ML18032A704

Proposed Tech Specs Re Interpretation of Surveillance Instruction Operability Requirements Concerning Testing of Reactor Bldg,Reactor Zone & Refuel Zone Radiation Monitors Listed in Table 3.2.A

ML18032A704
Person / Time
Site:	Browns Ferry
Issue date:	01/22/1988
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18032A703	List: ML18032A702 ML18032A704
References
TAC-R00008, TAC-R00009, TAC-R00010, TAC-R10, TAC-R8, TAC-R9, NUDOCS 8801280454
Download: ML18032A704 (54)

Text

ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATIONS REVISIONS BROHNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 (TVA BFN TS 228) 8801280454 880i22 PDR RDDCK 05000259 t P .. PDR

~ 4 Section ~Pa e No.

D. Reactivity Anomalies 3.3/4.3-11 E. Reactivity Control 3.3/4.3-12 F. Scram Discharge Volume 3.3/4.3-12 3.4/4.4 Standby Liquid Control System. 3.4/4.4-1 A. Normal System Availability . 3.4/4.4-1 B. Operation with Inoperable Components . 3.4/4.4-2 C. Sodium Pentaborate Solution. 3.4/4.4-3 3.5/4.5 Core and Containment Cooling Systems . 3.5/4.5-1 A. Core Spray System (CSS). 3.5/4.5-1 B. Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling) 3.5/4.5-4 C. RHR Service Water System and Emergency Equipment Cooling Water System (EECWS) 3.5/4.5-10 D. Equipment Area Coolers . 3.5/4.5-13 E. High Pressure Coolant Injection System (HPCIS) . 3.5/4.5-13 F. Reactor Core Isolation Cooling System (RCICS) . 3.5/4.5-14 G. Automatic Depressurization System (ADS). 3.5/4.5-15 H. Maintenance of Filled Discharge Pipe . 3.5/4.5-17 I. Average Planar Linear Heat Generation Rate 3.5/4.5-18 J. Linear Heat Generation Rate (LHGR) 3.5/4.5-18 K. Minimum Critical Power Ratio (MCPR). 3.5/4.5-19 "

L. APRM Setpoints 3.5/4.5-20 3.6/4.6 Primary System Boundary.

3.6/4.6-1'.6/4.6-1 A. Thermal and Pressurization Limitations B. Coolant Chemistry. 3.6/4.6-5 BFN-Uni t 1

NOTES FOR TABLE 3.2.C

1. The minimum number of operable channels for each trip function is detailed for the startup and run positions of the reactor mode selector switch. The SRM, IRM, and APRM (startup mode),

blocks need not be. operable in "run" mode, and the APRN (flow biased) rod blocks need not be operable in "startup" mode.

Nith the number of OPERABLE channels less than required by the minimum OPERABLE channels per trip function requirement, place at least one inoperable channel in the tripped condition within one hour.

2. N is the recirculation loop flow in percent of design. Trip level setting is in percent of rated power (3293 MNt).

A ratio of FRP/CNFLPD <1.0 is permitted at reduced power. See Specification 2.1 for APRM control rod block setpoint.

3. IRM downscale is bypassed when it is on its lowest range.

4. SRMs A and C downscale functions are bypassed when IRNs A, C, E, and G are above range 2. SRNs B and D downscale function is bypassed when IRNs B, D, F, and 'H are above range 2.

SRM detector not in startup position is bypassed when the count rate is >100 CPS or the above condition is satisfied.

5. .During repair or calibration of equipment, not more than one SRM or RBM channel nor more than two APRM or IRM channels may be bypassed. Bypassed channels are not counted as operable channels to meet the minimum operable channel requirements.

Refer to section 3.10.B for SRM requirements during core alterations.

6. IRM channels A, E, C, G all in range 8 or above bypasses SRM channels A and C functions.

IRM channels B, F, D, H all in range 8 or above bypasses SRM channels B and D functions.

7. The following operational restraints apply to the RBM only.

a. Both RBM channels are bypassed when reactor power is <30 percent and when a peripheral control rod is selected.

b. The RBM need not be operable in the "startup" position of the reactor mode selector switch.

C. Two RBM channels are provided and only one of these may be bypassed from the console. If the inoperable channel cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel shall be placed in the tripped condition within one hour.

d. Hith both RBM channels inoperable, place"at least one inoperable rod block monitor channel in the tripped condition within one hour.

BFN 3.2/4.2-26 Unit 1

1 NOTES FOR TABLES 4.2.A THROUGH 4.2.H exce t 4.2.D (Continued)

14. (Deleted) 15 The flow bias comparator will be tested by putting one flow unit in "Test" (producing 1/2 scram) and adjusting the test input to obtain comparator rod block. The flow bias upscale will be verified by observing a local upscale trip light during operation and verified

'6. that it will produce a rod block during. the operating cycle.

Performed during operating cycle. Portions of the logic is checked more frequently during functional tests of the functions that produce a rod block.

17. This calibration consists of removing the function from service and performing an electronic calibration of the channel.

18. Functional test is limited to the'condition where secondary containment integrity is not required as specified in Sections 3 '.C.2 and 3.7.C.3.

19. Functional test is limited to the time where the SGTS is required to meet the requirements of Section 4.7.C.l.a.

20. Calibration of the comparator requires the inputs from both recirculation loops to be interrupted, thereby removing the flow bias signal to the APRM and RBM and scramming the reactor. This calibration can only be performed during an outage.

21. Logic test is limited to the time where actual operation of the equipment is permissible.

22. One channel of either- the reactor zone or refueling zone Reactor Building Ventilation Radiation Monitoring System may be administratively bypassed for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for functional testing and calibration.

