam line isolation valves close on the same condition except for main steam line high radiation.

The reactor water sample line valves isolate only on reactor low water level at~ 398" or main steam line high radiation.

~Grou 2 Isolation valves are closed by reactor vessel low water level (538")

or high drywell pressure.

The Group 2 isolation signal also "isolates" the reactor building and starts the standby gas treatment system.

It is not desirable to actuate the Group 2 isolation signal by a transient or spurious signal.

~Grou Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from nonsafety related causes.

To protect the reactor from a possible pipe break BFN Unit 2 3.7/4.7-34

PROPOSED TECHNICALSPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 3 (TVA BFNP TS 322)

UNIT 3 EFFECTIVE PAGE LIST REMOVE 3.1/4.1-3 3.1/4.1-5 3.1/4.1-8 3.1/4.1-10 3.1/4.1-11 3.1/4.1-14 3.2/4.2-8 3.2/4.2-13 3.2/4.2-39 3.2/4.2-60 3.2/4.2-65 3.2/4.2-66 3.7/4.7-33 INSERT 3.1/4.1-3 3.1/4.1-5 3.1/4.1-8 3.1/4.1-10 3.1/4.1-11 3.1/4.1-14 3.2/4.2-8 3.2/4.2-13 3.2/4.2-39 3.2/4.2-60 3.2/4.2-65 3.2/4.2-66 3.7/4.7-33

TABLE 3.1.A REACTOR PROTECTION SYSTEH (SCRAH)

INSTRUHENTATION REgUIREHENTS Hin. No. of Operable Instr.

Channels Per Trip Tri Fun tion Shut-Tri Level S ttin own Hodes in Which F n ion Hust Be 0 rabl Startup/

~Refuel 7

~Rn S andb

~Rn A~Ci n

1 lA 2

IlA High Water Level in West Scram Discharge Tank (LS-85-45A-D)

High Water Level in East Scram Discharge Tank (LS-85-45E-H)

Hain Steam Line Isolation Valve Closure Turbine Control Valve Fast Closure or Turbine Trip Turbine Stop Valve Closure Turbine First Stage Pressure Permissive

< 50 Gallons

< 50 Gallons

<10K Valve Closure

>550 psig

<10'4 Valve Closure not >154 psig X(2)

X(2)

X(2)

X(2)

X(18)

X(18) 1.A X

1.A X(6) 1.A or 1.C X(4) 1.A or 1.0 X(4) 1.A or 1.D X(18) 1.A or 1.D (19)

NOTES FOR TABLE 3 1.A (Cont'd) 8.

Not required to be OPERABLE when primary containment integrity is not required.

9.

DELETED 10.

Not required to be OPERABLE when the reactor pressure vessel head is not bolted to the vessel.

The APRM downscale trip function is only active when the reactor mode switch is in RUN.

12.

The APRM downscale trip is automatically bypassed when the IRM instrumentation is OPERABLE and not high.

~

13.

Less than 14 OPERABLE LPRMs will cause a trip system trip.

14.

Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Isolation Control System.

A channel failure may be a channel failure in each system.

15.

The APRM 15 percent scram is bypassed in the RUN Mode.

16.

Channel shared by Reactor Protection System and Reactor Manual Control System (Rod Block Portion).

A channel failure may be a channel failure in each system.

If a channel is allowed to be inoperable per Table 3.1.A, the corresponding function in that same channel may be inoperable in the Reactor Manual Control System (Rod Block).

17.

Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt.

18.

This function must inhibit the automatic bypassing of turbine control valve fast closure or turbine trip scram and turbine stop valve closure scram whenever turbine first stage pressure is greater than or equal to 154 psig.

19.

Action 1.A or 1.D shall be taken only if the permissive fails in such a

manner to prevent the affected RPS logic from performing its intended function.

Otherwise, no action is required.

20.

DELETED 21.

The APRM High Flux and Inoperative Trips do not have to be OPERABLE in the REFUEL Mode if the Source Range Monitors are connected to give a noncoincidence, High Flux scram, at 5 x 10 cps.

The SRMs shall be 5

OPERABLE per Specification 3.10.B.1.

