Text

License Amendments 212, 227 & 185, Eliminating Requirements in Technical Specifications (TS) for Automatic Actuation of Several Functions Upon Msrm Detection of High Radiation Conditions in Main Steam Lines
	ML020030431
Person / Time
Site:	Browns Ferry
Issue date:	09/27/1994
From:	Williams J; Office of Nuclear Reactor Regulation
To:	Kingsley O; Tennessee Valley Authority
References
	-RFPFR, TAC M86095, TAC M86096, TAC M86097 NUDOCS 9410040091
	Download: ML020030431 (92)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-O00i September 27, 1994 Mr. Oliver D. Kingsley, Jr.

President, TVA Nuclear and Chief Nuclear Officer Tennessee Valley Authority 6A Lookout Place 1101 Market Street

-I Chattanooga, Tennessee 37402-2801

Dear Mr. Kingsley:

SUBJECT:

ISSUANCE OF TECHNICAL SPECIFICATION AMENDMENTS FOR THE BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 (TAC NOS.

M86095, M86096, AND M86097)

(TS 322)

The Commission has issued the enclosed Amendment Nos.

212, 227, and 185 to Facility Operating Licenses Nos.

DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant (BFN),

Units 1, 2, and 3, respectively.

These amendments are in response to your application dated April 4, 1994, regarding the reactor scram and system isolation functions generated by the Main Steamline Radiation Monitors (MSRMs).

That application superseded an earlier amendment request on this subject dated March 25, 1993.

The amendments eliminate the requirements in the Technical Specifications (TS) for automatic actuation of the following functions upon MSRM detection of a high radiation condition in the main steam lines:

(1)

(2)

(3)

(4)

(5) reactor scram main steam isolation valve closure main steam line drain valve closure reactor recirculation sample line valve closure main condenser mechanical vacuum pump isolation and trip 9~,I1 la.

1 941 0040091 9409~27 PDR ADOCK 13500025 P

AMENDMENT NO.

212 FOR BROWNS AMENDMENT NO.

227 FOR BROWNS AMENDMENT NO.

185 FOR BROWNS DATED:

September 27, 1994 Distribution w/enclosure FERRY UNIT I - DOCKET NO. 50-259 FERRY UNIT 2 - DOCKET NO.

50-260 FERRY UNIT 3 - DOCKET NO. 50-296 Docket File PUBLIC BFN Reading S. Varga G. Hill (6)

C. Grimes ACRS (10)

OPA OC/LFDCB L. Dudes J. Wermiel F. Paulitz L. Cunningham J. Minns J. Donoghue B. Boger M. Lesser T-5-C-3 0-11-E-22 T-9-E10 O-8-H-3 O-8-H-3 O-10-D-4 O-10-D-4 O-8-E-23 RII RII A copy of the NRC's Safety Evaluation is enclosed.

A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, J~oseph FWilliams, Project Manager Project Directorate 11-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket Nos.

Enclosures:

50-259, 50-260 and 50-296

1.

Amendment No.

212 to License No.

DPR-33

2.

Amendment No.

227 to License No.

DPR-52

3.

Amendment No.

185 to License No.

DPR-68

4.

Safety Evaluation cc w/enclosures:

See next page A copy of the NRC's Safety Evaluation is enclosed.

A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, ORIGINAL SIGNED:

Joseph F. Williams, Project Manager Project Directorate 11-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket Nos. 50-259, 50-260 and 50-296

Enclosures:

1. Amendment No. 212 to License No.

DPR-33

2.

Amendment No.

227 to License No.

DPR-52

3.

Amendment No.

1i5 to License No.

DPR-68

4.

Safety Evaluation cc w/enclosures:

See next page DOCUMENT NAME:

G:\\BFN\\TS322.AMD 'SEE PREVIOUS CONCURRENCE To receive a copy of this document, indicate In the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE PDII-4/LA

/-

PDII-4/PM

,*/II SRXB*

OGC f

141 PD11-4/1)

IC NAME BCItaytonDiC(X.M" JWi tiams,;"

I RJones R 6,,ee h/t-ql4i FlHebdon DATE 08/:-L 94 '

/./94 08/02/94 5I7/4

/94

,915/,*.Z/94 OFFICIAL RECORD COPY 01

Mr. Oliver D. Kingsley, Jr.