23. (Deleted)

24. This instrument .check consists of comparing the thermocouple readings for all valves for consistence and for nominal expected values (not required during refueling outages).

25. During each refueling outage, all acoustic monitoring channels shall be calibrated. This calibration includes verification of accelerometer response due to mechanical exci tation in the vicinity of the sensor.

BFN 3.2/4.2-60 I Ink t

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE R:. '.REMENTS 3.5 Core and Containment Coolin S stems

~Ss tems i ~C

1. Whenever the core thermal FRP/CMFLPD shall be power is > 25% of rated, the determined daily when ratio of FRP/CMFLPD shall the reactor is > 25% of be > 1.0, or the APRM scram rated thermal power.

and rod block setpoint equations listed in Sections 2.1.A and 2.1.8,shall be multiplied by FRP/CMFLPD as follows:

S< (0.66W + 54%)

CMFLPD Spa< (0.66W + 42%) (" " )

CMFLPD

2. When it is determined that 3.5.L.l is not being met, 6 hours is allowed to correct the condition.

3. If 3.5.L.l and 3.5.L.2, cannot be met, the reactor power shall be reduced to

< 25% of rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

BFH 3.5/4.5-20 Unit 1

3.5 BASES (Cont'd) 3.5.M. References

1. "Fuel Densification Effects on General Electric Boiling Hater Reactor Fuel," Supplements 6, 7, and 8, NEIM-10735, August 1973.

2. Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 14, 1974 (USA Regulatory Staff).

3. Communication: V. A. Moore to I. S. Mitchell, "Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974.

4. Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.

5. Letter from R. H. Buchholz (GE) to P. S. Check (NRC), "Response to NRC Request For Information On ODYN Computer Model," September 5, 1980.

BFN 3.5/4.5-34 Unit 1

) ~

k I,

3.6/4.6 PRIMARY SYSTEM BOUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.H. Seismic Restraints Su orts, 4.6.H. Seismic Restraints Su orts, and Snubbers and Snubbers During all modes of operation The surveillance requirements all seismic restraints, of paragraph 4.6.G are the snubbers, and supports shall, only requirements that apply be OPERABLE except as noted to any seismic restraint or in 3.6.H.l. All safety- support other than snubbers.

related snubbers are listed in Surveillance Instruction Each safety-related snubber BF SI 4.6.H-l and BF SI 4.6.H-2. shall be demonstrated OPERABLE by performance of the following

1. With one or more seismic augmented inservice inspection restraint, support, or program and the requirements snubber INOPERABLE on a of Specification 3.6.H/4.6.H.

system that is required These snubbers are listed in to be OPERABLE in the Surveillance Instruction current plant condition, BF SI 4.6.H-l and within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or BF SI 4.6.H-2.

restore the INOPERABLE seismic restraint(s), l. Ins ection Grou s support(s), or snubber(s) to OPERABLE status and The snubbers may be perform an engineering categorized into two major evaluation on the attached groups based on whether the component or declare, the snubbers are accessible or attached system INOPERABLE inaccessible during reactor and follow the appropriate operation. These major Limiting Condition statement groups may be further for that system. subdivided into groups based on design, environment, or other features which may be expected to affect the operability of the snubbers within the group. Each group may be inspected independently in accordance with 4.6.H.2 throu'gh 4.6.H.9.

Schedul e and Lot Si ze The first inservice visual inspection of snubbers not previously included in these technical specifications and whose visual inspection has BFN 3.6/4.6-15 Unit 1

The following are pages requested for UNIT 2

Section ~Pa e No.

D. Reactivity Anomalies 3.3/4.3-11 E. Reactivity Control 3..3/4.3-12 F. Scram Discharge Volume . 3.3/4.3-12 3.4/4.4 Standby Liquid Control System. 3.4/4.4-1 A. Normal System Avai labi 1 i ty . 3.4/4.4-1 B. Operation with Inoperable Components 3.4/4.4-2 C. Sodium Pentaborate Solution. 3.4/4.4-3 3.5/4.5 Core and Containment Cooling Systems 3.5/4.5-1 A. Core Spray System (CSS). 3.5/4.5-1 B. Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling) 3.5/4.5-4 RHR Service Hater System and Emergency Equipment Cooling Nater System (EECWS) 3.5/4.5-9 D. Equipment Area Coolers 3.5/4.5-13 High Pressure Coolant Injection System (HPCIS) . 3.5/4.5-13 F. Reactor Core Isolation Cooling System (RCICS). 3.5/4.5-14 G. Automatic Depressurization System (ADS). 3.5/4.5-15 H. Maintenance of Filled Discharge Pipe . 3.5/4.5-17 Average Planar Linear Heat Generation Rate 3.5/4.5-18 Linear Heat Generation Rate (LHGR) . 3.5/4.5-18 K. Minimum Critical Power Ratio (MCPR). 3.5/4.5-19 L. APRM Setpoints . 3.5/4.5-20 3.6/4.6 Primary System Boundary. 3.6/4.6-1 A. Thermal and Pressurization Limitations . 3,6/4.6-1 Coolant Chemistry. 3.6/4.6-5 BFN Unit 2

J ~

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flux Tri

~Settin s 2.1.A.l.a (Cont'd)

S<(0.66W + 54%)

where:

S = Setting in percent of rated thermal power (3293 MWt)

W = Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals 34.2xl0'b/hr)

b. For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.

BFN Unit 2 1.1/2.1-2

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIt1IT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flux Tri Settin s 2.1.A.l.b. (Cont'd)

NOTE: These settings assume operation within the basic thermal hydraulic design criteria. These criteria are LHGR <13.4 kW/ft and thCPR within limits of Specification 3.5.K.

