Text

Proposed TS Re COLR
	ML18036A819
Person / Time
Site:	Browns Ferry
Issue date:	08/20/1992
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML18036A818	List: ML18036A817; ML18036A819; ML18036A820;
References
	TVA-BFN-TS-309, NUDOCS 9208280164
	Download: ML18036A819 (158)
	v • d • e

ENCLOSURE 1 PROPOSED TECHNICALSPECIFICATION CHANGE BROWNS kVVRYNUCLEAR PLANT UNITS 1, 2, AND 3 (TVABFN TS 309) 0

'P208280ih4 920820 PDR ADOCK 05000289 P

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PROPOSED TECHNICALSPECIFICATION CHANGE BROWNS FERRY NUCLEAR PLANT UNIT 1 (TVABFN TS 309)

UNIT 1 EFFECTIVE PAGE LIST REMOVE INSERT vvii viii 1.0-5 1.0-7 1.1/2.1-3 1.1/2.1-11 1.1/2.1-12 1.1/2.1-16 3.1/4.1-20 3.3/4.3-17 3.5/4.5-18 3.5/4.5-19 3.5/4.5-20 3.5/4.5-21 3.5/4.5-22 3.5/4.5-23 3.5/4.5-24 3.5/4.5-25 3.5/4.5-32 3.5/4.5-33 3.5/4.5-35 5.0-1 vvii vii 1.0-5 1.0-7 1.0-12a 1.0-12b 1.1/2.1-3 1.1/2.1-11 1.1/2.1-12 1.1/2.1-16 3.1/4.1-20 3.3/4.3-17 3.5/4.5-18 3.5/4.5-19 3.5/4.5-20 3.5/4.5-21 3.5/4.5-22 3.5/4.5-23 3.5/4.5-24 3.5/4.5-25 3.5/4.5-32 3.5/4.5-33 3.5/4.5-35 5.0-1 6.0-26a 6.0-26b

ADMINISTRATIVE CO ROLS

~SECTIO PAGE RESPONSIBILITY........o.....................

~ ~ ~..

~

ORGANIZATION........-.................-.......

6.0-1 6.0-1 6.2.1 6.2.2 lant Staff..............................................

P 6.0-2 Offsite and Onsite Organizations.........................

6.0-1 PLANT STAFF UALIFICATIONS...............................

6.0-5 TRAINING.................................................

PLANT REVIEW AND AUDIT...................................

6.0-5 6.0-5 6.5.1 6.5.2 6.5.3 6.6

~67 REPORTABLE EVENT ACTIONS.................................

SAFETY LIMITVIOLATION...................................

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

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

6.0-11 Technical Reviev and Approval of Procedures..............

6.0-17 PROCEDURES INSTRUCTIONS AND PROGRAMS.....................

6.0-20 6.8.1 6.8.2 6.8.3 6.8.4 Procedures...............................................

rills ~

~ ~ ~ ~

~ ~ ~

~ ~ ~ ~ i ~

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

~ ~ ~ o ~ o ~ ~

D 6.0-20 6.0-21 Radiation Control Procedures.............................

6.0-22 Quality Assurance Procedures Effluent and 6.9.1 Environmental Monitoring..............................

REPORTING RE UIREMENTS..................,............,...

outine Reports..........................................

R Startup Reports..........................................

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

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

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

6.0-23 6.0-24 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 ource Tests.............................................

S Core Operating Limits Report.............................

Special Reports..........................................

6.0-26 6.0-26a 6.0-27 STATION OPERATING RECORDS AND RETENTION..................

6.0-29 6.11 PROCESS CONTROL PROGRAM................,,................

6.0-32 OFFSITE DOSE CALCULATIONMANUAL.......................... 6.0-32 6.13 RADIOLOGICAL EFFLUENT MANUAL............................. 6.0-33 BFN Unit 1 v

LIST OF TABLES (Cont'd)

Tablet 4.2.E Title Minimum Test and Calibration Frequency for Drywell Leak Detection Instrumentation

~Pa e

No 3.2/4.2-53 4.2.F Minimum Test and Calibration Frequency for Surveillance Instrumentation 3.2/4.2-54 4.2.G Surveillance Requirements for Control Room Isolation Instrumentation.

