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:
Shared Package
ML18036B190	List: ML18036B189 ML18036B191
References
NUDOCS 9303230034
Download: ML18036B191 (86)

Text

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 ii V

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

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

P 3.4/4.4 D.

Reactivity Anomalies E.

Reactivity Control F.

Scram Discharge Volume Standby Liquid Control System A.

Normal System Availability.

3.3/4.3-11 3.3/4.3-12 3.3/4.3-12 3.4/4.4-1 3.4/4.4-1 B.

Operation with Inoperable Components 3.4/4.4-3 3'/4.5 C.

Sodium Pentaborate Solution.

D.

Standby Liquid Control System Requirements Core and Containment Cooling Systems.

A.

Core Spray System (CSS).

3.4/4.4-3 3.4/4.4-4 3.5/4.5-1 3.5/4.5-1 B.

Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling)

C.

RHR Service Water and Emergency Equipment Cooling Water Systems (EECWS).

D.

Equipment Area Coolers 3.5/4.5-4 3.5/4.5-9 3.5/4.5-13 E.

High Pressure Coolant Injection System (HPCIS) ~

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3.5/4.5-13 F.

Reactor Core Isolation Cooling System (RCICS) ~

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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 3.6/4.6 J.

Linear Heat Generation Rate (LHGR)

K.

Minimum Critical Power Ratio (MCPR).

L.

APRM Setpoints Primary System Boundary A.

Thermal and Pressurization Limitations B.

Coolant Chemistry.

C.

Coolant Leakage.

D.

Relief Valves.

3.5/4.5-18 3.5/4.5-19 3.5/4.5-20 3.6/4;6-1 3.6/4.6-1 3.6/4.6-5 3.6/4.6-9 3.6/4.6-10 BFH Unit 1

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6 ~ 0 1

6.2.1 6.2.2 Offsite and Onsite Organizations.........................

lant Staff..............................................

P

6. 0-1 6.0-2

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6 ~ 0 5 6.5

~ 1 6.5.2 6.5.3 Technical Review and Approval of Procedures..............

Deleted)................................................

(

6.0-17 6.0-,18 Plant Operations Review Committee (PORC).................

6.0-5 Nuclear Safety Review Board (NSRB).......................

6.0-11

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6 ~ 0 19 6.8.1 6-8.2 6.8.3 6.8.4

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Procedures

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6.0-20 6.0-20 Drills...................................................

6.0-21 Radiation Control Procedures.............................

6.0-22 Quality Assurance Procedures Effluent and Environmental Monitoring...........................

6.0-23

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6.0-24 6.9.1 Routine Reports..........................................

Startup Reports..........................................

Annual Operating Report..................................

Monthly Operating Report.................................

Reportable Events........................................

Radioactive Effluent Release Report......................

6.0-24 6.0-24 6.0-25 6.0-26 6.0-26 6.0-26 6.9.2 2

Source Tests

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N 6.0-26 6.0-27 6.0-29

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6 ~ 0 32 Specxal Reports...........................................

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6 ~ 0 33 BPN Unit 1 v

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.

(2)(3)(

4.

• "'"'"' ""'"'I"I

'"'eactor 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.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 Unit 1 1.0-4

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 Unit 1 1.1/2.1-13

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 Unit 1 1.1/2.1-16

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

'aterials 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 Unit 1 3.1/4.1-15

333.E I-If Specifications 3.3.C and

.D above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.3.E v

Surveillance requirements areL as specified in 4.3.C and

.D above.

D'.3.F 1.

The scram discharge volume drain and vent valves shall be OPERABLE any time that the reactor protection system is required to be OPERABLE except as specified in 3.3.F.2.

l.a.

The scram discharge volume drain and vent valves shall be verified open PRIOR TO STARTUP and monthly thereafter.

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.NM.

2.

In the event any SDV drain or vent valve becomes inoperable, REACTOR POWER OPERATION may continue provided the redundant drain or vent valve is OPERABLE.

2 ~

When it is determined that any SDV drain or vent valve is inoperable, the redundant drain or vent valve shall be demonstrated OPERABLE immediately and weekly thereafter.

3.

If redundant drain or vent valves become inoperable, the reactor shall be in HOT STANDBX CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

No additional surveillance required.

