ML18036B191

Proposed Administrative Ts,Reflecting Deletion of Unit 2 Cycle 6 Only Requirement,Correction of Administrative Errors in Previous TS & Discrepancies Between Spec Bases & FSAR

ML18036B191
Person / Time
Site:	Browns Ferry
Issue date:	03/18/1993
From:	TENNESSEE VALLEY AUTHORITY
To:
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ML18036B190	List: ML18036B189 ML18036B191
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NUDOCS 9303230034
Download: ML18036B191 (86)

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ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE BROMNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 (TVA BFN TECHNICAL SPECIFICATION NO. 331) 9303230034 930318 PDR ADQCK 05000259 .

P PDR

e PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 1 (TVA BFN TECHNICAL SPECIFICATION NO. 331)

UNIT 1 EFFECTIVE PAGE LIST RENOVE INSERT ii ii V

V 1.0-4 1.0-4 1.1/2.1-13 1.1/2.1-13 1.1/2.1-16 1.1/2.1-16 3.1/4.1-15 3.1/4.1-15 3.3/4.3-12 3.3/4.3-12 3,4/4.4-4 3.4/4.4-4 3.6/4.6-6 3.6/4.6-6 3.6/4.6-10 3.6/4.6-10 3.7/4.7-3 3.7/4.7-3 3.7/4.7-23 3.7/4.7-23 3.9/4.9-20 3.9/4.9-20 6.0-18 6.0-18 6.0-26a 6.0-26a

P 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 . . . . . 3.4/4.4-3 C. Sodium Pentaborate Solution. 3.4/4.4-3 D. Standby Liquid Control System Requirements 3.4/4.4-4 3 '/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 and Emergency Equipment Cooling Water Systems (EECWS). 3.5/4.5-9 D. Equipment Area Coolers 3.5/4.5-13 E. High Pressure Coolant Injection System (HPCIS) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ s ~ 3.5/4.5-13 F. Reactor Core Isolation Cooling System (RCICS) ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ 3.5/4.5-14 G. Automatic Depressurization System (ADS). . . . 3.5/4.5-16 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 A. Thermal and Pressurization Limitations 3.6/4.6-1 B. Coolant Chemistry. 3.6/4.6-5 C. Coolant Leakage. 3.6/4.6-9 D. Relief Valves. 3.6/4.6-10 BFH Unit 1

~ \

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6. 0-1

$,2 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ 6 ~0 1 6.2.1 Offsite and Onsite Organizations......................... 6. 0-1 6.2.2 P lant Staff.............................................. 6.0-2

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6~0 5 32 ggg o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ s ~ ~ ~ ~ ~ 6. 0-5

~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~0 5 6.5 ~ 1 Plant Operations Review Committee (PORC)................. 6.0-5 6.5.2 Nuclear Safety Review Board (NSRB)....................... 6.0-11 6.5.3 Technical Review and Approval of Procedures.............. 6.0-17

( Deleted)................................................ 6.0-,18

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~ 0 19

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-20 6.8.1 Procedures ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ a ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-20 6-8.2 Drills................................................... 6.0-21 6.8.3 Radiation Control Procedures............................. 6.0-22 6.8.4 Quality Assurance Procedures Effluent and Environmental Monitoring........................... 6.0-23

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-24 6.9.1 Routine Reports.......................................... 6.0-24 Startup Reports.......................................... 6.0-24 Annual Operating Report.................................. 6.0-25 Monthly Operating Report................................. 6.0-26 Reportable Events........................................ 6.0-26 Radioactive Effluent Release Report...................... 6.0-26 Source Tests ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-26 6.9.2 S pecxal Reports........................................... 6.0-27 N ~ I~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-29

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ '6 ~ 0 32 2 L. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~ 0 32

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~0 33 BPN v Unit 1

1.0 M. t The reactor mode switch position determines the Mode of Operation of the reactor when there is fuel in the reactor vessel, except that the Mode of Operation may remain unchanged when the reactor mode switch is temporarily moved to another position as permitted by the notes. When there is no fuel in the reactor vessel, the reactor is considered not to be in any Mode of Operation or operational condition. The reactor mode switch may then be in any position or may be inoperable. I t t n The reactor is in the STARTUP/HOT STANDBY MODE when the reactor mode switch is in the "STARTUP/HOT STANDBY" position. This is often referred to as just the STARTUP MODE.

2~ Rg~~ The reactor is in the Run Mode when the reactor mode switch is in the "Run" position.

3~ The reactor is in the Shutdown Mode when the reefer mode switch is in the "Shutdown" position.

• "'"'"' ""'"'I"I (2)(3)(

4.

mode switch is in the "Refuel" position. '"'eactor The reactor mode switch may be placed in any position to perform required tests or maintenance authorized by the shift operations supervisor, provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

) The reactor mode switch may be placed in the "Refuel" position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.10.A.5 provided that reactor coolant temperature is equal to or less than 212'.

) The reactor mode switch may be placed in the "Refuel" position while a single control rod is being recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE.

(4) The reactor mode switch may be placed in the "Startup/Hot Standby" position and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.

M5 1.0-4 Unit 1

2.1 ~QgQ (Cont'd) ~

Analyses of the limiting transients show that no scram adjustment is required to assure MCPR ) 1.07 when the transient is initiated from MCPR limits specified in Specification 3.5.k.

2~

For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRN system would be more than adequate to assure a scram before the power could exceed the safety limit. The 15 percent APRN scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater than 850 psig.

3.

The IRM System consists of 8 chambers, 4 in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size.

The IRN scram setting of 120 divisions is active in each range of the IRN. For example, if the instrument were on range 1, the scram setting would be at 120 divisions for that range; likewise if the instrument was on range 5, the scram setting would be 120 divisions on that range.

BEN 1.1/2.1-13 Unit 1

2. 1 MgQ (Cont 'd)

F. (Deleted)

G. 6( H.

The low pressure isolation of the main steam lines at 825 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch be in the STARTUP position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase.

I.J.& K. t w t v t t t t These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures. The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.

L- Refmxucca

1. Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 1 (applicable cycle-specific document).

2. GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (applicable amendment specified in the CORE OPERATING LIMITS REPORT).

3. "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactor," NED0-24154-P, October 1978.

4. Letter from R. H. Buchholz (GE) to P. S- Check (NRC), "Response to NRC Request For Information On ODXH Computer Model,"

September 5, 1980.

BFN 1.1/2.1-16 Unit 1

3.1 /ASIA (Cont'd) ~

Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel. This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration. Additional IRM channels have also been provided to allow for bypassing of one such channel. The bases for the scram setting for the IRM, APRM, high reactor pressure, reactor low water level, MSIV closure, turbine control valve fast closure, and turbine stop valve closure are discussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switches) for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of .coolant accident and to prevent return to criticality. This instrumentation is a backup to the reactor vessel water level instrumentation.

High radiation levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds three times normal background. The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent release of radioactive material to the turbine. An alarm is initiated whenever the radiation level exceeds 1.5 times normal background to alert the operator to possible serious radioactivity spikes due to abnormal core behavior.

The air ejector off-gas monitors serve to back up the main steam line monitors to provide further assurance against release of radioactive to site environs by isolating the main condenser off-gas line

'aterials to the main stack.

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 Unit 1

333.E 4.3.E v I- If Specifications 3.3.C and .D Surveillance requirements areL above cannot be met, an orderly as specified in 4.3.C and .D shutdown shall be initiated and above.

D'.3.F the reactor shall be in the SHUTDOWN CONDITION within 24 hours.

1. The scram discharge volume l.a. The scram discharge drain and vent valves shall volume drain and vent be OPERABLE any time that valves shall be verified the reactor protection open PRIOR TO system is required to be STARTUP and monthly OPERABLE except as thereafter. The valves specified in 3.3.F.2. may be closed intermittently for testing not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in any 24-hour period during operation.

l.b. The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with Specification 1.0.NM.

2. In the event any SDV drain 2~ When it is determined or vent valve becomes that any SDV drain or inoperable, REACTOR POWER vent valve is inoperable, OPERATION may continue the redundant drain or provided the redundant vent valve shall be drain or vent valve is demonstrated OPERABLE OPERABLE. immediately and weekly thereafter.

3. If redundant drain or vent 3. No additional valves become inoperable, surveillance required.

the reactor shall be in HOT STANDBX CONDITION within 24 hours.

BFN 3.3/4.3-12 Unit 1

.4

a. Calculate the enrich-ment within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. Verify by analysis within 30 days.

s 4.4.D khan~ imisLGa~tnl t

The Standby Liquid Control Verify that the equation System conditions must satisfy given in Specification the following equation. 3.4.D is satisfied at least y 1 once per month and within (13 wt.X)(86 gpm)(19.8 atom%) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> anytime water or boron is added to the where, solution.

sodium pentaborate solution concentration (weight percent)

Determined by the most recent performance of the surveillance instruction required by Specification 4.4.C.2.

Q = pump flow rate (gpm)

Determined by the most recent performance of the surveillance instruction required by Specification 4.4.A.2.b.

E = Boron-10 enrichment (atom percent Boron-10)

Determined by the most recent performance of the surveillance instruction required by Specification 4.4.C.4.

If Specification 3.4.A through No additional 3.4.D cannot be met, make at surveillance required.

least one subsystem OPERABLE within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the reactor shall be placed in a SHUTDOWN hours'.