Tennessee Valley Authority cc:

Mr. Craven Crowell, Chairman Tennessee Valley Authority ET 12A 400 West Summit Hill Drive Knoxville, TN 37902 Mr. W. H. Kennoy, Director Tennessee Valley Authority ET 12A 400 West Summit Hill Drive Knoxville, TN 37902 Mr. Johnny H. Hayes, Director Tennessee Valley Authority ET 12A 400 West Summit Hill Drive Knoxville, TN 37902 Mr. 0. J. Zeringue, Sr. Vice President Nuclear Operations Tennessee Valley Authority 3B Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Dr. Mark 0. Medford, Vice President Engineering & Technical Services Tennessee Valley Authority 3B Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. D. E. Nunn, Vice President New Plant Completion Tennessee Valley Authority 3B Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. R. D. Machon, Site Vice President Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35602 General Counsel Tennessee Valley Authority ET 11H 400 West Summit Hill Drive Knoxville, TN 37902 BROWNS FERRY NUCLEAR PLANT Mr. Roger W. Huston, Manager Nuclear Licensing and Regulatory Affairs Tennessee Valley Authority 4G Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Mr. T. D. Shriver Nuclear Assurance and Licensing Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35602 Mr. Pedro Salas Site Licensing Manager Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35602 TVA Representative Tennessee Valley Authority 11921 Rockville Pike, Suite 402 Rockville, MD 20852 Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW.,

Suite 2900 Atlanta, GA 30323 Mr. Charles Patterson Senior Resident Inspector Browns Ferry Nuclear Plant U.S. Nuclear Regulatory Commission Route 12, Box 637 Athens, AL 35611 Chairman Limestone County Commission P.O. Box 188 Athens, AL 35611 State Health Officer Alabama Department of Public Health 434 Monroe Street Montgomery, AL 36130-1701

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO.

50-259 BROWNS FERRY NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 212 License No.

DPR-33

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated April 4, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

941(040096 940927 PDR ADOCK 05000259 P

PDR

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-33 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

212, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Frederick J. Heb n, Director Project Directorate 11-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 27, 1994

ATTACHMENT TO LICENSE AMENDMENT NO. 212 FACILITY OPERATING LICENSE NO.

DPR-33 DOCKET NO. 50-259 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Overleaf* pages are provided to maintain document completeness.

REMOVE INSERT 3.1/4.1-3 3.1/4.1-4 3.1/4.1-5 3.1/4.1-6 3.1/4.1-8 3.1/4.1-9 3.1/4.1-11 3.1/4.1-12 3.1/4.1-14 3.1/4.1-15 3.2/4.2-7 3.2/4.2-8 3.2/4.2-12 3.2/4.2-13 3.2/4.2-40 3.2/4.2-41 3.2/4.2-67 3.2/4.2-68 3.7/4.7-33 3.7/4.7-34 3.8/4.8-3 3.8/4.8-4 3.8/4.8-9 3.8/4.8-10 3.1/4.1-3*

3.1/4.1-4 3.1/4.1-5*

3.1/4.1-6 3.1/4.1-8*

3.1/4.1-9 3.1/4.1-11 3.1/4.1-12 3.1/4.1-14*

3.1/4.1-15 3.2/4.2-7*

3.2/4.2-8 3.2/4.2-12*

3.2/4.2-13 3.2/4.2-40 3.2/4.2-41*

3.2/4.2-67 3.2/4.2-68*

3.7/4.7-33*

3.7/4.7-34 3.8/4.8-3*

3.8/4.8-4 3.8/4.8-9 3.8/4.8-10*

TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENTATION REQUIREMENTS Min. No. of Operable Instr.

Channels Per Trip System (1)(23) 1 1

3 3

2 2

Trio Function Mode Switch in Shutdown Manual Scram IRM (16)

High Flux Inoperable APRM (16)(24)(25)

High Flux (Flow Biased)

High Flux (Fixed Trip)

High Flux Inoperative Downscale High Reactor Pressure High Drywell Pressure (14)

Reactor Low Water Level (14)

Trio Level Setting Shut down X

X

_120/125 Indicated on scale X(22)

See Spec. 2.1.A.1 1 120%

$ 15% rated power (13) 2 3 Indicated on Scale 1 1055 psig

. 2.5 psig 2538" above vessel zero Modes in Which Function Must Be Operable Startup/

Refuel (7)

Hot Standby X

X X

X X(22)

X X(21)

X(21)

(11)

X(lO)

X(8)

X X

X X(17)

X(17)

(11)

X X(8)

X rt I.-

I I-Action (1) 1.A l.A 1.A 1.A l.A or I.B l.A or 1.B l.A l.A l.A or 1.B 1.A Run X

X (5)

(5)

X X

(15)

X X(12)

K K

X 0

0

"- 0 M : 00 W:

l.A l.A

/ (

TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENTATION REQUIREMENTS Min.

No.

of Operable Instr.