3.2/4.2-56 4.2.H Minimum Test and Calibration Frequency for Flood Protection Instrumentation 3.2/4.2-57 4.2eJ Seismic Monitoring Instrument Surveillance Requirements 3.2/4.2-58 4.2eK Radioactive Gaseous Effluent Instrumentation Surveillance 3.2/4.2-62 4.2.L ATWS-Recirculation Pump Trip (RPT)

Instrumentation Surveillance 3.2/4.2-63a 3.5-1 Minimum RHRSW and EECW Pump Assignment 3.5/4.5-11 3.7.A Primary Containment Isolation Valves t

3.7.B Testable Penetrations with Double 0-Ring Seals 3.7.C Testable Penetrations with Testable Bellows 3.7/4.7-25 3.7/4.7-32 3.7/4.7-33 3.7.D Air Tested Isolation Valves 3.7/4.7-34 3.7.E Primary Containment Isolation Valves which Terminate below the Suppression Pool Water Level 3.7/4.7-37 3.7.F 3.7.H Testable Electrical Penetrations 4.9.A Auxiliary Electrical Systems Primary Containment Isolation Valves Located in Water Sealed Seismic Class 1 Lines 3.7/4.7-38 3.7/4.7-39 3.9/4.9-16 4.9.A.4.C Voltage Relay Setpoints/Diesel Generator Start 3.9/4.9-18 3.11.B Spray/Sprinkler Systems 3.11.C Hose Stations 3.11.D Yard Fire Hydrants t

6.2.A Minimum Shift Crew BFN Unit 1 and Fire Hose Houses.

Requirements.

vii 3.11.A Fire Detection Instrumentation 3.11/4.11-14 3.11/4.11-18 3.11/4.11-20 3.11/4.11-22 6.0-4

LIST OF ILLUSTRATIONS

~Pi ere Title

~Pe e Ne.

APRM Flow Reference Scram and APRM Rod Block Settings 1.1/2.1-6 2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow 1.1/2.1-7 4.1-1 Graphical Aid in the Selection of an Adequate Interval Between Tests 3.1/4.1-13 4.2-1 System Unavailability.

3.2/4.2-64 3.6-1 Minimum Temperature F Above Change in Transient Temperature.

3.6/4.6-24 3.6-2 Change in Charpy V Transition Temperature Vs.

Neutron Exposure.,'.

3.6/4.6-25 4.8.1.a Gaseous Release Points and Elevations.

3.8/4.8-10 4.8.l.b Land Site Boundary 3.8/4.8-11 BFN Unit 1 viii

DEFINITIONS (Cont'd)

N.

Rated Power Rated power refers to operation at a reactor power of 3,293 MWt; this is also termed 100 percent power and is the maximum power level authorized by the operating license.

Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters when the reactor is at rated power.

0.

Prima Containment Inte rit Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:.

1.

All nonautomatic containment isolation valves on lines connected to the reactor coolant systems or containment which are not required to be open during accident conditions are closed.

These valves may be opened to perform necessary operational activities.

2.

At least one door in each airlock is closed and sealed.

3.

All automatic containment isolation valves are OPERABLE or each line which contains an inoperable isolation valve is isolated as required by Specification 3.7.D.2.

4.

All blind flanges and manways are closed.

P.

Seconda Containment Inte rit 1.

Secondary containment integrity means that the required unit reactor zones and refueling zone are intact and the following conditions are met:

a)

At least one door in each access opening to the turbine building, control bay and out-of-doors is closed.

b)

The standby gas treatment system is OPERABLE and can maintain 0.25 inches of water negative pressure in those areas where secondary containment integrity is stated to exist.

c)

All secondary containment penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.

Closed by at least one secondary containment automatic isolation valve deactivated in the isolated position.

2.

Reactor zone secondary containment integrity means the unit reactor building is intact and the following conditions are met:

a)

At least one door between any opening to the turbine building, control bay and out-of-doors is closed.

BFN Unit 1

1. &-

a DEFINITIONS (Cont'd)

Q.

0 eratin C cle Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit.

R.

Refuelin Outa e Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling.

For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled outage;

however, where such outages occur within 8 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage.

S.

CORE ALTERATION The addition, removal, relocation, or movement of fuel, sources, in-core instruments, or reactivity controls within the reactor pressure vessel with the head removed and fuel in the vessel.

Normal movement of in-core instrumentation and the traversing in-core probe is not defined as a Core Alteration.

Suspension of Core Alterations shall not preclude completion of the movement of a component to a safe conservative position.

T.

Reactor Vessel Pressure Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.

U.

Thermal Parameters 1.

Minimum Critical Power Ratio MCPR Minimum Critical Power Ratio (MCPR) is the value of the critical power ratio associated with the most limiting assembly in the reactor core.

Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated to cause some point in the assembly to experience boiling transition, to the actual assembly operating power.

2.

Transition Boilin Transition boiling means the boiling regime between nucleate and film boiling.

Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

3.

Core Maximum Fraction of Limitin Power Densit CMFLPD The highest ratio, for all fuel assemblies and all axial locations in the core, of the maximum fuel rod power density (kW/ft) for a given fuel assembly and axial location to the limiting fuel rod power density (kW/ft) at that location.

4.

Avera e Planar Linear Heat Generation Rate APLHGR The Average Planar Heat Generation Rate is applicable to a specific planar height and is equal to the sum of the linear heat generation rates for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

BFN Unit 1

1.0 DEFINITIONS (Cont'd)

NN. Core 0 eratin Limits Re ort COLR The COLR is the unit-specific document that provides the core operating limits for the current operating cycle.