If it is determined that either of these design criteria is being violated during operation, action shall be initiated within 15 minutes to restore operation within prescribed limits.

Surveillance requirements for APRM scram setpoint are given in Specification 4.5.L.

C. The APRM Rod Block trip setting shall be:

Sg8< (0.66W + 42%)

where:

SRB Rod Block setting in percent of rated thermal power (3293 tiWt)

Loop recirculation flow rate in percent of rated (rated loop recirculati'on flow rate equals 34.2 x 10',

lb/hr)

BFN 1.1/2.1-3 Unit 2

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1.A Thermal Power Limits 2.1.A Neutron Flux Tri

~Settin s <Cont'd)

d. Fixed High Neutron Flux Scram Trip

Setting Hhen the mode switch is in the RUN position, the APRM fixed high flux scram trip setting shall be:

S<120% power.

2. Reactor Pressure <800 psia 2. APRM and IRM Trip Settings or Core Flow <10% of rated. (Startup and Hot Standby Modes).

Nhen the reactor pressure a. APRM Hhen the is <800 psia or core flow reactor mode switch is <10% of rated, the core is in the STARTUP thermal power shall not position, the APRM exceed 823 MNt ( 25% of scram shall be set at rated thermal power). less than or equal to 15% of rated power.

b. IRM The IRM scram shall be set at less than or equal to 120/125 of full scale.

BFN Unit 2 1.1/2.1-4

4

'P l

~?

NOTES FOR TABLE 3.2.C

1. The minimum number of OPERABLE channels for each trip fu -.ion is detailed for the STARTUP and RUN positions of the reacto ,ode selector switch. The SRM, IRM, and APRM (STARTUP mode), 'ocks need not be OPERABLE. in "RUN" mode, and the APRM (flow biased , od blocks need not be OPERABLE in "STARTUP" mode.

Hith the number of OPERABLE channels less than required by the minimum OPERABLE channels per trip function requirement, place at least one INOPERABLE channel in the tripped condition within one hour.

2. N is the recirculation loop flow in percent of design. Trip level setting is in percent of rated power (3293 MNt).

3. IRM downscale is bypassed when it is on its lowest range.

4. SRMs A and C downscale functions are bypassed when IRMs A, C, E, and G are above range'2. SRMs B and D downscale function is bypassed when IRMs B, D, F, and H are above range 2.

SRM detector not in startup position is bypassed when the count rate is >100 CPS or the above condition is satisfied.

5. During repair or calibration of equipment, not more than one SRM or RBM channel nor more than two APRM or IRM channels may be bypassed.

Bypassed channels are not counted as OPERABLE channels to meet the minimum OPERABLE channel requirements. Refer to section 3.10.B for SRM requirements during core alterations.

6. IRM channels A', E, C, G all in range '8 or above bypasses SRM channels A and C functions.

IRM channels B, F, D,= H all in range 8 or above bypasses SRM channels B and D functions.

7.- The following operational restraints apply to the RBM only.

a.. Both RBM channels are bypassed when reactor power is <30 percent and when a peripheral control rod is selected.

b. The RBM need not be OPERABLE in the "startup" position of the reactor mode selector switch.

c. Two RBM channels are provided and only one of these may be bypassed from the console. If the INOPERABLE channel cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the INOPERABLE channel shall be placed in the tripped condition within one hour.

d. Nith both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within one hour.

BFN 3.2/4.2-26 Unit 2

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t

NOTES FOR TABLES t

4.2.A THROUGH 4.2.H exce t t

4.2.D (Cont'd)

14. (Deleted)

15. The flow bias comparator will be tested by putting one flow unit in "Test" (producing 1/2 scram) and adjusting the test input to obtain comparator rod block. The flow bias upscale will be verified by observing a local upscale trip light during operation and verified that it will produce a rod block during the operating cycle.

16 ~ Performed during operating cycle. Portions of the logic is checked more frequently during functional tests of the functions that produce a rod block..

17. This calibration consists of removing the function from service and performing an electronic calibration of the channel.

18. Functional test is limited to the condition where secondary containment integrity is not required as specified in Sections 3.7.C.2 and 3.7.C.3.

19. Functional test is limited to the time where the.SGTS is required to meet the requirements of Section 4.7.C.l.a.

20. Calibration of the comparator requires the inputs from both recirculation loops to be interrupted, thereby removing the flow bias signal to the APRM and RBM and scramming the reactor. This calibration can only be performed during an outage.

,21. Logic test is limited to the time where actual operation of the equipment is permissible.

22. One channel of either the reactor zone or refueling zone Reactor Building Ventilation Radiation Monitoring System may be administratively bypassed for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for functional testing and calibration.

23. (Deleted)

24. This instrument check consists of comparing the thermocouple readings for all valves for consistence and for nominal expected values (not required during refueling outages).

25. During each refueling outage, all acoustic monitoring channels shall be calibrated. This calibration includes verification of accelerometer response due to mechanical excitation in the vicinity of the sensor.