Channels Per Trip System (1)(23) 2 2

4 2

Trip Function High Water Level in West Scram Discharge Tank (LS-85-45A-0)

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

Main Steam Line Isolation Valve Closure Turbine Control Valve Fast Closure or Turbine Trip Turbine Stop Valve Closure Turbine First Stage Pressure Permissive Trip Level Setting 1 50 Gallons j 50 Gallons slO% Valve Closure Modes in Which Function Must Be Operable Shut-Startup/

down Refuel (7) Hot Standby X(2)

X(2)

X(2) x x

X(3)(6)

>550 psig 110% Valve Closure not >154 psig X(18)

X(08)

Run x

x X(6)

X(4)

Action (1)

I.A 1.A 1.A or l.C 1.A or 1.D 1.A or 1.0 X(18) l.A or 1.D (19)

Ft 0J M.i

-I

(

e I

4:-

NOTES FOR TABLE 3.1.A

1.

There shall be two operable or tripped trip systems for each function.

If the minimum number of operable instrument channels per trip system cannot be met for one trip system, trip the inoperable channels or entire trip system within one hour, or, alternatively, take the below listed action for that trip function.

If the minimum number of operable instrument channels cannot be met by either trip system, the appropriate action listed below (refer to right-hand column of Table) shall be taken.

An inoperable channel need not be placed in the tripped condition where this would cause the trip function to occur.

In these cases, the inoperable channel shall be restored to operable status within two hours, or take the action listed below for that trip function.

A.

Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

In refueling mode, suspend all operations involving core alterations and fully insert all operable control rods within one hour.

B.

Reduce power level to IRN range and place mode switch in the STARTUP/HOT Standby position within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

C.

Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

D.

Reduce power to less than 30 percent of rated.

2.

Scram discharge volume high bypass may be used in shutdown or refuel to bypass scram discharge volume scram with control rod block for reactor protection system reset.

3.

Bypassed if reactor pressure is less than 1055 psig and mode switch not in RUN.

4.

Bypassed when turbine first stage pressure is less than 154 psig.

5.

IRMs are bypassed when APRMs are onscale and the reactor mode switch is in the RUN position.

6.

The design permits closure of any two lines without a scram being initiated.

7.

When the reactor is subcritical and the reactor water temperature is less than 212 0 F, only the following trip functions need to be operable:

A.

Mode switch in shutdown B.

Manual scram C.

High flux IRM D. Scram discharge volume high level E.

APRM 15 percent scram BFN 3.1/4.1-5 Unit 1

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.

11.

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 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 I.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 105 cps.

The SRMs shall be 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 3.1/4.1-6 A10-M107 NO. 212 Unit 1

TABLE 4.1.A REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENTATION FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTR.

AND CONTROL CIRC "Group (2)

Functional Test n

A Place Mode Switch in Shutdown A

Trip Channel and Alarm C

C Inoperative APRM High Flux (15% Scram)

High Flux (Flow Biased)

High Flux (Fixed Trip)

Inoperative Downscale Flow Bias High Reactor Pressure High Drywell Pressure Reactor Low Water Level C

B B

A A

A Mode Switch in Shutdow Manual Scram IRM High Flux rt S

UITS Minimum Frequencv(3)

Each Refueling Outage Every 3 Months Once/Week During Refueling and Before Each Startup Once/Week During Refueling and Before Each Startup Before Each Startup and Weekly When Required to be Operable Once/Week Once/Week Once/Week Once/Week (6)

Once/Month (1)

Trip Channel and Alarm (4)

Trip Output Relays (4)

(6)

Trip Channel and Alarm Trip Channel and Alarm Trip Channel and Alarm

(

0-.

I Oo

Group (2)

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

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

A Main Steam Line Isolation Valve Closure A

Turbine Control Valve Fast Closure or turbine trip A

Turbine First Stage Pressure Permissive (PT-1-81A and B, PT-1-91A and B)

B Turbine Stop Valve Closure A

TABLE 4.1.A (Continued)

Functional 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 Frequency(3)

Once/Month Once/Month Once/3 Months (8)

Once/Month (1)

Every three months Once/Month (1)

-I:

I-(

0*

rt I-.

t

Instrument Channe IRM High Flux TABLE 4.1.0 REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENTATION CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS 1

Group (1)

Calibration C

Comparison to APRM on Controlled Note Startups (6)

APRM High Flux Output Signal Flow Bias Signal LPRM 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, 6, H)

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

Main 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 B

B B

A A

B A

A 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)

Minimum Frequencv(2)

(4)

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

Note (5)

Once/Operating Cycle (9)

Once/Operating Cycle Note (5) rt I

-I C

C

NOTES FOR TABLE 4.I.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 IRNs 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 LPRN signals to the process computer.

