ML19327B747

Amends 172,175 & 143 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Adding Restrictions to Core Addition Activities by Requiring Continuous Core Monitoring During Refueling

ML19327B747
Person / Time
Site:	Browns Ferry
Issue date:	11/03/1989
From:	Black S Office of Nuclear Reactor Regulation
To:
Shared Package
ML19327B748	List: ML19327B747
References
NUDOCS 8911130046
Download: ML19327B747 (37)

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TENNESSEE VALLEY AUTHORITY DOCKET N0.50-25g SRWNS FERRY NUCLEAR PLANT. UNIT 1 AMENOMENT TO FACILITY OPERAT!NG LICEN$E Amendueat No.172 License No. DPR-33 1.

The Nuclesr Mogulatory Cossission (the Comission) hss found thet:

A.

The application for amendment by Tennessee' Valley Authority (the licensee) dated June 20, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act).

and the Comission's rules and regulations set forth in 10 CFR l

Chapter is 8.

The facility will operate in confomity with the application, the l-provisions of the Act and the rules and regulations of the i

Commission:

C.

There is reasonable assurance (1) that the activities authorized by this amenenent can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in coupliance with the Commission's regulations; D.

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 E.

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

8911130046 E:91303 POR ADOCK 05C40259

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7 2-2.

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 Seecifications The Technical Specifications contained in Appendices A and B. as revised through Amendment No.172 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60, days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMNIS$10N M

Suzanne Black. Assistant Director for Projects

,TVA Projects Division Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: November 3. 1989 l

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ATTACHMENT TO LICENSE AMENDMENT N0.172 FAc!LITY OPERATING LICENSE N0. DPR-33 DOCKET NO.50-25g l

Revise the Appndix A Technical Specifications by removing the pages l

identified below and inserting the enclosed pages. The revised pages t

are identified by the captioned amendment number and contain margir.a1 lines indicating the area of change.

RQgi INSERT

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1.0-7 1.0-7 l

1.0-8 1.0-8*

l 3.3/4.3-11 3.3/4.3-11

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3.3/4.3-12 3.3/4.3-12*

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3.10/4.10-3 3.10/4.10-3*

f 3.10/4.10-4 3.10/4.10-4 3.10/4.10 5 3.10/4.10-5

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3.10/4.10-6 3.10/4.10-6 l

3.10/4.10-13 3.10/4.10-13 3.10/4.10-14 3.10/4.10-14*

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........ -. -. -.. -. ~. _ - _,. _...... _., _ _ _ - - _ _. _.. _ _ - - -. - -. - -.. _ - - -, - -

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,1. 0 DErlE111951 (C et'd)

J q.

omaratina cvela - Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage I

for the same unit.

l R.

Infuelina Outama - 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 :sean a regularly j

schedpled outaget however, where such outages occur within 8 months of the completion of the previous refueling outase, the required

(

surveillance testius need not be performed until the next regularly l

acheduled outage.

3.

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

i Notsal movement of in-core instrumentation and the traversing in-core 1

l probe is not defined as a Core Alteration. Suspension of Core Alterations shall not preclude completion of the movement of a l

l component to a safe conservative position.

i l

l T.

Sameter Vaanal Praamura - Unless otherwise indicated, reactor vessel l

pressuree listed in the Technical Specifications are those sessured by j

l the reactor vessel steam space detectors.

l, U.

Tharmal Paramatara f

l 1.

Pthinum Critical Power Ratie (MCPR) - Minimum Critical Power Ratio (MCPR) is the value of the critical power ratio assceinted with the most limiting asceably in the reactor core. Critical Power

[

Ratio (CPR) is the ratis of that power in a fudi assembly, which l

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

i i

5 2.

Tramattien mailina - Transitiwa boiling means the boiling regime l

between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur l

intermittently with neither type being completely stable.

l The i

3.

Cara M*wi-Fraction of Limittaa Pawar Dammity (CNFLPD) highest ratio, for all fuel types in the core, of the nazimum fuel l

rod power density (kWft) for a given fuel type to the limiting fuel rod power density (kWft) for that fuel type.

4.

