ML18036B239

Amends 194,209 & 166 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Revising Surveillance Requirements Associated W/Certain Refueling Equipment & Limiting Conditions for Operations

ML18036B239
Person / Time
Site:	Browns Ferry
Issue date:	04/09/1993
From:	Hebdon F Office of Nuclear Reactor Regulation
To:
Shared Package
ML18036B240	List: ML18036B239 ML18036B241
References
DPR-33-A-194, DPR-52-A-209, DPR-68-A-166 NUDOCS 9304160172
Download: ML18036B239 (88)

Text

E

~r

~+*<<+

UNITED STATES NUCLEAR REGULATORY COMMiSSION WASHINGTON. O.C. 20555 0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT UNIT 1

AH NDHENT TO FACI ITY OP RAT NG IC NS Amendment No.

194 License No.

DPR-33 The Nuclear Regulatory Commission (the Commission) has found that:

A.

B.

C.

D.

E.

The application for amendment by Tennessee Valley Authority (the licensee) dated October 9,

1992, as supplemented on March 31,

1993, complies with the standards and requirements of the Atomic Energy Act of

1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facil,.ity will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9304ih0172 930409'PDR ADOCK 05000259,'

PDR

1

'I

2.

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

DPR-33 is hereby amended to read as follows:

(2) Technical S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 194

, are hereby incorporated in the license.

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

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Project Directorate II-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Date of Issuance:

April 9, 1993

I 0

4I L.

I

'i I

C I

ATTACHMENT TO LIC NSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO.

DPR-33 DOCKET NO. 50-259 Revise the Appendix A Technical Specifications

'by removing the pages identified below and inserting the enclosed pages.

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

Overleaf* and spillover** pages are provided to maintain document completeness.

REMOV ill iv 1.0-7 1.0-8 3.10/4.10-1 3..10/4.10-2 3.10/4.10-3 3.10/4.10-4 3.10/4.10-5 3.10/4.10-6 3.10/4.10-11 3.10/4.10-12 3.'10/4.10-13 3'.10/4.10-14 3.10/4.10-15 6.0-3

6. 0'-4 INSERT ill iv*

1.0-7 1.0-8*

3.10/4.10-1 3.10/4.10-2 3.10/4.10-3 3.10/4.10-4 3.10/4.10-5 3.10/4.10-6*

3.10/4.10-11 3.10/4.10-12**

3.10/4.10-13 3.10/4.10-14*

3.10/4.10-'5 3.10/4.10-16 6.0-3 6.0-4*

E Je t Pumps

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

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 3 '/4.6-11 3.6/4.6-12'.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 G.

Containment Atmosphere Dilution System (CAD).

3.7/4.7-22 H.

Containment Atmosphere Monitoring (CAM)

System H2 Analyzer 3.8/4.8 Radioactive Materials A.

Liquid,'Effluents B.

Airborne Effluents C.

Radioactive Effluents Dose D.

'Mechanical Vacuum Pump

'3.7/4.7-24 3.8/4.8-1'.8/4.8-1 3.8/4.8-3 3.8/4.8-6 3.8/4.8-6 E.

Miscellaneous Radioactive Materials Sources..

3.8/4.8-7 F.

Solid Radwaste A.

Auxili.ary 'Electrical Equipment

'B.

Operation with Inoperable Equipment.

C.

Operation in Cold Shutdown

~

~

~

~

~

~

3.10/4.10 Core Al-terato;ons A.

.Refueling Interlocks B.

Core Monitoring.

3.9/4.9 Auxiliary Electrical System 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 3.10/4.10-1 3.10/4.10-1 3.10/4.10-5 BFN Unit 1

Amendment No. 194

C.

D.

F.

Spent Fuel Pool Water.

Reactor Building Crane Spent Fuel Cask.

Spent Fuel Cask Handling-Refuel'ing Floor 3.10/4.10-7 3.10/4.10-8 F 10/4.10-9 3.10/4.10-10 3.11/4.11 Deleted...

3..11/4.11-1 5.0 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 5.0-1 5.0-1 5.0-1 5.0-1 5.0-1 5.0-1 5.0-2 BFN Unit 1 iv hkICgpggpgg y g

~

~

I 1.0 (Cont'd)

Q.

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.

S.

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.

CORE ALTERATION shall be the movement of any fuel,

sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel.

Movement of source range monitors, intermediate range monitors, traversing in-core probes, or special movable detectors (including undervessel replacement) is not considered a

CORE ALTERATION.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe location.

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

1. ~~gg~Q,~Q w

t

- 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 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 ~

'~~

w The highest ratio, for all fuel types in the core, of the maximum fuel rod power density (kW/ft) for a given fuel type to the limiting fuel rod power density (kW/ft) for that fuel type.

4. ~vugg~~

IAMKl-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 1.0-7 Amendment No. 194

1.0 (Cont'd) 0 V ~

t ~ An instrument calibration means the adjustment of an instrument signal'utput so that it corresponds, within acceptable

range, and accuracy, to a known value(s) of the: parameter which the instrument monitors.

2.

~ggapJ, A channel is an arrangement of the sensor(s) and associated components used to evaluate plant variables and produce discrete outputs used in logic.

A channel terminates and loses its identity where individual channel outputs are combined in logic.

the injection of a simulated signal into the instrument primary sensor to veri;fy the proper instrument channel

response, alarm and/or initiating action.

4.

An instrument check is qualitative determination of acceptable operability by observation of instrument behavior during operation.

This determination shall

include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

5 ~

Lagi~~m~uo~lipmL1 ~t A logic system functional test means a test of 'all relays and contacts of a logic circuit to insure all components are operable per design intent.

Where practicable, action will go to completion; i.e.,

pumps will be started and valves operated.

6.

channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function.

A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action.

Initiation of protective action may require the tripping of a single trip system or ~ne coincident tripping of two 'trip systems.

7 ~

t v

An action initiated by the protection system when a limit is reached.

A protective action can be at a channel or system level.

'8.

t t'v t'

A system protective action which results from the protective action of the channels monitoring a particular plant condition.

9.

t t'tg~t' Simulated automatic actuation means applying a s&nulated signal to the sensor to actuate the circuit in question.

BFN Unit 1

3.10 4.10 Applies to the fuel handling and core reactivity limitations.

Applies to the periodic testing of those refueling equipment interlocks and instrumentation required during refueling and CORE ALTERATIONS.

Qk.hS~t' To ensure that, core reactivity is within 'the capability of the control rods and to prevent criticality during refueling.

To verify the OPERABILITY of instrumentation and refueling equipment inter-locks required during refueling and CORE ALTERATIONS.

A.

A.

1.

The reactor mode switch shall be locked in the REFUEL position during CORE ALTERATIONS.

The required refueling equipment interlocks shall be OPERABLE during in-vessel fuel movement with equipment associated with the interlocks except as specified in 3.10.A.6 and 3.10.A.7 below.

1. Prior to any fuel handling with the

'head off the

vessel, the following required refueling equipment interlocks shall be functionally tested:

a.

All,rods inserted b.

Refueling platform positioned near or over the core c.

Refueling platform main hoist is fuel loaded d.

Fuel grapple is not ful'1 up e.

One rod withdrawn BFN Unit 1 3.10/4.10-1 Amendment No. 194

3.10.A.

4.10.A.

4.10.A.l (Continued)

• f ~

Refueling platform frame-mounted hoist is fuel loaded Monorail ho'ist is fuel loaded

• h.

Service platform hoist is fuel loaded They shall be tested at weekly intervals thereafter until no longer required.

They shall also be tested following any repair work associated with the interlocks.

• NOTE:

These interlocks are required to be OPERABLE only when the associated equipment is used for in-vessel

'fuel movement.

2.

Fuel shall not be loaded.

into the reactor core unless al'1 control rods are fully inserted.

2.

No additional

,surveil'lance required.

3.

The fuel'rapple hoist load switch shall be set at g 1,000 lbs.

3.

No additional surveillance required.

4.

If the frame-mounted auxiliary hoist, the monorail-mounted'uxiliary hoist, or 'the service platform hoist is to be used for handling fuel with the head

,off the reactor vessel, the load limit switch on the hoist to be used shall be set at

< 400 lbs.

4.

No additional surveillance required.

BFN Unit 1 3.10/4.10-2 Amendment No.

194

~

~

3..10.A.

4.10.A.

5.

Maintenance may be performed on a single control rod or control rod drive without removing the fuel in the control cell if the following conditions are met:

a.

The requirements of Specification 3.10.A.1 are met, and b.

All control rods diagonally and face adjacent to the maintenance rod are fully inserted and have their directional control'alves electrically disarmed.

5.

Prior to 'performing control rod or control rod drive maintenance on a control cell without removing fuel assemblies the surveillance require-ments of Specification 4.10.A.1 shall be performed and all rods face adjacent and diagonally adjacent to the maintenance rod shall be electrically disarmed per Specification 3.10.A.5.b.

6.

A maximum of two non-adjacent control rods may simultaneously be withdrawn from the core for the purpose of performing control rod and/or control rod drive maintenance without removing the fuel from the cells provided the following conditions are satisfied:

6.