BFN Unit 1 3.3/4.3-12

.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 t

4.4.D khan~ imisLGa~tnl The Standby Liquid Control System conditions must satisfy the following equation.

y 1 (13 wt.X)(86 gpm)(19.8 atom%)

where, Verify that the equation given in Specification 3.4.D is satisfied at least once per month and within 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 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 3.4.D cannot be met, make at 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 />'.

No additional surveillance required.

BPN Unit 1 3.4/4.4-4

N 3.6.B 4.6.B 3.

At steaming rates greater than 100,000 lb/hr, the reactor water quality may exceed Specification 3.6.B.2 only for the time limits specified below.

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

a.

Conductivity time above 1 Pmho/cm at 25'C 2 weeks/year.

Maximum Limit 10 Pmho/cm at 25'C 3.

Whenever the reactor is operating (including HOT STANDBY CONDITION) measurements of reactor water quality shall be performed according to the following schedule'.

a.

Chloride ion content shall be measured at least once every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

b.

Chlorid'e ion content shall be measured at least every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> whenever reactor conductivity is

>1.0 pmho/cm at 25'C.

b.

Chloride concentration time above 0.2 ppm 2 weeks/year.

Maximum Limit 0.5 ppm.

c.

A sample of primary coolant shall be measured for pH at least once every 8

hours whenever the reactor coolant 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 Unit 1 3.6/4.6-6

4 3.6.C t

k 4.6.C nt 2.

Anytime 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.

With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.6.D 1.

When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

1.

Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.

All 13 valves will have been checked or replaced upon the completion of every 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 Unit 1 3.6/4.6-10

4 NG 3.7.A.

t mnt

4. 7.A.

t nm nt 2.a.

Primary containment integrity shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the reactor vessel except while performing "open vessel" physics tests at power levels not to exceed 5 m(t).

b. Primary containment integrity is confirmed if the maximum allowable integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the 49.6 psig design basis accident pressure,'a.

c. If N2 makeup to the primary containment averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for

pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced 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 placed in Hot Shutdown within the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

2 ~

t'n Primary containment nitrogen consumption shall be monitored to determine the average daily nitrogen consumption for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Excessive leakage is indicated by a N2 consumption rate of

> 2X of the primary containment free 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 temperature,

pressure, and venting operations) at 49.6 psig.

Corrected to normal drywell operating pressure of 1.1 psig, this value is 542 SCFH.

If this value is exceeded, the action specified in 3.7.A.2.C shall be taken.

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.

a.

Three type A tests (overall integrated 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 Unit 1 3.7/4.7-3

4.

LSt

.h t

t D

4.7.G.

t't t

t' 2 0 The Containment Atmosphere Dilution (CAD) System shall be OPERABLE whenever the reactor is in the RUN NODE.

2.

When FCV 84-8B is inoper-

able, each solenoid operated air/nitrogen valve of System B shall 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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 Unit 1 3.7/4.7-23

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 Unit 1 3.9/4.9-20

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 Unit 1 6.0-18,

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 Unit 1 6.0-26a

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

UNIT 2 EFFECTIVE PAGE LIST REMOVE iiiii V

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

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

>'<Denotes Spi11-Over Pages

D.

Reactivity Anomalies E.

Reactivity Control F,.

Scram Discharge Volume 3.3/4.3-11 3.3/4.3-12 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)

K.

Minimum Critical Power Ratio (MCPR).

3.5/4.5-18 3.5/4.5-19 L.

APRM Setpoints 3.5/4.5-20 3.6/4.6 M.

Core Thermal-Hydraulic Stability Primary System Boundary A.

Thermal and Pressurization Limitations 3.5/4.5-20 3.6/4.6-1 3.6/4.6-1 B.

Coolant Chemistry.

3-6/4.6-5 BPN Unit 2

t t

C.

Coolant Leakage.

D.

Relief Valves.

F.

Recirculation Pump Operation G.

Structural Integrity H.

Snubbers 3.7/4.7 Containment Systems A.

Primary Containment.

B.

Standby Gas Treatment System C.

Secondary Containment.

D.

Primary Containment Isolation Valves E.

Control Room Emergency Ventilation F.