The removal of eight (8) shorting links is required to provide noncoincidence high-flux scram protection from the Source Range Monitors.

BFN Unit 3.1/4.1-5

~GrOr 2

High Water Level in Scram Discharge lA Tank Float Switches (LS-85-45C-F)

A Electronic Level Switches (LS-85-45A, B, G, H)

Hain Steam Line Isolation Valve Closure Turbine Control Valve Fast Closure or turbine trip Turbine First Stage Pressure Permissive Turbine Stop Valve Closure TABLE 4.1.A (Continued)

Fun tional Tes Trip Channel and Alarm Trip Channel and Alarm (7)

Trip Channel and Alarm Trip Channel and Alarm Trip Channel and Alarm Trip Channel and Alarm Hinimum Fre uenc Once/Month Once/Honth Once/3 Honths (8)

Once/Honth (1)

Every three months Once/Honth (1)

~

C rt IRH High Flux harm 1

Grou 1

~Ca1 i bra i on Comparison to APRH on Controlled Startups (6)

Hinimum Fre uenc 2

Note (4)

TABLE 4.1.B REACTOR PROTECTION SYSTEH (SCRAH)

INSTRUMENT CALIBRATION HINIHUH CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUHENT CHANNELS I

I C)

APRH High Flux Output Signal Flow Bias Signal LPRH Signal High Reactor Pressure High Drywell Pressure Reactor Low Water Level High Mater Level in Scram Discharge Volume Float Switches (LS-85-45C-F)

Electronic Lvl Switches (LS-85-45-A, 8, G, H)

Hain Steam Line Isolation Valve Closure A

Turbine Fi rst Stage Pressure Permissive Turbine Control Valve Fast Closure or Turbine Trip Turbine Stop Valve Closure Heat Balance Calibrate Flow Bias Signal (7)

TIP System Traverse (8)

Standard Pressure Source Standard Pressure Source Pressure Standard Calibrated Mater Column (5)

Calibrated Mater Column Note (5)

Standard Pressure Source Standard Pressure Source Note (5)

Once Every 7 Days Once/Operating Cycle Every 1000 Effective Full Power Hours Every 3 Honths Every 3 Honths Every 3 Honths Note (5)

Once/Operating Cycle (9)

Note (5)

Every 6 Honths Once/Operating Cycle Note (5)

OTES FOR TABLE 4 1

B l.

A description of three groups is included in the Bases of this specification.

2.

Calibrations are not required when the systems are not required to be OPERABLE or are tripped. If calibrations are missed, they shall be performed prior to returning the system to an OPERABLE status.

l 3.

DELETED 4.

Required frequency is initial startup following each refueling outage.

5.

Physical inspection and -actuation of these position switches will be performed once per operating cycle.

6.

On controlled startups, overlap between the IRMs and APRMs will be verified.

7.

The Flow Bias Signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.

The instrumentation is an analog type with redundant flow signals that can be compared.

The flow comparator trip and upscale will be functionally tested according to Table 4.2.C to ensure the proper operation during the operating cycle.

Refer to 4.1 Bases for further explanation of calibration frequency.

8.

A complete TIP system traverse calibrates the LPRM signals to the process computer.

The individual LPRM meter readings will be adjusted as a

minimum at the beginning of each operating cycle before reaching 100 percent power.

9.

Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.

BFN Unit 3 3.1/4.1-11

3.1 BASES (Cont'd)

Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel.

This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration.

Additional IRM channels have also been provided to allow for bypassing of one such channel.

The bases for the scram setting for the IRM, APRM, high reactor pressure, reactor low water level, MSIV

closure, turbine control valve fast closure, turbine stop valve closure and loss of condenser vacuum are discussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switches) for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment.

A high drywell pressure scram-is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality.

This instrumentation is a backup to the reactor vessel water level instrumentation.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

Reference Section 7.2.3.7 FSAR.

The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

The IRM system (120/125 scram) in conjunction with the APRM system (15 percent scram) provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges.

The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping.

The discharge volume tank accommodates in excess of 50 gallons of water and is the low point in the piping.

No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram.