Prior to performing control rod or control rod drive maintenance on two control cells simultaneously without removing the fuel from the cells, two SROs shall verify that the requirements of Specification 3.10.A.6 are satisfied.

a.

The reactor mode switch shall be locked in the REFUEL position.

The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed for one of the control rods on which maintenance is being performed.

All other required refueling equipment interlocks shall be OPERABLE.

BFN Unit 1

3.10/4.10-3 Amendment No. 194

3.10'.A.

4.10.A.

3.10.A.6. (Cont"d) b.

All directional control valves for remaining control rods shall be disarmed electrically except as specified in 3.10.A.7 and sufficient margin to criticality shall be demonstrated.

c.

The two maintenance cells must be separated by more than two control cel'ls in any direction.

d.

An appropriate number of SRMs are available as defined in Specification 3.10.B.,

7.

Any number of control rods may be withdrawn or removed from the reactor core providing the, following conditions are satisfied:

a.

The reactor mode switch is locked in the REFUEL position.

The refueling interlock which

,prevents more than one control rod from being withdrawn may be bypassed on a withdrawn control rod after the fuel assemblies in the cell containing (controlled by) that control rod have been removed from the reactor core.

All other required refueling equipment interlocks shall be OPERABLE.

7.

With the mode selection switch in the REFUEL or SHUTDOWN mode, no more than:one control rod may be withdrawn without first removing fuel from the cell except as specified in 4.10.A.6.

Any number of rods may be withdrawn once verified by two licensed operators that the fuel has been removed from each cell.

BFN

,Unit 1

3.10/4.10-4 Anendment No. 194

~

~

~

3.10.B.

4.10.B.

1.

During CORE ALTERATIONS, except as specified in 3 '0.B.2, two SRMs (FLCs) shall be OPERABLE.

For an SRM (FLC) to be considered OPERABLE, the following shall be satisfied:

a.

The SRM shall be inserted to the normal operating level.

(Use of special'oveable, dunking type detectors during initial fuel loading and major

CORE, ALTERATIONS in place of normal detectors is permissible as long as the detector is connected to the normal SRM.circuit.)

1.

Prior to making any CORE ALTERATIONS, the SRMs (FLCs) shall be functionally tested and checked for neutron response.

2.

Note:

Not required to, be met with less than or equal to four fuel assemblies adjacent to the SRM- (FLC) and no other fuel assemblies in the associated core quadrant.

Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS, verify that the associated SRM (FLC) is reading y 3 cps with a signal-to-noise ratio ~ 3:l.

b.

Verify an OPERABLE SRM ('FLC) is located in:'.

The fueled region; 2.

The quadrant where CORE ALTERATIONS're being performed, w'hen the associated

,SRM (FLC) is included in the fueled region;,

and'.

A core quadrant adjacent to where CORE ALTERATIONS are being performed',

when the associated SRM (FL'C) is included in the fueled region.

Note:

One SRM (FLC) may be used to sati'sfy more than one.of the above.

. BFN Unit 1

3.10/4.10-5 Amendment No.

194

li

. ~

~

~

~

LIMITING CONDITIONS FOR OPERATION 3.10.B.

SURVEILLANCE RE UIREMENTS 4.10.B 2.

During a complete core

removal, the SRMs shall have an initial

.minimum count rate of 3 cps prior to fuel removal.

With all'ods fully inserted and rendered electrically disarmed and inoperable, once the SRM count rate decreases below 3 cps, the 'SRMs will no longer be required to be OPERABLE during fuel removal.

Individual control rods outside the periphery of the then existing fuel matrix may be electrically armed and moved for maintenance after all fuel in the cell containing (controlled by) that control rod have been removed from the reactor core.

BFN Unit 1 3.10/4.10-6 QMENDMM'O. 1 '7 2

A.

The refueling interlocks are designed to back up procedural core reactivity controls during refueling operations.

The interlocks prevent an inadvertent criticality during, refueling operations when the reactivity potential of the core is being altered.

To minimize the possibility of loading fuel into a cell containing no control rod, it is requ'ired that al'1 control rods are fully inserted when fuel is being loaded into the reactor core.

This requirement assures that during ref'ueling the refuel'ing interlocks, as designed, will prevent inadvertent criticality.

.The refueling interlocks reinforce operational procedures that prohibit taking the reactor critical under certain situations encountered, during refueling operations by restricting the movement of control rods and the operation of refueling equipment.

The refueling interlocks include circuitry which senses the condition of the refueling equipment and the control'ods.

Depending on, the sensed condition, interlocks are actuated'hich prevent the movement of the refueling equipment or withdrawal of control rods (rod block)'.

Circuitry is provided which senses the following. conditions.

l.

All rods inserted 2.

Refueling platform positioned near or over the core 3.

Refueling platform main hoist is fuel loaded 4.

Fuel grapple not full up 5.

.One rod withdrawn 6.

Refueling platform frame-mounted hoist is fuel loaded 7.

Refueling platform monorail hoist.is fuel loaded 8.

Service platform hoist is fuel-loaded When the mode switch is in the "Refuel" position, interlocks prevent the refueling platform from being moved over the core if a control rod is withdrawn and fuel is on a hoist.

Likewise, if the refueling platform is over the core with fuel on a hoist, control rod motion is blocked by the interlocks.

When the mode switch is in the refuel posit'ion only one control rod can be withdrawn.

The refuel'ing interlocks, in combination with core nuclear design and refueling procedures, limit the probability of an inadvertent criticality.

The nuclear characteristics of the core assure that the reactor is The refueling platform frame-mounted, monorail and the service platform fuel-loaded hoist interlocks are required to be OPERABLE only when utilized for in-vessel fuel movements.

BFN Unit 1

3.10/4.10-11 Amendment No. 194

~

~

3.10 ~ (Cont'd) subcritical even when the highest worth control rod is fully withdrawn.

The combination of refueling interlocks for control rods and the refueling platform provide redundant methods of preventing inadvertent criticality even after procedural violations.

The interlocks on hoists provide yet another method of avoiding inadvertent criticality.

Fuel handling is normally conducted with the fuel grapple hoist.

The total load on this hoist when the interlock is required consists of the weight of the fuel grapple and the fuel assembly.

This total is approximately 1,500 lbs, in comparison to the load-trip setting of 1,000 lbs.

Provisions have also been made to allow fuel handling with either of the three auxiliary hoists and still maintain the refueling interlocks.

The 400-ib load-trip setting on these hoists is adequate to trip the interlock when one of the more than 600-lb fuel bundles is being handled.

During certain periods, it is desirable to perform maintenance on two control rods and/or control rod drives at the same time without removing fuel from the cells.

The maintenance is performed with the mode switch in the refuel position to provide the refueling interlocks normally available during refueling operations.

In order to withdraw a second control rod after withdrawal of the first rod, it is necessary to bypass the refueling interlock on the first control rod which prevents more than one control rod from being withdrawn at the same time.

The requirement that an adequate shutdown margin be demonstrated and that all remaining control rods have their directional control valves electrically disarmed ensures that inadvertent criticality cannot occur during this maintenance.

The adequacy of the shutdown margin is verified by demonstrating that at least 0.38 percent Dk shutdown margin is available.

Disarming the directional control valves does not inhibit control rod scram capability.

Specification 3.10.A.7 allows unloading of a significant portion of the reactor core.

This operation is performed with the mode switch in the refuel position to provide the refueling interlocks normally available during refueling operations.

In order to withdraw more than one control rod, it is necessary to bypass the refueling interlock on each withdrawn control rod which prevents more than one control rod from being withdrawn at a time.

The requirement that the fuel assemblies in the cell controlled by the control rod be removed from the reactor core before the interlock can be bypassed ensures that withdrawal of another control rod'oes not result in inadvertent criti.cality.

Each control rod provides primary reactivity control for the fuel assemblies in the cell as'sociated with that control rod.

Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core.

The requirements for SRN OPERABILITY during these CORE ALTERATIONS assure sufficient core monitoring.

BFN Unit 1

3.10/4.10-12 Amendment No. 194

3.10 ~ (Cont'd) 3.10.A (Cont'd) 1.

Refueling interlocks (BFNP FSAR Subsection 7.6)

B.

The SRMs are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and station startup.

Requiring two OPERABLE SRMs (FLCs) one in and one adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations.

Each SRM (FLC) is not required to read Z 3 cps until after four fuel assemblies have 'been loaded adjacent to the SRM (FLC) if no other fuel assemblies are in the associated core quadrant.

These four locations are adjacent to the SRM dry tube.

When utilizing FLCs, the FLCs will be located such that the required count rate is achieved without exceeding the SRM upscale setpoint.

With four fuel assemblies or fewer loaded around each SRM, even with a control rod withdrawn, the configuration will not be critical.

Under the special condition of removing the full core with all control rods inserted and electrically disarmed, it is permissible to allow SRM count rate to decrease below three counts per second.

All fuel moves during core unloading will reduce reactivity.

It is expected that the SRMs will drop below three counts per second before all of the fuel is unloaded.

Since there will be no reactivity additions during this

period, the low number of counts will not present a hazard.

When sufficient fuel has been removed to the spent fuel storage pool to drop the SRM count rate ibelow 3 cps, SRMs will no longer be required to be OPERABLE.