CONDITION with all operable control rods fully inserted within the following 12 BPN 3.4/4.4-4 Unit 1

N 3.6.B 4.6.B

3. At steaming rates 3. Whenever the reactor greater than 100,000 is operating (including lb/hr, the reactor HOT STANDBY CONDITION) water quality may measurements of reactor exceed Specification water quality shall be 3.6.B.2 only for the performed according to time limits specified the following schedule'.

below. Exceeding these time limits or the following a. Chloride ion content maximum quality limits shally shall be measured be cause for placing at least once the reactor in the every 96 hours.

COLD SHUTDOWN CONDITION. b. Chlorid'e ion content shall be

a. Conductivity measured at least time above every 8 hours 1 Pmho/cm at 25'C whenever reactor 2 weeks/year. conductivity is Maximum Limit >1.0 pmho/cm 10 Pmho/cm at 25'C at 25'C.

c. A sample of primary

b. Chloride coolant shall be concentration time measured for pH at above 0.2 ppm least once every 8 2 weeks/year. hours whenever the Maximum Limit reactor coolant 0.5 ppm. conductivity is >1.0 Pmho/cm at 25'C.

c. The reactor shall be placed in the SHUTDOWN CONDITION if pH (5.6 or

>8.6 for a 24-hour period.

BEN 3.6/4.6-6 Unit 1

4 3.6.C t k 4.6.C nt

2. Anytime irradiated fuel is in 2. With the air sampling the reactor vessel and reactor system inoperable, grab coolant temperature is above samples shall be 212'F, both the sump and air obtained and analyzed sampling systems shall be at least once every 24 OPERABLE. From and after the hours.

date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the sump system or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the air sampling system.

The air sampling system may be removed from service for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for calibration, function testing, and maintenance without providing a temporary monitor.

3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours. 4.6.D

1. Approximately one-half of all relief valves

1. When more than one relief valve shall be bench-checked is known to be failed, an or replaced with a orderly shutdown shall be bench-checked valve initiated and the reactor each operating cycle.

depressurized to less than 105 All 13 valves will have psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The been checked or relief valves are not required replaced upon the to be OPERABLE in the COLD completion of every SHUTDOWN CONDITION. second cycle.

2. In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.

BFN 3.6/4.6-10 Unit 1

4 NG 3.7.A. t mnt 4. 7.A. t nm nt 2.a. Primary containment 2~ t'n integrity shall be maintained at all times Primary containment nitrogen when the reactor is critical consumption shall be or when the reactor water monitored to determine the temperature is above 212'F average daily nitrogen and fuel is in the reactor consumption for the last vessel except while 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Excessive leakage performing "open vessel" is indicated by a N2 physics tests at power consumption rate of > 2X of levels not to exceed the primary containment free 5 m(t). volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for drywell

b. Primary containment temperature, pressure, and integrity is confirmed if venting operations) at the maximum allowable 49.6 psig. Corrected to integrated leakage rate, normal drywell operating La, does not exceed the pressure of 1.1 psig, this equivalent of 2 percent of value is 542 SCFH. If this the primary containment value is exceeded, the volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the action specified in 49.6 psig design basis 3.7.A.2.C shall be taken.

accident pressure,'a.

The containment leakage rates

c. If N2 makeup to the primary shall be demonstrated at the containment averaged over following test schedule and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for shall be determined in pressure, temperature, and accordance with Appendix J venting operations) exceeds to 10 CFR 50 as modified 542 SCFH, it must be reduced by approved exemptions.

to ( 542 SCFH within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the reactor shall be a. Three type A tests placed in Hot Shutdown (overall integrated within the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. containment leakage rate) shall be conducted at 40'~ 10-month intervals during shutdown at Pa, 49.6 psig, during each 10-year plant inservice inspection.

BFN 3.7/4.7-3 Unit 1

't 4.

LSt .h 4.7.G. t t t t D t' 20 The Containment Atmosphere 2. When FCV 84-8B is inoper-Dilution (CAD) System shall able, each solenoid be OPERABLE whenever the operated air/nitrogen reactor is in the RUN valve of System B shall NODE. be cycled through at least one complete cycle of full travel and each manual valve in the flow path of System B shall be verified open at least once per week.

3. If one system is inoperable, the reactor may remain in operation for a period of 30 days provided all active components in the other system are OPERABLE.

4. If Specifications 3.7.G.1 and 3.7.G.2, or '3.7.G.3 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the COLD SHUTDOWN CONDITION within 24 hours.

5. Primary containment pressure shall be limited to a maximum of 30 psig during repressurization following a loss of coolant accident.

6. System A may be considered OPERABLE with FCV 84-8B inoperable provided that all active components in System B and all other active components in System A are OPERABLE.

7 ~ Specifications 3.7.G.6 and 4.7.G.2 are in effect until the first Cold Shutdown of unit 1 after July 20, 1984 or until January 17, 1985 whichever occurs first.

BFN 3.7/4.7-23 Unit 1

Each 250-V dc shutdown board control power supply can receive power from its own battery, battery charger, or from a spare charger. The chargers are powered from normal plant auxiliary power or from the standby diesel-driven generator system. Zero resistance short circuits between the control power supply and the shutdown board are cleared by fuses located in the respective control power supply. Each power supply is located in the reactor building near the shutdown board it supplies.

Each battery is located in its own independently ventilated battery room.

The 250-V dc system is so arranged, and the batteries sized so that the loss of any one unit battery will not prevent the safe shutdown and cooldown of all three units in the event of the loss of offsite power and a design basis accident in any one unit. Loss of control power to any engineered safeguard control circuits is annunciated in the main control room of the unit affected. The loss of one 250-V shutdown board battery affects normal control power for the 480-V and 4,160-V shutdown board which it supplies. The station battery supplies loads that are not essential for safe shutdown and cooldown of the nuclear system.

This battery was not considered in the accident load calculations.

There are two 480-Volt ac RMOV boards that contain MG sets in their feeder lines. These 480-Volt ac RMOV boards have an automatic transfer from their normal to alternate power source (480-Volt ac shutdown boards). The MG sets act as electrical isolators to prevent a fault from propagating between electrical divisions due to an automatic transfer. The 480-Volt ac RMOV boards involved provide motive power to valves associated with the LPCI mode of the RHR system. Having an MG set out of service reduces the assurance that full RHR (LPCI) capacity will be available when required. Since sufficient equipment is available to maintain the minimum complement required for RHR (LPCI) operation, a 7-day servicing period is justified. Having two MG sets out of service can considerably reduce equipment availability; therefore, the affected unit shall be placed in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The offsite power source requirements are based on the capacity of the respective lines. The Trinity line is limited to supplying two operating units because of the load limitations of CSST's A and B. The Athens line is limited to supplying one operating unit because of the load limitations of the Athens line. The limiting conditions are intended to prevent the 161-kV system from supplying more than two units in the event of a single failure in the offsite power system.

BFN 3.9/4.9-20 Unit 1

6.5.3.3 Individuals responsible for reviews performed in accordance with 6.5.3.1 shall be members of the site supervisory staff previously designated by the Plant Manager. Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary. If deemed necessary, such review shall be performed by review I

personnel of the appropriate discipline.

6.5.3.4 The Plant Manager shall approve all administrative procedures requiring PORC review prior to implementation.

6-6 (Deleted)

BFN 6.0-18, Unit 1

6.9.1.7 CORE OPERATING'IMITS REPORT

a. Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remaining portion of an operating cycle, for the following:

(1) The APLHGR for Specification 3.5.I (2) The LHGR for Specification 3.5.J (3) The MCPR Operating Limit for Specification 3.5.K/4.5.K

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (applicable amendment specified in the CORE OPERATING LIMITS REPORT).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin limits, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The CORE OPERATING LIMITS REPORT shall be provided wi.thin 30 days after cycle STARTUP for each reload cycle or within 30 days of issuance of any mid-cycle revision to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

BFN 6.0-26a Unit 1

PROPOSED TECHNICAL SPECIFICATION CHANGE BROMNS FERRY NUCLEAR PLANT UNIT 2 (TVA BFN TECHNICAL SPECIFICATION NO. 331)

UNIT 2 EFFECTIVE PAGE LIST REMOVE INSERT ii iii iii*

V V 1.0-4 1.0-4 1.1/2.1-13 1.1/2.1-13 1.1/2.1-16 1.1/2.1-16 3.2/4.2-26 3.2/4.2-26 3.2/4.2-27 3.2/4.2-27 3.2/4.2-68 3.2/4.2-68 3.3/4.3-12 3.3/4.3-12 3.4/4.4-4 3.4/4.4-4 3.5/4.5-19 3.5/4.5-19 3.5/4.5-27 3.5/4.5-27 3.6/4.6-6 3.6/4.6-6 3.6/4.6-10 3.6/4.6-10 3.7/4.7-3 3.7/4.7-3 3.7/4.7-19 3.7/4.7-19 3.7/4.7-23 3.7/4.7-23 6.0-18 6.0-18 6.0-26a 6.0-26a

>'<Denotes Spi11-Over Pages

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 . . . . . 3.4/4.4-3 C.'odium Pentaborate Solution. . . . . . . . . . 3.4/4.4-3 D. Standby Liquid Control System Requirements . . 3.4/4.4-4 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 and Emergency Equipment Cooling Water Systems (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). . . . 3.5/4.5-16 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 M. Core Thermal-Hydraulic Stability . 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 BPN Unit 2

C.

t Coolant Leakage.