The individual LPRN 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 3.1/4.1-12 A"kIM NO.

212 Unit 1

3.1 BASES The reactor protection system automatically initiates a reactor scram to:

1.

Preserve the integrity of the fuel cladding.

2.

Preserve the integrity of the reactor coolant system.

3.

Minimize the energy which must be absorbed following a loss of coolant accident, and prevents criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periods when instrument channels may be out of service because of maintenance.

When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

The reactor protection trip system is supplied, via a separate bus, by its own high inertia, ac motor-generator set.

Alternate power is available to either Reactor Protection System bus from an electrical bus that can receive standby electrical power.

The RPS monitoring system provides an isolation between nonclass 1E power supply and the class 1E RPS bus.

This will ensure that failure of a nonclass 1E reactor protection power supply will not cause adverse interaction to the class 1E Reactor Protection System.

The reactor protection system is made up of two independent trip systems (refer to Section 7.2, FSAR).

There are usually four channels provided to monitor each critical parameter, with two channels in each trip system.

The outputs of the channels in a trip system are combined in a logic such that either channel trip will trip that trip system.

The simultaneous tripping of both trip systems will produce a reactor scram.

This system meets the intent of IEEE-279 for Nuclear Power Plant Protection Systems.

The system has a reliability greater than that of a 2-out-of-3 system and somewhat less than that of a 1-out-of-2 system.

With the exception of the Average Power Range Monitor (APRM)

channels, the Intermediate Range Monitor (IRM) channels, the Main Steam Isolation Valve closure and the Turbine Stop Valve closure, each trip system logic has one instrument channel.

When the minimum condition for operation on the number of operable instrument channels per untripped protection trip system is met or if it cannot be met and the effected protection trip system is placed in a tripped condition, the effectiveness of the protection system is preserved; i.e.,

the system can tolerate a single failure and still perform its intended function of scramming the reactor.

Three APR' instrument channels are provided for each protection trip system.

BFN 3.1/4.1-14 Unit 1

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 BFN 3.1/4.1-15 A2M'IL-07, NIO.

212 Unit 1

TABLE 3.2.A PRIMARY CONTAINMENT AND REACTOR BUILDING Minimum No.

Instrument Channels Operable Per Trio Svs(1)(11) 2 2

2 rt Trio Level Setting

> 538" above vessel zero 100 +/- 15 psig 2 378" above vessel zero

( 2.5 psig ISOLATION INSTRUMENTATION Action (1)

A or (B and E)

D A

A or (B and E)

Remarks

1.

Below trip setting does the following:

a.

Initiates Reactor Building Isolation

b.

Initiates Primary Containment Isolation (Groups 2, 3, and 6)

c.

Initiates SGTS

1. Above trip setting isolates the shutdown cooling suction valves of the RHR system.

1. Below trip setting initiates Main Steam Line Isolation

1.

Above trip setting does the following:

a.

Initiates Reactor Building Isolation

b.

Initiates Primary Containment Isolation

c.

Initiates SGTS Instrument Channel Reactor Low Water Level(6)

Instrument Channel Reactor High Pressure (P$-68-93 and 94)

Instrument Channel Reactor Low Water Level (LIS-3-56A-D, SW #1)

Instrument Channel High Orywell Pressure (6)

(PS-64-56A-D) 4-li N

N

-.4

(

p

(

TABLE 3.2.A (Continued)

PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.

Instrument Channels Operable Per Trip Svs(l)(11) 2 2(3) 2(12) 2(14) 2 1(05) ii rt Trip Level Setting

> 825 psig (4)

< 140% of rated steam flow

< 2006F Action (1)

B 8

B C

160 - 1806F Function Instrument Channel Low Pressure Main Steam Line Instrument Channel High Flow Main Steam Line Instrument Channel Main Steam Line Tunnel High Temperature Instrument Channel Reactor Water Cleanup System Floor Drain High Temperature Instrument Channel Reactor Water Cleanup System Space High Temperature Instrument Channel Reactor Building Ventilation High Radiation - Reactor Zone 1 100 mr/hr or downscale C

G Remarks

1.

Below trip setting initiates Main Steam Line Isolation

1.

Above trip setting initiates Main Steam Line Isolation

1.

Above trip setting initiates Main Steam Line Isolation.

1.

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

1.

Same as above

1.

1 upscale channel or 2 downscale channels will

a.

Initiate SGTS

b.

Isolate reactor zone and refueling floor.

c.

Close atmosphere control system.