Requiring the SRMs to be functionally tested prior to fuel removal assures that the SRMs will be OPERABLE at the start of fuel removal.

The once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY until the count rate diminishes due to fuel removal.

Control rods in cells from which all fuel has been removed and which are outside the periphery of the then existing fuel matrix may be armed, electrically and moved for maintenance purposes during full core removal, provided all rods that control fuel are fully inserted and electrically disarmed.

1.

Neutron Monitoring System (BFNP FSAR Subsection 7.5) 2.

Morgan, W. R., "In-Core Neut;ron Monitoring System for General Electric Boiling Water Reactors,"

General Electric Company, Atomic Power Equipment Department, November 1968, revised April 1969 (APED-5706)

BFN Unit 1 3 10/4 10-13 Amendment No. 194

C ~

The design of the spent fuel storage pool provides a storage location for approximately 140 percent of the full core load of fuel assemblies in the reactor building which ensures adequate shielding, cooling, and reactivity control of irradiated fuel.

An analysis has been performed which shows that a water level at or in excess of eight and one-half feet over the top of the stored assemblies will provide shielding such that the maximum calculated radiological doses do not exceed the limits of 10 CFR 20.

The normal water level provides 14-1/2 feet of additional water shielding.

The capacity of the skimmer surge tanks is available to maintain the water level at its normal height for three days in the absence of additional water input from the condensate storage tanks.

All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the normal water level more than one-half foot.

The fuel pool cooling system is designed to maintain the pool water temperature less than 125'F during normal heat loads.

If the reactor core is completely unloaded when the pool contains two previous discharge

batches, the temperature may increase to greater than 125'F.

The RHR system supplemental fuel pool cooling mode will be used under these conditions to maintain the pool temperature to less than 125'F.

4.

The reactor building crane and 125-ton hoist are required to be operable for handling of the spent fuel in the reactor building.

The controls for the 125-ton hoist are located in the crane cab.

The five-ton has both cab and pendant controls.

A visual inspection of the load-bearing hoist wire rope assures detection of signs of distress or wear so that corrections can be promptly made if,needed.

I The testing of the various limits and interlocks assures their proper operation when the crane is used.

The spent fuel cask design incorporates removable lifting trunnions.

The visual inspection of the trunnions and fasteners prior to attachment to the cask assures that no visual damage has occurred during prior handling.

The trunnions must be properly attached to the cask for lifting of the cask and the visual inspection assures correct installation.

BFN Unit 1

3.10/4.10-14

Although single failure protection has been provided in the design of the 125-ton hoist drum shaft, wire ropes, hook and lower block assembly on the reactor building crane, the limiting of lift height of a spent fuel cask controls the amount of energy available in a dropped cask accident when the cask is over the refueling floor.

An analysis has been made which shows that the floor and support members in the area of cask entry into the decontamination facility can satisfactorily sustain a dropped cask from a height of three feet.

The yoke safety links provide single failure protection for the hook and lower block assembly and limit cask rotation.

Cask rotation is necessary for decontamination,and the safety links are removed during decontamination.

A.

Complete functional testing of all required refueling equipment interlocks before any refueling outage wil'1 provide positive indication that the interlocks operate in the situations for which they were designed.

By loading each hoist with a weight equal to the fuel

assembly, positioning the refueling platform, and withdrawing control

rods, the interlocks can be subjected to valid operational tests.

Where redundancy is provided in the logic circuitry, tests can be performed to assure that each redundant logic element can independently perform its function.

B.

Requiring the SRNs to,be functional'ly tested prior to any CORE ALTERATION assures that the SRNs will be OPERABLE at the start of that alteration.

The once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verification of the SRN count rate and signal-to-noise ratio ensures; their continued OPERABILITY.

1.

Fuel Pool Cooling and Cleanup System (BFNP FSAR Subsection 10.5)'.

Spent Fuel Storage,(BFNP FSAR Subsection 10.3)

BFN Unit 1 3.10/4.10-15 Amendment No.

194

0

~

~

~

~

r THIS PAGE INTENTIONALLYLEFT BLANK BFN Unit 1 3.10/4.10-16 Amendment No.

194

6.2.2 (Cont.)

d.

Two licensed reactor operators shall be in the control room during any cold startups, while shutting down the reactor, and during recovery from unit trip.

In addition, a person holding a senior operator license shall'e in the control room for that unit whenever it is in an operational mode other than cold shutdown or refueling.

e.

A Health Physics Technician* shall be present at 'the facility at all times when there is fuel in the reactor.

f.

Either a licensed SRO or licensed SRO limited to fuel handling who has no concurrent responsibilities during this operation shall be present during fuel handling and shal'l directly supervise all CORE ALTERATIONS.

g.

Deleted.

• The Health Physics: Technician may be absent from the facility for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided immediate action is taken to fillthe required position.

BFN Unit l 6.0-3 Amendment No. 194

0 0

I

Xah3.~&

t w

t b Senior Operator Senior Operator Licensed Operators Additional 'Licensed Operators Assistant Unit Operators (AUO)

Shift Technical Advisor (STA)

Health Physics Technician 1

2 1

1 1

1 0'

2 2

3 3

3, 3

0 1

2 2

4 4

5 5

0 1

1 1

1 1

1 1

SRO SRO RO or 'SRO RO or SRO None None None a.

A senior operator will be assigned responsibility for overall plant operation at all times there is fuel in, any unit.

b.

Except for the senior operator discussed in note "a", the shift crew composition may be one less than the minimum, requirements of Table 6.2.A for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2.A.

This'rovision does not permit any shift crew position to.be unmanned upon shift change due to an oncoming shift crewman being late or absent.

c.

One of the Additional Licensed Operators must be assigned to each control room with an operating unit.

d.

The number of required licensed personnel, when the operating units are controlled from a common control room, are two senior operators and four operators.

BFN Unit 1

6.0-4

,AMENOMERTRU.I4 9

~gR +f0g(~rr I

}

0

<<**<<+

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555.0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT UNIT 2 AMENDMENT TO FACILITY OPERATING LICENS Amendment No.

209 License No.

DPR-52 The Nuclear Regulatory Commission (the Commissi'on) has found that:

A'.

B.

C.

D.

The application for amendment by Tennessee Valley Authority (the, licensee) dated October 9,

1992, as supplemen'ted on March 31,

1993, complies with the standards and requirements of the Atomic Energy.Act of

1954, as amended (the Act), and the Commission's rules and.regulations, set forth in 10 CFR Chapter, I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that..the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR.Part 51 of the Commission's regulations,and all applicable requirements have been satisfied.

~ s

~

II'

2.

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

DPR-52 is hereby amended to read as follows 3.

(2) Technical S ecifications The Technical Specifications contained-in Appendices A and B,

as revised through Amendment No. 209, are hereby incorporated in the license.

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

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

'FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifi'cations Date of Issuance:

April 9, 1993 Project Directorate II-4 Division. of Reactor Projects - I/II Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE'MENOMENT NO

,FACI ITY OPERATING ICENSE NO.

OPR-52 DOCKET NO. 50-260 Revise the Appendix A Technical Specifications by'emovi'ng the pages identified below and inserting the enclosed pages.

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

Overleaf* and spillover** pages are provided to maintain document completeness.

REMOVE 111 iv 1.0-7 1.0-8 3.10/4.10-1 3.10/4.10-2 3.10/4.10-3 3.10/4.10-4 3.10/4.10-5 3.10/4.10-6 3.10/4.10-11 3.10/4.10-12 3.10/4.10-13 3.10/4.10-14 3.10/4.10-15 6.0-3 6.0-4 INSERT iii*

iv 1.0-7 1.0-8*

3.10/4.10-1 3.10/4.10-2 3.10/4.10-3 3.10/4'.10-4 3.10/4.10-5 3.10/4.10-6*

3.10/4.10-11 3.10/4.10-12**

3.10/4.10-13 3.10/4.10-14*

3.10/4.10-15 3.10/4.10-16 6.0-3 6.0-4*

3.7/4.7 E.

F.

Recirculation Pump Operation G.

Structural Integrity H.

Snubbers 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 e t Pumps

~

~

~

~

~

~

~

o

~

~

~

~

~

~

~

~

~

~

J 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 G.

Containment Atmosphere Dilution System (CAD) 3.7/4.7-22 3.8/4.8 3.,9/4.9'.

.Containment. Atmosphere Monitoring (CAM)

System H2 Analyzer Radioactive Materials 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 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 B.

Operation with Inoperable Equipment.

3.9/4.9-8 C.

Operation in Cold Shutdown 3.9/4.9-15 D.

Unit 3 Diesel Generators Required for Unit 2 Operation 3.9/4.9-15a BFN:

Unit2'MENOMENTNO. I 8 6

l

~

~

3.10/4.10 Core Alterations 3'.10/4.10-1 A.

B.

C.

D.

Refueling Interlocks Core Monitoring,.

Spent Fuel Pool Water Reactor Building Crane Spent Fuel Cask 3.10/4.10-1 3.10/4.10-5 3.10/4.10-7 3.10/4.10-8 3.10/4.10-9 F.