Primary Containment Purge System G.

H.

Containment Atmosphere Dilution System (CAD)

Containment Atmosphere Monitoring (CAN)

System H2 Analyzer

3. 8/4. 8

'adioac tive Materials A.

Liquid Effluents B.

Airborne Effluents C.

Radioactive Effluents Dose D.

Mechanical Vacuum Pump E.

Miscellaneous Radioactive Materials Sources.

F.

Solid Radwaste 3-9/4.9 Auxiliary Electrical System A.

Auxiliary Electrical Equipment B..

Operation with Inoperable Equipment.

C.

Operation in Cold Shutdown D.

Unit 3 Diesel Generators Required for Unit 2 Operation E.

Jet Pumps

~

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~

~

o

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~

~

~

~

~

~

~

~

~

~

~

3.6/4.6-9 3.6/4.6-10 3.6/4.6-11 3.6/4.6-12 3.6/4.6-13 3.6/4.6-15 3.7/4.7-1 3.7/4.7-1 3.7/4.7-13 3.7/4.7-16 3.7/4.7-17 3.7/4.7-19 3.7/4.7-21 3.7/4.7-22 3.7/4.7-24 3.8/4.8-1 3.8/4.8-1 3.8/4.8-3 3.8/4.8-6 3.8/4.8-6 3.8/4.8-7 3.8/4.8-9 3.9/4.9-1 3.9/4.9-1 3.9/4.9-8 3.9/4.9-15 3.9/4.9-15a BPK Unit 2

SZXXQH

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6 ~ 0 1

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6 ~ 0 1

6.2.1 6.2.2 lant Staff..............................................

P 6.0-2 Offsite and Onsite Organizations...

~.....................

6.0-1

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6 ~ 0 5

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6. 0-5

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6 ~ 0 5 6.5.1 6.5.2 6.5.3 Technical Review and Approval of Procedures..............

Deleted)

(

~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 ~ ~ ~ ~

~ ~

~ ~ ~ ~

6. 0-17 6.0-18 6.0-19 Plant Operations Review Committee (PORC).................

6.0-5 Nuclear Safety Review Board (NSRB).......................

6.0-11

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~

6 ~ 0 20 6.8.1 6.8.2 6.8.3 Procedures.....-........-.-..........-

Drills~ ~ ~ ~

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

Radiation Control Procedures.............................

6.0-20 6.0-21 6.0-22 6.8.4 guality Assurance Procedures Effluent and Environmental Monitoring............................

~. 6.0-23 6.8.5 Programs

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~ 6.0-23

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6 '

24 6.9.1 Routzne Reports...............................;..........

Startup Reports..........................................

Annual Operating Report..................................

Monthly Operating Report.................................

Reportable Events.......

6.0-24 6.0-24 6.0-25 6.0-26 6.0-26 Radioactive Effluent Release Report................

~..... 6.0-26 6-9.2 Source Tests.............................................

Special Reports.............-............................

6.0-26 6.0-27

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6 ~ 0 29

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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 Unit 2 1.0-4

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 Unit 2 1.1/2.1-13

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 Unit 2 1.1/2.1-16

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 Unit 2 3.2/4.2-26

(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 Unit 2 3.2/4.2-27

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 Unit 2 3.2/4.2-68

4.3.E.

t'v't If Specifications 3.3.C and

.D above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Surveillance requirements are as specified in 4.3.C and

.D above.

3.3.F.

4.3.F.

1.

The scram discharge volume drain and vent valves shall be OPERABLE any time that the reactor protection system is required to be OPERABLE except as specified in 3.3.F.2.

l.a.

The scram discharge volume drain and vent valves shall be verified open PRIOR TO STARTUP and monthly thereafter.

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 or vent valve becomes inoperable, REACTOR POWER OPERATION may continue provided the redundant drain or vent valve is OPERABLE.

2 ~

When it is determined that any SDV drain or vent valve is inoperable, the redundant drain or vent valve shall be demonstrated OPERABLE immediately and weekly thereafter.

3.

If redundant drain or vent valves become inoperable, the reactor shall be in HOT STANDBY CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

No additional surveillance required.

BFN Unit 2 3.3/4.3-12

.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 System conditions must satisfy the following equation.