160 -

180OF N

N, I

(

-1 C-I 0

C

NOTES FOR TABLE 3.2.A

1.

Whenever the respective functions are required to be OPERABLE there shall be two OPERABLE or tripped trip systems for each function.

If the first column cannot be met for one of the trip systems, that trip system or logic for that function shall be tripped (or the appropriate action listed below shall be taken).

If the column cannot be met for all trip systems, the appropriate action listed below shall be taken.

A.

Initiate an orderly shutdown and have the reactor in Cold Shutdown in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B.

Initiate an orderly load reduction and have Main Steam Lines isolated within eight hours.

C.

Isolate Reactor Water Cleanup System.

D.

Administratively control the affected system isolation valves in the closed position within one hour and then declare the affected system inoperable.

E.

Initiate primary containment isolation within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

F.

The handling of spent fuel will be prohibited and all operations over spent fuels and open reactor wells shall be prohibited.

G.

Isolate the reactor building and start the standby gas treatment system.

H.

Immediately perform a logic system functional test on the logic in the other trip systems and daily thereafter not to exceed 7 days.

I.

Deleted J.

Withdraw TIP.

K.

Manually isolate the affected lines.

Refer to Section 4.2.E for the requirements of an inoperable system.

L.

If one SGTS train is inoperable take actions H or A and F.

If two SGTS trains are inoperable take actions A and F.

2.

Deleted

3.

There are four sensors per steam line of which at least one sensor per trip system must be OPERABLE.

BFN 3.2/4.2-12 AIDMEIT NO. 18 5 Unit 1

INOTES 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 & 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.

For the Reactor Building Ventilation system, one channel may be inoperable for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for functional testing or for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for calibration and maintenance, as long as the downscale trip of the inoperable channel is placed in the tripped condition.

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.

Deleted BFN 3.2/4.2-13 AMEaD1,MN NO.

212 Unit 1

TABLE 4.2.A SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Function Functional Test Calibration Frequency Instrument Check C"

Instrument Channel -

(1)

(5) once/day Reactor Low Water Level (LIS-3-203A-D, SW 2-3)

Instrument Channel -

(31) once/18 months None Reactor High Pressure (PS-68-93 & -94)

Instrument Channel -

(1) once/3 month once/day Reactor Low Water Level (LIS-3-56A-D, SW #1)

Instrument Channel -

(1)

(5)

N/A High Drywell Pressure (PS-64-56A-D)

Instrument Channel -

once/3 months (27) (29) once/operating cycle (28)

None Low Pressure Main Steam Line (PT-1-72,

-76. -82, -86)

Channel -

once/3 months (27)

(29) once/operating cycle (28) once/day I'-

High Flow Main Steam Line (dPT-1-13A-D,

-25A-D.

-36A-D.

-5OA-D)

-I C

t,.(

rt LJ*

I-TABLE 4.2.A (Cont'd)

FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Functional Test Calibration Frequency Instrument Check once/3 months (27) once/operating cycle None SURVEILLANCE REQUIREMENTS Function Instrument Channel Main Steam Line Tunnel High Temperature Instrument Channel Reactor Building Ventilation High Radiation - Reactor Zone Instrument Channel Reactor Building Ventilation High Radiation - Refueling Zone Instrument Channel SGTS Train A Heaters Instrument Channel SGTS Train B Heaters Instrument Channel SGTS Train C Heaters Reactor Building Isolation Timer (refueling floor)

Reactor Building Isolation Timer (reactor zone) once/18 months once/18 Months (4)

(4)

(9)

(9) once/operating cycle once/operating cycle once/day (8) once/day (8)

(

N/A N/A N/A N/A N/A

(

R CA cii (1) (30)

(1) (30)

3.2 BASES (Cont'd)

The setting of 2000F 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 2000F.

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.

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 200OF 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 200OF 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.

BFN 3.2/4.2-67 ANUIDi1EN NO.

212 Unit 1

3.2 BASES (Cont'd)

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 IRNs, 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 RB? 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 KCPR, 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 IRN 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.

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 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.

BFN 3.2/4.2-68 AMENDMW K 205 Unit 1

3.7/4.7 BASES (Cont'd) containment is opened for refueling.

Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system performance capability.

The test frequencies are adequate to detect equipment deterioration prior to significant defects, but the tests are not frequent enough to load the filters, thus reducing their reserve capacity too quickly.

That the testing frequency is adequate to detect deterioration was demonstrated by the tests which showed no loss of filter efficiency after two years of operation in the rugged shipboard environment on the US Savannah (0__

7).

Pressure drop across the combined HEPA filters and charcoal adsorbers of less than six inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter.