These cycle-specific core operating limits shall be determined for each operating cycle in accordance with Specification 6.9.1.7.

Plant operation. within these limits is addressed in individual specifications.

BFN Unit 1 1.0-12a

THIS PAGE INTENTIONALLYLEFT BLANK BFN Unit 1 1.0-12b

2 1

FUEL CLADDI G INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flux Tri Settin s

2.1.A.l.b (Cont'd)

NOTE:

These settings assume operation within the basic thermal hYdraulic design criteria.

These criteria are LHGR within the limits of Specification 3.5.J and MCPR within the limits of Specification 3.5.K. If it is determined that either of these design criteria is being violated during operation, action shall be initiated within 15 minutes to restore operation within prescribed limits.

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

c ~

The APRM Rod Block trip setting shall be:

SR~ (0'66W + 42%)

where:

SRB =

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

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

BFN Unit 1 1.1/2.1-3

2.1'ASES:

LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed in support of planned operating conditions up to the maximum thermal power of'293 MWt.

The analyses were based upon plant operation in accordance with Reference l.

In addition, 3293 MWt is the licensed maximum power level for each Browns Ferry Nuclear Plant unit, and this represents the maximum steady-state power which shall not be knowingly exceeded.

The transient analyses performed for each reload are described in Reference 2.

Models and model conservatisms are also described in this reference.

BFN Unit 1 1.1/2.1-11

~ II

2. 1'ASES (Cont'd)

The bases for individual setpoints are discussed below:

A.

Neutron Flux Scram 1.

APRM Flow-Biased Hi h Flux Scram Tri Settin Run Mode The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3,293 MWt).

Because fission chambers provide the basic input signals, the APRM system responds directly to core average neutron flux.

During transients, the instantaneous fuel surface heat flux is less than the instantaneous neutron flux by an amount depending upon the duration of the transient and the fuel time constant.

For this reason, the flow-biased scram APRM flux signal is passed through a filtering network with a time constant which is representative of the fuel time constant.

As a result of this filtering, APRM flow-biased scram will occur only if the neutron flux signal is in excess of the setpoint and of sufficient time duration to overcome the fuel time constant and result in an average fuel surface heat flux which is equivalent to the neutron flux trip setpoint.

This setpoint is variable up to 120 percent of rated power based on recirculation drive flow according to the equations given in Section 2.1.A.l and the graph in Figure 2.1-2.

For the purpose of licensing transient analysis, neutron flux scram is assumed to occur at 120 percent of rated power.

Therefore, the flow biased provides additional margin to the thermal limits for slow transients such as loss of feedwater heating.

No safety credit is taken for flow-biased scrams.

Unit 1 1.1/2.1-12

2.1'ASES (Cont'd)

F.

(Deleted) t G.

& H.

Main Steam Line Isolation *on Low Pressure and Main Steam Line Isolation Scram 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. Reactor Low Water Level Set oint for Initiation of HPCI and RCIC Closin Main Steam Isolation Valves and Startin LPCI and Core 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.

References 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 (latest approved version).

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 ODYN Computer Model,"

September 5, 1980.

Unit 1 1.1/2.1-16

4.1'ASES (Cont'd)

The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate.

The APRM system, which uses the LPRM readings to detect a change in thermal power, will be calibrated every seven days using a heat balance to compensate for this change in sensitivity.

The RBM system uses the LPRM reading to detect a localized change in thermal power. It applies a correction factor based on the APRM output signal to determine the percent thermal power and therefore any change in LPRM sensitivity is compensated for by the APRM calibration.

The technical specification limits of CMFLPD,

CPR, and APLHGR are determined by the use of the process computer or other backup methods.

These methods use LPRM readings and TIP data to determine the power distribution.

Compensation in the process computer for changes in LPRM sensitivity will be made by performing a full core TIP traverse to update the computer calculated LPRM correction factors every 1000 effective full power hours.

As a minimum the individual LPRM meter readings will be adjusted at the beginning of each operating cycle before reaching 100 percent power.

BFN Unit 1 3.1/4.1-20

"3.3/4.3 BASES (Cont'd) 5.

The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation.

Two RBM channels are provided, and one of these may be bypassed from the console for maintenance and/or testing.

Automatic rod withdrawal blocks from one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.

The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.

A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit, (i.e.,

MCPR given by Specification 3.5.K or LHGR given by Specification 3.5.J).

During use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods to assure its OPERABILITY will assure that improper withdrawal does not occur. It is normally the responsibility of the nuclear engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns.

Other personnel qualified to perform these functions may be designated by the plant superintendent to perform these functions.

C.

Scram Insertion Times The control rod system is designated to bring the reactor subcritical at the rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than 1.07.

The limiting power transient is given in Reference 1.

Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification provide the required protection, and MCPR remains greater than 1.07.

On an early BWR, some degradation of control rod scram performance occurred during plant startup and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter.