BFN I In i i'.2/4.2-60

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3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 Core and Containment Coolin S stems 4.5 Core and Containment

1. Whenever the core thermal FRP/CMFLPD shall be power is > 25% of rated, the determined daily when ration of FRP/CMFLPD shall the reactor is > 25% of be > 1.0, or the APRM scram rated thermal power.

and rod block setpoint equations listed in Sections F 1.A and 2.1.B shall be multiplied by FRP/CMFLPD as follows:

S< (0.66W + 54%)

CMFLPD Sgg< (0.66W + 42%) ('" )

CMFLPD

2. When it is determined that 3.5.L.1 is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to correct the condition.

3. If 3.5.L.l and 3.5.L.2 cannot be met, the reactor power shall be reduced to

< 25% of rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

BFN 3.5/4.5-20 Unit 2

3.5 BASES (Cont'd) 3.5.M. References

1. Loss-of-Coolant Accident Analysis for Brogans Ferry Nuclear Plant Unit 2, NEDO 24088-1 and Addenda.

2. "BNR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

3. Generic Reload Fuel Application, Licensing Topical Report, NEDE 24011-P-A and Addenda.

BFN 3.5/4.5-32 Unit 2

3.6/4.6 PRIMARY SYSTEM BOUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.H. Seismic Restraints Su orts, 4.6.H. Seismic Restraints Su orts, and Snubbers and Snubbers During all modes of operation The surveillance requirements all seismic restraints, of paragraph 4.6.G are the snubbers, and supports shall only requirements that apply be OPERABLE except as noted to any seismic restraint or in 3.6.H.l. All safety-related support other than snubbers.

snubbers are listed in Surveillance Instructions Each safety-related snubber BF SI 4.6.H-l and BF SI 4.6.H-2. shall be demonstrated OPERABLE by performance of the following

1. Hith one or more seismic augmented inservice inspection restraint, support, or snubber program and the requirements INOPERABLE on a system that is of Specification 3.6.H/4.6.H.

required to be OPERABLE in the These snubbers are listed in current plant condition, within Surveillance Instructions 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore the BF SI 4.6.H-l and INOPERABLE seismic restraint(s), BF SI 4.6.H-2.

support(s), or snubber(s) to OPERABLE status and perform an Ins ection Grou s engineering evaluation on the attached component or declare The snubbers may be the attached system INOPERABLE categorized into two major and follow the appropriate groups based on whether the Limiting Condition statement snubbers are accessible or for that system. inaccessible during reactor operation. These major groups may be further subdivided into groups based on design, environment, or other features which may be expected to affect the operability of the snubbers within the group. Each group may be inspected independently in accordance with 4.6.H.2 through 4.6.H.9.

Visual Ins ection Schedule and Lot Size The first inservice visual inspection of snubbers not previously .included in these technical. specifications and whose visual inspection has BFN 3.6/4.6-15 Unit 2

The following are pages requested for Unit 3 Section Pacae No.

D. Reactivity Anomalies 3. 3/4. 3-11 E. Reactivity Control". 3.3/4.3-12 F. Scram Discharge Volume . 3.3/4.3-12 3.4/4.4 Standby Liquid Control System. 3.4/4.4-1 A. Normal System Availability .=. 3.4/4.4-1 B. Operation with Inoperable Components 3.4/4.4-2 C. Sodium Pentaborate Solution. 3.4/4.4-3 3.5/4.5 Core and Containment Cooling Systems . 3.5/4.5-1 A. Core Spray System (CSS). 3.5/4.5-1 B. Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling) 3.5/4.5-4 C. RHR Service Water System and Emergency Equipment Cooling Hater System (EECWS) 3.5/4.5-9 D. Equipment Area Coolers . 3.5/4.5-13 E. High Pressure Coolant Injection System (HPCIS) . 3.5/4.5-13 F. Reactor Core Isolation Cooling System (RCICS). 3.5/4.5-14 G. Automatic Depressurization System (ADS). 3.5/4.5-15 H. Maintenance of Filled Discharge Pipe . 3.5/4.5-17 I. Average Planar Linear Heat Generation Rate 3-5/4.5-18 J. Linear Heat Generation Rate (LHGR) 3.5/4.5-18 1

K. Minimum Critical Power Ratio (MCPR). 3.5/4.5-19 L. APRM Setpoints . 3.5/4.5-20 3.6/4.6 Primary System Boundary. 3.6/4.6-1 A. Thermal and Pressurization Limitations . 3.6/4.6-1 B. Coolant Chemistry. 3.6/4.6-5 BFN-Unit 3

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSl '1 SETTING 2.1.A Neutron Flux Tri

~Settin n 2.1.A.l.a (Cont'd)

S<(0.66N + S4%)

where:

S = Setting in percent of rated thermal power (3293 MHt)

N = Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals 34.2xl0'b/hr)

b. For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.

8FN Unit 3 1.1/2.1-2

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flux Tri Settin s 2.1.A.l.b (Cont'd)

NOTE: These settings assume operation wi thin the basic thermal hydraulic design criteria. These criteria are LHGR <13.4 kW/ft and MCPR within limits of Specification 3.5.K.

If it is determined that either of these design criteria is being violated during operation, action shall be initiated within 15 minutes to restore operation within the prescribed limits.

Surveillance requirements for APRM scram setpoint are given in Specification 4.5.L.

c ~ The APRM Rod Block trip setting shall be:

Sg8 <(0.66W + 42%)

where:

SR8 Rod Block setting in percent of rated thermal power (3293 MWt)

Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate 'equals 34.2 x 10'b/hr)

BFN 1. 1/2. 1-3 Unit 3

1.1/2.1 FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1.A Thermal Power Limits 2.1.A Neutron Flux Tri Settin s

d. Fixed High Neutron Flux Scram Trip Setting When the mode switch is in the RUN position, the APRM fixed high flux scram trip setting shall be:

S<120% power.

2. Reactor Pressure <800 psia 2 . APRM and IRM Trip Settings or Core Flow <10% of rated. (Startup and Hot Standby Modes).

lJhen the reactor pressure a. APRM When the is <800 psia or core flow reactor mode switch is <10% of rated, the core is in the STARTUP thermal power shall not position, the APRM exceed 823 MWt ('25% of scram shall be set at rated thermal power). less than or equal to 15% of rated power.

b.

IRM The IRM scram shal 1 be set at less than or equal to 120/125 of full scale.