Avarmaa Pl==ar Limmar Nant a - catian Rata (APtmen) - The Average l

Planar Erst Generation Rate is applicable to a specific planar height and is equal to the sum of the 11asar 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.

l 1

BTN 1.0-7 Amend.nent No. 158. 172 Unit 1 P

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1.0 DEFINIT 10ER (Cont'd)

V.

Inmirumentattag 1.

Instr e ang calibra1133 - An instraent calibration means the

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adjustaant of an instruent signal output so that it i

corresponds, within acceptable range, and accuracy, to a known l

value(s) of the parameter which the instrument monitors, t

i I"s 2.

Channal - A channel is an arrangement of the sensor (s) ar.d associated components used to evaluate plant variables and

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produce discrete outputs used in logic. A channel terminates

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and loses its identiuy where individual channel outputs are combined in logic.

t 3.

Instrument Functional Tant - An instrument factional test means the injection of a simulated signal into the instru ent primary l

senser to verify the proper instrument channel response, alars and/or initiating action.

l 4.

Instrument check An instrument check is qualitative determination of acceptable operability by observation of instrument bahavior during eparation. This determination shall l

include, where possible, comparison of the instruent with other independent instruments measuring the same variable, f

5.

Lapie swataa runctional Tant - A logie system factional 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 completioni.i.e., p u ps will be i

started and velves operated.

1 6.

Trin svates - A trip system means an arrangement of instrument channel trip signals and au111ary equipment required to I

initiate action to accomplish a protective trip function. A l

trip system may require one or more instrsament channel trip signals related to one or more plant parameters in order to it.itiate trip system action. Initiation of protective action any require the tripping of a single trip systen or the coincident tripping of two trip systems.

i 7.

Prataativa Antion - An action initiated'by the protection system when a limit is reached. A protective action can be at a chamael or system level.

8.

Pratantiva Fjaf4133 - A system protective action which results from the protective action of the channels monitoring a j

particular plant condition.

i 9.

Simulated Automatic Actuation - Simulated automatic actuation i

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

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.BFN 1.0-8 Unit 1

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3,3fa 3 DeAeyIVITT PamThat

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'tftsfTfaC C8MBITInna 19R OPenATION SURVRIttAmet REQUIREMENTS f

3.3.C.

Stres Inmartian Timaa 4.3.C.

seras innartion Times 2.

The average of the screa inser-2.

At 16-week intervals, 10%

tion times for the three fastest of the operable control operable control rods of all red drives shall be scram-groups of four control rods in timed above 400 pais.

a two-by two arrey shall be no Whenever auch scraa time i

greater thant measurements are made, an evaluation shall be made 5 Inserted From Avs. Scram Inser-to provide reasonable j

Fully Withdraus tien Times faae) assurance that proper j

control rod drive 5

0.398 performance is being 20 0.954 maintained.

50 2.120 90 3.800 l

3.

The naminius scram insertion time for 905 insertion of any operable control rod shall not ascoed 7.00 seconds.

D.

Raaetivity Ancmaliam D.

Reactivity Anomalig, The reactivity equivalent of During the startup test the dif ference betweer. the program and startup followins actuel crittsal rod refueling outages, the i

configuration and the expected critical rod configurations i

configuration during power will be compared to the operation shall not exceed 15 Ak.

expected configurations at If this limit is exceeded, the selected operating conditions.

reactor will be placed in the These comparisons will be j

SHUTDOWN CONDITION until the cause used as base data for J

has been determined and corrective teactivity monitoring durins actions have been taken as subsequent power operation i

appropriate.

throughout the fuel cycle.

At specific power operating l

conditions, the critical rod j

configuration will be compared to the configuration expected based upon

'j appropriately corrected past data. This comparison will i

be made at least every full power month.

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l BFN 3.3/4.3-11 Amendment No. 133. 172 Unit 1 9

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l 3.1/4.1 DRAT""Vfff CnM E, TING CONDITION 8 POS OPERATION SURVEILLANCE REQUIREMENTS 3..i.L If Speettications 3.3.C and.D 4.3.E.