Spent Fuel Cask Handling-Refueling Floor 3.10/4.10-10 3.11'/4.11 Deleted.

3.11/4.11-1 5.0 Maj'or Design Features 5.1 Site Features 5.2 Reactor 5.3 Reactor Vessel 5.'4 Containment 5.5 Fuel Storage.

5.6 Seismic Design 5.0-1 5.0-1 5.0-1 5.0-1 5.0-1 5'. 0-1 5.0-2 BFN Unit 2 Amendment No. 209

Ih ll

~

~

1.0 Q.

a particular unit and the end of the next subsequent refueling outage for the same unit.

R. Qe~glj~ggt~

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 shall be the movement of any fuel,

sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel.

Movement of source range monitors, intermediate range monitors, traversing in-core probes, or special movable detectors (including undervessel replacement) is not considered a

CORE ALTERATION.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe location.

T. ~t~~~gz~ge Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.

U.

5!

(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 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. ~~~~

t

'ighest ratio, for all fuel types in the core, of the maximum fuel rod power density (kW/ft) for a given fuel type to the limiting fuel rod power density (kW/ft) for that fuel type.

s" ~'a 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 2 1.0-7 Amendment No. 209

1.0 (Cont'd)

V ~

adjustment of an instrument signal output so that it corresponds, within acceptable

range, and accuracy, to a known value(s)'f the parameter which the instrument monitors.

2.

QhglgngJ, A channel is an arrangement of the sensor(s) and associated components used to evaluate plant variables and produce discrete outputs used in logic.

A channel terminates and loses its identity where individual channel outputs are combined in logic.

3.

j~iggggt ~~to~~

An instrument functional test means the injection of a simulated signal into the instrument primary sensor to verify the proper instrument channel

response, alarm and/or initiating action.

4. ~g~~t~

An instrument check is qualitative determination of acceptable operability by observation of instrument behavior during operation.

This determination shall

include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

5.

t t~~~ A logic system functional test means a test of all relays and contacts of a logic circuit to insure all components are operable per design intent.

Where practicable, action will go to completion; i.e.,

pumps will be started and valves operated.

6.

channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function.

A trip system may require one or more instrument channel trip signal's related to one or more plant parameters in order to initiate trip system action.

Initiation of protective action may require the tripping of a single tri.p system or the coincident tripping of two trip systems.

7.

An action initiated by the protection system when a limit is reached.

A protective action can be at a channel or system level.

8.

t t v A system protective action which results from the protective action of the channels monitoring a particular plant condition.

9. 5jzgg~~gg~~~tt~

Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.

BW Unit 2 AMENDMENTNO. T 85

3.. 10 4.10 Applies to the fuel handling and core reactivity limitations.

Applies to the periodic testing of those refueling equipment interlocks and instrumentation required during refueling and CORE ALTERATIONS.

Q~~t' To ensure that core reactivity is within the capability of the control rods and to prevent criticality during refueling.

To verify the OPERABILITY of instrumentation and refueling equipment inter-locks required during refueling and CORE ALTERATIONS.

A.

A.

1.

The reactor mode switch shall be locked in.the REFUEL posit'ion during CORE ALTERATIONS.

The required refueling equipment interlocks shall be OPERABLE'uring in-vessel fuel movement with equipment

,associated with the interlocks except as specified in 3.10.A.6 and 3.10.A'.7 below.

1. Prior to any fuel handling with the head off the

vessel, the following required refueling equipment interlocks shall be functionally tested:

a.

All rods inserted b.

Refueling platform positioned near or oyer the core c.

Refueling platform main, hoist is fuel loaded d.

Fuel grapple is not full up

,e.

One rod withdrawn BFN Unit 2 3 ~ 10/4 10-1'mendment Np. 209

Il

~

11

~5

I G

N 3.10.A.

4.10.A.

4.10.A.l (Continued)

• f.

Refueling platform frame-mounted hoist is fuel loaded

• g.

Monorail hoist is fuel loaded

• h.

'Service platform hoist is fuel loaded They shall be tested at weekly intervals thereafter until no longer required.

They shall also be tested foll'owing any repair work associated with the interlocks.

• NOTE:

These interlocks are required to be OPERABLE only when the associated equipment is used for in-vessel fuel movement.

2.

'Fuel shall not be loaded into the reactor core unless all control rods are ful'ly inserted.

2.

No additional surveillance required.

3.

The fuel grapple hoist load'.switch shall be set

-at g 1,000 lbs.

3.

No additional surveillance required.

4.

If the frame-mounted auxiliary hoist, the monorail-mounted auxiliary hoist, or the service platform hoist is to be, used for.

handling fuel with the.head off the reactor vessel, the load limit switch on the hoist to be used shall be set at

< 400 lbs.

4.

No additional surveillance required.

BFN Unit 2 3.10/4.10-2 Amendment No. 209

II

~, II

3;10.A.

4.10.A.

5.

'Haintenance may be performed on a single control rod or control rod drive without removing the fuel in the control cell if the following conditions are met:

a.

The requirements of.

Specification 3.10.A.1 are met,, and b.

All control rods diagonally and face adjacent to the-maintenance rod are fully inserted and have their directional control valves electrically disarmed.

Prior to performing control rod or control rod drive maintenance on a control cell without removing fuel assemblies the surveillance require-ments of Specification 4.10.A.1 shall be performed and all rods face adjacent and diagonally adjacent to the maintenance rod shall be electrically disarmed per Specification 3.10.A.5.b.

6.

A maximum of two non-adjacent control rods may simultaneously be withdrawn from the core for the purpose of performing control rod and/or control rod drive maintenance without removing the fuel from the cells provided the, following conditi'ons are satisfied:

6.

Prior to performing control rod or control rod.dr'ive maintenance on two control cells simultaneously without removing the fuel from the cells, two SROs shall verify that the requirements of

.Specification 3.10.A.6 are satisfied.

a.

The reactor mode switch shall be locked in the REFUEL position.

The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed for one of the control rods on which maintenance is being performed.

All other, required refueling equipment interlocks shall be OPERABLE.

BFN Unit 2 3.10/4.10-3 Amendment No. 209

3.10.A.

4.10.A.

3.10.A.6 (Cont'd) b.

All directional control valves for remaining control rods shall be disarmed electrically except as specified in 3.10.A.7 and sufficient marg'in to criticality shall be demonstrated.

c.

The two maintenance cells must be separated by more than two control cells in any direction.

d.

An appropriate number of SRMs are available as defined in Specification 3.10.B.

7.

Any number of control rods may be withdrawn or removed from the reactor core providing the following conditions are satisfied:

a.

'The reactor mode switch is locked in the REFUEL position.

The refueling interlock whi'ch prevents more than one control rod from being withdrawn may be bypassed on a withdrawn control rod after the fuel assemblies in the cell containing (controlled by) that control rod have been removed from the reactor core.

All other required refueling equipment interlocks shall be OPERABLE.

7.

With the mode selection switch in the REFUEL or SHUTDOWN mode, no more than one control rod may be withdrawn without first removing fuel from the cell except as specified in 4.10.A.6.

Any number of rods may be withdrawn once verified by '"wo licensed operators that the fuel has been removed from each cell.

BFN Unit 2 3.10/4.10-4 Amendment No. 209

3.10.B.

4.10.B.

1.

During CORE'LTERATIONS, except as specified in 3.10.B.2, two SRMs (FLCs) shal'1 be OPERABLE.

For

,an SRM (FLC) to be considered OPERABLE, the following shall be satisfied:

a.

The SRM shall be inserted to the normal. operating level.

(Use of special

moveable, dunking type detectors during initial fuel loading and major CORE ALTERATIONS in place of normal detectors is permissible as long as the detector is connected to the normal SRM circuit.)

b.

Verify an OPERABLE SRM (FLC) is located in:

2.

Note:

Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM (FLC) and no other fuel assemblies in the associated core quadrant.

.Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS, verify that the associated SRM (FLC) is reading y 3 cps with a signal'-to-noise ratio 2 3:l.

1.

Prior to making any CORE ALTERATIONS, the SRMs (FLCs) shall be functionally tested and checked for neutron response.

1.

The fueled region; 2.

The quadrant where CORE ALTERATIONS are being performed, when the associated SRM (FLC) is included in the fueled region; and

'3. A core quadrant adjacent to where CORE ALTERATIONS are 'being performed, when the associated SRM (FLC) is included in the fueled region.

Note:

One SRM (FLC) may be used to satisfy more than one of the above.,

BFN Unit 2 3.10/4.10-5 Amendment No. 209

LIMITING CONDITIONS FOR'PERATION SURVEILLANCE REQUIREMENTS 3.10.B.

4.10.B 2.

During a complete core

removal, the SRNs shall 'have, an initial minimum count rate of 3 cps prior to fuel removal.

With all rods fully inserted and rendered electrically disarmed and inoperable, once the SRN count rate decreases below 3 cps, the SRNs will no longer be required to be OPERABLE during

'fuel removal.

Individual control rods outside the periphery of the then existing fuel matrix may be electrically armed and moved for maintenance after all fuel in the cell containing (controlled by) that control rod have been removed from the reactor core.