BFN Unit 3 1.1/2.1-4

l4

,h l

NOTES FOR TABLE 3.2.C 0

1. The minimum number of operable channels for each trip f .ion 1 s detailed for the startup and run positions of the react< lode selector switch. The SRM, IRM, and APRM (startup mode), ocks need not be operable in "run" mode, and the APRM (flow biasec od blocks need not be operable in "startup" mode.

With the number of OPERABLE channels less than required by the minimum OPERABLE channels per trip function requirement, place at least one inoperable channel in the tripped condition within one hour.

2. W is the recirculation loop flow in percent of design. Trip level setting is in percent of rated power (3293 MWt).

See Specification 2.1 for APRM control rod block setpoint

3. IRM downscale is bypassed when it is on its lowest range

4. SRMs A and C downscale functions are bypassed when IRMs A, C, E, and G are above range 2. SRMs B and D downscale function is bypassed when IRMs B, D, F, and H are above range 2.

SRM detector not in.startup position is bypassed when the count rate is >100 counts per second or the above condition is satisfied.

5. During repair or calibration of equipment, not more than one SRM or RBM channel nor more than two APRM or IRM channels may be bypassed.

Bypassed channels are not counted as operable channels to meet the minimum operable channel requirements. Refer to section 3.10.B for SRM requirements during core alterations.

6. IRM channels A, E, C, G all in range 8 or above bypasses SRM channels A and C functions.

IRM channels B, F, D, H all in range 8 or above bypasses SRM channels B and D functions.

7. ,The following operational restraints apply to the RBM only.

a ~ Both RBM channels are bypassed when reactor power is <30 percent and when a peripheral control rod is selected.

b. The RBM need not be operable in the "startup" position of the reactor mode selector switch'wo RBM channels are provided and only one of these may be bypassed from the console. If the inoperable channel cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel shall be placed in the tripped condition within one hour.

d. With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition wi'thin one hour.

BFN 3.2/4.2-25 Unit 3

I ~

,f A,

NOTES FOR TABLES 4.2.A THROUGH 4.2.H exce t 4.2.D (Continued)

14. (Deleted)

15. The flow bias comparator will be tested by putting one flow u. t in

'"Test" (producing 1/2 scram) and adjusting the test input to ~ tain comparator rod block. The flow bias upscale will be verified observing a local upscale trip light during operation and veri -ied that it will produce a rod block during the operating cycle.

16. Performed during operating cycle. Portions of the logic is checked more frequently during functional tests of the functions that produce a rod block..

17. This calibration consists of removing the function from service and performing an electronic calibration of the channel.

18. Functional test is limited to the condition where secondary containment integrity is not required as specified in Sections 3.7.C.2 and 3.7.C.3.

19. Functional test is limited to the time where the SGTS is required to meet the requirements of Section 4.7 C.l.a.

~

20. Calibration of the comparator requires the inputs from both recirculation loops to be interrupted, thereby removing the flow bias signal to the APRM and RBM and scramming the reactor. This calibration can only be performed during an outage.

.'1. Logic test's limited to the time where actual operation of the equipment is permissible.

22. One channel of either the reactor zone or refueling zone Reactor Building Ventilation Radiation Monitoring System may be administratively bypassed for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for functional testing and calibration.

23. (Deleted)

24. This instrument check consists of comparing the thermocouple readings for all valves for consistence and for nominal expected values (not required during refueling outages).

25. During each refueling outage, all acoustic monitoring channels shall be calibrated. This calibration includes verification of accelerometer response due to mechanical excitation in the vicinity of the sensor.

BFN 3.2/4.2-59 line 0

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 Core and Containment Coolin S stems 4.5 Core and Containment Coolin

~sstems

1. Whenever the core thermal FRP/CMFLPD shall be power is > 25% of rated, the determined daily when ratio of FRP/CMFLPD shall the reactor is > 25% of be > 1.0, or the APRM scram rated thermal power.

and rod block setpoint equations listed in Sections 2.1.A and 2.1.B shall be multi plied by FRP/CMFLPD as follows:

S<<O.66W ~ 54%)

'

CMFLPD SRB< (0.66W + 42%) ( )

CMFLPD

2. When it is determined that 3.5.L.1 is not being met, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to correct the condition.

3. If 3.5.L.1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to

< 25% of rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

BFN 3.5/4.5-20 Unit 3

3.5 BASES (Cont'd) 3.5.M References

1. Loss-of-Coolant Accident Analysis for Brogans Ferry Nuclear Plant Unit 3, NEDO-24194A and'ddenda.

2. "BHR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

3. Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.

BFN 3.5/4.5-35 Unit 3

I P

3. 6/4. 6 PRIMARY SYSTEM BOUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.H. Seismic Restraints Su orts, 4.6.H. Seismic Restraints Su orts, and Snubbers and Snubbers During all modes of operation The surveillance requirements all seismic restraints, of paragraph 4.6.G are the snubbers, and supports shall, only requirements that apply be OPERABLE except as noted to any seismic restraint or in 3.6.H.1. All safety- support other than snubbers.

related snubbers are listed in Surveillance Instruction Each safety-related snubber BF SI 4.6.H-l and BF SI 4.6.H-2. shall be demonstrated OPERABLE by performance of the following

1. Hi th one or more sei smi c augmented inservice inspection restraint, support, or program and the requirements snubber INOPERABLE on a of Specification 3.6.H/4.6.H.

system that is required These snubbers are listed in to be OPERABLE in the Surveillance Instruction current plant condition, BF SI 4.6.H-l and within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or BF SI 4.6.H-2.

restore the INOPERABLE seismic restraint(s), l. Ins ection Grou s support(s), or snubber(s) to OPERABLE status and The snubbers may be perform an engineering categorized into two major evaluation on the attached groups based on whether the component or declare the snubbers are accessible or attached system INOPERABLE inaccessible during reactor and follow the appropriate operation. These major Limiting Condition statement groups may be further for that system. subdivided into groups based on design, environment, or other features which may be expected to affect the operability of the snubbers within the group. Each group may be inspected independently. in accordance with 4.6.H.2 throUgh 4.6.H.9.