During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not BFH Unit 3 3.1/4.1-14

TABLE 3.2.A (Continued)

PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Hinimum No.

Instrument Channels Operable Per Tri S

s 1 ll Fun tion Tri Level e

in A tion 1

Remarks Instrument Channel 3 times normal rated High Radiation Hain Steam full power background (13)

Line Tunnel l.

Above trip setting initiates Hain Steam Line drain and reactor water sample line Isolation 2(3) 2(12) 2(14)

Instrument Channel-Low Pressure Hain Steam Line Instrument Channel High Flow Hain Steam Line Instrument Channel Hain Steam Line Tunnel High Temperature Instrument Channel Reactor Water Cleanup System Floor Drain High Temperature

) 825 psig (4)

< 140K of rated steam flow

< 2000F 160 180'F 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.

1.

Above trip setting initiates Isolation of Reactor Water Cleanup Line from Reactor and Reactor Water Return Line.

Instrument Channel Reactor Water Cleanup System Space High Temperature Instrument Channel-Reactor Building Ventilation High Radiation Reactor Zone 160 - 1800F

< 100 mr/hr or downscale 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.

~

~

~

~

NOTES FOR TABLE 3.2.A (Cont'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.

DELETED 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 isolation (1.5 and 3.0 times background, respectively) are established based on normal full power background radiation levels (setpoint may be adjusted based on a calculated value of the radiation level expected).

The allowable setpoints for alarm and isolation are 1.2-1.8 and 2.4-3.6 times background, respectively.

e 14.

Requires two independent channels from each physical location; there are two locations.

BFN Unit 3 3.2/4.2-13

TABLE 4.2.A SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Function Instrument Channel Reactor Low Water Level (LIS-3-203A-D, SW 2-3)

Instrument Channel Reactor High Pressure Instrument Channel Reactor Low Water Level (LIS-3-56A-D, SW ¹1)

Instrument Channel High Drywell Pressure (PS-64-56A-0)

Instrument Channel High Radiation Main Steam Line Tunnel Instrument Channel-Low Pressure Main Steam Line Instrument Channel-High Flow Main Steam Line Functional Te t once/3 months (31) once/3 months (27) once/3 months (27) alibration Fre u nc (5) once/3 months once/3 month (5) once/18 months (30) once/3 months once/3 months In trument Check once/day None once/day N/A once/day None once/day

OTES FOR TABLES 4 2 A THROUGH 4 2 L exce t 4 2

D AND 4 2 K (Continued) 26.

This instrument check consists of comparing the background signal levels for all valves for consistency and for nominal expected values (not required during refueling outages).

27.

Functional test frequency decreased to once/3 months to reduce the challenges to relief valves per NUREG-0737, Item II.K.3.16.

28.

Functional test consists of the injection of a simulated signal into the electronic trip circuitry in place of the sensor signal to verify OPERABILITY of the trip and alarm functions.

29.

Calibration consists of the adjustment of the-primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of the electronic trip circuitry, so its output relay changes state at or more conservatively than the analog equivalent of the trip level settings.

30.

Calibration consists of using a current source to provide an instrument channel alignment of the monitor electronics and the radiation source provides a calibration of the primary sensor.

I 31.

This instrumentation is exempted from the instrument channel test definition.

This instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels.

BFN Unit 3 3.2/4.2-60

3.2 BASES (Cont'd)

The low reactor water level instrumentation that is set to trip when reactor water level is 378 inches above vessel zero (Table 3.2.B) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation, and starts the diesel generators.

These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated.

For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a -diverse signal -to the water level instrumentation and, in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves.

For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation;

thus, the results given above are applicable here also.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.

The primary function of the instrumentation is to detect a break in the main steam line.

For the worst case accident, main steam line break outside the drywell, a trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000'F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines.

Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas.

Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.

The setting of 200'F for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks.

For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.

In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200'F.

The temperature increases can cause an unnecessary main steam line isolation and reactor scram.

Permission is provided to bypass the temperature trip for four hours to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure.