BFN Unit 2 3.10/4.10-6 AMENDMENTNO. g 7Q

Il Il I

A.

The refueling interlocks are designed to back up procedural core reactivity controls during refueling operations.

The interlocks prevent an inadvertent criticality during refueling operations when the reactivity potential of the core is being altered.

To minimize the possibil'ity of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core.

This requirement assures that during refueling the refueling interlocks, as designed, will prevent inadvertent criticality.

The refueling interlocks reinforce operational procedures that prohibit taking the reactor critical under certain situations encountered during refueling operations by restricting the movement of control rods and the operation of refueling equipment.

The refueling interlocks include circuitry which senses the condition of the refueling equipment and the control rods.

Depending on the sensed condition, interlocks are actuated which prevent the movement of the refueling equipment or withdrawal of control rods (rod block).

Circuitry is provided which senses the following conditions.

l.

All rods inserted 2.

Refueling platform positioned near or over the core 3.

Refueling platform main hoist is fuel loaded 4.

Fuel grapple not full up 5.

One rod withdrawn 6.

Refueling platform frame-mounted hoist is fuel loaded 7.

Refueling platform monorail hoist is fuel loaded 8.

Service platform hoist i's fuel loaded When the mode switch is in the REFUEL position, interlocks prevent the refueling platform from being moved over the core if a control rod is withdrawn and'uel is on a hoist.

Likewise, if the refueling platform is over the core with fuel on a hoist, control rod motion is blocked by the interlocks.

When the mode switch is in the refuel position only one control rod can be withdrawn.

The refueling interlocks, in combination with core nuclear design and refueling procedures, limit the probability of an inadvertent criticality.

The nuclear characteristics of the core assure that the reactor is The refueling platform frame-mounted, monorail and the service platform fuel-loaded hoist interlocks are required to be OPERABLE only when utilized for in-vessel'uel movements.

BFN Unit 2 3.10/4.10-11 Amendment iito. 209

subcritical even when the highest worth control rod is fully withdrawn.

The combination of refueling interlocks for control rods and the refueling platform provide redundant methods of preventing inadvertent criticality even after procedural violations.

The interlocks on hoists provide yet another method of avoiding inadvertent criticality.

Fuel handling is normally conducted with the fuel grapple hoist.

The total load on this hoist when the interlock is required consists of the weight of the fuel grapple and the fuel assembly.

This total is approximately le500 lbs, in comparison to the load-trip setting of 1,000 lbs.

Provisions have also been made to allow fuel handling with either of the three auxiliary hoists and still maintain the refueling interlocks.

The 400-1b load-trip setting on these hoists is adequate to trip the interlock when one of the more than 600-lb fuel bundles is being handled.

During certain periods, it is desirable to perform maintenance on two control rods and/or control rod drives at the same time without removing fuel from the cells.

The maintenance is performed with the mode switch in the refuel position to provide the refueling interlocks normally available during refueling operations.

In order to withdraw a second control rod after withdrawal of the first rod, it is necessary to bypass the refueling interlock on the first control rod which prevents more than one control rod from being withdrawn at the same time.

The requirement that an adequate shutdown margin be demonstrated and that all remaining control rods have their directional control valves electrically disarmed ensures that inadvertent criticality cannot occur during this maintenance.

The adequacy of the shutdown margin is verified by demonstrating that at least 0.38 percent ilk shutdown margin is available.

Disarming the directional control valves does not inhibit control rod scram capability.

Specification 3.10.A.7 allows unloading of a significant portion of "he reactor core.

This operation is performed with the mode switch in the REFUEL position to provide the refueling interlocks normally available during refueling operations.

In order to withdraw more than one control rod, it is necessary to bypass the refueling interlock on each withdrawn control rod which prevents more than one control rod from being withdrawn at a time.

The requirement that the fuel assemblies in 'the cell controlled by the control rod be removed from the reactor core before the interlock can be bypassed ensures that withdrawal of another control rod does not result in inadvertent criticality.

Each control rod provides primary reactivity control for the fuel assemblies in the cell associated with that control rod.

Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core.

The requirements for SRN OPERABILITY during these CORE ALTERATIONS assure

)

sufficient core monitoring.

BFN Unit 2 3 >o/4 to-t>

~

Amendment No. 209

0

~

II

3.10 ~ (Cont'd) 3.10.A (Cont'd)

JMXEKN~

1.

Refueling interlocks (BFNP FSAR Subsection 7.6)

B.

The SRMs are provided.to monitor the core during periods of station shutdown and to guide the operator during refueling operations and station startup.

Requiring two OPERABLE SRMs (FLCs) one in and one adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations.

Each SRM (FLC) is not required to read Z 3 cps until after four fuel assemblies have been loaded adjacent to the SRM (FLC) if no other fuel assemblies are in the associated core quadrant.

These four locations are adjacent to the SRM dry tube.

When utilizing FLCs, the FLCs vill be located such that the required count rate is achieved without exceeding the SRM upscale setpoint.

With four fuel assemblies or fewer loaded around each SRM, even with a control rod withdrawn, the configuration will not be critical.

Under the special condition of removing the full core with all control rods inserted and electrically disarmed, it is permissible to allow SRM count rate to decrease below three counts per second.

All fuel moves during core unloading will reduce reactivity. It is expected that the SRMs will drop below three counts per second before all of the fuel is unloaded.

Since there will be no reactivity additions during this

period, the low number of counts will not present a hazard.

When sufficient fuel has been removed to the spent fuel storage pool to drop the SRM count rate below 3 cps, SRMs will no longer be required to be OPERABLE.

Requiring the SRMs to be functionally tested prior to fuel removal assures that the SRMs will be OPERABLE at the start of fuel removal.

The once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY until the count rate diminishes due to fuel removal.

Control rods in cells from which all fuel has been removed and which are outside the periphery of the then existing fuel matrix may be armed electrically and moved for maintenance purposes during full core removal, provided all rods that control fuel are fully inserted and electrically disarmed.

1.

Neutron Monitoring System (BFNP FSAR Subsection 7.5) 2.

Morgan, W. R., "In-Core Neutron Monitoring System for General Electric Boiling Water Reactors,"

General Electric Company, Atomic Power Equipment Department, November

1968, revi:sed April 1969 (APED-5706)

BFN Unit 2 3.10/4.10-13 Amendment No. 209

3.10 ~ (Cont'd)

C.

The design of the spent 'fuel storage pool 'provides a storage location for approximately 140 percent of the full core load of fuel assemblies in the reactor building which ensures adequate shielding, cooling, and reactivity control of irradiated fuel.

An analysis has been performed which shows that a water level at or in excess:of eight and one-half feet over the top of the stored assemblies will provide shielding such that the maximum calculated. radiological doses do not exceed the limits of 10 CFR 20.

The normal water level provides 14-1/2 feet of additional water shielding.

The capacity of the skimmer surge tanks is available to maintain the water level at its normal height for three days in the absence of additional water input 'from the condensate storage tanks.

All penetrations of the fuel pool have been install'ed at such a height that their presence does not provide a possible drainage route that could 1'ower the normal water level more than one-half foot.

The fuel pool cooling, system is designed to maintain the pool water temperature less than 125'F during normal heat loads.

If the reactor core is completely unloaded when the pool contains two previous discharge

batches, the temperature may increase to greater than 125'F.

The RHR system supplemental fuel pool cooling mode will be used under these conditions to maintain the pool temperature to less than 125'F.

D.

The reactor building crane and 125-ton hoist are required to be operable for handling of the spent fuel in the reactor building.

The controls for the 125-ton hoist are 'located in the crane cab.

The five-ton has both cab and pendant controls.

A visual inspection of the load-bearing hoist wire rope assures detection of signs of distress or wear so that corrections can be promptly made if needed.

The testing of the various limits and i'nterlocks assures their proper operation when the crane is used.

E.

The spent fuel cask design incorporates removable lifting trunnions.

The visual inspection of the trunnions and fasteners prior to attachment to the cask assures that no visual damage has occurred during prior handling.

The trunnions must be properly attached to the cask for lifting of the cask and.the visual inspection assures correct.

installation.

BFN, Unit 2 3.10/4.10-14

Although single failure protection has been, provided in the design of the 125-ton hoist drum shaft, wire ropes, hook and lower block assembly on the reactor building crane, the limiting of lift height of a spent fuel'ask controls the amount of energy available in, a dropped cask accident when the cask is over the refueling floor.

An analysis has been. made which shows that the floor and support members in the area of cask entry into the decontamination facility can satis'factorily sustain a dropped cask from a height of three feet.

The yoke safety links provide single fai'lure protection for the hook and lower block assembly and limit cask rotation.

Cask rotation is, necessary, for decontamination and the safety links are removed during decontamination.

4. 1O A.

Complete functional testing of all required refueling equipment interlocks before any refueling outage will provide positive indication that the interlocks operate in the situations for which they were designed.

By loading each hoist with a weight equal to the fuel

assembly, positioning the refueling platform, and withdrawing control

rods, the interlocks can be subjected to valid operational tests.

Where redundancy is provided in the logic circuitry, tests can be performed to assure that each redundant logic element can independently perform its function.

B.