Z 1 (13 wt.X)(86 gpm)(19.8 atom%)

where, Verify that the equation given in Specification 3.4.D is satisfied at least

'once per month and within 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 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 3.4.D cannot be met, make at 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 />.

1.

No additional surveillance required.

BFN Unit 2 3.4/4.4-4

3.5.K ~Q.

4.5.K r

The minimum critical power ratio (MCPR) as a function of scram time and core flow, shall be equal to or greater than shown in Figure 3.5.K-l multiplied by the Kf shown in Figure 3.5.2, where:

= 0 or W V

, whichever is 7

=

A 7 B greater

~A = 0-90 sec (Specification 3.3.C.1 scram time limit to 20K insertion from fully withdrawn) 1.

MCPR shall be determined daily during reactor power operation at y 25K rated thermal power and following any change in power level or distribution that would cause operation with a limiting control rod pattern as described in the bases for Specification ~

2.

The MCPR limit shall be determined for each fuel type 8X8, 8XSR,

PSXSR, from Figure 3.5.K-1, respectively, using:

7 B = 0.710+1.65 n

2 (0.053) [Ref.2]

7 ave

~G

7g number of surveillance rod tests performed to date in cycle (including BOC test).

Scram time to 20K insertion from fully withdrawn of the it rod.

~t Ltt number of active rods measured in Specification 4.3.C.l at BOC.

If at any time during steady-state operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 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.

a. 7

= 0.0 prior to initial scram time measurements for the cycle, performed in accordance with Specification 4.3.C.l.

b.

~as defined in Specifi-cation 3.5.K following the conclusion of'ach scram-time surveillance test re-quired by Specifications 4.3.C.1 and 4.3.C.2.

The determination of the limit must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of each scram-time surveillance required by Specification 4.3.C.

BFN Unit 2 3.5/4.5-19

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 Unit 2 3.5/4.5-27

4.

N 3.6.B.

4.6.B.

3 ~

a ~

Conductivity time above 1 Nmho/cm at 25'C 2 weeks/year.

Maximum Limit 10 Pmho/cm at 25'C At steaming rates greater than 100,000 lb/hr, the reactor water quality may exceed Specification 3.6.B.2 only for the time limits specified below.

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

3 ~

Whenever the reactor is operating-(including HOT STANDBX CONDITION) measurements of reactor water quality shall be performed according to the following schedule:

a.

Chloride ion content shall be measured at least once every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

b.

Chloride ion content shall be measured at least every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> whenever reactor conductivity is

>1.0 Pmho/cm at 25'C.

b.

Chloride concentration time above 0.2 ppm 2 weeks/year.

Maximum Limit 0.5 ppm.

c.

A sample of primary coolant shall be measured for pH at least once every 8

hours whenever the reactor coolant 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 Unit 2 3.6/4.6-6

N D

3.6.C 4.6.C 2.

Anytime 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.

With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1.

When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

4.6.D.

1.

Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.

All 13 valves will have been checked or replaced upon the completion of every 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 Unit 2 3.6/4.6-10

4.

3.7.A.

'm 4.7.A.

t '

2 ~ ao b.

c ~

Primary containment integrity shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the reactor vessel except while performing "open vessel" physics tests at power levels not to exceed 5 mW(t).

Primary containment integrity is confirmed if the maximum allowable integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the 49.6 psig design basis accident pressure, Pa.

If N2 makeup to the primary containment averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for

pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced 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 placed in Hot Shutdown within the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

2.

t t

Primary containment nitrogen consumption shall be monitored to determine the average daily nitrogen consumption for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Excessive leakage is indicated by a N2 consumption rate of 2X of the primary containment" free 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 temperature,

pressure, and venting operations) at 49.6 psig.

Corrected to normal drywell operating pressure of 1.1 psig, this value is 542 SCFH.

If this value is exceeded, the action specified in 3.7.A.2.C shall be taken.

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.

a.

Three type A tests (overall integrated containment 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 Unit 2 3.7/4.7-3

4 G

N 3.7.E.

4.7.E.

l.

Except as specified in Specification 3.7.E.3 below, both control room emergency pressurization systems shall be OPERABLE at all times when any reactor vessel contains irradiated fuel.