t 3.6/4.6-9 D. Relief Valves. 3.6/4.6-10 E. Jet Pumps ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3.6/4.6-11 F. Recirculation Pump Operation 3.6/4.6-12 G. Structural Integrity 3.6/4.6-13 H. Snubbers 3.6/4.6-15 3.7/4.7 Containment Systems 3.7/4.7-1 A. Primary Containment. 3.7/4.7-1 B. Standby Gas Treatment System . 3.7/4.7-13 C. Secondary Containment. 3.7/4.7-16 D. Primary Containment Isolation Valves 3.7/4.7-17 E. Control Room Emergency Ventilation 3.7/4.7-19 F. Primary Containment Purge System 3.7/4.7-21 G. Containment Atmosphere Dilution System (CAD) 3.7/4.7-22 H. Containment Atmosphere Monitoring (CAN)

System H2 Analyzer 3.7/4.7-24

3. 8/4. 8 'adioac tive Materials 3.8/4.8-1 A. Liquid Effluents 3.8/4.8-1 B. Airborne Effluents 3.8/4.8-3 C. Radioactive Effluents Dose 3.8/4.8-6 D. Mechanical Vacuum Pump .' 3.8/4.8-6 E. Miscellaneous Radioactive Materials Sources. 3.8/4.8-7 F. Solid Radwaste 3.8/4.8-9 3-9/4.9 Auxiliary Electrical System . 3.9/4.9-1 A. Auxiliary Electrical Equipment 3.9/4.9-1 B.. Operation with Inoperable Equipment. 3.9/4.9-8 C. Operation in Cold Shutdown 3.9/4.9-15 D. Unit 3 Diesel Generators Required for Unit 2 Operation . 3.9/4.9-15a BPK Unit 2

SZXXQH

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~0 1

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~0 1 6.2.1 Offsite and Onsite Organizations... ..................... ~ 6.0-1 6.2.2 P lant Staff.............................................. 6.0-2

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~0 5 xpjggg o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6. 0-5

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~0 5 6.5.1 Plant Operations Review Committee (PORC)................. 6.0-5 6.5.2 Nuclear Safety Review Board (NSRB)....................... 6.0-11 6.5.3 Technical Review and Approval of Procedures.............. 6. 0-17

( Deleted) 6.0-18

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-19

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~ 0 20 6.8.1 Procedures.....-........-.-..........- 6.0-20 6.8.2 Drills ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-21 6.8.3 Radiation Control Procedures............................. 6.0-22 6.8.4 guality Assurance Procedures Effluent and Environmental Monitoring............................ ~ . 6.0-23 6.8.5 Programs ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-23

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ' 24 6.9.1 R outzne Reports...............................;.......... 6.0-24 Startup Reports.......................................... 6.0-24 Annual Operating Report.................................. 6.0-25 Monthly Operating Report................................. 6.0-26 Reportable Events....... 6.0-26 Radioactive Effluent Release Report................ ..... ~ 6.0-26 Source Tests............................................. 6.0-26 6-9.2 Special Reports.............-............................ 6.0-27

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~ 0 29

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-32 6.0-32

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~0 33 BFN Unit 2

1.0 Mode of Operation of the reactor when there is fuel in the reactor vessel, except that the Mode of Operation may remain unchanged when the reactor mode switch is temporarily moved to another position as permitted by the notes. When there is no fuel in the reactor vessel, the reactor is considered not to be in any Mode of Operation or operational condition. The reactor mode switch may. then be in any position or may be inoperable.

1~ t t t t The reactor is in the STARTUP/HOT STANDBY MODE when the reactor mode switch is in the "STARTUP/HOT STANDBY" position. This is often referred to as just the STARTUP MODE.

2~ Rg~~ The reactor is in the Run Mode when the reactor mode switch is in the "Run" position.

3.

the reefer mode switch is in the "Shutdown" position.

(2)(3)(

4~

reactor mode switch is in the "Refuel" position.

The reactor mode switch may be placed in any position to perform required tests or maintenance authorized by the shift operations supervisor, provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

) The reactor mode switch may be placed in the "Refuel" position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.10.Ae5 provided that reactor coolant temperature is equal to or less than 212'.

) The reactor mode switch may be placed in the "Refuel" position while a single control rod is being recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE.

) The reactor mode switch may be placed in the "Startup/Hot Standby" position and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.

BFN 1.0-4 Unit 2

2.1 lhSIK (Cont'd) ~

Analyses of the limiting transients show that no scram adjustment is required to assure MCPR > 1.07 when the transient is initiated from MCPR limits specified in Specification 3.5.k.

2~

For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram 1'evel, the rate of power rise is no more than five percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater than 850 psig.'.

The IRM System consists of eight chambers, four in each of the reactor protection system logic channels. The IRM is a five-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The five decades are covered by the IRM by means of a range switch and the five decades are broken down into 10 ranges, each being one-half of a decade in size. The IRM scram setting of 120 divisions is active in each range of the IRM. For example, if the instrument were on range 1, the scram setting would be at 120 divisions for that range; likewise if the instrument was on range 5, the scram setting would be 120 divisions on that range.

BP5 1.1/2.1-13 Unit 2

2.1 &RED (cont'd) ~

F. (Deleted)

G. 6( H.

The low pressure isolation of the main steam lines at 825 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. The scram feature that occurs when the main steamline isolation valves close shuts down the reactor so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch be in the STARTUP position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutron flux scrams.

Thus, the combination of main steamline low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure.

With the scrams set at 10 percent of valve closure, neutron flux does not increase.

I.J.& K. w t v t t These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures. The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.

L.

1. Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 2 (applicable cycle-specific document).

2. GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (applicable amendment specified in the CORE OPERATING LIMITS REPORT).

BFN 1.1/2.1-16 Unit 2

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 SRN, IRM, and APRM (STARTUP mode), blocks need not be OPERABLE in "RUN" mode, and the APRM (flow biased) rod 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).

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

4. SRNs 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 IRN channels B, F, D, H all in range 8 or above bypasses SRN channels B and D functions.

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

a. Both RBN channels are bypassed when reactor power is ~30 percent or when a peripheral (edge) 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 with the console selector. If the inope'rable 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.

BFH 3.2/4.2-26 Unit 2

(Cont'd)

7. (Continued)

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

8. This function is bypassed when the mode switch is placed in RUN.

9. This function is only active when the mode switch is in RUN. This function is automatically bypassed when the IRM instrumentation is OPERABLE and not high.

10. The inoperative trips are produced by the following functions:

a. SRM and IRM (1) Local "operate-calibrate" switch not in operate.

(2) Power supply voltage low.

(3) Circuit boards not in circuit.

b. APRM (1) Local "operate-calibrate" switch not in operate.

(2) Less than 14 LPRM inputs.

(3) Circuit boards not in circuit.

c. RBM (1) Local "operate-calibrate" switch not in operate.

(2) Circuit boards not in circuit.

(3) RBM fails to null.

(4) Less than required number of LPRM inputs for rod selected.

ll. Detector traverse is adjusted to 114 + 2 inches, placing detector lower position 24 inches below the lower core plate.

BFN 3.2/4.2-27 Unit 2

3.2 MAES (Cont'd)o The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed. An exception to this is when logic functional testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 1.07. The trip logic for this function is 1-out-of-n: e.g., any trip on one of six APRMs, eight IRMs, or four SRMs will result in a rod block.

When the RBM is required, the minimum instrument channel requirements apply. These requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This does not significantly increase the risk of an inadvertent control rod withdrawal, as the other channel is available, and the RBM is a backup system to the written sequence for withdrawal of control rods.

The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at reduced flow. The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than 1.07.

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

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

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

A downscale indication is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.

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

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 BFN 3.2/4.2-68 Unit 2

4.3.E. t'v't If Specifications 3.3.C and .D Surveillance requirements above cannot be met, an orderly are as specified in 4.3.C shutdown shall be initiated and and .D above.

the reactor shall be in the SHUTDOWN CONDITION within 24 hours.

3.3.F. 4.3.F.

1. The scram discharge volume l.a. The scram discharge drain and vent valves shall volume drain and vent be OPERABLE any time that valves shall be the reactor protection verified open PRIOR TO system is required to be STARTUP and monthly OPERABLE except as thereafter. The valves specified in 3.3.F.2. may be closed intermittently for testing not to exceed 1 hour in any 24-hour period during operation.

l.b. The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with Specification 1.0.MM.

2. In the event any SDV drain 2~ When it is determined or vent valve becomes that any SDV drain or inoperable, REACTOR POWER vent valve is OPERATION may continue inoperable, the provided the redundant redundant drain or drain or vent valve is vent valve shall be OPERABLE. demonstrated OPERABLE immediately and weekly thereafter.

3. If redundant drain or vent 3. No additional valves become inoperable, surveillance required.

the reactor shall be in HOT STANDBY CONDITION within 24 hours.

BFN 3.3/4.3-12 Unit 2

.4 ~

a. Calculate the enrich-ment within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. Verify by analysis within 30 days.

4.4.D The Standby Liquid Control Verify that the equation System conditions must satisfy given in Specification the following equation. 3.4.D is satisfied at least Z 1 'once per month and within (13 wt.X)(86 gpm)(19.8 atom%) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> anytime water or boron is added to the where, solution.

C = sodium pentaborate solution concentration (weight percent)

Determined by the most recent performance of the surveillance instruction required by Specification 4.4.C.2.

Q = pump flow rate (gpm)

Determined by the most recent performance of the surveillance instruction required by Specification 4.4.A.2.b.

E = Boron-10 enrichment (atom percent Boron-10)

Determined by the most recent performance of the surveillance instruction required by Specification 4.4.C.4.