Heater capability, pressure drop and air distribution should be determined at least once per operating cycle to show system performance capability.

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated.

Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with USAEC Report DP-1082.

Iodine removal efficiency tests shall follow ASTM D3803.

The charcoal adsorber efficiency test procedures shoul allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples.

Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.

If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified according to Table 1 of Regulatory Guide 1.52.

The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality.

Tests of the HEPA filters with DOP aerosol shall be performed in accordance to ANSI N510-1975.

Any HEPA filters found defective shall be replaced with filters qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52.

All elements of the heater should be demonstrated to be functional and OPERABLE during the test of heater capacity.

Operation of each filter train for a minimum of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months each month will prevent moisture buildup in the filters and adsorber system.

With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal.

Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated.

If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use.

The determination of significance shall be made by the operator on duty at the time of the incident.

Knowledgeable staff members should be consulted prior to making this determination.

BFN 3.7/4.7-33 I AMENDMENT NO. 189 Unit 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 Primary 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 Specifictions.

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.

Group 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 valves in Group 1, except the reactor water sample line valves, are also closed when process instrumentation detects excessive main steam line flow, low pressure, or main steam space high temperature.

The reactor water sample line valves isolate only on reactor low water level at 378".

Group 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.

Group 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 3.7/4.7-34 AHRID T NO.

212 Unit 1

-3.8/4.8 RADIOACTIVE MATERIALS Y1WT'l.Td V

Ul'Yf AD MD'LAITA QTTD1I'ITTT-A f11P* DrnTTT'D*W*rQ

-.U.AAAAM.LL UU.KIAX1 X

M&.LUJL M

W

, JJ*,-

h.t al*,

3.8.B.

Airborne Effluents

1.

(Deleted)

2.

(Deleted)

3.

(Deleted)

4.

(Deleted)

5.

(Deleted)

6.

(Deleted)

7.

(Deleted)

8.

(Deleted)

9.

Whenever the SJAE is in service, the concentration of hydrogen in the offgas downstream of the recombiners shall be limited to 14% by volume.

10. With the concentration of hydrogen exceeding the limit of 3.8.B.9 above, restore the concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

BFN 3.8/4.

Unit 1 4.8.B.

Airborne Effluents

1.

(Deleted)

2.

(Deleted)

3.

(Deleted)

4.

(Deleted)

5.

The concentration of hydrogen downstream of the recombiners shall be determined to be within the limits of 3.8.B.9 by continuously monitoring the off-gas whenever the SUAE is in service using instruments described in Table 3.2.K.

Instrument surveillance requirements are specified in Table 4.2.K.

AMENDMENT NO. 19 9 I

tr 0

lu r-I 4j

-r4

3.,8 BASES (Deleted) 3.8.A LIOUID HOLDUP TANKS Specification 3.8.A.5 includes any tanks containing radioactive material that are not surrounded by liners, dikes, or walls capable of holding the contents and that do not have overflows and surrounding area drains connected to the liquid radwaste treatment system.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

3.8.B EXPLOSIVE GAS MIXTURE Specification 3.8.B.9 and 10 is provided to ensure that the concentration of potentially explosive gas mixtures contained in the offgas system is maintained below the flammability limits of hydrogen.

Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

4.8.A and 4.8.B BASES (Deleted) 3.8.C and 4.8.C BASES (Deleted) 3.8.D and 4.8.D BASES (Deleted)

BFn Unit 1 3.8/4.8-9 AMEND=ELN NO. 212

-I

3.8.E and 4.8.E BASES The limitations on removable contamination for sources requiring leak testing, including alpha emitters, based on 10 CFR 70.39(c) limits for plutonium.

This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group.

Those sources which are frequently handled are required to be tested more often than those which are not.

Sealed sources which are continuously enclosed within a shielded mechanism (i.e.,

sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

BFN 3.8/4.8-10 I AMENDMENT NO. 19 9 Unit 1 I

SUNITED STATES NUCLEAR REGULATORY COMMISSION "WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO.

50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 227 License No.

DPR-52

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated April 4, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-52 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 227, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Frederick J. Hebdbn, Director Project Directorate 11-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 27, 1994

ATTACHMENT TO LICENSE AMENDMENT NO.

227 FACILITY OPERATING LICENSE NO.

DPR-52 DOCKET NO.

50-260 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Overleaf* pages are provided to maintain document completeness.