1.

At least once every 18 months, the pressure drop across the combined HEPA filters and charcoal adsorber banks shall be demonstrated to be less than 6 inches of water at system design flow rate (g 10K).

2 ~

ao The results of the inplace cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show p99X DOP removal and >99K halogenated hydrocarbon removal when tested in accordance with ANSI N510-1975.

2.

a.

The tests and sample analysis of Specification 3.7.E.2 shall be performed at least once per operating cycle or once every 18 months, whichever occurs first for standby service or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

b.

The results of laboratory carbon sample analysis shall show y90X radioactive methyl iodide removal at a velocity when tested in accordance with ASTM D3803 (130'C, 95K R.H.).

b.

Cold DOP testing shall be performed after each complete or partial replacement of the HEPA filter bank or after any structural maintenance 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.

BFH Unit 2 3.7/4.7-19

3.7.G.

tt t

t 4.7.G.

t t

h D

m 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5.

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

BFH Unit 2 3.7/4.7-23

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 Unit 2 6.0-18

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 Unit 2 6.0-26a

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

UNIT 3 EFFECTIVE PAGE LIST RENOVE iiiii iv V

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

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

iv+

V 1.0-4 1.1/2.1-13 1.1/2.1-16 3.1/4.1-14 3.3/4.3-12 3.4/4.4-4 3.5/4.5-30 3.6/4.6-6 3.6/4.6-10 3.7/4.7-3 3.7/4.7-19 3.7/4.7-23 6.0-18 6.0-26a

>'<Denotes Spill-Over Page

5rzJtun D.

Reactivity Anomalies E-Reactivity Control F.

Scram Discharge Uolume 3.3/4.3-11 3.3/4.3-12 3.3/4.3-12 3.4/4 '

3-5/4.5 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 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)

K.

Minimum Critical Power Ratio (MCPR).

L-APRM Setpoints 3.5/4.5-18 3.5/4.5-19 3.5/4.5-20 3'/4.6 B.

Coolant Chemistry.

C.

Coolant Leakage.

Primary System Boundary A.

Thermal and Pressurization Limitations 3.6/4.6-1 3.6/4.6-1 3.6/4.6-5 3.6/4.6-9 BFN Unit 3

D.

Relief Valves.

F.

Recirculation Pump Operation G.

Structural Integrity.

H.

Snubbers E

Je 't Pumps

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~

~

~

~

~

~

o

~

~

~

~

~

~

~

~

~

~

3.6/4.6-10 3.6/4.6-11 3.6/4.6-12 3.6/4.6-13 3.6/4.6-15 3.7/4.7 Containment Systems A.

Primary Containment.

3.7/4.7-1 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 3.8/4.8 3-9/4.9 H.

Containment Atmosphere Monitoring (CAM)

System H2 Analyzer Radioactive Materials I

A.

Liquid Effluents B.

Airborne Effluents C.

Radioactive Effluents Dose D.

Mechanical Vacuum Pump E.

Miscellaneous Radioactive Materials Sources F.

Solid Radwaste Auxiliary Electrical System A.

Auxiliary Electrical Equipment B.

Operation with Inoperable Equipment.

C.

Operation in Cold Shutdown Condition 3.7/4.7-23a 3.8/4.8-1 3.8/4.8-1 s.ai~.s-~

3.8/4.8-6 3.8/4.8-6 3.8/4.8-7 3.8/4.8-9 3.9/4.9-1 3.9/4.9-1 3.9/4.9-8 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 A.

Refueling Interlocks B.

Core Monitoring C.

Spent Fuel Pool Water D.

Reactor Building Crane E.

Spent Fuel Cask 3.10/4.10-1 3.10/4.10-1 3.10/4.10-4 3.10/4.10-7 3.10/4.10-8 3.10/4.10-9 F.

Spent Fuel Cask Handling-Refueling Floor.

3.10/4.10-9 B.

Fire Pumps and Water Distribution Mains C.

Spray and/or Sprinkler Systems D.

C02 System E.

Fire Hose Stations.

F.

Yard Fire Hydrants and Hose Houses G.

Fire-Rated Assemblies

~

~

3.11/4.11 Fire Protection Systems A.