1. If Specification 3.4.A through 1. No additional 3.4.D cannot be met, make at surveillance required.

least one subsystem OPERABLE within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the reactor shall be placed in a SHUTDOWN CONDITION with all OPERABLE control rods fully inserted within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

BFN 3.4/4.4-4 Unit 2

3.5.K 4.5.K

~Q.

r The minimum critical power ratio 1. MCPR shall be determined daily (MCPR) as a function of scram time during reactor power operation and core flow, shall be equal to or at y 25K rated thermal power greater than shown in Figure 3.5.K-l and following any change in multiplied by the Kf shown in power level or distribution Figure 3.5.2, where: that would cause operation with a limiting control rod or W V, whichever is pattern as described in the 7 = 0 A 7 B greater bases for Specification ~

~A = 0-90 sec (Specification 3.3.C.1 2. The MCPR limit shall be scram time limit to 20K insertion determined for each from fully withdrawn) fuel type 8X8, 8XSR, PSXSR, from Figure 3.5.K-1, respectively, using:

2 (0.053) [Ref.2]

7 B = 0.710+1.65 n

7 ave = ~G= a. 7 = 0.0 prior to initial scram time measurements for the cycle, performed number of surveillance rod in accordance with tests performed to date in Specification 4.3.C.l.

cycle (including BOC test).

b. ~as defined in Specifi-7g Scram time to 20K insertion from cation 3.5.K following the fully withdrawn of the it rod. conclusion of'ach scram-time surveillance test re-

~t Ltt number of active rods quired by Specifications measured in Specification 4.3.C.1 and 4.3.C.2.

4.3.C.l at BOC.

The determination of the If at any time during steady-state limit must be completed operation it is determined by normal within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of each surveillance that the limiting scram-time surveillance value for MCPR is being exceeded, required by Specification action shall be initiated within 4.3.C.

15 minutes to restore operation to within the prescribed limits. If the steady-state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the COLD SHUTDOWN CONDITION within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

BFN 3.5/4.5-19 Unit 2

3.5 (Cont'he RHR Service Water System was designed as a shared system for three units. The specification, as written, is conservative when consider-ation is given to particular pumps being out of service and to possible valving arrangements. If unusual operating conditions arise such that more pumps are out of service than allowed by this specification, a special case request may be made to the NRC to allow continued operation if the actual system cooling requirements can be assured.

Should one of the two RHRSW pumps normally or alternately assigned to the RHR heat exchanger header supplying the standby coolant supply connection become inoperable, an equal capability for long-term fluid makeup to the unit reactor and for cooling of the unit containment remains OPERABLE. Because of the availability of an equal makeup and cooling capability, a 30-day repair period is justified. Should the capability to provide standby coolant supply be lost, a 10-day repair time is justified based on the low probability for ever needing the standby coolant supply. Verification that the LPCI subsystem cross-tie valve is closed and power to its operator is disconnected'ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem.

With only one unit fueled, four RHRSW pumps are required to be OPERABLE for indefinite operation to meet the requirements of Specification 3.5.B.l (RHR system). If only three RHRSW pumps are OPERABLE, a 30-day LCO exists because of the requirement of Specification 3.5.B.5 (RHR system).

3.5.D There is an equipment area cooler for each RHR pump and an equipment area cooler for each set (two pumps, either the A and C or B and D pumps) of core spray pumps. The equipment area coolers take suction near the cooling air discharge of the motor of the pump(s) served and discharge air near the cooling air suction of the motor of the pump(8) served. This ensures that cool air is supplied for cooling the pump motors.

The equipment area coolers also remove the pump, and equipment waste heat from the basement rooms housing the engineered safeguard equipment. The various conditions under which the operation of the equipment air coolers is required have been identified by evaluating the normal and abnormal operating transients and accidents over the full range of planned operations. The surveillance and testing of the equipment area coolers in each of their various modes is accomplished during the testing of the equipment served by these coolers. This testing is adequate to assure the OPERABILITY of the equipment area coolers.

gg~gN~Q

1. Residual Heat Removal System (BFN FSAR Section 4.8)

2. Core Standby Cooling System (BFN FSAR Section 6)

BFN 3.5/4.5-27 Unit 2

4.

N 3.6.B. 4.6.B.

3~ At steaming rates 3~ Whenever the reactor greater than 100,000 is operating-(including lb/hr, the reactor HOT STANDBX CONDITION) water quality may measurements of reactor exceed Specification water quality shall be 3.6.B.2 only for the performed according to time limits specified the following schedule:

below. Exceeding these time limits or the following a. Chloride ion content maximum quality limits shall shall be measured be cause for placing at least once the reactor in the every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

COLD SHUTDOWN CONDITION. b. Chloride ion content shall be a~ Conductivity measured at least time above every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1 Nmho/cm at 25'C whenever reactor 2 weeks/year. conductivity is Maximum Limit >1.0 Pmho/cm 10 Pmho/cm at 25'C at 25'C.

c. A sample of primary

b. Chloride coolant shall be concentration time measured for pH at above 0.2 ppm least once every 8 2 weeks/year. hours whenever the Maximum Limit reactor coolant 0.5 ppm. conductivity is >1.0

. Pmho/cm at 25'C.

c. The reactor shall be placed in the SHUTDOWN CONDITION if pH <5.6 or

>8.6 for a 24-hour period.

BFN 3.6/4.6-6 Unit 2

N D 3.6.C 4.6.C

2. Anytime irradiated fuel is in 2. With the air sampling the reactor vessel and reactor system inoperable, grab coolant temperature is above samples shall be obtained 212'F, both the sump and air and analyzed at least sampling systems shall be once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

~ OPERABLE. From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the sump system or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the air sampling system.

The air sampling system may be removed from service for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for calibration, function testing, and maintenance without providing a temporary monitor.

3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 4.6.D.

24 hours.

1. Approximately one-half of all relief valves shall be bench-checked or

1. When more than one relief replaced with a valve is known to be failed, bench-checked valve each an orderly shutdown shall be operating cycle. All 13 initiated and the reactor valves will have been depressurized to less than 105 checked or replaced upon psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The the completion of every relief valves are not required second cycle.

to be OPERABLE in the COLD SHUTDOWN CONDITION. 2. In accordance with Specification 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.

BFN 3.6/4.6-10 Unit 2

4.

'm t '

3.7.A. 4.7.A.

2 ~ ao Primary containment 2. t t integrity shall be maintained at all times Primary containment nitrogen when the reactor is critical consumption shall be or when the reactor water monitored to determine the temperature is above 212'F average daily nitrogen and fuel is in the reactor consumption for the last vessel except while 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Excessive leakage performing "open vessel" is indicated by a N2 physics tests at power consumption rate of > 2X of levels not to exceed the primary containment" free 5 mW(t). volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for drywell

b. Primary containment temperature, pressure, and integrity is confirmed if venting operations) at the maximum allowable 49.6 psig. Corrected to integrated leakage rate, normal drywell operating La, does not exceed the pressure of 1.1 psig, this equivalent of 2 percent of value is 542 SCFH. If this the primary containment value is exceeded, the volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the action specified in 49.6 psig design basis 3.7.A.2.C shall be taken.

accident pressure, Pa.

The containment leakage rates c~ If N2 makeup to the primary shall be demonstrated at the containment averaged over following test schedule and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for shall be determined in pressure, temperature, and accordance with Appendix J to venting operations) exceeds 10 CFR 50 as modified by 542 SCFH, it must be reduced approved exemptions.

to < 542 SCFH within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the reactor shall be a. Three type A tests (overall placed in Hot Shutdown integrated containment within the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. leakage rate) shall be conducted at 40 g 10-month intervals during shutdown at Pa, 49.6 psig, during each 10-year plant inservice inspection.

BFN 3.7/4.7-3 Unit 2

4 G N 3.7.E. 4.7.E.

l. Except as specified in 1. At least once every 18 months, Specification 3.7.E.3 below, the pressure drop across the both control room emergency combined HEPA filters and pressurization systems charcoal adsorber banks shall shall be OPERABLE at all be demonstrated to be less times when any reactor than 6 inches of water at vessel contains irradiated system design flow rate fuel. (g 10K).

2~ ao The results of the inplace 2. a. The tests and sample cold DOP and halogenated analysis of Specification hydrocarbon tests at design 3.7.E.2 shall be performed flows on HEPA filters and at least once per operating charcoal adsorber banks cycle or once every shall show p99X DOP removal 18 months, whichever occurs and >99K halogenated first for standby service hydrocarbon removal when or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of tested in accordance with system operation and ANSI N510-1975. following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

b. The results of laboratory b. Cold DOP testing shall be carbon sample analysis shall performed after each show y90X radioactive methyl complete or partial iodide removal at a velocity replacement of the HEPA when tested in accordance filter bank or after any with ASTM D3803 structural maintenance on (130'C, 95K R.H.). the system housing.

CREVS is considered inoperable only because it does not meet its design basis for essentially zero unfiltered inleakage. REACTOR POWER OPERATION and fuel movement are acceptable until just PRIOR TO STARTUP for unit 2 cycle 7.

During cycle 6, CREVS must be demonstrated to be functional by performing all applicable surveillances. In the event that the applicable surveillances are not successfully performed, the actions required by the LCO's must be complied with.

BFH 3.7/4.7-19 Unit 2

3.7.G. t t

t t 4.7.G.

D t t m

h A

2. The Containment Atmosphere Dilution (CAD) System shall be OPERABLE whenever the reactor is in the RUN MODE.

3. If one system is inoperable, the reactor may remain in operation for a period of 30 days provided all active components in the other system are OPERABLE.