A trip setting of three times normal full-power background is established to close the main steam line drain isolation valves, reactor water sample line isolation valves, and trip BFN Unit 3 3.2/4.2-65

3.2 BASES (Cont' the mechanical vacuum pump.

For changes in the background radiation level, the setpoint may be adjusted based on a calculated value of the radiation level expected.

An alarm with a nominal setpoint of 1;5 x normal full-power background is provided also.

Pressure instrumentation is provided to close the main steam isolation valves in RUN Node when the main steam line pressure drops below 825 psig.

W The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping.

Tripping of this instrumentation results in actuation of HPCI isolation valves.

Tripping logic for the high flow is a 1-out-of-2 logic, and all sensors are required to be OPERABLE.

High temperature in the vicinity of the HPCI equipment is sensed by four sets of four bimetallic temperature switches.

The 16 temperature switches are arranged in two trip systems with eight temperature switches in each trip system.

The HPCI trip settings of 90 psi for high flow and 200'F for high temperature are such that core uncovery is prevented and fission product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI.

The trip setting of 450" water for high flow and 200'F for temperature are based on the same criteria as the HPCI.

High temperature at the Reactor Cleanup System floor drain could indicate a break in the cleanup system.

When high temperature

occurs, the cleanup system is isolated.

The instrumentation which initiates CSCS action is arranged in a dual bus system.

As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.

An exception to this is when logic functional testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 1.07.

The trip logic for this function is 1-out-of-n:

e.g.,

any trip on one of six APRMs, eight IRMs, or four SRMs will result in a rod block.

The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met.

The minimum instrument channel requirements for the RBM may be reduced by one for maintenance,

testing, or calibration.

This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.

BFN Unit 3 3.2/4.2-66

3.7/4.7 BASES (Cont'd)

Demonstration of the automatic initiation capability and OPERABILITY of filter cooling is necessary to assure system performance capability. If one standby gas treatment system is inoperable, the other systems must be tested daily.

This substantiates the availability of the OPERABLE systems and thus reactor operation and refueling operation can continue for a limited period of time.

3.7.D/4.7.D Prima Containment Isolation Valves The Browns Ferry Containment Leak Rate Program and Procedures contains the list of all the Primary Containment Isolation Valves for which the Technical Specification requirements apply.

The procedures are subject to the change control provisions for plant-procedures in the administrative controls section of the Technical Specifications.

The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations:

(1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment.

Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.

Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA.

~Grot 1 Process lines are isolated by reactor vessel low water level (378")

in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems.

The main steam line drain isolation valves in Group 1 are also closed when process instrumentation detects excessive main steam line flow, high radiation, low

pressure, or main steam space high temperature.

The main steam line isolation valves close on the same condition except for main steam line high radiation.

The reactor water sample line valves isolate only on reactor low water level at 378" or main steam line high radiation.

~Grou 2 Isolation valves are closed by reactor vessel low water level (538")

or high drywell pressure.

The Group 2 isolation signal also "isolates" the reactor building and starts the standby gas treatment system.

It is not desirable to actuate the Group 2 isolation signal by a transient or spurious signal.

~Gros Process lines are normally in use, and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from nonsafety related causes.

To protect the reactor from a possible pipe break BFN Unit, 3 3.7/4.7-33

~ ~ V 0

ENCLOSURE 2

DESCRIPTION AND JUSTIFICATION BROGANS FERRY NUCLEAR PLANT (BFN)

REASON FOR CHANGE The proposed changes to the technical specifications are being made to reflect changes introduced by the implementation of a design change which eliminates"the-reactor..-scram-..and-main-steam line isolation valve (MSIV) closure reguirements associated with the main steamline radiation monitors (MSRM).

The proposed change is based upon the General Electric Licensing Topical Report NEDO-31400 and the NRC Safety Evaluation dated May 15, 1991 on NED0-31400.

The NRC safety evaluation accepted referencing NEDO-31400 for Technical Specification changes provided the following conditions were met:

1 ~

2 ~

Demonstrate that the assumptions with regard to input values(including power per assembly, Chi/Q, and decay times) that are made in the generic analysis bound those for the plant.

Include sufficient evidence (which could be implemented or proposed operating procedure or ecpxivalent commitments) to provide reasonable assurance that increased significant levels of radioactivity in the main steam lines will be controlled expeditiously to limit both occupational doses and environmental releases.