The design of the present control rod drive (Model 7RDB144B) is grossly improved by the relocation of the filter to a location out of the scram drive path; i.e., it can no longer interfere with scram performance, even if completely blocked.

The degraded performance of the original drive (CRD7RDB144A) under dirty operating conditions and the insensitivity of the redesigned drive (CRD7RDB144B) has been demonstrated by a series of engineering tests under simulated reactor operating conditions.

The successful performance of the new drive under actual operating conditions has also been demonstrated by consistently good in-service test results for plants using the new drive and may be inferred from plants using the older model BFN Unit 1 3.3/4.3-17

5'.

CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION t

3.5.I Avera e Planar Linear Heat Generation Rate During steady-state power operation, the Average Planar Linear Heat Generation Rate (APLHGR) of any fuel assembly at any axial location shall not exceed the appropriate APLHGR limit provided in the CORE OPERATING LIMITS REPORT. If at any time during steady state operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR 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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

SURVEILLANCE RE UIREMENTS 4.5.I Avera e Planar Linear Heat Generation Rate

~APLHGR I

The APLHGR shall be checked daily during reactor operation at g 25% rated thermal power.

J.

Linear Heat Generation Rate LHGR J.

Linear Heat Generation Rate

~LHGR During steady-state power operation, the linear heat generation rate (LHGR) of any rod in any fuel assembly at any axial location shall not exceed the appropriate LHGR limit provided in the CORE OPERATING LIMITS REPORT.

The LHGR shall be checked daily during reactor operation at g 25% rated thermal power.

BFN Unit 1 3.5/4.5-18

5 CORE AND CONTAINMENT COOLING SYSTE MS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 3.5.J

~

(Cont'd)

If at any time during steady-state operation it is determined by normal surveillance that the limiting value for LHGR is being

exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR 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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

3.5.K Minimum Critical Power Ratio

~MCPR 4.5.K Minimum Critical Power The minimum critical power ratio (MCPR) shall be equal to or greater than the operating limit MCPR (OLMCPR) as provided in the CORE OPERATING LIMITS REPORT.

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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

MCPR shall be checked daily during reactor power operation at g 25/ 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 ~3.

2.

The MCPR limit at rated flow and rated power shall be determined as provided in the CORE OPERATING LIMITS REPORT using:

a.

Q as defined in the CORE OPERATING LIMITS REPORT prior to initial scram time measurements for the cycle, performed in accordance with Specification 4.3.C.1.

BFN Unit 3.5/4.5-19

5/4 5

CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS 4.5.K.2 (Cont'd) as defined in the CORE OPERATING LIMITS REPORT following the conclusion of each scram-time surveillance test required by Speci-fications 4.3.C.l and 4.3.C.2.

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

L.

APRM Set pints L.

APRM Set pints 1.

Whenever the core thermal power is g 25% of rated, the ratio of FRP/CMFLPD shall be g 1.0, or the APRM scram and rod block setpoint equations listed in Sections 2.1.A and 2.1.B shall be multiplied by FRP/CMFLPD as follows:

FRP/CMFLPD shall be determined daily when the reactor is g 25% of rated thermal power.

SZ (O.66W + 54%)

CMFLPD SR~ (0.66W + 42%) (

)

CMFLPD 2.

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

3. If 3.5.L.l and 3.5.L.2 cannot be met, the reactor power shall be reduced to g 25/ of rated thermal power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

BFN Unit 1 3.5/4.5-20

Table 3.5.I-l DELETED Table 3.5.I-2 DELETED Unit 1 3.5/4.5-21

Table 3.5.I-3 DELETED Table 3.5.I-4 DELETED BFN Unit l 3.5/4.5-22

0

0 Table 3.5.I-5 DELETED Table 3.5.I-6 DELETED BFN Unit 1 3.5/4.5-23

Figure 3.5.K-1 DELETED BFN Unit 1 3.5/4.5-24

Figure 3.5.2 DELETED BFN Unit 1 3.5/4.5-25

3.5'ASES (Cont'd)

Because the automatic depressurization system does not provide makeup to the reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further conservatism to the CSCS.

With two ADS valves known to be incapable of automatic operation, four valves remain OPERABLE to perform their ADS function.

The ECCS loss-of-coolant accident analyses for small line breaks assumed that four of the six ADS valves were OPERABLE.

Reactor operation with three ADS valves inoperable is allowed to continue for seven days provided that the HPCI system is OPERABLE.

Operation with more than three of the six ADS valves inoperable is not acceptable.

H.

Maintenance of Filled Dischar e Pi e

If the discharge piping of the core spray, LPCI, HPCIS, and RCICS are not filled, a water hammer can develop in this piping when the pump and/or pumps are started.

To minimize damage to the discharge piping and to ensure added margin in the operation of these

systems, this Technical Specification requires the discharge lines to be filled whenever the system is in an OPERABLE condition. If a discharge pipe is not filled, the pumps that supply that line must be assumed to be inoperable for Technical Specification purposes.