Visual Ins ection Schedule and Lot Size The first inservice visual inspection of snubbers not previously included in these technical specifications and whose visual inspection has BFN 3.6/4.6-15 Unit 3 j4 f J 4 0 $ y~

TABLE 4.2.A SURVEILLANCE REI0UIREHENTS FOR PRIHARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Instrument Channel - once/3 months (29) once/operating cycle None Hain Steam Line Tunnel High Temperature Instrument Channel- ( 1) (22) once/3 months once/day (8)

Reactor Building Ventilation High Radiation Reactor Zone Instrument Channel - (1) (22) once/3 Honths once/day (8)

Reactor Building Ventilation High Radiation - Refueling Zone Instrument Channel - (4) (9) N/A SGTS Train A Heaters Instrument Channel - (4) (9) N/A SGTS Train B Heaters Instrument Channel - (4) (9) N/A SGTS Train C Heaters Reactor Building Isolation once/operating cycle N/A Timer (refueling floor)

Reactor Building Isolation once/operating cycle N/A Timer (reactor zone)

BFN-Unit 1

TABLE 4.2.A SURVEILLANCE REQUIREHENTS FOR PRIHARY CONTAINHEWT AND REACTOR BUILDING ISOLATION IWSTRUHEWTATIOW Instrument Channel - (29) Once/operating cycle None Hain Steam Line Tunnel High Temperature Instrument Channel- (1) (22) Once/3 months Once/day (8)

Reactor. Building Ventilation High Radiation - Reactor Zone

-Instrument Channel - (1) (22) Once/3 Honths Once/day (8)

Reactor Building Ventilation High Radiation - Refueling Zone Instrument Channel - (4) (9) W/A SGTS Train A Heaters Instrument Channel - (4) (9) N/A SGTS Train B Heaters Instrument Channel - (4) (9) W/A SGTS Train C Heaters Reactor Building Isolation (4) Once/operating cycle N/A Timer (refueling floor)

Reactor Building Isolation Once/operating cycle N/A Timer (reactor zone) 1 BFN-Unit 2

TABLE 3.2.A (Continued)

PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION Minimum No.

Instrument Channels Operable Instrument Channel- 3 times normal rated l. Above trip setting High Radiation Hain Steam full power background ( 13) initiates Hain Steam Line Line Tunnel (6) Isolation Instrument Channel- 2 825 psig (4) 1. Below trip setting Low Pressure Hain Steam initiates Main Steam Line Line Isolation 2(3) Instrument Channel- S 140K of rated steam flow l. Above trip setting Nigh Flow Hain Steam Line initiates Hain Steam Line Isolation 2(12) Instrument Channel- S 2000F 1. Above trip setting Hain Steam Line Tunnel initiates Main Steam High Temperature Line Isolation.

2(14) Instrument Channel- 160 - 180~F l. Above trip setting Reactor Water Cleanup initiates Isolation System Floor Drain of Reactor Water High Temperature Cleanup Line from Reactor and Reactor Water Return Line.

Instrument Channel- 160 - 180~F l. Same as above Reactor Water Cleanup System Space High Temperature 1(9) Instrument Channel- Z 100 mr/hr or downscale l. 1 upscale or 2 downscale will Reactor Building a. Initiate SGTS Ventilation Nigh b. Isolate reactor zone and Radiation - Reactor Zone refueling floor.

c. Close atmosphere control system.

BFN-Unit 3

TABLE 3.2.A (Continued)

PRIHARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION Hinimum No.

Instrument Channels Operable r

1(9) Instrument Channel- S 100 mr/hr or downscale l. 1 upscale or 2 downscale will Reactor Building a. Initiate SGTS Ventilation High b. Isolate refueling floor.

Radiation Refueling Zone c. Close atmosphere control system 2(7) (8) Instrument Channel R.H. Heaters S 2000 cfm H and 1. Below 2000 cfm, trip setting SGTS Flow - Train A (A or F) R.H. Heaters will turn on.

Heater 2(7) (8) Instrument Channel R.H. Heaters g 2000 cfm H and 1. Below 2000 cfm, trip setting SGTS Flow Train B (A or F) R.H. Heaters will turn on.

Heater 2(7) (8) Instrument Channel R.H. Heaters S 2000 cfm H and 1. Below 2000 cfm, trip setting SGTS Flow - Train C (A or F) R.H. Heaters will turn on.

Heater Reactor Building Isolation 0 S t Z 2 secs. H or F 1. Below trip setting prevents Timer (refueling floor) spurious trips and system perturbations from initiating isolation Reactor Building Isolation 0 g t g 2 secs. G or A 1. Below trip setting prevents Timer (reactor zone) or H spurious trips and system perturbations from initiating isolation 2(10) Group 1 (Initiating) Logic N/A 1. Refer to Table 3.7.A for list of valves.