3.

Standardize the MSRM and off gas radiatiog monitor alarm setpoint to 1.5 times the nominal N

background dose rate at the monitor locations and commits to promptly sample the reactor coolant to determine possible contamination levels in the plant reactor coolant and the need for additional corrective actions, if the MSRM or offgas radiation monitors or both exceed their alarm setpoints.

i The proposed TS change has met the above criteria and benefits the plant in several ways:

~

Reduces the scram frequency.

~

Maintains availability of the condenser for scram recovery.

~

Maintains/increases operator control over radioactive releases.

~

~

ENCLOSURE 2

Continued Page 2

The proposed TS change revises the definition of full power background radiation level and revises the calibration frequency for MSRMs from 3 months to 18 months.

DESCRIPTION OP THE PROPOSED CHANGE PAGE 3.1/F 1-4 (U1

& 2) 3.1/F 1-3 (U3) 3.1/4.1-6 (U1

& 2) 3.1/4.1-5 (U3) 3 1/4. 1-9 (U1

& 2)

F 1/4.1-8 (U3) 3.1/4.1-11 (U1

& 2) 3.1/4.1-10 (U3) 3.1/4.1-12 (U1

& 2) 3 '/4.1-11 (U3) 3.1/4.1-15 (U1

& 2) 3.1/4 '-14 (U3) 3.2/4.2-8 (U1, 2,

& 3)

DESCRXPTXON.OP-THE-CHANGES-.

~

Removes Main Steamline (MS) Radiation scram from Table 3.1.A.

~

Removes corresponding notes 9 and 20.

~

Removes MS High Radiation functional test requirements from Table 4.1.A.

~

Removes MS High Radiation calibration requirements from Table 4.1.B.

~

Removes corresponding note 3.

~

Removes discussion of MS High Radiation scram from the Bases.

Table 3.2.A, under the "Function" column for the High Radiation Main Steam Line

Tunnel, removes note 6.

E Table 3.2.A, under the "Remarks" column, adds the MS drain and reactor water sample line to the section concerning what the "Above trip setting " initiates.

ENCLOSURE 2

Continued Page 3

DESCRIPTION OP THE PROPOSED CHANGE PAGE DESCRY'PTTON OP CHANGES 3.2/4.2-13 (Ul, 2)

& 3) 3.2/4.2-40 (U 1

& 2) 3.2/4.2-39 (U3) 3.2/4.2-61 (U1

& 2) 3.2/4.2-60 (U3) 3.2/4.2-61 (U1

& 2) 3.2/4.2-60 (U3) 3 2/4.2-66&67 (Ul) 3.2/4.2-67 (U2) 3.2/4.2-65&66 (U3) 3.7/4.7-34 (U1) 3'/4 '-34 (U2) 3.7/4.7-33 (U3)

Various Notes For Table 3.2.A, removes the reactor trip from note 13.

Adds "isolation" in place of -reactor-trip. -Adds -'-'full-..power.'.!background "radiation levels (setpoint may be adjusted based on a calculated value of the radiation level expected)."

Table 4.2.A, adds note 32 (U1), note 33(U2),

and note 31 (U3) to the High Radiation Main Steamline Tunnel "Functional Test" column.

Removes note 5 and revises with calibration frequency to "Once/18 months".

Removes note 29 for (U1

& U2) and note 27 (U3) and spells out the "functional test frequency" of once/3months on U2

~

Adds note 31 (Ul), note 32 (U2), and note 30 (U3) to the "Calibration Frequency".

Notes For Table 4.2.A Through 4.2.L Except.

4.2.D And 4.2.K, adds note 31 (Ul), note 32 (U2), and note 30 (U3) which explains the calibration sources.

Notes For Table 4.2.A Through 4.2.L Except 4.2.D And 4.2.K, adds note 32 (Ul), note 33 (U2), and note 31 (U3) which explains the instrument channel functional test.

Revises the Bases to reflect the setpoint adjustment based on a calculated value of the radiation level expected and to discuss the valves closed by the trip setting.