The core spray and RHR system discharge piping high point vent is visually checked for water flow once a month and prior to testing to ensure that the lines are filled.

The visual checking will avoid starting the core spray or RHR system with a discharge line not filled.

In addition to the visual observation and to ensure a filled discharge line other than prior to testing, a pressure suppression chamber head tank is located approximately 20 feet above the discharge line high point to supply makeup water for these systems.

The condensate head tank located approximately 100 feet above the discharge high point serves as a

backup charging system when the pressure suppression chamber head tank is not in service.

System discharge pressure indicators are used to determine the water level above the discharge line high point.

The indicators will reflect approximately 30 psig for a water level at the high point and 45 psig for a water level in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge lines are filled.

When in their normal standby condition, the suction for the HPCI and RCIC pumps are aligned to the condensate storage tank, which is physically at a higher elevation than the HPCIS and RCICS piping.

This assures that the HPCI and RCIC discharge piping remains filled.

Further assurance is provided by observing water flow from these systems'igh points monthly.

Avera e Planar Linear Heat Generation Rate APLHGR This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K.

BFN Unit 3.5/4.5-32

3.5'ASES (Cont'd)

The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod-to-rod power distribution within an assembly.

Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than g 20 F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 Appendix K limit.

3.5.J. Linear Heat Generation Rate LHGR This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densification is postulated.

The LHGR shall be checked daily during reactor operation at g 25 percent power to determine if fuel burnup, or control rod movement has caused changes in power distribution.

For LHGR to be a limiting value below 25 percent of rated thermal power, the largest total peaking would have to be greater than approximately 9.7 which is precluded by a considerable margin when employing any permissible control rod pattern.

3.5.K. Minimum Critical Power Ratio MCPR At core thermal power levels less than or equal to 25 percent, the reactor will be operating at m'inimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin.

With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.

The daily requirement for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

3.5.L.

APRM Set pints The fuel cladding integrity safety limits of Section 2.1 were based on a total peaking factor within design limits (FRP/CMFLPD y 1.0).

The APRM instruments must be adjusted to ensure that the core thermal limits are not exceeded in a degraded situation when entry conditions are less conservative than design assumptions.

BFN Unit 1 3.5/4.5-33

4.5 Core and Containment Coolin S stems Surveillance Fre uencies The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality.

The core cooling systems have not been designed to be fully testable during operation.

For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable.

Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory.

To increase the availability of the core and containment cooling system, the components which make up the system, i.e.,

instrumentation, pumps, valves, etc.,

are tested frequently.

The pumps and motor operated injection valves are also tested in accordance with Specification 1.0.MM to assure their OPERABILITY.

A simulated automatic actuation test once each cycle combined with testing of the pumps and injection valves in accordance with Specification 1.0.MM is deemed to be adequate testing of these systems.

Monthly alignment checks of valves that are not locked or sealed in position which affect the ability of the systems to perform their intended safety function are also verified to be in the proper position.

Valves which automatically reposition themselves on an initiation signal are permitted to be in a position other than normal to facilitate other operational modes of the system.

When components and subsystems are out-of-service, overall core and containment cooling reliability is maintained by OPERABILITY of the remaining redundant equipment.

Whenever a

CSCS system or loop is made inoperable, the other CSCS systems or loops that are required to be OPERABLE shall be considered OPERABLE if they are within the required surveillance testing frequency and there is no reason to suspect they are inoperable.

If the function, system, or loop under test or calibration is found inoperable or exceeds the trip level setting, the LCO and the required surveillance testing for the system or loop shall apply.

Avera e Planar LHGR LHGR and MCPR The APLHGR, LHGR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution.

Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate.

BFN Unit 3.5/4.5-35

5.0.

MAJOR DESIGN FEATURES 5.1 SITE FEATURES Browns Ferry unit 1 is located at Browns Ferry Nuclear Plant site on property owned by the United States and in custody of the TVA.

The site shall consist of approximately 840 acres on the north shore of Wheeler Lake at Tennessee River Mile 294 in Limestone County, Alabama.

The minimum distance from the outside of the secondary containment building to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 4,000 feet.

5.2 REACTOR A.

The reactor core may contain 764 fuel assemblies.

B.

The reactor core shall contain 185 cruciform-shaped control rods.

5.3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2-2 of the FSAR.

The applicable design codes shall be as described in Table 4.2-1 of the FSAR.

5.4 CONTAINMENT A.

The principal design parameters for the primary containment shall be as given in Table 5.2-1 of the FSAR.

The applicable design codes shall be as described in Section 5.2 of the FSAR.

B.

The secondary containment shall be as described in Section 5.3 of the FSAR.

C.

Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with the standards set forth in Section 5.2.3.4 of the FSAR.

5.5 FUEL STORAGE A.

The arrangement of fuel in the new-fuel storage facility shall be such that keff, for dry conditions, is less than 0.90 and flooded is less than 0.95 (Section 10.2 of FSAR).