REMOVE INSERT 3.1/4.1-3 3.1/4.1-4 3.1/4.1-5 3.1/4.1-6 3.1/4.1-8

3. 1/4. 1-9

3. 1/4. 1-11 3.1/4. 1-11a 3.1/4. 1-12 3.1/4. 1-14 3.1/4. 1-15 3.2/4.2-7 3.2/4.2-8 3.2/4.2-12 3.2/4.2-13 3.2/4.2-40 3.2/4.2-41 3.2/4.2-67 3.2/4.2-68 3.7/4.7-33 3.7/4.7-34 3.8/4.8-3 3.8/4.8-4 3.8/4.8-9 3.8/4.8-10

3. 1/4. 1-3*

3.1/4.1-4

3. 1/4. 1-5*

3.1/4.1-6

3. 1/4.1-8*

3.1/4.1-9 3.1/4.1-11 3.1/4.1-12 3.1/4.1-14*

3.1/4.1-15 3.2/4.2-7*

3.2/4.2-8 3.2/4.2-12*

3.2/4.2-13 3.2/4.2-40 3.2/4.2-41*

3.2/4.2-67 3.2/4.2-68*

3.7/4.7-33*

3.7/4.7-34 3.8/4.8-3*

3.8/4.8-4 3.8/4.8-9 3.8/4.8-10*

TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENTATION REQUIREMENTS Min. No. of Operable Instr.

Channels Per Trip System (1)(23)

Trio Function Mode Switch in Shutdown Manual Scram IRM (16)

High Flux Inoperable APRM (16)(24)(25)

High Flux (Flow Biased)

High Flux (Fixed Trip)

High Flux Inoperative Downscale Trio Level Setting

_120/125 Indicated on scale Shut down X

X X(22)

See Spec. 2.1.A.1

< 120%

< 15% rated power (13)

> 3 Indicated on Scale High Reactor Pressure

< 1055 pslg (PIS-3-22AA.6B,C.D)

High Drywell Pressure (14)

. 2.5 psig (PIS-64-56 A-O)

Reactor Low Water Level (14)

> 538" above (LIS-3-203 A-D) vessel zero Modes in which Function Must Be Operable Startup/

Refuel (7)

Hot Standby X

X X

X X(22)

X X(21)

X(21)

(11)

X(10)

X(8)

X X

X X(17)

X(17)

(11)

X X(8)

X Run X

X (5)

(5)

X X

(15)

X X(02)

X X

X Action (1) 1.A 1.A 1.A 1.A 1.A or 1.8 1.A or 1.B 1.A 1.A 1.A or 1.B l.A l.A l.A C__ txi rt I

-.o 1

3 3

2 2

0ti 0

fH M 0 00Z

-0 I

TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENTATION REQUIREMENTS Min. No. of Operable Instr.

Channels Per Trip System (1)(23) 2 2

4 2

2 wJ z

0-.

Modes in which Function Must Be Operable Trip Level Setting High Water Level in West Scram Discharge Tank (LS-85-45A-D)

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

Main 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-I-91A&B)

Low Scram Pilot Air Header Pressure

( 50 Gallons

_ 50 Gallons (10% Valve Closure Z550 psig Shut down Startup/

Refuel (7)

Hot Standby x

x X(2)

X(2)

Action (1) x x

X(6)

X(4)

(10% Valve Closure not 1154 psig 50 psig X(18)

X(2)

X(18) x 1.A 1.A 1.A or 1.C l.A or 1.D l.A or 1.D X(18)

I.A or 1.D (19) x 1.A TJj

'I..

rt t'3

(

-I

(

- NOTES FOR TABLE 3.1.A

1.

There shall be two OPERABLE or tripped trip systems for each function.

If the minimum number of OPERABLE instrument channels per trip system cannot be met for one trip system, trip the INOPERABLE channels or entire trip system within one hour, or, alternatively, take the below listed action for that trip function.

If the minimum number of OPERABLE instrument channels cannot be met by either trip system, the appropriate action listed below (refer to right-hand column of Table) shall be taken.

An INOPERABLE channel need not be placed in the tripped condition where this would cause the trip function to occur.

In these cases, the INOPERABLE channel shall be restored to OPERABLE status within two hours, or take the action listed below for that trip function.

A.

Initiate insertion of OPERABLE rods and complete insertion of all OPERABLE rods within four hours.

In refueling mode, suspend all operations involving core alterations and fully insert all OPERABLE control rods within one hour.

B.

Reduce power level to IRM range and place mode switch in the STARTUP/HOT Standby position within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

C.

Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

D.

Reduce power to less than 30 percent of rated.

2.

Scram discharge volume high bypass may be used in SHUTDOWN or REFUEL to bypass scram discharge volume scram and scram pilot air header low pressure scram with control rod block for reactor protection system reset.

3.

(Deleted)

4.