Adds an explanation in the Bases of the footnote added to Table 3.7.A concerning the Group 1 Primary Containment Isolation Valves.

~

Correct spelling of Specification in Ul only.

~

Operable and Operability is changed to upper case where appropriate.

Inoperable changed to lower case where appropriate.

5 ~ ~

ENCLOSURE 2

Continued Page 4

JUSTIFICATION FOR THE PROPOSED CHANGE Justification for removal of these functions is provided in a General Electric (GE) Licensing Topical Report, NED0-31400, "Safety Evaluation For Eliminating The Boiling Water Reactor Main Steam Line Radiation Monitor", dated May 1987, prepared by the Boiling Water, Reactor.. Owner',s. Group (BWROG)...,

The NRC Safety Evaluation Report (SER) dated May 15, 1991 accepted NEDO-31400 for use as a reference in the licensee application, provided the following conditions in the SER were satisfied:

Demonstrate that the assumptions with regard to input values(including power per assembly, Chi/Q, and decay times) that are made in the generic analysis bound those for the plant.

BROWNS FERRY POSITION We have determined that the assumptions made in the generic analysis bound Browns Ferry as demonstrated in the following Table:

COMPARISON OF KEY ANALYSIS INPUT VALUES*

NEDO-31400 PARAHETERS BRSNIS FERRY NEDO-31400 POWER FAILED FUEL RODS OPERATION RELEASES (Nonmelt)

(Helted)

CHI/O (Ground)

CHI/O (Elevated)

HOLDUP (Delay Time) 0.1~*HW/rod(105X) 330 (FSAR 14.6-8) 1000 Days (FSAR 14.6-8) 1.8X Noble &.32X Iodine NO FUEL HELTED (FSAR 14.6-8) 1.22 x 10'Sec/H'FSAR Table 14.6-7) 9.7 x 10'Sec/H'FSAR Table 14.6.7)

Delay time provided by 6 hr holdup pipe plus charcoal adsorbers provide a 7.3 day Xenon and a 9.7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Krypton hol (FSAR pg 9.5-5)'.12HW/rod(105X) 850 LONG TERH 10X Noble

& 10X Iodine 100X Noble

& 10X Iodine 2.5 x 10'ec/H'.0 x10'Sec/H*

(BASED ON GRAPHS)

~ The values in this table in regard to the Rod Drop Accident are based on a design limit enthalpy of 280 cal/gm uhich envelopes any value in subsequent refuel analyses.

(FSAR Page 14.6-7)

• • Calculated as:

(3293 x 1.05 x 1.5) / (764 x ( 8 x 8 - 2)) "-0.109 HW/ROD (FSAR Table 1.7-1)

~ ~ ~

Page 5

ENCLOSURE 2

Continued 2 ~

Include sufficient evidence(

which could be implemented or proposed operating procedure or equivalent commitments) to provide reasonable assurance that increased significant levels of radioactivity in the main steam lines would be controlled expeditiously to limit both occupational doses and environmental releases.

BROGANS FERRY POSITION We have procedures in place which address the actions required in the event of high radiation in the main steam line.

3 ~

Standardize the MSRM and off gas ra(iation monitor alarm setpoint to 1.5 times the nominal N

background dose rate at the monitor locations and commits to promptly sample the reactor coolant to determine possible contamination levels in the plant reactor coolant and the need for additional corrective actions, if the MSRM or offgas radiation monitors or both exceed their alarm setpoints.

BROWNS FERRY POSITION The MSRM will be set to alarm at 1.5 times normal fullpower background which may be adjusted based on a calculated value of the radiation level expected.

BFN has procedures for controlling the offgas monitor setpoints as part of the Offsite Dose Calculation Manual which implement 10 CFR Part 50, Appendix I.

These settings meet the intent of the above requirements.

The MSRM will continue to trip the following:

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Main Steam Drain Valves Reactor Water Sample Lines Mechanical Condenser Vacuum Pump Mechanical Vacuum Pump Suction Line Valve

'The present MSRM calibration frequency of once/3 months in Table 4.2.A will be changed to once/18 months based on a BFN Unit 2 specific calculation.