BFH Unit 1

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."

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 1 6.0-26a

THIS PAGE INTENTIONALLYLEFT BLANK BFN Unit 1 6.0-26b

PROPOSED TECHNICALSPECIFICATION CHANGE BROWNS FFWRY NVCLEARPLANT UNIT2 (TVABFN TS 309)

UNIT 2 EFFECTIVE PAGE LIST REMOVE INSERT vvii viii 1.0-5 1.0-7 1.0-12a 1.1/2.1-3 1.1/2.1-11 1.1/2.1-12 1.1/2.1-15 1.1/2.1-16 3.1/4.1-20 3.3/4.3-17 3.5/4.5-18 3.5/4.5-19 3.5/4.5-21 3.5/4.5-2la 3.5/4.5-22 3.5/4.5-23 3.5/4.5-30 3.5/4.5-31 3.5/4.5-33 5.0-1 vvii vii 1.0-5 1.0-7 1.0-12a 1.1/2.1-3 1.1/2.1-11 1.1/2.1-12 1.1/2.1-15 1.1/2.1-16 3.1/4.1-20 3.3/4.3-17 3.5/4.5-18 3.5/4.5-19 3.5/4.5-21 3.5/4.5-21a 3.5/4.5-22 3.5/4.5-23 3.5/4.5-30 3.5/4.5-31 3.5/4.5-33 5.0-1 6.0-26a 6.0-26b

ADMINISTRATIVE CONTROLS SECTION PAGE 6.2 6.2.1 6.2.2 ORGANIZATION.............................................

6.0-1 6.0-1 Offsite and Onsite Organizations............

Plant Staff.................................

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6 ~ 0 2 RESPONSIBILITY......,............,.......................

PLANT STAFF UALIFICATIONS...............................

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T PLANT REVIEW AND AUDIT...................................

6.0-5 6.0-5 6.5.1 6.5.2 6.5.3 Plant Operations Review Committee (PORC).................

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

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

6.0-17 REPORTABLE EVENT ACTIONS.................................

6.0-18 SAFETY LIMITVIOLATION...................................

6.0-19 PROCEDURES INSTRUCTIONS AND PROGRAMS.....................

6oO-20 6.8.1 6.8.2 6.8.3 6.8.4 rocedures...............................................

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

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

6.0-23 6.8.5 6.9.1 rograms

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P REPORTING RE UIREMENTS...................................

outine Reports..........................................

R tartup Reports..........................................

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

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

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

6.0-23 6.0-24 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 ource Tests.............................................

S Core Operating Limits Report.............................

pecial Reports..........................................

S 6.0-26 6.026a 6.0-27

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~1 STATION OPERATING RECORDS AND RETENTION..................

6.0-29 PROCESS CONTROL PROGRAM..................,...............

6.0-32 OFFSITE DOSE CALCULATIONMANUAL.......................... 6.0-32 RADIOLOGICAL EFFLUENT MANUAL...........~. ~......

~ ~ ~ ~ ~ ~... 6.0-33 BFN Unit 2 v

LIST OF TABLES (Cont'd)

Table Title

~Pa e Ne.

4.2.E Minimum Test and Calibration Frequency for Drywell Leak Detection Instrumentation.

3.2/4.2-53 4.2.F Minimum Test and Calibration Frequency for Surveillance Instrumentation.

3.2/4.2-54 4.2.G Surveillance Requirements for Control Room Isolation Instrumentation 3.2/4.2-56 4.2.H Minimum Test and Calibration Frequency for Flood Protection Instrumentation.

3.2/4.2-57 4.2eJ Seismic Monitoring Instrument Surveillance Requirements.

3.2/4.2-58 4.2eK Radioactive Gaseous Effluent Instrumentation Surveillance.

3.2/4.2-62 4.2.L ATWS-Recirculation Pump Trip (RPT)

Instrumentation Surveillance 3.2/4.2-63a 3.5-1 3.7.A 4.9.A Minimum RHRSW and EECW Pump Assignment.

Primary Containment Isolation Valves.

3.5/4.5-11 3.7/4.7-25 Diesel Generator Reliability.

3.9/4.9-16 3.11.A 3.11.B 3.11.C 3.11.D 6.2.A Fire Detection Instrumentation.

Spray/Sprinkler Systems Hose Stations Yard Fire Hydrants and Fire Hose Houses Minimum Shift Crew Requirements 4.9.A.4.C Voltage Relay Setpoints/Diesel Generator Start.

3.9/4.9-18 3.11/4.11-14 3.11/4.11-17 3.11/4.11-18 3.11/4.11-20 6.0-4 BFH Unit 2 vii

IST OF ILLUSTRATIONS

~Ti ere 2.1.1 Title

~Pa e

Ne APRM Flow Reference Scram and APRM Rod Block Settings

~

1.1/2.1-6 2 '

2 4.1-1 APRM Flow Bias Scram Vs. Reactor Core Flow.