BFt(+nit 3

TABLE 4.2.A SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION r i r n Instrument Channel- once/3 months (27) once/operating cycle None Main Steam Line Tunnel High Temperature Instrument Channel- (1) (22) once/3 months once/day (8)

Reactor Building Ventilation High Radiation - Reactor Zone Instrument Channel- ('I) (22) once/3 Honths once/day (8)

- Reactor Building Ventilation High Radiation - Refueling Zone Instrument Channel- (4) N/A SGTS Train A Heaters Instrument Channel- (4) (9) N/A SGTS Train B Heaters Instrument Channel- (4) N/A SGTS Train C Heaters Reactor Building Isolation (4) once/operating cycle N/A Timer (refueling floor)

Reactor Building Isolation (4) once/operating cycle N/A Timer (reactor zone)

BFN-Unit 3

Enclosure 2 Description and Justification Browns ferry Nuclear Plant (BFN)

Units 1, 2, and 3 Descri tion of Chan e The technical specifications are being revised to delete section 3.5.M, Reporting Requirements, the bases for it, and its reference in the index.

2. Note 7.d for table 3.2.C is being revised to clarify an ambiquity and provide an action whenboth Rod Block Monitor (RBM) channels are inoperable.

3. The technical specifications are being revised to make the Limiting Condition of Operation (LCO), 3.6.H.1, reflect the correct Surveillance Instruction (SI) number for the safety related snubber list.

Section 2.1.A.l.c is being revised to show the correct reference of specification 4.5.L for the Surveillance Requirement (SR) for APRM setpoints.

5. Table 4.2.A note (14) is deleted.

Reason for Chan e Section 3.5.M, Reporting Requirements, is to be deleted since it is redundant to 10 CFR 50.73 and requirements in the Administrative Controls section of the technical specifications..

2. Table 3.2.C requires both channels of the RBM to be operable except for its reference to note 7 which has four parts. The current note 7.d immediately prevents control rod movement if the conditions for the table are not met. 'However, note 7.c allows one channel to be bypassed and inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without having to prevent control rod movement. Note 7.d is being revised since it currently causes a with note 7.c. The new wording will be taken from Standard 'onflict Technical Specifications (STS). It will not produce any conflict and will address the possibility of both RBM channels being inoperable which i's not specifically addressed at present.

Recent amendments (Nos. 128, 123, and 99 for units 1, 2, and 3 respectively) revised the SR 4.6.H to reference the correct SI containing snubber lists. However, the corresponding LCO reference which was not changed should also be corrected.

4, The revision to the reference in section 2.1.A.l.c for APRM setpoints is needed to correct an error in units 2 and 5 technical-specifications. This same error was corrected'n the unit technical1 specific'ations by amendment No. 128.

I

5. There is no relationship between the surveillance testing required by table 4.2.A for the reactor zone and refueling zone radiation monitor instrumentation channels and either of the surveillance requirements referenced in footnote (14). Therefore, this footnote which ties performance of these surveillance requirements together should be deleted.

Justification for Chan e The proposed amendment to the technical specifications for units 1, 2, and 3 is justified on the basis that it will correct and/or clarify the current technical specification revision. Each change included in this package is proposed to either correct an error or to achieve consistency throughout the technical specifications. More specific reasoning is given below for each change.

1. Section 3.5.M requires that a written report be made within 30 days if any of the limiting values in specifications 3 .5 . I, J, K, or L.3 are exceeded and the remedial action is taken. The remedial action for specifications 3 . 5. I, J, and K is to bring the reactor to cold shutdown with 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> .

The remedial action of 3.S'IL.3 is to reduce thermal power to < 25 percent of rated within four hours. If 3.5.M is deleted, the technical specifications will continue to require reporting under the requirements of 10 CFR 50.73 which is referenced in section 6.6.l.a. The requirements of 10 CFR 50.73(a)(2)(i)(A) and (B) are that a Licensee Event Report (LER) be submitted .within 30 days of any nuclear plant shutdown required by technical specifications, or any operation or condition prohibited by the plant's technical specifictions ~ Therefore, this change will not result in a significant decrease in technical specification reporting requirements.

2. The note 7.d regarding RBM requirements in table 3.2.C should be changed since it currently presents an apparent conflict with note 7.c. The current note 7.d is also confusing since it is not apparent when the note is supposed to apply. The proposed revision to note 7.d is taken from STS and does not conflict with any other requirements. Furthermore, it clarifies the action to be taken in the event that both RBM channels fail. Since this change would remove an apparent conflict, clarify required actions, and is consistent with the requirements of STS, TVA believes that safety will be enhanced.

3. Correcting the reference to the SI that lists safety related snubbers is an administrative change that in no way affects technical specification requirements or operations and will not have an adverse effect on nuclear safety.

4. Correcting the reference describing where to find APRM setpoint requirements is an administrative correction of an error and will not change any technical specification requirements or operations and will not have an adverse effect on nuclear safety. This change was previously approved for unit 1 by amendment No. 128.

5.'his change involves three separate surveillance testing requirements.

The first requirement is to functionally test the reactor building isolation trip caused by high radiation in the reactor building refueling zone and reactor zone.'his is an instrumentation functional test required once per month by table 4.2.A.

The second test is performed once/year per SR 4.7.B.l.a. Its purpose is to show that the pressure drop across the combined HEPA filters and charcoal adsorber banks of the Standby Gas Treatment System (SGTS) is less than 6 inches of water at a flow of 9000 cfm (+10%).

The third test is performed before refueling and is to verify the capability of the secondary containment to maintain 1/4-inch of water vacuum with a system leakage rate of not more than 12000 cfm.

The only relation between these tests is that the high radiation trip signal will start the SGTS and isolate the secondary containment. Both of these functions are already tested as part of the instrument functional test as required by the technical specification definition of Instrument Functional test. Remote manual initiation is the method actually used in the other surveillance instructions. Finally, since each test has a different frequency requirement, it is not practical to perform the tests together.

For 'the reasons stated above, TVA has concluded that none of these proposed TS changes will reduce the margin -of nuclear safety.