Surveillance requirements are above commet be met, an orderly as specified in 4.3.C and.D shutdown shall be initiated and above.

the resetor shall be in the SNUTDOWN CONDITION with!.a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

f.

Scram D12A nna Val - (SDVi F.

Deram Discharma volume (sDv)

]

s 1.

The scram discharge volume 1.a. The screa discharge drain and vent valves shall volume drain and vent j

be operable any time that valves shall be verified l

the reactor protection open PRICE TO system is required to be STARTUP and monthly operable except as thereafter. The valves specitted in 3.3.P.2.

may be closed intermittently for i

testing not to exceed t

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in any 24-hour period during operation.

}

1.b. The scram discharge volume drain and vent valves shall be i

demonstrated OPERABL t

in accardance with Specification 1.0.15.

l 2.

In the event any 3DV drain 2.

When it is determined or vent valve becomes that any SDV drain or IN0PERABLE, PUCTOR POWER vent valve is inoperable, j

OPERATION nay continue the redundant drain or provided the redundant vent valve shall be drain or vent valve is demonstrated operable operable.

immediately and weekly I

thereafter.

3.

If redundant drain or vent 3.

No additional valves become IN0PERABLE, surveillance required.

the reactor shall be in ROT 6

I STANDSY CONDITION within 24 hosts.

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J 3FN 3,3/4.3 12 Amendment No. 133. 159 Unit 1

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• L lo/a. in cant Atm eAvinna LIMI?!NG CONDITIONS FOR OPg8ATION SUtyt!LIANCE REQUltstgNis 3.10.A.

Rafuelina Interlocka 4.10.A. Refuelina Interlocks

]

6. A maximum of two non-

6. Prior to performing adjacent control rods any control rod or control simultaneously be withdrawn rod drive maintenance j

from the cere for the purpose on two control cells of perforsing control rod simultaneously without and/or control tod drive removing the fuel from i

maintenance without removing the cells, two Stos the fuel from the cells shall verify that the provided the following requirements of conditions are satisfied:

Specification 3.10.A.6 are satisfied.

1 t

a.

The reactor mode evitch shall be locked in the RE WEL position. The refueling interlock which r

prevents more than one l

control rod from being i

withdrawn may be bypassed for one of the control l

I rods on which maintenance is being performed. All l

other refueling interlocks shall be OPERAELI.

b.

All directional control valves for remaining l

control rods shall be

[

I disarmed electrically except as specified in l.

3.10.A.7 and sufficient margin to criticality l

shal.1 be demonstrated.

l c.

The two maintenance cells i

j sont be separated by more j

than two control cells in any direction.

f d.

An appropriate number of SENs are available as defined in Specification 3.10.8.

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SF11 3.10/4.10-3 Unit 1 l

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L10/4.10 C0kE ALTERAT10R2 l

l t.tutefan enunivfa=* FOR OPenATION SUtvrit t Amer aroUfitstrere 3.10.A.

natualina intarineks.

4.10.A.

Refuelina Interiecka 7.

Amy number of control rode 7.

With the mode selection may be withdrawn er removed switch in the RE NEL or from the reacter core SNUTDOWN mode, no more providias the following than one control rod l

conditions are satisfied may be withdrawn without first removing a.

The reactor mode switch fuel from the cell is locked in the except as specified in REFUEL position. The 4.10.A.6.

Any number i

refueling interlock which of rods may be i

prevents more than one withdrawn once l

control red from iting verified by two j

withdrawn may be bypassed licensed operators j

ca a withdrawn control that the fuel has been red after the fuel removed from each cell.

l assemblies in the cell containing (controlled i

by) that control rod

~

have been removed fron l

the reactor core. All i

othe'! refueltag interlocks l

aks11 be OPERASLE.

l S.

Cora Manitarint, 5.

Cara Monitorina 1.

During core alterations, Prior to making any except as specified in alterations to the 1

3.10.8.2, two SRMs (FLCs) core, the SRMs (FLCs) l l

shall be OPERASLE, one in shall be functionally l

l and one adjacent to any tested and checked for quadrant where fuel or neutron response.

control rods are being hereafter, while i

l l

moved. For an SRM (FLC) required to be to be considered OPERARLE, OPERA &LE, the SRMs

}

the following shall be will be checked daily satisfied for resposse.

a.