4. If Specifications 3.7.G.l and 3.7.G.2, or 3.7.G.3 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the COLD SHUTDOWN CONDITION within 24 hours.

5. Primary containment pressure shall be limited to a maximum of 30 psig during repressurization following a loss of coolant accident.

BFH 3.7/4.7-23 Unit 2

6.5.3.3 Individuals responsible for reviews performed in accordance with 6.5.3.1 shall be members of the site supervisory staff previously designated by the Plant Manager. Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary. If deemed necessary, such review shall be performed by review personnel of the appropriate discipline.

6.5.3.4 The Plant Manager shall approve all administrative procedures requiring PORC review prior to implementation.

6.6 (Deleted)

BFN 6.0-18 Unit 2

6.9.1.7 CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remaining portion of an operating cycle, for the following:

(1) The APLHGR for Specification 3.5.I (2) The LHGR for Specification 3.5.J (3) The MCPR Operating Limit for Specification 3.5.K/4.5.K

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" '(applicable amendment specified in the CORE OPERATING LIMITS REPORT).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin limits, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle STARTUP for each reload cycle or within 30 days of issuance of any mid-cycle revision to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

BFN 6.0-26a Unit 2

PROPOSED TECHNICAL SPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 3 (TVA BFN TECHNICAL SPECIFICATION NO. 331)

UNIT 3 EFFECTIVE PAGE LIST RENOVE INSERT ii ii iii iv iii*

iv+

V V 1.0-4 1.0-4 1.1/2.1-13 1.1/2.1-13 1.1/2.1-16 1.1/2.1-16 3.1/4.1-14 3.1/4.1-14 3.3/4.3-12 3.3/4.3-12 3.4/4.4-4 3.4/4.4-4 3.5/4.5-30 3.5/4.5-30 3.6/4.6-6 3.6/4.6-6 3.6/4.6-10 3.6/4.6-10 3.7/4.7-3 3.7/4.7-3

'.7/4.7-19 3.7/4.7-19 3.7/4.7-23 3.7/4.7-23 6.0-18 6.0-18 6.0-26a 6.0-26a

>'<Denotes Spill-Over Page

5rzJtun D. Reactivity Anomalies 3.3/4.3-11 E- Reactivity Control 3.3/4.3-12 F. Scram Discharge Uolume 3.3/4.3-12 3.4/4 ' Standby Liquid Control System . . . . . . . . . . . 3.4/4.4-1 A. Normal System Availability . . . . . . . . . . 3.4/4.4-1 B. Operation with Inoperable Components . . . . . 3.4/4.4-3 C. Sodium Pentaborate Solution. . . . . . . . . . 3.4/4.4-3 D. Standby Liquid Control System Requirements .. 3.4/4.4-4 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 and Emergency Equipment Cooling Water Systems (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). . . . 3.5/4.5-16 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 '/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 C. Coolant Leakage. 3.6/4.6-9 BFN Unit 3

D. Relief Valves. 3.6/4.6-10 E Je 't Pumps ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3.6/4.6-11 F. Recirculation Pump Operation 3.6/4.6-12 G. Structural Integrity . 3.6/4.6-13 H. Snubbers 3.6/4.6-15 3.7/4.7 Containment Systems 3.7/4.7-1 A. Primary Containment. 3.7/4.7-1 B. Standby Gas Treatment System . . . . . . . . . 3.7/4.7-13 C. Secondary Containment. 3.7/4.7-16 D. Primary Containment Isolation Valves . . . . . 3.7/4.7-17 E. Control Room Emergency Ventilation . . . . . . 3.7/4.7-19 F. Primary Containment Purge System . 3.7/4.7-21 G. Containment Atmosphere Dilution System (CAD) . 3.7/4.7-22 H. Containment Atmosphere Monitoring (CAM)

System H2 Analyzer . 3.7/4.7-23a 3.8/4.8 Radioactive Materials I 3.8/4.8-1 A. Liquid Effluents 3.8/4.8-1 B. Airborne Effluents s.ai~.s-~

C. Radioactive Effluents Dose 3.8/4.8-6 D. Mechanical Vacuum Pump 3.8/4.8-6 E. Miscellaneous Radioactive Materials Sources 3.8/4.8-7 F. Solid Radwaste 3.8/4.8-9 3-9/4.9 Auxiliary Electrical System . 3.9/4.9-1 A. Auxiliary Electrical Equipment 3.9/4.9-1 B. Operation with Inoperable Equipment. 3.9/4.9-8 C. Operation in Cold Shutdown Condition 3.9/4.9-14 D. Unit 3 Diesel Generators Required for Unit 2 Operation 3.9/4.9-14a BFN Unit 3

3.10/4.10 Core Alterations 3.10/4.10-1 A. Refueling Interlocks 3.10/4.10-1 B. Core Monitoring 3.10/4.10-4 C. Spent Fuel Pool Water . 3.10/4.10-7 D. Reactor Building Crane 3.10/4.10-8 E. Spent Fuel Cask 3.10/4.10-9 F. Spent Fuel Cask Handling-Refueling Floor. . . . 3.10/4.10-9 3.11/4.11 Fire Protection Systems 3.11/4.11-1 A. Fire Detection Instrumentation 3.11/4.11-1 B. Fire Pumps and Water Distribution Mains 3.11/4.11-2 C. Spray and/or Sprinkler Systems 3.11/4.11-7 D. C02 System 3.11/4.11-8 E. Fire Hose Stations. 3.11/4.11-9 F. Yard Fire Hydrants and Hose Houses 3.11/4.11-11 G. Fire-Rated Assemblies ~ ~ 3.11/4.11-12 H. Open Flames, Welding and Burning in the Cable Spreading Room. ~ ~ ~ ~ 3.11/4.11-13 5 ~0 Major Design Features 5.0-1 5.1 . Site Features 5.0-1 5.2 Reactor . 5.0-1 5.3 Reactor Vessel 5.0-1 5.4 Containment 5.0-1 5.5 Fuel Storage 5.0-1 5.6 Seismic Design

'.0-2 BFN iv Unit 3

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6oO 1

~ ~ ~ ~ ~ ~ ~ ~ s ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ 6 ~0 1 6.2.1 Offsite and Onsite Organizations......................... 6.0-1 6.2.2 P lant Staff.............................................. 6.0-2

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-5 XRh Qg o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-5

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6~0 5 6.5.1 Plant Operations Review Committee (PORC)................. 6.0-5 6.5.2 Nuclear Safety Review Board (NSRB)...................,.... 6.0-11 6.5.3 Technical Review and Approval of Procedures.............. 6.0-17

( Deleted)................................................ 6.0-18 6.0-19

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ s ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~ 0 20 6.8.1 Procedures............................................... 6.0-20 6-8.2 Dri1 ls 6.0-21

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

6.8.3 Radiation Control Procedures............................. 6.0-22 6.8.4 Quality Assurance Procedures Effluent and Environmental Monitorang.............................. 6.0-23

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~ 0 24 6.9.1 Routine Reports.......................................... 6.0-24 Startup Reports.......................................... 6.0-24 Annual Operating Report.................................. 6.0-25 Monthly Operating Report................................. 6.0-26 Reportable Events...... 6.0-26 Radioactive Effluent Release Report...................... 6.0-26 Source Tests............................................. 6.0-26 6.9.2 S pecxal Reports.......................................... 6.0-27 N D ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 ~ 0 29

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o 6 '-32 2 L. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-32 ALa ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ 6.0-33 BFN v Unit 3

1.0 M. The reactor mode switch position determines the Mode of Operation of the reactor when there is fuel in the reactor vessel, except that the Mode of Operation may remain unchanged when the reactor mode switch is temporarily moved to another position as permitted by the notes. When there is no fuel in the reactor vessel, the reactor is considered not to be in any Mode of Operation or operational condition. The reactor mode switch may then be in any position or may be inoperable. l

1. ~t~t t The reactor is in the STARTUP/HOT STANDBY MODE when the reactor mode switch is in the "STARTUP/HOT STANDBY" position. This is often referred to as just the STARTUP MODE.

2~ gggLJ5~ The reactor is in the Run Mode whe'n the reactor mode switch is in the "Run" position.

The reactor is in the Shutdown Mode when the re~ger mode switch is in the "Shutdown" position. ~

(2)(3)

4. Rgggg~~ The reactor is in the Refuel Mode ~h~n the reactor mode switch is in the "Refuel" position.

The reactor mode switch may be placed in any position to perform required tests or maintenance authorized by the shift operations supervisor, provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

) The reactor mode switch may be placed in the "Refuel" position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.10.Ae5 provided that reactor coolant temperature is equal to or less than 212'.

The reactor mode switch may be placed in the "Refuel" position while a single control rod is being recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE.

) The reactor mode switch may be placed in the "Startup/Hot Standby" position and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.

BFN 1.0-4 Unit 3

2.1 g~Q (Cont'1) ~

Analyses of the limiting transients show that no scram adjustment is required to assure MCPR > 1.07 when the transient is initiated from MCPR >***.

2~ t n For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Worth of individual rods is very low in a uniform .rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater than 850 psig.

3.

The IRM System consists of 8 chambers, 4 in each of the reactor protection system logic channels, The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size.

The IRM scram setting of 120 divisions is active in each range of the IRM. For example, if the instrument was on range 1, the scram setting would be 120 divisions for that range; likewise if the instrument was on range 5, the scram setting would be 120 divisions on that range.