)>>

'I

Enclosure 3 Determination of No Significant Hazards Consideration Browns Ferry Nuclear Plant (BFN)

Units 1, 2, and 3 Descri tion of Amendment Re uest The proposed amendment would modify the technical specifications of BFN units 1, 2, and 3 to incorporate the following corrections and clarifications.

1. Delete section 3.5.,M on reporting requirements for core thermal limits since it is redundant to reporting requirements specified elsewhere in technical specifications and 10 CFR 50.73.

2. Revise note 7.d of table 3.2.C since it is in conflict with note 7.c of the same table. These notes deal with the requirements for the Rod Block Monitor (RBM) and the revised note wi 1 1 be consi stent with Standard Technical Specifications (STS). Note 7.c allows that one of the two RBM channels may be bypassed from the console and that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> can be used to restore an inoperable channel before placing it in the tripped condition and thereby preventing control rod withdrawal. The current note 7.d, without the provisions of 7.c, requires that control rod withdrawal be immediately stopped if either RBM channel is inoperable.

The new note taken from STS would require that one channel be placed in the tripped condition within one hour if both RBM channels are

>inoperable, thus removing any conflict.

3. Change the references to the lists of safety related snubbers from "Surveillance Instruction BF SI 4.6.H" to "Surveillance Instruction BF SI 4.6.H-l and 2." This change would reflect reissued plant procedures.

4. Correct a reference to the surveillance requirement in the unit 2 and 3 Limiting Safety System Setting specification for'he Average Power Range Monitor (APRM). The present references to section 4.5.8 which specify surveillance requirements for the Reactor Protection System Power Monitoring System (RPSPMS) would be replaced by a reference to section 4.5.L whi.ch specifies surveillance-requirements for the Reactor Protection System (RPS) and is the correct reference.

5. Change the technical specifications to delete the erroneous note (14) of table 4.2.A. It infers that the upscale functional test of the refuel and reactor zone radiation monitors is conducted during execution of two other surveillance tests; however, no apparent relationship exists.

Basis for ro osed No Si nificant Hazards Consideration Determination:

The Commission 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 considerations 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, (2) create the possibility of a new or different kind of accident form an accident previously evaluated, or (3) involve a significant reduction in a margin of safety. Except for Item Nos. 1 and 2, the proposed amendments correct errors or eliminate inconsistencies. For Item No.l, the proposed change will only remove a requirement that is redundant to the reporting requirements in section 6 of the technical specifications and in 10 CFR 50.73. Furthermore, because no operability or surveillance requirements for systems, structures, or components used to terminate or mitigate accidents would be reduced, the amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.

Hhile Item-No. 2 removes an inconsistency, it also adopts the requirements of STS for the RBM when both channels are inoperable. Therefore, it has been evaluated and determined not to cause a significant reduction in a safety margin. Finally, since this correction will not change any surveillance requirements or modes of operation, it will not involve a significant increase in the,probability or. consequences, of an accident or create the possibility of a new kind of accident.

Item Nos. 3, 4, and 5 are administrative in nature in that only clarifications and corrections are made which do not affect the actual TS requirements.

These technical specification changes will not eliminate or modify any protective functions, surveillance requirements, nor permit any new operational conditions . Therefore, they do not create the possibility of a new kind .of accident or significantly increase the probability or consequences of an accident. Because the changes will clarify requirements and make corrections, the margin of safety will not be reduced.

Since the application for amendment involves proposed changes that by the criteria for which no significant hazards considerationare'ncompassed TVA has made a proposed determination that the application involves 'xists, no significant hazards consideration.

IV 1, TO: W. H. Hannum, Chairman, Nuclear Safety Review Board, BR 1N "77B-C FROM  : M. J. May, Manager of Site Licensing, Browns Ferry Nuclear Plant DATE

SUBJECT:

BROWNS FERRY NUCLEAR PLANT (BFN) TECHNICAL SPECIFICATION 228 MISCELLANEOUS NOTE CORRECTIONS REVISIONS TO DESCRIPTION AND JUSTIFICATION AND DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION

References:

(1) Memorandum from R. W. Cantrell to S. A. White dated October 1, 1986, "Nuclear Safety Review Board (NSRB) Disposition of Proposed Technical Specification Change" (L42 861002 802)

(2) Letter from R. Gridley to U.S. Nuclear Regulatory Commission (NRC) dated April 03, 1987, "Browns Ferry Nuclear Plant (BFN) TVA BFN TS 228 (L44 870403 803)

Browns Ferry Technical Specification 228 was approved by the NSRB by reference 1 and sent to NRC by reference 2. One of the changes approved by NSRB moved a footnote from table 4.2.A to table 3.2.A in response to,a Resident Inspector's concern. It has subsequently been decided that the Resident Inspector's concern would be better resolved through a procedural change to the surveillance test; this does not require a technical specification change. The procedural change will be tracked under the original Inspector Followup Item and annotated to ensure it is not subsequently deleted. A resubmittal to NRC of the technical specification change is required to delete this change from the original request for revision. The deletion was requested by the Browns Ferry Nuclear Plant NRC Project Manager. Additional nontechnical changes were made to the enclosures .to be more explicit and easy to read.

Enclosed is the revised Description and Justification (enclosure 2) and a No Significant Hazards Consideration (enclosure 3). These enclosures are designed to be substituted for the originals on a page by page basis. The change to enclosure 1 (the technical specification pages) deletes some of the change pages.

M. J. May PPC:JEM'SJL Enclosures cc (Enclosures):

'.

RIMS, MR 4N Gridley,.LP 72A-C 5N 157B-C This was prepared principally by J. E. McCarthy.

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