Requiring the SRMs to be functionally tested prior to any CORE ALTERATION assures that the SRNs will be OPERABLE at the start of that alteration.

The once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY.

1.

Fuel. Pool Cooling and Cleanup System (BFNP FSAR Subsection 10.5')

2.

'Spent Fuel Storage (BFNP FSAR Subsection 10.3)

BFN

,Unit 2 3.10/4.10-15 Amendment No. 209

THIS PAGE, INTENTIONALLYLEFT BLANK BFN Unit 2 3.10/4.10-16 Amendment No. 209

~

1

6.2.2 (Cont.)

d.

Two licensed reactor operators shall be in the control room during any cold startups, while shutting down the reactor, and during recovery from unit trip.

In addition, a person holding a senior operator license shall be in the control room for that unit.whenever it is in an operational mode other than cold shutdown or refueling.

e.

A Health Physics Technician* shall be present at the facility at all'imes when there is fuel in the reactor.

f..

Either a licensed SRO or licensed SRO limited to fuel handl'ing who has no concurrent responsibilities during this operation shall be present during fuel handling and shall directly supervise all CORE ALTERATIONS.

g.

Deleted.

• The Health Physics Technician ma9 be absent from the facili.ty for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided 'immedi.ate action is taken to fillthe required position.

BFN Unit 2 6.0-3 Amendment No. 209

I I

t b Senior Operator Senior Operator Licensed Operators Additional Licensed Operators Assistant Unit Operators (AUO)

Shift Technical Advisor (STA)

Health Physics Technician Q

3.

1 1

1 1

0 1

2 2

3' 3

3 0

1 2

2 4

4 5

5 0

1 1

1 1

1 1

1 SRO SRO RO or SRO RO or SRO None None None a..

A senior operator w'ill be, assigned responsibility for overall plant operation at all times there is fuel in any unit.

b.

Except for the senior operator discussed in note "a", the shift crew composition may be one less than the minimum requirements of Table 6.2.A for a period of time not to exceed two hours in order to accommodate unexpected, absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within,the minimum requirements of Table 6.2.A.

This provision does not "ermit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

c.

One of the Additional Licensed Operators must be assigned to each control room with an operating unit.

d.

The number of required licensed personnel, when the operating units are controlled from a common control room, are two senior operators and four operators.

BFN Unit 2 6.0-4 AMENOMEN NO. y g g

0

,0

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 0001 TENNESSEE VAL Y AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR'LANT UNIT 3 AM NDM NT TO FACILITY OP RATING ICENS Amendment No.

166 License No.

DPR-68 The Nuclear Regulatory Commission (the Commission) has found that:

A.

B.

C.

D.

E.

The application for amendment by Tennessee*Valley Authority (the licensee) dated October 9,

1992, and supplemented on March 31,

1993, complies wi,th the standards and requirements of the, Atomic Energy Act of 1954',

as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR'hapter I; The facility will operate in conformity with, the application, the provisions of the Act, and the rules and regulations of the Commission;.

There is reasonable assurance (i) that the activiti'es authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will.be conducted 'in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and.

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

,2.

Accordingly, the license is amended by changes to the Technical Speci.fications as indi'cated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No.

DPR-68 is hereby

'amended'o read as follows:

3.

(2) Technical S ecifications The Technical Specifications contained in Appendices A and B,

as revised.

through Amendment'o.

166, are hereby incorporated in the license.

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

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 9, 1993 rederi cPOO.

Hebdon, irectorM~

Project Directorate II-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT 'NO. 166 FACILITY OPERATING LICENSE NO.

DPR-68 DOCKET NO. 50-296 Revise the Appendix A Technical Specifications by removing the pages identi.fied below and inserting the enclosed pages.

The revised pages are

.identi'fied by the. captioned amendment number and contain marginal lines indicating the area of change.

Overleaf* and spillover** pages are provided to maintain document completeness.

REMOV ill 1V 1.0-7 1.0-8 3.10/4.10-1 3.10/4.10-2 3.10/4.10-3 3.10/4.10-4 3.10/4.10-5 3.10/4.10-6 3.10/4.10-9 3.10/4.10-10 3.10/4.10-11 3.10/4.10-12 3.-10/4.10-13 3.10/4.10-14 6.0-3 6.0-4 INS RT ill 1 V*'.0-7 1.0-8*

3.10/4.10-1 3.10/4.10-2 3.10/4.10-3 3.10/4.10-4 3.10/4.10-5 3.10/4.10-6*

3.10/4.10-9*

3.10/4.10-10 3.10/4.10-1'1**

3.10/4.10-12 3.10/4.10-13*

3.10/4.10-14 6.0-3 6.0-4*

Ci 4l

~ '

F.

Recirculation Pump Operation G.

Structural Integrity H.

Snubbers

3. 7/4. 7 Containment Sys tems 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 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'.

Containment Atmosphere Dilution System (CAD) 3.7/4.7-22 H.

Containment Atmosphere. Monitoring (CAM)

System H2 Analyzer 3.8/4.8 Radioactive Materials A.

Liquid Effluents B.

Airborne Effluents C.

Radioactive Effluents Dose D.

Mechanical Vacuum Pump 3'7/4.7-23a 3.8/4.8-1 3.8/4.8-1 3 '/4.8-2 3.8/4.8-'6 3.8/4.8-6 E.

Miscellaneous Radioactive Materials Sources 3.8/4.8-7 F.

Solid Radwaste 3.9/4.9 Auxiliary Electrical System A.

Auxiliary Electrical Equipment 3.8/4.8-9 3.9/4.9-1 3.9/4.9-1 B.

Operation with Inoperable Equipment.

3.9/4.9-8 C.

Operat'ion in Cold Shutdown 3.10/4.10 Core Alterations A.

Refueling Interlocks B.

Core Monitoring C.

'Spent Fuel Pool Water 3.9/4.9-14 3.10/4.10-1 3.10/4.10-1 3.10/4.10-5 3.10/4.10-7 BFN Uni;t3'mendment No. 166

E.

Spent Fuel Cask D.

Reactor Building Crane 3.10/4.10-8 3.10/4.10-9 F'pent Fuel Cask Handling-Refueling Floor.

3.10/4.10-9 3.11/4.11 Deleted.

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

'5. 0-1 5.0-2 BFN Unit 3 iv AMENDMENT'NO; X, 6 f.

1.0

( ont'd) g.

Qggz~~~~~

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. Egf~~~ta~

Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of'he 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'he next regularly scheduled outage.

S'.

CORE 'ALTERATION shall be the movement of any fuel,

sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel.

Movement of source range monitors, intermediate range monitors,. traversing in-core probes, or special movable detectors (including undervessel replacement) is not considered a

CORE ALTERATION.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe location.

pressures listed in the Technical Specifications.are those measured by the reactor vessel steam space detectors.

U.

l.

g'

't' t

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 bo'iling 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 ~

The highest ratio, for all fuel types in the core-, of the maximum fuel rod power density (kW/ft) for.a given fuel type to the limiting fuel rod power density (kW/ft) for that fuel type.

6 ~lI'~

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 3 1.0-7 Amendment No.

166

0 Il I

1.0 (Cont'd)

V.

t t

t t'gg An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable

range, and accuracy, to a known value(s) of the parameter

.which the instrument monitors.

2.

Chyle~

A channel is an arrangement of the sensor(s) and associated components used to evaluate plant, variables and produce discrete outputs used in logic.

A channel terminates and loses its identity where individual channel outputs are combined in 1'ogic.

3. Lgs~gzy~tpslgt~~~t An instrument functional test means the injection of a simulated signal into the instrument primary sensor to verify the proper instrument channel

response, alarm and/or initiating action.

4. ~times~~

An instrument check is qualitative determination of acceptable operab'ility by, observation of instrument behavior during operation.

This determination shall

include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

5.

means a test of all relays and contacts of a logic circuit to insure all components are operable per design intent.

Where

,practicable, action will go to completion; i.e.,

pumps will be started and valves operated.

6. ~~~t A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to

.initiate action to accomplish a protective trip function.

A trip system may require one or more instrument channel trip signals related to one. or.more plant parameters in order to initiate trip system. action.

Initiation of protective action may,require the tripping of a single trip. system or the coincident tripping of two trip -systems.

'7. ~~tjv~t~ An action initiated by the protection system when a limit is reached.

A protect'ive action can be at a channel or. system level.

8.

t

<jgn A, system protective action which results from the protective action of the channels monitoring a particular plant condition.,

9.

t' t'

.Simulated automatic actuation means applying a -simulated signal to the sensor to actuate the circuit in question.

BFN'nit 3

3.. 10 4.10

'Appl'ies to the fuel handling and core reactivity limitations.

Applies to the periodic testing of those refueling equipment interlocks and instrumentati.on required during refueling and CORE ALTERATIONS.

Q~~v To ensure that core reactivity is within the capability of the control rods and to prevent criticality during refueling.

Q]ggg~v To verify the OPERABILITY.of instrumentation and refueling equipment inter-locks required during refueling and CORE ALTERATIONS.

A.

1.

The reactor mode switch shall be locked in the REFUEL position during CORE ALTERATIONS.