<<>'<>'<See Section 3.5.K BFH 1.1/2.1-13 Unit 3

2.1 $ 3~5 (Cont'd F. (Deleted)

G. 6 H.

The low pressure isolation of the main steam lines at, 850 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the. vessel. Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 850 psig requires that the reactor mode switch be in the STARTUP position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRN high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase.

These systems maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures. The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and initiation setpoints. Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.

L.

1. Supplemental Reload Licensing Report of Browns Ferry Nuclear Plant, Unit 3 (applicable cycle-specific document).

2. GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (applicable amendment specified in the CORE OPERATING LINITS REPORT).

BFN 1.1/2.1-16 Unit 3

3.1 ~~ (Cont'd)o Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel. This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration. Additional IRM channels have also been provided to allow for bypassing of one such channel. The bases for the scram setting for the IRM, APRM, high reactor pressure, reactor low water level, MSIV closure, turbine control valve fast closure, and turbine stop valve closure are discussed in Specifications 2.1 and 2.2.

Instrumentation (pressure switches) for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality. This'nstrumentation is a backup to the reactor vessel water level instrumentation.

High radiation levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel.' scram is initiated whenever such radiation level exceeds three times normal background. The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent release of radioactive material to the turbine. An alarm is initiated whenever the radiation level exceeds 1.5 times normal background to alert the operator to possible serious radioactivity spikes due to abnormal core behavior.

The air ejector off-gas monitors serve to back up the main steam line monitors to provide further assurance against release of radioactive materials to site environs by isolating the main condenser off-gas line to the main stack.

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 3.1/4.1-14 Unit 3

4 L E

)

3.3.E. 4.3.E.

If Specifications 3.3.C and Surveillance requirements 3.3.D above cannot be met, are as specified in 4.3.C an orderly shutdown shall be and 4.3.D above.

initiated and the reactor shall be in the SHUTDOWN CONDITION within 24 hours.

[

3.3.F. 4.3.F.

1. The scram discharge volume 1 oa ~ The scram discharge

, drain and vent valves shall volume drain and vent be OPERABLE any time that valves shall be the reactor protection verified open PRIOR system is required to be TO STARTUP and OPERABLE except as monthly thereafter.

specified in 3.3.F.2. The valves may be closed intermittently for testing not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in any 24-hour period during operation.

l.b. The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with Specification 1.0.MM.

2. In the event any SDV drain 2~ When it is determined or vent valve becomes that any SDV drain or inoperable, REACTOR POWER vent valve is OPERATION may continue inoperable, the provided the redundant redundant drain drain or vent valve is or vent valve shall OPERABLE. be demonstrated OPERABLE immediately and weekly thereafter'.

3. If redundant drain or vent No additional valves become inoperable, surveillance the reactor shall be in HOT required.

STANDBY CONDITION within 24 hours.

BFN 3.3/4.3-12 Unit 3

.4 4 4

a. Calculate the enrich-ment within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. Verify by analysis within 30 days.

]

3.4.D 4.4.D The Standby Liquid Control Verify that the equation System conditions must satisfy given in Specification the following equation. 3.4.D is satisfied at least g 1 once per month and within (13 wt.X)(86 gpm)(19.8 atom%) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> anytime water or boron is added to the where, solution.

C = sodium pentaborate solution concentration (weight percent)

Determined by the most recent performance of the surveillance instruction required by Specification 4.4.C.2.

Q = pump flow rate (gpm)

Determined by the most recent performance of the surveillance instruction required by Specification 4.4.A.2.b.

E = Boron-10 enrichment (atom percent Boron-10)

Determined by the most recent performance of the surveillance instruction required by

,Specification 4.4.C.4.

l. If Specification 3.4.A through 1. No additional )

3.4.D cannot be met, make at surveillance required.

least one subsystem OPERABLE within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the reactor shall be placed in a SHUTDOWN CONDITION with all OPERABLE control rods fully inserted within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

BPN 3.4/4.4-4 Unit 3

3.5 355KB (Cont'he RHR Service Water System was designed as a shared system for three units. The specification, as written, is conservative when consider-ation is given to particular pumps being out of service and to possible valving arrangements. If unusual operating conditions arise such that more pumps are out of service than allowed by this specification, a special case request may be made to the NRC to allow continued operation if the actual system cooling requirements can be assured.

Should one of the two RHRSW pumps normally or alternately assigned to the RHR heat exchanger header supplying the standby coolant supply connection become inoperable, an equal capability for long-term fluid makeup to the unit reactor and for cooling of the unit containment remains OPERABLE. Because of the availability of an equal makeup and cooling capability, a 30-day repair period is justified. Should the capability to provide standby coolant supply be lost, a 10-day repair time is justified based on the low probability for ever needing the standby coolant supply. Verification that the LPCI subsystem cross-tie valve is closed and power to its operator is disconnected ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem.

With only one unit fueled, four RHRSW pumps are required to be OPERABLE for indefinite operation to meet the requirements of Specification 3.5.B.1 (RHR system). If only three RHRSW pumps are OPERABLE, a 30-day LCO exists because of the requirement of Specification 3.5.B.5 (RHR system).

3-5.D There is an equipment area cooler for each RHR pump and an equipment area cooler for each set (two pumps, either the A and C or B and D pumps) of core spray pumps. The equipment area coolers take suction near the cooling air discharge of the motor of the pump(s) served and discharge air near the cooling air suction of the motor of the pump(s) served. This ensures that cool air is supplied for cooling the pump motors.

The equipment area coolers also remove the pump, and equipment waste heat from the basement rooms housing the engineered safeguard equipment. The various conditions under which the operation of the equipment air coolers is required have been identified by evaluating the normal and abnormal operating transients and accidents over the full range of planned operations. The surveillance and testing of the equipment area coolers in each of their various modes is accomplished during the testing of the equipment served by these coolers. This testing is adequate to assure the OPERABILITY of the equipment area coolers.

~~E~ENl~~

1. Residual Heat Removal System (BFN FSAR Section 4.8)

2. Core Standby Cooling System (BFN FSAR Section 6)

BFN 3.5/4.5-30 Unit 3

4 N

3.6.B. 4.6.B.

3. At steaming rates 3. Additional coolant samples greater than 100,000 shall be taken whenever the lb/hr, the reactor reactor activity exceeds one water quality may percent of the equilibrium exceed Specification concentration specified in 3.6.B.2 only for the 3.6.B.5 and one of the time limits specified following conditions are met.

below. Exceeding these time limits or the following a. During the STARTUP maximum quality limits shall CONDITION be cause for placing the reactor in the b. Following a significant COLD SHUTDOWN power change<<<<.

CONDITION.

c. Following an increase in the equilibrium off-gas

a. Conductivity level exceeding 10,000 time above puci/sec (at the steam 2 Pmho/cm at 25'C jet air ejector) within 4 weeks/year. a 48-hour period.

Maximum Limit 10 pmho/cm at 25'C d. Whenever the equilibrium iodine limit specified

b. Chloride in 3.6.B.5 is exceeded.

concentration time above 0.2 ppm 4 weeks/year.

Maximum Limit **For the purpose of this section on 0.5 ppm. sampling frequency, a significant power exchange is defined as a change exceeding 15K of rated power in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

BFN 3.6/4.6-6 Unit 3

4.

2. Anytime irradiated fuel is in 2. With the air sampling the reactor vessel and reactor system inoperable, grab coolant temperature is above samples shall be 212'F, both the sump and air obtained and analyzed sampling systems shall be at least once every 24 OPERABLE. From and after the hours.

date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the sump system or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the air sampling system.

The air sampling system may be removed from service for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for calibration, function testing, and maintenance without providing a temporary monitor.

3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 4.6.D.

24 hours.

1. Approximately one-half i

3.6.D. of all relief valves shall be bench-checked

1. When more than one relief or replaced with a valve is known to be failed, bench-checked valve an orderly shutdown shall be each operating cycle.

initiated and the reactor All 13 valves will have depressurized to less than 105 been checked or psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The replaced upon the relief valves are not required completion of every to be OPERABLE in the COLD second cycle.

SHUTDOWN CONDITION.

2. In accordance with Specification. 1.0.MM, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.

BFN 3.6/4.6-10 Unit 3

' t '

3.7.A. r m t 4.7.A.

2.a. Primary containment 2~ t t t t integrity shall be maintained at all times Primary containment nitrogen when the reactor is critical consumption shall be or when the reactor water monitored to determine the temperature is above 212'F average daily nitrogen and fuel is in the reactor consumption for the last vessel except while 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Excessive leakage performing "open vessel" is indicated by a N2 physics tests at power consumption rate of > 2X of levels not to exceed the primary containment free 5 m(t). volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for drywell

b. Primary containment temperature, pressure, and integrity is confirmed if venting operations) at the maximum allowable 49.6 psig. Corrected to integrated leakage rate, normal drywell operating La, does not exceed the pressure of 1.1 psig, this equivalent of 2 percent of value is 542 SCFH. If this the primary containment value is exceeded, the volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the action specified in 49.6 psig design basis 3.7.A.2.c shall be taken.

accident pressure, Pa.

The containment leakage rates

c. If N2 makeup to the primary shall be demonstrated at the containment averaged over following test schedule and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for shall be determined in pressure, temperature, and accordance with Appendix J to venting operations) exceeds 10 CFR 50 as modified by 542 SCFH, it must be reduced approved exempti'ons.

to ( 542 SCFH within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the reactor shall be a. Three type A tests placed in Hot Shutdown (overall integrated within the next 16 hours. containment leakage rate) shall be conducted at 40 p 10-month intervals during shutdown at psig, during each Pa,'9.6 10-year plant inservice inspection.