Fire Detection Instrumentation 3.11/4.11-1 3.11/4.11-1 3.11/4.11-2 3.11/4.11-7 3.11/4.11-8 3.11/4.11-9 3.11/4.11-11 3.11/4.11-12 5 ~ 0 H.

Open Flames, Welding and Burning in Spreading Room.

Major Design Features 5.1

. Site Features 5.2 Reactor 5.3 Reactor Vessel 5.4 Containment 5.5 Fuel Storage 5.6 Seismic Design the Cable

~

~

~

~

3.11/4.11-13 5.0-1 5.0-1 5.0-1 5.0-1 5.0-1 5.0-1

'.0-2 BFN Unit 3 iv

~ ~ ~

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

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

6oO 1

~ ~ ~ ~ ~ ~ ~ ~ s

~ ~

~

~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~

~ ~ ~

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

6 ~ 0 1

6.2.1 6.2.2 lant Staff..............................................

P

~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~

Qg o ~ ~ ~

~ ~

~ ~ ~ ~ ~ ~

~ ~

~ ~ ~ ~ ~

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

XRh 6.0-2 6.0-5 6.0-5 Offsite and Onsite Organizations.........................

6.0-1

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

6 ~ 0 5 6.5.1 6.5.2 Plant Operations Review Committee (PORC).................

6.0-5 Nuclear Safety Review Board (NSRB)...................,....

6.0-11 6.5.3 Technical Review and Approval of Procedures..............

Deleted)................................................

(

6.0-17 6.0-18 6.0-19

~ ~ ~ ~

~

~ ~ ~ ~ ~ s

~ ~ ~ ~ ~ ~

~ ~ ~ ~

6 ~ 0 20 6.8.1 6-8.2 6.8.3 6.8.4 Procedures...............................................

ri1ls ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

D Radiation Control Procedures.............................

Quality Assurance Procedures Effluent and 6.0-20 6.0-21 6.0-22 Environmental Monitorang..............................

6.0-23

~

~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

6 ~ 0 24 6.9.1 Routine Reports..........................................

Startup Reports..........................................

Annual Operating Report..................................

Monthly Operating Report.................................

Reportable Events......

6.0-24 6.0-24 6.0-25 6.0-26 6.0-26 Radioactive Effluent Release Report......................

6.0-26 6.9.2 2

Specxal Reports..........................................

6.0-26 6.0-27 N

D

~

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~

~ ~ ~ ~ ~

~ ~ ~ ~

~ ~

6 ~ 0 29

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~

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~ ~

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~ o 6 '-32 L.

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~ ~ ~ ~ ~

~ ~

~ ~

La ~

~ ~

~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~

~ ~ ~ ~ o ~ ~ ~ ~ ~

~ ~

~ ~

A 6.0-32 6.0-33 Source Tests.............................................

BFN Unit 3 v

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 Unit 3 1.0-4

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 Unit 3 1.1/2.1-13

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 Unit 3 1.1/2.1-16

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 Unit 3 3.1/4.1-14

4 L

E

) 3.3.E.

4.3.E.

If Specifications 3.3.C and 3.3.D above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Surveillance requirements are as specified in 4.3.C and 4.3.D above.

[ 3.3.F.

1.

The scram discharge volume

, drain and vent valves shall be OPERABLE any time that the reactor protection system is required to be OPERABLE except as specified in 3.3.F.2.

4.3.F.

1 oa ~

The scram discharge volume drain and vent valves shall be verified open PRIOR TO STARTUP and monthly thereafter.

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 or vent valve becomes inoperable, REACTOR POWER OPERATION may continue provided the redundant drain or vent valve is OPERABLE.

2 ~

When it is determined that any SDV drain or vent valve is inoperable, the redundant drain or vent valve shall be demonstrated OPERABLE immediately and weekly thereafter'.

3.

If redundant drain or vent valves become inoperable, the reactor shall be in HOT STANDBY CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

No additional surveillance required.

BFN Unit 3 3.3/4.3-12

.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 System conditions must satisfy the following equation.

g 1

(13 wt.X)(86 gpm)(19.8 atom%)

where, Verify that the equation given in Specification 3.4.D is satisfied at least once per month and within 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 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 3.4.D cannot be met, make at 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 />.