The required refueling equipment interlocks shall be OPERABLE during in-vessel fuel movement with equipment associated with the interlocks except as specified in 3.10.A'.6 and 3.10.A.7 below.

A.

1. Prior to, any fuel handling with the

'head off the

vessel, the following required refueling equipment interlocks shall be functionally tested:

a.

All rods inserted b.

Refueling platform positioned near or over the core c.

Refueling platform main, hoist is fuel loaded d.

Fuel grapple is not full up e.

One rod withdrawn BFN Unit 3 3.10/4.10-1 Amendment No. 166

Ib 4

II

3.10.A.

4.10.A.

4.10.A.1 (Continued)

• f.

Refueling. platform frame-mounted hoist is fuel loaded

• g.

Monorail hoist is fuel loaded

• h.

Service platform hoist is fuel loaded They shall be tested at weekly intervals thereafter until no longer required.

They shall also be tested following any repair work associated with the interlocks.

• NOTE:

These interlocks are required to be OPERABLE only when the associated equipment is used for in-vessel fuel movement.

2.

Fuel shall'ot be loaded into the reactor core unless all control rods are fully inserted.

2.

No additional surveillance required.

3.

The fuel grapple hoist load switch shall be.set at g 1,000 lbs.

3.

No additional surveillance required.

4.

If the frame-mounted auxiliary hoist, the monorail-mounted auxiliary hoist, or the service platform hoist is to be used for handling fuel with the head off the reactor vessel, the load limit switch on the hoist to be used shall be set at

( 400 lbs.

'4.

No additional surveillance required.

BFN,

'Unit 3 3.10/4.10-2 Amendment No. 166

3.10.A.

4. 10.A.

5.

Naintenance may be, performed on a single control rod or control rod drive without removing the fuel in,the control cell if the foll'owing conditions are met:

a.

The requirements of Specification 3.10.A.1 are met, and b.

All control rods diagonally and face adjacent to the maintenance rod are fully inserted and have had their directional control valves electrically disarmed.

5.

Prior to performing control rod or control rod drive maintenance on a control cell without removing fuel assemblies the surveillance require-ments of Specification 4.10.A.1 shall be performed and all rods face adjacent and diagonally adjacent to the maintenance rod shall be electrically disarmed per Specification 3.10.A.5.b.

6.

A maximum of two non-adjacent control rods may be simultaneously withdrawn from the core for, the purpose of performing control'od and/or control rod drive maintenance without removing the fuel from the cells provided the following conditions are satisfied:

6.

Prior to performing control rod or control rod drive maintenance on two control cells simultaneously without removing the fuel from the cells, two SROs shall verify that the requirements. of Specification 3.10.A.6 are satisfied.

a.

The reactor mode switch shal'1 be locked in the REFUEL position.

The refueling interlock which prevents more than one control rod from. being withdrawn may be bypassed for one of the. control rods on which maintenance is being performed.

All other required refueling equipment interlocks shal'1 be OPERABLE.

BFN Unit 3, 3 10/4 10-3

.Amendment No.

166

0

~

~ II

3.10.A.

4.10.A.

'3.10.A.6 (Cont'd) b.

All directional control valves for remaining control rods shall be disarmed electrically except, as specified in 3.10.A.7 and sufficient margin to criticality shall be, demonstrated.

c.

The two maintenance, cells must be separated'by more than two.control cells in any direction.

d.

An, appropriate number of SRNs are available as defined in Specification 3'.10.B.

7.

Any.number of control rods may be withdrawn or removed from the reactor, core provid'ing the following conditions are satisfied:

a.

The reactor mode switch is locked in the REFUEL position.

The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed on a withdrawn control rod after the fuel assemblies in the cell containing (controlled

,by) that control rod have been removed from the reactor core.

All other required refueling equipment interlocks shall be OPERABLE.

7.

With the mode selector switch in the REFUEL or SHUTDOWN mode, no more than one control rod may be withdrawn without first removing fuel from the, cell except as specified in 4.10.A.6.

Any number of rods may be withdrawn once verified by two licensed operators that the fuel has been removed from each cell.

BPN Unit 3 3.10/4.10-4 Amendment No. 166

3.10.B.

1.

During CORE ALTERATIONS, except as specified in 3.10.B.2, two SRMs (FLCs) shall be OPERABLE.

For an SRM (FLC) to be considered OPERABLE, the following shall be satisfied:

a.

The SRM shall be inserted to the normal operating level.

(Use of special

moveable, dunking type detectors during initial fuel loading and major CORE ALTERATIONS in place of normal detectors is permissible as long as the detector is connected to the normal SRM circuit.)

b.

Verify an OPERABLE SRM (FLC) is located in:

4.10.B.

1.

Prior to making any CORE'LTERATIONS, the SRMs (FLCs) shall be functionally tested and checked for neutron response.

2.

Note:

Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM (FLC) and'o other fuel assemblies in the associated core quadrant.

Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS, verify that the associated SRM (FLC) is reading ~ 3 cps with a signal-to-noise ratio Z 3:l.

1.

The fueled region; 2.

The quadrant where CORE ALTERATIONS are being performed, when the associated SRM (FLC) is included in the fueled region; and

3. A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM (FLC) is included in the fueled region.

Note:

One SRM (FLC) may be used to satisfy more than one.of the above.

BFN Unit 3 3.10/4.10-5 Amendment No. 166

~

~

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.10.B.

4.10.B 2'.

During a complete core

removal, the SRMs shall have an initial minimum count rate of 3 cps prior to fuel removal.

With all rods fully inserted and rendered electri.'cally disarmed and inoperable, once the SRM count rate decreases below 3 cps, the SRMs will no longer be required to be OPERABLE during fuel removal.

Individual control rods outside the periphery of the then existing fuel matrix may be electrically armed and moved for maintenance after all fuel in the cell containing (controlled by) that control rod have been removed from the reactor core.

BFN Unit 3 3.10/4.10-6 AMENDMENTHO. I4 8

~

~

LIMITING CONDITIONS FOR OPERATION

,SURVEILLANCE REQUIREMENTS 3.10.D.

4.10.D.

Upon receipt, an empty fuel cask shall not be lifted until a visual inspection is made of the cask-lifting trunnions and fastening connection has been conducted.

1.

Prior to attachment and 1'ifting of an empty spent fuel cask from the shipping trailer, a

visual inspection shall be conducted on the lifting trunnions and the fasteners used to connect the trunnion to the cask.

2.

A visual inspection shall be made of the assembled trunnion on the empty cask to insure proper assembly.

R'LZ LO l.

Administrative control shall be exercised to limit the height the spent fuel cask is rai'sed above the refueling floor by the reactor building crane to 6 inches, except for entry into the cask decontamination chamber where height above the floor will'be, approximately 3 feet.

2.

'The spent fuel cask yoke safety links shall be properly positioned at all times except when the cask is in the decontamination chamber.

BFN Unit 3 3.10/4.10-9

~

I

'A.

The refueling interlocks are designed'o back up procedural core reactivity controls during refueling operations.

The interlocks prevent an inadvertent criticality during refueling operations when the reactivity potential of the core is being altered.

To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core.

This requirement assures

.that during refueling the refueling interlocks, as designed, will prevent inadvertent criticality.

The refueling interlocks reinforce operational procedures that prohibit taking the reactor criti'cal under certain situations encountered during refueling operations by restricting the movement of control rods and the operation of refueling equipment.

The refueling interlocks include circuitry which, senses the condition of the refueling equipment and the control rods.

Depending on the sensed condition, interlocks are actuated which prevent the movement of the refueling equipment or withdrawal of control rods (rod block).

Circuitry is provided which senses the following conditions.

l.

All rods inserted 2.

Refueling platform positioned near or over the core 3.

Refueling platform main hoist is fuel'oaded 4.

Fuel grapple not full'p 5.

,One rod withdrawn 6.

Refueling platform frame-mounted hoist is fuel loaded 7.

Refueling platform monorail hoist is fuel loaded 8.

Service platform hoist is fuel loaded When the mode switch is in the REFUEL position, interlocks prevent the refueling platform from 'being moved over the core if a control rod is withdrawn and fuel is on a hoist.

Likewise, if the refueling platform is over the core with fuel on.a hoist, control rod motion is blocked by the interlocks.

When the mode switch is in the refuel position only one control rod can be withdrawn.

The refueling interlocks, in combination with core nuclear design and refueling procedures, limit the probabil'ity of an inadvertent criticality.

The nuclear characteristi'cs of the core assure that the reactor is The refueling platform frame-mounted, monorail and the service platform fuel-loaded hoist interlocks are required to be OPERABLE only when utilized for in-vessel'uel movements.

BFN Unit 3 Amendment No. 166

ih il

~ 'l

subcritical even when the highest worth control rod is fully withdrawn.

The combination of refueling interlocks for control rods and the refueling platform provide redundant methods of preventing inadvertent criticality even after procedural violations.

The interlocks on hoists provide yet another method of avoiding inadvertent criticality.

Fuel handling is normally conducted with the fuel grapple hoist.

The total load on this hoist when the interlock is required consists of the weight of the fuel grapple and the fuel assembly.