BFN 3.7/4.7-3 Unit 3

4.

3.7.E. t t'.7.E.

l. Except as specified in l. At least once every 18 months, Specification 3.7.E.3 below, the pressure drop across the both control room emergency combined HEPA filters and

'pressurization systems charcoal adsorber banks shall shall be OPERABLE at all be demonstrated to be less times when any reactor than 6 inches of water at vessel contains irradiated system design flow rate fuel. (~ 10K).

2~ a~ The results of the inplace 2. a. The tests and sample

.cold DOP and halogenated analysis of Specification hydrocarbon tests at design 3.7.E.2 shall be performed flows on HEPA filters and at least once per charcoal adsorber banks operating cycle or once shall show y99X DOP removal every 18 months, and y99X halogenated whichever occurs first hydrocarbon removal when for- standby service tested in accordance with or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> ANSI N510-1975. of system operation and following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

b. The results of laboratory b. Cold DOP testing shall be carbon sample analysis shall performed after each show y90X radioactive methyl complete or partial iodide removal at a velocity replacement of the HEPA when tested in accordance filter bank or after any with ASTM D3803 structural maintenance (130'C, 95K R.H.). on the system housing.

CREVS is considered inoperable only because it does not meet its design basis for essentially zero unfiltered inleakage. REACTOR POWER OPERATION and fuel movement are acceptable until just PRIOR .TO STARTUP for unit 2 cycle 7.

During cycle 6, CREVS must be demonstrated to be functional by performing all applicable surveillances. In the event that the applicable surveillances are not successfully performed, the actions required by the LCO's must be complied with.

BPN 3.7/4.7-19 Unit 3

.7 4.

3.7.G. t in t tm t'

2. The Containment Atmosphere Dilution (CAD) System shall be OPERABLE whenever the reactor is in the. RUN NODE.

3. If one system is inoperable, the reactor may remain in operation for a period of 30 days provided all active components in the other system are OPERABLE.

4. If Specifications 3.7.G.l and 3.7.G.2, or 3.7.G.3 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the COLD SHUTDOWN CONDITION within 24 hours.

5. Primary containment pressure shall be limited to a maximum of 30 psig during repressurization following a loss of coolant accident.

BPN 3.7/4.7-23 Unit 3

6.5.3.3 Individuals responsible for reviews performed in accordance with 6.5.3.1 shall be members of the site supervisory staff previously designated by the Plant Manager. Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary. If deemed necessary, such review shall be performed by review personnel of the appropriate discipline.

6.5.3.4 The Plant Manager shall approve all administrative procedures requiring PORC review prior to implementation.

6.6 (Deleted)

BFN 6.0-18 Unit 3

6.9.1.7 CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each operating cycle, or prior to any remaining portion of an operating cycle, for the following:

(1) The APLHGR for Specification 3.5.I (2) The LHGR for Specification 3.5.J (3) The MCPR Operating Limit for Specification 3.5.K/4.5.K

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (applicable amendment specified in the CORE OPERATING LIMITS REPORT).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin limits, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The, CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle STARTUP for each reload cycle or within 30 days of issuance of any mid-cycle revision to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

BFN 6.0-26a Unit 3

ENCLOSURE 2 BROMNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 (TVA BFN TECHNICAL SPECIFICATION NO. 331)

REASO FOR HE HA GE DESCRIPTIO A D JUSTIFICATIO REASON These proposed changes to the BFN technical specifications are administrative in nature. The proposed changes are needed to delete a Unit 2 cycle 6 only requirement, to correct administrative errors in previous technical specifications and to correct discrepancies between specification bases and the BFN Final Safety Analysis Report (FSAR). In addition, the proposed changes include clarification of some requirements to ensure consistent application throughout the specifications.

DESCRI ION OF THE PROPOSED CHANGE

1. For Units 1, 2, and 3 Specification 3.6.C.2 reads:

"Both the sump and air sampling systems shall be OPERABLE during REACTOR POWER OPERATION. From and after the date that one these systems is made or found to be inoperable for any reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the sump syst: em or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the air sampling system."

The revised specification reads:

"Any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F, both the sump and air sampling systems shall be OPERABLE. From and after the date that one of these systems is made or found to be inoperable for any reason, the reactor may remain in operation during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the sump system or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the air sampling system."

2. For Unit 1-in Bases 3.9, page 3.9/4.9-20, the sentence reads" "The loss of one 250V shutdown board battery affects normal control power only for the 4160V shutdown board which it supplies."

The revised sentence reads:

"The loss of one 250V shutdown board battery affects normal control power for the 480V and 4160V shutdown board which it supplies.

ENCLOSURE 2 (CONTINUED)

Page 2 of 6 For Unit 2 in note 7 to Table 3.2.C the asterisks, the associated footnotes and items 7.e and 7.f are deleted, In item 7.c, the sentence "The other channel may also be defeated only if the conditions of "e" or "f" are met" and the phrase "and the conditions of "e" and "f" are not met" are deleted. In item 7.d, the phrase "and the conditions of "e" or "F" are not met" is deleted. In the Bases 3.2, page 3.2/4.2-68, the first two sentences of the third paragraph are deleted. In Specification 3.5.K, the phrase "Except when the provisions of Note 7 of Table 3.2.C are being employed due to the inoperability of the Rod Block Monitor" is deleted.

In Specification 4.5.K.2, the phrase "Except as provided by Note 7 of Table 3.2.C" is deleted.

For Units 1, 2, and 3, the entire Specification 6.6 "Reportable Event Action" is deleted and Table of Contents is revised to delete "Reportable Event Actions . . . 6.0-18."

For Units 1, 2, and 3, Specification 4.7.A.2 reads:

"The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J to 10 CFR 50 using the methods and provisions of ANSI N45 4 (1972) n The revised specification reads:

"The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in accordance with Appendix J to 10 CFR 50 as modified by approved exemptions."

For Units 1, 2, and 3, in note 4 to Definition 1.M, the phrase "RSCS and" is deleted. In the Bases 2.1.2, page 1.1/2.1-13, APRM Flux Scram Trip Setting (Refuel or Start and Hot Standby Mode), the phrase "and the Rod Sequence Control System" is deleted.

For Units 1 and 3, in Bases 3.1, pages 3.1/4.1-15 and 3.1/4.1-14, the sentence reads; "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."

The revised sentence reads:

The bases for the scram setting for the IRM, APRM, high reactor pressure, reactor low water level, MSIV closure, turbine control valve fast closure and turbine stop valve closure are discussed in Specifications 2.1 and 2.2."

For Units 2 and 3, in references to Bases 3.5, pages 3.5/4.5-27 and 3.5/4 '-30, the "BFN FSAR subsection 6.7" is revised to "BFN FSAR Section 6."

ENCLOSURE 2 (CONTINUED)

Page 3 of 6

9. For Units 1, 2, and 3, Specification 3.6.B.3 reads:

"Exceeding these time limits of the following maximum quality limits shall be cause for placing the reactor in the COLD SHUTDOWN CONDITION."

The revised Specification reads:

"Exceeding these time limits or the following maximum quality limits shall be cause for placing the reactor in the COLD SHUTDOWN CONDITION."

10. For Units 2 and 3, in Specification 4.7.E.l the second "to" in the fifth line is deleted.

ll. For Units 1, 2 and 3, the Table of Contents, section 3.4/4.4, Standby Liquid Control System is revised to add items "D. Standby Liquid Control System Requirements . . . 3.4/4.4-4".

For Units 1, 2, and 3, Specifications 3.3.E and 4.3.E are revised to add a heading of "3.3.E Reactivity Control" and "4.3.E Reactivity Control."

For Units 1, 2, and 3, the heading for Specification 3.3.F, "F. Scram Discharge Volume (SDV)" is revised to "3.3.F Scram Discharge Volume (SDV)." The heading for Specification 4.3.F, "F. Scram Discharge Volume (SDV)" is revised to "4.3.F Scram Discharge Volume (SDV)."

For Units 1, 2, and 3, Specifications 3.4.D and 4.4.D are revised to add a heading of "3.4.D Standby Liquid Control System Requirements" and "4.4.D Standby Liquid Control System Requirements." Specifications 3.4.E and 4.4.E are renumbered to 3.4.D.1 and 4.4.D.l.

For Units 1, 2, and 3, the heading for Specification 3.6.D, "D. Relief Valves" is revised to "3.6.D Relief Valves." The heading for Specification 4.6.D, "D. Relief Valves" is revised to "4.6.D Relief Valves."

For Units 1, 2, and 3, the heading for Specification "4.7.F Containment Atmosphere Dilution System (CAD)" is revised to "4.7.G Containment Atmosphere Dilution System (CAD)."

12. For Units 1, 2, and 3, "INOPERABLE" is changed to lowercase on pages 1.0-4 and 3.3/4.3-12. "Operable is changed to uppercase on pages 1.0-4, 3.3/4.3-12 (Ul only) and 3.4/4,4-4.

For Units 1, 2, and 3, "S" in specification is capitalized on page 1.0-4.

For Units 1, 2, and 3, "Operability" is changed to uppercase on page 1.0-4.

For Units 1, 2, and 3, "Refuel or Start and Hot Standby Mode" is changed to uppercase on page 1.1/2.1-13.

P C

I

ENCLOSURE 2 (CONTINUED)

Page 4 of 6 For Units 1, 2, and 3, "Shutdown condition" is changed to uppercase on page 3.4/4.4-4.