1.1/2.1-7 Graphical Aid in the Selection of an Adequate Interval Between Tests 3.1/4.1-13 4.2-1 3.5.M-1 3.6-1 System Unavailability.

BFN Power/Flow Stability Regions Minimum Temperature F Above Change in Transient Temperature.

3.2/4.2-64 3.5/4.5-22a 3.6/4.6-24 3.6-2 Change in Charpy V Transition Temperature Vs.

Neutron Exposure 3.6/4.6-25 4.8.l.a Gaseous Release Points and Elevations 3.8/4.8-10 4.8.1.b Land Site Boundary 3.8/4.8-11 BFN Unit 2 viii

1'. 0 DEFINITIONS (Cont'd)

N.

Rated Power Rated power refers to operation at a reactor power of 3, 293 MWt; this is also termed 100 percent power and is the maximum power level authorized by the operating license.

Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters when the reactor is at rated power.

0.

Prima Containment Inte rit Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied :

1.

All nonautomati c containment isolation valves on lines connected to the reactor coolant systems or containment which are not required to be open during accident conditions are closed.

These valves may be opened to perform necessary operational activities.

2.

At least one door in each airlock is closed and sealed.

3.

All automatic containment isolation valves are OPERABLE or each line which contains an inoperable isolation valve is isolated as required by Specification 3. 7. D. 2.

4.

All blind flanges and manways are closed.

P.

S econda Containment Inte rit 1.

Secondary containment integrity means that the required unit reactor zones and refueling zone are intact and the following conditions are met:

a)

At least one door in each access opening to the turbine building, control bay and out-of-doors is closed.

b )

The standby gas treatment system is OPERABLE and can maintain 0. 25 inches of water negative pressure in those areas where secondary containment integrity is stated to exist.

c ) All secondary containment penetrations requi red to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.

Closed by at least one secondary containment automatic isolation valve deactivated in the isolated position.

2.

Reactor zone secondary containment integrity means the unit reactor building is intact and the following conditions are met:

a)

At least one door between any opening to the turbine building, control bay and out-o f-doors is closed.

BFN Unit 2

1'. 0 Q.

0 erat in C cle Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit.

R.

Re fuelin Outa e Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling.

For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a

regularly scheduled outage;

however, where such outages occur within 8 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage.

S.

CORE ALTERATION The addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls within the reactor pressure vessel with the head removed and fuel in the vessel.

Normal movement of in-core instrumentation and the traversing in-core probe is not defined as a Core Alteration.

Suspension of Core Alterations shall not preclude compl et ion of the movement of a component to a safe conservative position.

T.

Reactor Vessel Pressure Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.

U.

Thermal Parameters 1.

Minimum Critical Power Ratio MCPR Minimum Critical Power Ratio (MCPR) is the value of the critical power ratio associated with the most limiting assembly in the reactor core.

Critical Power Ratio ( CPR) is the ratio of that power in a fuel assembly,

which is calculated to cause some point in the assembly to experience boiling transi tion, to the actual assembly operating power.

2.

Transition Boilin Transition boiling means the boiling regime between nucleate and film boiling.

Trans ition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

3.

Core Maximum Fraction of Limitin Power Dens it CMFLPD The highest ratio, for all fuel assembli es and all axial locations in the core, of the maximum fuel rod power density (kW/ft) for a given fuel assembly and axial location to the limiting fuel rod power density (kW/ft) at that location.

4.

Avera e Planar Linear Heat Generation Rate APLHGR The Average Planar Heat Generation Rate is applicable to a speci fic planar height and is equal to the sum of the linear heat generat ion rates for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

BFN Unit 2

1. 0-7

DEFI ITIONS Cont'd NN.

A endix R Safe Shutdown Pro ram BFN has developed an Appendix R Safe Shutdown Program.

This Program is to ensure that the equipment required by the Appendix R Safe Shutdown Analysis is maintained and demonstrated functional as follows:

1.

The functional requirements of the Safe Shutdown systems and equipment, as well as appropriate compensatory measures should these systems/components be unable to perform their intended function are outlined in Section III of the Program.

2.

Testing and monitoring of the Appendix R Safe Shutdown systems and equipment are defined in Section V of the Program.

3.

Changes made to the BFN Appendix R Safe Shutdown Program will be processed in accordance with License Condition 2.C.5.(a).

00.

CORE OPERATI G LIMITS REPORT COLR The COLR is the unit-specific document that provides the core operating limits for the current operating cycle.

These cycle-specific core operating limits shall be determined for each operating cycle in accordance with Specification 6.9.1.7.

Plant operation within these limits is addressed in individual specifications.

BFN Unit 2 1.0-12a

i'. 1 2 1

FUEL CLADDING INTEGRITY SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A Neutron Flux Tri Settin s

2.1.A.l.b. (Cont'd)

NOTE:

These settings assume operation within the basic thermal hydraulic design criteria.