This total is approximately 1,500 lbs, in comparison to the load-trip setting of 1,000 lbs.

Provisions have also been made to allow fuel handling with either of the three auxiliary hoists and still maintain the refueling interlocks.

The 400-lb load-trip setting on these hoists is adequate to trip the interlock when one of. the more than 600-lb fuel bundles is. being handled.

During certain periods, it is desirable to perform maintenance on two control rods and/or control rod drives at the same time without removing fuel from the cells.

The maintenance is performed with the mode switch in the refuel position to provide the refueling interlocks normally available during refueling operations.

In order to withdraw a second control rod after withdrawal of the first rod, it is necessary to bypass the refueling interlock on the first control rod which prevents more than one control rod from being withdrawn at the same time.

The requirement that an adequate shutdown margin be demonstrated and that all remaining control rods have their directional control valves electrically disarmed ensures that inadvertent criticality cannot occur during this maintenance.

The adequacy of the shutdown margin is verified by demonstrating that at least 0.38 percent hk shutdown margin is available.

Disarming the directional control valves does not inhibit control rod scram capability.

Specification 3.10.A.7 allows unloading of a significant portion of the reactor core.

This operation is performed with the mode switch in the REFUEL position to provide the refueling interlocks normally available during refueling operations.

In order to withdraw more than one control rod, it is necessary to bypass the refueling interlock on each withdrawn control rod which prevents more than one control rod from being withdrawn at a time.

The requirement that the fuel assemblies in the cell controlled by the control rod be removed from the reactor core before the interlock can be bypassed ensures that withdrawal of another control rod does not result in inadvertent criticality.

Each control rod provides primary reactivity control for the fuel assemblies in the cell associated with that control rod.

Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core.

The requirements for SRM OPERABILITY during these CORE ALTERATIONS assure sufficient core monitoring.

BFN Unit 3 3.10/4.10-11 Amendment No.

166

0 0

J

<<y

3.10 ~ (Cont'd) 3.1O.A (Cont'd)

EEKEEHQFD 1.

Refueling interlocks (BFNP FSAR Subsection 7.6)

B.

The SRMs are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and station startup.

Requiring two OPERABLE SRMs (FLCs) one in and one adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations.

Each SRM (FLC) is not required to read > 3 cps until after four fuel assemblies have been loaded adjacent to the SRM (FLC) if no other fuel assemblies are in the associated core quadrant.

These four locations are adjacent to the SRM dry tube.

When utilizing FLCs, the FLCs will be located such that the required count rate is achieved without exceeding the SRM upscale setpoint.

With four fuel assemblies or fewer loaded around each SRM, even with a control rod withdrawn, the configuration will not be critical.

Under the special condition of removing the full core with all control rods inserted and electrically disarmed, it is permissible to allow SRM count rate to decrease below three counts per second.

All fuel moves during core unloading will reduce. reactivity. It is expected that the SRMs will drop below three counts per second before all of the fuel is unloaded.

Since there will be no reactivity additions during this

period, the low number of counts will not present a hazard.

When sufficient fuel has been removed to the spent fuel storage

.pool to drop the SRM count rate below 3 cps, SRMs will no longer be required to be OPERABLE.

Requiring the SRMs to be functionally tested prior to fuel removal assures that the SRMs will be OPERABLE at the start of fuel removal.

The once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verification of the SRM count rate and signal-to-noise ratio ensures their continued OPERABILITY until the count rate diminishes due to fuel removal.

Control rods in cells from which all fuel has been removed may be armed electrically and moved for maintenance purposes during full core removal, provided all rods that control fuel are fully inserted and electrically disarmed.

1.

Neutron Monitoring System (BFNP FSAR Subsection 7.5) 2.

Morgan, W. R., "In-Core Neutron Monitoring System for General Electric Boiling Water Reactors,"

General Electric Company, Atomic Power Equipment Department, November 1968, revised April 1969 (APED-5706)

BFN Unit 3 3.10/4.10-12 Amendment No.

166

3.10 ~'Cont'd)

C ~

The design of the spent fuel'torage pool provides a storage location for approximately 140 percent of the full core load of fuel assemblies in the reactor building which ensures adequate shielding, cooling, and reactivity control of irradiated fuel.

An analysis has been performed which shows that a water level at or in excess of eight and one-'half feet over the top of.the stored assemblies w'ill provide shielding such that the maximum calculated radiol'ogical doses do not exceed the limits of 10 CFR 20.

The normal water level provides 14-1/2 feet of additional water shielding.

The capacity of the skimmer surge tanks is avail'able to,maintain the water level at its normal height for three days in the absence of additional water input 'from the condensate storage tanks.

All penetrations of the fuel pool have been installed at such a height that their presence does not provide a possible drainage route that could lower the normal water level more than one-hal'f foot.

The fuel pool cooling system is designed to maintain the pool water temperature less than 125'F during normal heat loads.

If the reactor core is completely unloaded when the pool contains two,previous discharge

batches, the temperatures may increase to greater than 125'F.

The RHR system supplemental fuel'ool cooling mode will be used under these conditions to maintain the pool temperature to less than 125'F.

The reactor building crane and'25-ton hoist are required to be OPERABLE for handling of the spent.fuel in the reactor building.

The controls for the 125-ton hoist are located in the crane cab.

The five-ton has both cab and pendant controls.

A visual inspection of the load-bearing hoist wire rope assures, detection of signs of distress or wear so that correctiohs can be promptly made if needed.

The testing of the various limits and interl'ocks assures their proper operation when the crane is used'.

The spent fuel cask design incorporates removable lifting trunnions.

The visual inspection of the trunnions and fasteners prior to attachment to the cask assures that no visual damage has occurred during prior handling.

The trunnions must be properly, attached to the cask for lifting of the cask and the visual inspection assures correct installation.

BFN Unit 3 3.10/4.10-13

Although single failure protection has been provided in the design of the 125-ton hoist drum shaft, wire ropes, hook and lower block assembly on the reactor building crane, the limiting of lift height of a spent fuel cask controls the amount of energy available in a dropped cask accident when the cask is over the refueling floor.

An analysis

.has been made which shows that the floor and support members in the area of cask entry into the decontamination facility can satisfactorily sustain a dropped cask from a height of three feet.

The yoke safety links provide single failure protection for the hook and. 1'ower block assembly and limit cask rotation.

Cask rotation is necessary for decontamination and the safety links are removed during decontamination.

4 'O 26KÃ A.

Complete functional testing of all required refueling equipment interlocks before any refueling outage will provide positive indication

'that, the interlocks operate in the situations for which they were designed.

By loading each hoist with a weight equal to the fuel

assembly, positioning the refueling platform, and withdrawing control rods, 'the interlocks can be subjected 'to valid operational'ests.

Where redundancy, is provided in the logic circuitry, tests can be performed to assure that each redundant logic element can independently perform its function.

B.

Requiring the SRMs to -be functionally tested prior to any CORE ALTERATION assures that the SRMs will be OPERABLE.at the start of that alteration.

The once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verification of the SRM count rate and signal-to-noise ratio, ensures their continued OPERABILITY.

.1.

Fuel Pool Cooling and Cleanup System (BFNP FSAR Subsection 10.5) 2.

Spent Fuel Storage (BFNP.FSAR Subsection 10.3)

BFN Unit 3 3.10/4.10-14 Amendment No.,$ 66

6.2.2 (Cont.)

d.

Two licensed reactor operators shall be in the control room during any cold startups, while shutting down the reactor.,

and during recovery from -unit trip.

In addition, a person holding a senior operator license shall be in the control room for that unit whenever, it i:s in an operational mode other than cold shutdown or refuel'ing.

e.

A Health Physics Technician* shall be present at the facility at all times when there is fuel in. the reactor.

f.

Either a licensed SRO or licensed SRO limited to fuel handling who has no concurrent responsibilities during this operation shall be present during fuel handling and shall directly supervise all CORE ALTERATIONS.

g.

Deleted.

• The Health Physics Technician may be absent from the facility for a period, of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided immediate action is taken to fillthe required position.

BFN Unit 3 6.0-3 Amendment No.

166

4

.I

~ttg Senior Operator Senior Operator Licensed Operators Additional Licensed Operators Assistant Unit Operators (AUO)

Shift Technical Advisor (STA)

Health Physics Technician 3

3 3'

0 1

2 2

4

'4 5

5 0

1 1

1 1

1 1

1 a

1 1

1

'1 0

1 2

2 SRO SRO RO or SRO RO or SRO None None None.

a.

A senior operator will be assigned responsibility for overall plant operation at all times -there is fuel in any unit.

b.

Except for the senior operator discussed. in note "a", the shift crew composition may be.one less than the minimum requirements of.Table 6.2.A for a period of time not to exceed two -hours in order to accommodate unexpected absence of on-duty shift crew members provi.ded immediate action is taken 'to restore the shift crew compositi'on to within the minimum requirements of Table 6.2.A.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

c.

One of the Additional Licensed Operators must be assigned to each control room wi,th an operating unit.

d.

The number of required licensed personnel, when the operating units are controlled from a common.control room, are two senior operators and four operators.

BFN Unit 3

6. 0-4 AMENDMENTNO. T 2'0

lb P