Bypassed when turbine first stage pressure is less than 154 psig.

5.

IRMs are bypassed when APRMs are onscale and the reactor mode switch is in the RUN position.

6.

The design permits closure of any two lines without a scram being initiated.

7.

When the reactor is subcritical and the reactor water temperature is less than 2120 F, only the following trip functions need to be OPERABLE:

A.

Node switch in SHUTDOWN B.

Manual scram C.

High flux IRM D.

Scram discharge volume high level E.

APR.

15 percent scram F.

Scram pilot air header low pressure BFN 3.1/4.1-5 Unit 2

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.

11.

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 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 l.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 105 cps.

The SRMs shall be 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 3.1/4.1-6 A=-TEENT NO.

227 Unit 2

Mode Switch in Shutd Manual Scram IRH High Flux Inoperative TABLE 4.1.A REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENTATION FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CI Group (2)

Functional Test own A

Place Mode Switch in Shutdown A

Trip Channel and Alarm APRM High Flux (15% Scram)

High Flux (Flow Biased)

High Flux (Fixed Trip)

Inoperative Downscale Flow Bias High Reactor Pressure (PIS-3-22AA, BB. C, D)

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

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

Trip Channel and Alarm (4)

Trip Output Relays (4)

Trip Trip Trip Trip (6)

Trip B

Output Relays (4)

Channel and Alarm (7)

Trip Channel and Alarm (7)

B Trip Channel and Alarm (7)

Ft I.J RCUITS Minimum Frequencv(3)

Each Refueling Outage Every 3 Months Once/Week During Refueling and Before Each Startup Once/Week During Refueling and Before Each Startup Before Each Startup and Weekly When Required to be Operable Once/Week Once/Week Once/Week Once/Week (6)

Once/Month Once/Month Once/Month I-.

I-'

(.

(

Group (2)

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

A Electronic Level Switches (LS-85-45A, B, 6, H)

B Main Steam Line Isolation Valve Closure A

Turbine Control Valve Fast Closure or turbine trip A

Turbine First Stage Pressure Permissive (PIS-1-81A and B, PIS-1-91A and B) 8 Turbine Stop Valve Closure A

Low Scram Pilot Air Header Pressure (PS 85-35 Al, A2, Bl,

& B2)

A 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 Minimum Frequencv(3)

Once/Month Once/Month Once/3 Months (8)

Once/Month (1)

Every three months Once/Month (1)

Once/6 Months rt K)

I I-D

-I

t'3 IJ)

I-.

I z

0 TABLE 4.1.B REACTOR PROTECTION SYSTEM (SCRAM)

INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Group (1)

Calibration Minimum Frequencv(2)

IRM High Flux C

Comparison to APRM on Controlled Note (4)

Startups (6)

APRM High Flux Output Signal B

Heat Balance Once/7 Days Flow Bias Signal B

Calibrate Flow Bias Signal (7)

Once/Operating Cycle LPRM Signal B

TIP System Traverse (8)

Every 1000 Effective Full Power Hours High Reactor Pressure B

Standard Pressure Source Once/6 Months (9)

(PIS-3-22 AA, BB, C, D)

High Drywell Pressure B

Standard Pressure Source Once/18 Months (9)

(PIS-64-56 A-D)

Reactor Low Water Level B

Pressure Standard Once/18 Months (9)

(LIS-3-203 A-D)

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

A Calibrated Water Column Once/18 Months Electronic Level Switches (LS-85-45 A, B. G, H)

B Calibrated Water Column Once/18 Months (9)

Main Steam Line Isolation Valve Closure A

Note (5)

Turbine First Stage Pressure Permissive (PIS-1-81 A&B, PIS-1-91 A&B)

B Standard Pressure Source Once/18 Months (9)

Turbine Stop Valve Closure A

Note (5)

Turbine Control Valve Fast Closure on Turbine Trip A

Standard Pressure Source Once/Operating Cycle Low Scram Pilot Air Header Pressure (PS 85-35 Al.

A Standard Pressure Source Once/18 Months A2.

01, & B2)

-I

(

NOTES FOR TABLE 4.1.B

1.

A description of three groups is intluded 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 3.1/4.1-12 AIMMIERU NO. 227 Unit 2

-31 BASES The reactor protection system automatically initiates a reactor scram to:

1.

Preserve the integrity of the fuel cladding.

2.

Preserve the integrity of the reactor coolant system.

3.

Minimize the energy which must be absorbed following a loss of coolant accident, and prevents criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periods when instrument channels may be out of service because of maintenance.

When necessary, one channel may be made INOPERABLE for brief intervals to conduct required functional tests and calibrations.