These criteria are LHGR within the limits of Specification 3.5.J and MCPR within the limits of Specification 3.5.K. If it is determined that either of these design criteria is being violated during operation, action shall be initiated within 15 minutes to restore operation within prescribed limits.

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

c.

The APRM Rod Block trip setting shall be:

SR~ (0.58W + 50%)

where:

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

W

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

BFN Unit 2 1.1/2.1-3

BASES:

LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed in support of planned operating conditions up to the maximum thermal power of 3293 MWt.

The analyses were based upon plant operation in accordance with Reference 1.

In addition, 3293 MWt is the licensed maximum power level for each Browns Ferry Nuclear Plant unit, and this represents the maximum steady-state power which shall not be knowingly exceeded.

The transient analyses performed for each reload are described in Reference 2.

Models and model conservatisms are also described in this reference.

BFN Unit 2 1.1/2.1-11

2.1 BASES (Cont'd)

The bases for individual setpoints are discussed below:

A.

eutron Flux Scram 1.

APRM Flow-Biased Hi h Flux Scram Tri Settin RUN Mode The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3,293 MWt).

Because fission chambers provide the basic input signals, the APRM system responds directly to core average neutron flux.

During power increase transients, the instantaneous fuel surface heat flux is less than the instantaneous neutron flux by an amount depending upon the duration of the transient and the fuel time constant.

For this reason, the flow-biased scram APRM flux signal is passed through a

filtering network with a time constant which is representative of the fuel time constant.

As a result of this filtering, APRM flow-biased scram will occur only if the neutron flux signal is in excess of the setpoint and of sufficient time duration to overcome the fuel time constant and result in an average fuel surface heat flux which is equivalent to the neutron flux trip setpoint.

This setpoint is variable up to 120 percent of rated power based on recirculation drive flow according to the equations given in Section 2.1.A.1 and the graph in Figure 2.1-2.

For the purpose of licensing transient analysis, neutron flux scram is assumed to occur at 120 percent of rated power.

Therefore, the flow biased scram provides additional margin to the thermal limits for slow transients such as loss of feedwater heating.

No safety credit is taken for flow-biased scrams.

BFN Unit 2 1.1/2.1-12

2'. 1 BASES (Cont'd) including above the rated rod line (Reference 1).

,The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 108 percent of rated thermal power because of the APRM rod block trip setting.

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the incore LPRM system.

C.

Reactor Water Low Level Scram and Isolation Exce t Main Steam lines The setpoint for the low level scram is above the bottom of the separator skirt.

This level has been used in transient analyses dealing with coolant inventory decrease.

The results reported in FSAR Subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than 1.07 in all

cases, and system pressure does not reach the safety valve settings.

The scram setting is sufficiently below normal operating range to avoid spurious scrams.

D.

Turbine Sto Valve Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves.

With a trip setting of 10 percent of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed.

(Reference 2)

E.

Turbine Control Valve Fast Closure or Turbine Tri Scram Turbine control valve fast closure or turbine trip scram anticipates the pressure, neutron flux, and heat flux increase that could result from control valve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capability.

The reactor protection system initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load rejection or control valve closure due to turbine trip.

This scram is achieved by rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system.

This trip setting, a nominally 50 percent greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve.

No significant change in MCPR occurs.

Relevant transient analyses are discussed in References 2

and 3 of the Final Safety Analysis Report.

This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by turbine first state pressure.

BFN Unit 2 1.1/2.1-15

hr

2'.1 BASES (Cont'd)

F.

(Deleted)

G.

& H.

ain Steam line Isolation on Low Pressure and Main Steam Line Isolation Scram 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.Ec K. Reactor Low Water Level Set oint for Initiation of HPCI and RCIC Closin Main Steam Isolation Valves and Startin LPCI and Core 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.

References 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 (latest approved version).

BFN Unit 2 1.1/2.1-16

4.1 BASES (Cont'd)

The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate.

The APRM system, which uses the LPRM readings to detect a change in thermal power, will be calibrated every seven days using a heat balance to compensate for this change in sensitivity.

The RBM system uses the LPRM reading to detect a localized change in thermal power. It applies a correction factor based on the APRM output signal to determine the percent thermal power and therefore any change in LPRM sensitivity is compensated for by the APRM calibration.

The technical specification limits of CMFLPD, CPR, and APLHGR are, determined by the use of the process computer or other backup methods.

These methods use LPRM readings and TIP data to determine the power distribution.

Compensation in the process computer for changes in LPRM sensitivity will be made by performing a full core TIP traverse to update the computer calculated LPRM correction factors every 1000 effective full power hours.

As a minimum the individual LPRM meter readings will be adjusted at the beginning of each operating cycle before reaching 100 percent power.

BFN Unit 2 3.1/4.1-20

3'.3/4.3 BASES (Cont'd) 5.