A unit specific calculation will be performed for both Units 1 and 3 prior to their restart to confirm their calibration frequency.

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ENCLOSURE 3

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION BROGANS FERRY NUCLEAR PLANT (BFN)

UNITS 1p 2g AND 3 DESCRIPTION OF THE PROPOSED TS CHANGE The proposed TS changes affects Units 1, 2, and 3 at BFN.

The TS change proposes to eliminate the reactor scram and main steam line isolation.valve.(MSIV), closure.requirements. associated with the main steamline radiation monitors (MSRM).

The proposed change is based upon the General Electric Licensing Topical Report NEDO-31400 and the NRC staffs'ay 15, 1991 Safety Evaluation on NED0-31400.

The NRC safety evaluation accepted the referencing of NEDO-31400 provided the following conditions were met:

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Demonstrate that the assumptions with regard to input values(including power per assembly, Chi/Q, and decay times) that are made in the generic analysis bound those for the plant.

Include sufficient evidence(which could be implemented or proposed operating procedure or equivalent commitments) to provide reasonable assurance that increased significant levels of radioactivity in the main steam lines will be controlled expeditiously to limit both occupational doses and environmental releases.

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Standardize the MSRM and off gas radiatioIII, monitor alarm setpoint to 1.5 times the nominal N

background dose rate at the monitor locations and commits to promptly sample the reactor coolant to determine possible contamination levels in the plant reactor coolant and the need for additional corrective actions, if the MSRM or offgas radiation monitors or both exceed their alarm setpoints.

The proposed TS change revises the definition of full power background radiation level and revises the calibration frequency for MSRMs from 3 months to 18 months.

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ENCL08URE 3 Continued BASIS FOR PROPOSED NO 8IGNXFXCANT HAZARDS CONSXDERATION DETERMXNATXON 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 the margin, of safety.

The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The probability of occurrence of these accidents is based on initial conditions and assumptions which are not dependent on the use of or interactions with the MSBM system.

Elimination of the scram and isolation function on a high radiation signal will not affect operation of other Reactor Protection System or primary containment isolation functions.

The analysis of the control rod drop accident is described in Section 14.6.2 of the BFN Updated Final Safety Analysis Report (UFSAR).

This analysis takes credit for closure of the MSIVs upon receipt of a MSRM high radiation signal.

This closure signal limits the release of radioactivity via the condenser.

Removal of the MSRM high radiation trip signal will delay the MSIV closure, allowing more radioactivity to reach the condenser and eventually be released.

Although the resulting offsite doses calculated in the BWROG report are higher than those previously reported in the BFN UFSAR, they are not a significant increase and remain well below the limits of 10 CFR Part 100.

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ENCLOSURE 3

Continued Page 3

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In addition the Unit 2 specific calculation supports the change to the MSRMs calibration frequency of once/18 months.

Therefore, the proposed amendment does not involve a significant increase in the consequences of an accident previously evaluated.

The proposed amendment does not create the possibility of a new or different kind of accident from an accident previously evaluated.

This amendment affects the trip functions of the MSRMs.

The sole purpose of these trip functions is to mitigate the consequences of a control rod drop accident (CRDA), a previously analyzed event.

Removal of the high radiation trip signal was justified by NEDO-31400 wh'ich has been reviewed and accepted by the NRC.

The Unit 2 specific calculation supports the change to the MSRMs calibration frequency of once/18 months.

Therefore, the possibility of an accident of a new or different type is not created by this change.

The proposed amendment does not involve a significant reduction in the margin of safety.

The BFN Technical Specification Bases state that these monitors were provided to detect gross fuel failure resulting from the CRDA and provide MSIV closure to maintain radiological releases below 10 CFR Part 100 limits.

As discussed in the NRC's SER approving NED0-31400, the calculated radiological release consequences of the bounding CRDA are well within the acceptable dose limits as specified in 10 CFR Part 100.

The Unit 2 specific calculation supports the change to the MSRMs calibration frequency of once/18 months.

Unit 1 and 3 calculations will be performed prior to their restart to confirm their calibration frequency.

Therefore, these changes will not result in a significant reduction in the margin of safety.

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