For Unit 2, "Cold shutdown condition" is changed to uppercase on page 3.5/4.5-19.

13. For Units 1, 2, and 3, reference 2 to Bases 2.1, page 1.1/2.1-16 reads:

"GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved version)."

The revised reference 2 reads:

"GE Standard Application for Reactor Fuel, NEDE-24011-P-A and NEDE-24011-P-A-US (applicable amendment specified in the CORE OPERATING LIMITS REPORT).

For Units 1, 2, and 3, Specification 6.9.1.7.b reads:

"The analytical methods used to determine the core operating limits shall be previously reviewed and approved by the NRC, specifically those described in General Electric Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel."

The revised specification reads:

"The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (applicable amendment specified in the CORE OPERATING LIMITS REPORT)."

JUSTIFICATION R THE PROPOSED CHANGE The justification for each change is provided below in order in which it appears in the description of the proposed change section of this enclosure.

1. Technical Specification 3.6.C.1 limits the reactor coolant system leakage to 5 gpm unidentified and 25 gpm total, any time irradiated fuel is in the reactor vessel and reactor coolant temperature is above 212'F. The leak detection system which measures this leakage is not required to be operable until reactor power operation in accordance with Specification 3.6.C.2. The proposed change is a clarification that requires the leak detecti.on systems to be operable when the leakage rate limits are required to be met. This clarification ensures system operability requirements of Technical Specification 3.6.C are consistent through the specification.

Requiring the leak detection systems to be operable at lower temperatures and pressures will not affect the safety function of the system and is conservative and consistent with safe operation of the plant.

4 II

ENCLOSURE 2 (CONTINUED)

Page 5 of 6 The Bases 3.9 for Unit 1 is revised to reflect a change resulting from the implementation of an engineering change during the previous outage. The revision to bases reflects the fact that the loss of 250V shutdown board battery affects the control power for both 480V and 4160V shutdown boards.

A similar change was submitted as part of TS-283 dated July 13, 1990, and approved by Amendment No. 186 dated January 9, 1991.

For Unit 2, Amendment 202 dated July 2, 1992, approved changes to Note 7 for Table 3.2.C, the Bases 3.2 and Specifications 3.5.K and 4.5.K.2 which were applicable during Unit 2 cycle 6 only. Unit 2 cycle 6 has been completed and these changes are deleted since they are no longer applicable.

The proposed change deletes Specification 6.6, since these requirements are covered by other specifications. The reporting requirement of Specification 6.6.1.2 is duplicated in Specification 6.9.1.4. The review requirement for PORC in Specification 6.6.l.b is duplicated in Specification 6.5.1.6.1. The review requirements for the NSRB in Specification 6.6.1.b is duplicated in Specification 6.5.2.7.g. The review requirement for the Site Director is covered since Reportable Events are submitted the Site Vice President (formerly the Site Director).

The proposed change to Specification 4.7.A.2 is a clarification that revises the specification to reflect the latest 10 CFR 50 Appendix J requirements and any NRC approved exemptions. On November 15, 1988, 10 CFR 50 Appendix J was amended to permit the use of a new statistical data analysis technique that the NRC considers to be an acceptable method of calculating containment leakage rates. This change to Appendix J has not been incorporated into BFN Technical Specification. Thus, the proposed change revises Specifications 4.7.A.2 to be consistent with the latest 10 CFR 50 Appendix J requirements and revised the specification to reflect that Appendix J requirements may be altered by approved exemptions.

The proposed change deletes two references to the Rod Sequence Control System (RSCS) that were not deleted in TS-310 submittal on the RSCS dated July 20, 1992. The evaluation and no significant hazards consideration provided by the July 20, 1992, submittal are applicable to these changes.

The proposed change deletes the reference to the "loss of condenser vacuum" in the Bases 3.1 for Units 1 and 3. The loss of condenser vacuum scram feature was deleted by Amendment Nos. 118, 113, and 89. However, the Bases for Units 1 and 3 were not revised to reflect the deletion while the Unit 2 Bases were issued correctly. Since the scram function has been deleted from the technical specifications, the Bases is revised to be consistent with the specification.

For Units 2 and 3, the reference in the Bases 3.5 to "BFN FSAR subsection 6.7" is incorrect. The correct FSAR reference is to "Section 6.0."

I' 4

ENCLOSURE 2 (CONTINUED)

Page 6 of 6

9. Specification 3.6.B.3 should reflect the two requirements (i.e., time limits and maximum quality limits) that could result in a reactor shutdown. The changing of the word "of" to "or" corrects the specification to accurately reflect the two requirements.

10. For Units 2 and 3, Specification 4.7.E.l is revised to correct a typographical error of two "to" in the fifth line.

11. An editorial change is made to the Table of Contents to include titles and page number of specifications 3.4.D/4.4.D as found in the Specification. Editorial changes are made to add/correct the heading for specifications 3.3.E/4.3.E, 3.3.F/4.3.F, 3.4.D/4.4.D, 3.6.D/4.6.D and 4.7.F and renumber specifications 3.4.E/4.4.E,

12. Editorial changes to upper or lowercase are made for consistence to words such as inoperable, operable, operability, etc.

13. The proposed changes clarify the requirement to use the applicable amendment of NEDE-24011-P-A as specified in the CORE OPERATING LIMITS REPORT. This clarification applies to TS-309 submitted on August 20, 1992. The evaluation and no significant hazards consideration provided by the August 20, 1992, submittal are applicable to this change.

A

\

V V

ENCLOSURE 3 BROMNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 (TVA BFN TECHNICAL SPECIFICATION NO. 331)

PROPO ED SIG IFICA T HAZARDS CO SIDERATIO S DETER I ATIO DESCRIPTION 0 E 0 OS D ECHNICAL SPECI CATION CHA G The BFN technical specifications are being revised as follows:

1. For Units 1, 2, and 3, revise Specification 3.6.C.2 to require the leak detection systems to be 'operable when the leak rate limits are required to be met.

2. For Unit 1, revise the Bases 3,9 to reflect that the loss of a 250V shutdown board battery affects the control power for both 480V and 4160V shutdown boards.

3. For Unit 2, delete Unit 2 cycle 6 only requirements in Table 3.2.C Note 7, Bases 3.2, and Specifications 3.5.K and 4.5.K.2.

4. For Units 1, 2, and 3, delete Specification 6.6, "Reportable Events Action."

5. For Units 1, 2, and 3, revise Specification 4.7.A.2 to reflect the latest 10 CFR 50 Appendix J requirements and any NRC approved exemptions.

6. For Units 1, 2, and 3, delete reference to Rod Sequence Control System (RSCS) from definition 1.M Note 4 and Bases 2.1.2.

7. For Units 1 and 3, delete reference to "loss of condenser vacuum" in Bases 3.1,

8. For Units 2 and 3, revise FSAR reference to "Section 6.0" in Bases 3.5.

9.'or Units 1, requirements 2, and 3, revise Specification 3.6.BE (i.e., time limits and maximum 3 to reflect the two quality limits) that could result in a reactor shutdown.

10. For Units 2 and 3, correct typographical error of two "to" in Specification 4.7.E.1.

11. For Units 1, 2, and 3, revise the headings for specifications 3.3.E/4.3.E, 3.3.F/4.3.F, 3.4.D/4.4.D, 3.6.D/4.6.D and 4,7.F, add title and page number of Specifications 3.4.D/4.4.D to Table of Contents and renumber specifications 3.4.E/4.4.E.

ENCLOSURE 3 (CONTINUED)

Page 2 of 3

12. For Units 1, 2, and 3, editorial changes to upper or lowercase to words such as inoperable, operable.

13. For Units 1, 2, and 3, revise Specification 6.9.1.7.b and reference 2 in Bases 2.1 to include the requirement to use the applicable amendment of NEDE-24011-P-A.

BASES FOR RO OSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMIN ION NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.91(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, or (2) create the possibility of a new or different kind of accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The proposed TS change is fudged to involve no significant hazards considerations based on the following:

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

The proposed changes are administrative in nature. They are being made to delete Unit 2 cycle 6 only requirement, to correct administrative errors in previous technical specifications, and to correct discrepancies between technical specification bases and the BFN FSAR. They also include the clarification of some requirements to ensure consistent application throughout the specifications. These changes do not affect any of the design basis accidents. They do not involve an increase in the probability or consequences of an accident previously evaluated.

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

The proposed changes are administrative in nature. They are being made to delete a Unit 2 cycle 6 only requirement, to correct administrative errors in previous technical specifications, and to correct discrepancies between technical specification bases and the BFN FSAR. They also include the clarification of some requirements to ensure consistent application throughout the specifications. No modifications to any plant equipment are involved, There are no effects on system interactions made by these changes. The changes will correct the technical specifications so that they are more accurate and more closely reflect actual plant condition.

They do not create the possibility of a new or different kind of accident from an accident previously evaluated.

ENCLOSURE 3 (CONTINUED)

Page 3 of 3

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

The proposed changes are administrative in nature. They delete a Unit 2 cycle 6 only requirement,'orrect administrative errors in previous technical specifications, and correct discrepancies between technical specification bases and the BFN FSAR. They also include the clarification of some requirements to ensure consistent application through the specifications. No safety margins are affected by these changes.

CONCLUSION TVA has evaluated the proposed amendment described above against the criteria given in 10 CFR 50.92(c) in accordance with the requirements of 10 CFR 50.91(a)(l). This evaluation has determined that the proposed amendment will got (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility for a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. Thus, TVA has concluded that the proposed amendment does not involve a significant hazards consideration.

~ - < s -, r l

~t