1.

No additional

)

surveillance required.

BPN Unit 3 3.4/4.4-4

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 Unit 3 3.5/4.5-30

4 N

3.6.B.

4.6.B.

3.

At steaming rates greater than 100,000 lb/hr, the reactor water quality may exceed Specification 3.6.B.2 only for the time limits specified below.

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

3.

Additional coolant samples shall be taken whenever the reactor activity exceeds one percent of the equilibrium concentration specified in 3.6.B.5 and one of the following conditions are met.

a.

During the STARTUP CONDITION b.

Following a significant power change<<<<.

a.

Conductivity time above 2 Pmho/cm at 25'C 4 weeks/year.

Maximum Limit 10 pmho/cm at 25'C b.

Chloride concentration time above 0.2 ppm 4 weeks/year.

Maximum Limit 0.5 ppm.

c.

Following an increase in the equilibrium off-gas level exceeding 10,000 puci/sec (at the steam jet air ejector) within a 48-hour period.

d.

Whenever the equilibrium iodine limit specified in 3.6.B.5 is exceeded.

• • For the purpose of this section on 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 Unit 3 3.6/4.6-6

4.

2.

Anytime 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.

With the air sampling system inoperable, grab samples shall be obtained and analyzed at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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.

i 3.6.D.

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1.

When more than one relief valve is known to be failed, an orderly shutdown shall be initiated and the reactor depressurized to less than 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION.

4.6.D.

1.

Approximately one-half of all relief valves shall be bench-checked or replaced with a bench-checked valve each operating cycle.

All 13 valves will have been checked or replaced upon the completion of every 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 Unit 3 3.6/4.6-10

3.7.A.

r m t '

4.7.A.

t '

2.a.

Primary containment integrity shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the reactor vessel except while performing "open vessel" physics tests at power levels not to exceed 5 m(t).

b. Primary containment integrity is confirmed if the maximum allowable integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the 49.6 psig design basis accident pressure, Pa.

c. If N2 makeup to the primary containment averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for

pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced 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 placed in Hot Shutdown within the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

2 ~

t t

t t

Primary containment nitrogen consumption shall be monitored to determine the average daily nitrogen consumption for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Excessive leakage is indicated by a N2 consumption rate of 2X of the primary containment free 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 temperature,

pressure, and venting operations) at 49.6 psig.

Corrected to normal drywell operating pressure of 1.1 psig, this value is 542 SCFH.

If this value is exceeded, the action specified in 3.7.A.2.c shall be taken.

a.

Three type A tests (overall integrated containment leakage rate) shall be conducted at 40 p 10-month intervals during shutdown at Pa,'9.6 psig, during each 10-year plant inservice inspection.

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 exempti'ons.

BFN Unit 3 3.7/4.7-3

4.

3.7.E.

t t'.7.E.

l.

Except as specified in Specification 3.7.E.3 below, both control room emergency

'pressurization systems shall be OPERABLE at all times when any reactor vessel contains irradiated fuel.

l.

At least once every 18 months, the pressure drop across the combined HEPA filters and charcoal adsorber banks shall be demonstrated to be less than 6 inches of water at system design flow rate

(~ 10K).

2 ~

a ~

The results of the inplace

.cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show y99X DOP removal and y99X halogenated hydrocarbon removal when tested in accordance with ANSI N510-1975.

2.

a.

The tests and sample analysis of Specification 3.7.E.2 shall be performed at least once per operating cycle or once every 18 months, whichever occurs first for-standby service or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and following significant painting, fire, or chemical release in any ventilation zone communicating with the system.

b.

The results of laboratory carbon sample analysis shall show y90X radioactive methyl iodide removal at a velocity when tested in accordance with ASTM D3803 (130'C, 95K R.H.).

b.

Cold DOP testing shall be performed after each complete or partial replacement of the HEPA filter bank or after any structural maintenance 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 Unit 3 3.7/4.7-19

.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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5.

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

BPN Unit 3 3.7/4.7-23

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 Unit 3 6.0-18

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 Unit 3 6.0-26a

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."

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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.

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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, 2, and 3, revise Specification 3.6.BE 3 to reflect the two requirements (i.e., time limits and maximum 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.

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