The 8RM shall be l

l inserted to the normal operating level.

(Use of special moveable, dunking type detectors during initial fuel i

loading and major core s

BFN 3.10/4.10-4 Amendment No. 172 Unit 1

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3_. _10__/ 4.18 Cott At_PERAffama i

' LIMITING CO*DIT10Ns rot CPERATION SUtytittAm0E REQUIRIMENTS t

l 3.10.8.

Cara monitarina 3.10.5.1.a.

(Cont'd) alterations in place i

of normal detectors is permissible as long as the detector is connected to the normal SEN circuit.)

l i

b.

When one or more fuel assemblies are in the core, ascept as specified in i

3.10.3.2, the SRM i

(FLC) shall have a minimum indicated

[

reading of 3 eps l

vhile monitoring the loaded assembly

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i (assemblies) with all rods fully inserted in the core.

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l BrN 3.10/4.10-5 Amendment No.172 Unit 1 L

s 3.10/4.10 c0RE ALTERATIONS i

LIMITING CONDIfl958 FOR OPEt& TION SURVIIt useg REQUIREMENTS l

t 3.10.5.

cara Manitaring 4.10.8 Core Monitorina I

2.

During a complete core removal, the FRMs shall have an initial l

minimum count rate of

(

3 cpe prior to fuel removal. With all l

rods fully inserted and rendered i

electrically disarmed and inoperable, once j

the SIM count rate i

decreases below 3 eps, the SRMs will no longer be required to be OPERABLE during i

fuel removal.

j Individual control

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rods outside the l

periphery of the then J

existing fuel matrix i

may be electrics.11y armed and moved for l

maintenance after all i

fuel in the cell containing (controlled J

by) that control rod

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have been removed from i

the reactor core.

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I BFN 3.10/4.10-6 Unit 1

-Amendment No. 172 i

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'3.10 -R&Igt (Cont'd) l 1

B.

Cara Manitarina l

f.

The SRMs are provided to monitor the core during periods of station shutdown and to guide the operrtor during refueling operations and station startup. Requiring two operable SRMs (FLCs) one in and one l

djacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations. The requirement of three counts per second provides assurance that neutron flux is being monitored and ensures that startup i

is conducted only if the source range flux level is above the minimum i

j; assumed in the control rod drop accident. During a full core reload, i

the fuel vill be loaded in control cells that are continuous to l

previously loaded control celle. This provided coupling of the loaded fuel matrix which is being monitored by the SRMs (FLCs).

l t

Under the special condition of removing the full core with all control rods inserted and electrically disarmed, it is persissible to allow SRM I

count rate to decrease below three counts per second. All fuel moves during core unloading will reduce reactivity.

It is espected that the l

SENs will drop below three counts per second before all of the fuel is j

unloaded. Since there will be no reactivity additions during this period, the low number of counts will not present a hasard. When i

sufficient fuel has been removed to the spent fuel storage pool to drop i

the SRM count rate below 3 eps, SRMs will no longer be required to be l

operable. Requiring the SIMs to be functionally tested prior to fuel i

removal assures that the SRMs will be operable at the start of fuel l

rezeval. The daily response check of the SIMs ensures their continued l

operability until the count rate diminishes due to fuel removal.

j Control rede in cells from which all fuel has been removed and which are outside the periphery of the then existir.g fuel matriz may be armed electrically and moved for maintenance purposes during full core i

removal, provided all rods that control fuel are fully inserted and electrically disarmed.

j i

REFERENCES 1.

Neutron Monitoring System (BFNP F8AR Subsection 7.5) 2.

Morgan, W.

R., "In-Core Neutron Monitoring System for General Electric Soiling Water Reactors," General Electric Company, Atomic Power Equipment Department, November 1968, revised April 1969 i

(APRD-5706) i l

1 r

Amendment No. 172 8FN 3.10/4,10 13 Unit 1

o 3.10 &gggg (Cont'd) l C.

Inant Fuel Pool Water The design of the spent fuel storage pool provides a storage location for apprezimately 140 percent of the full core load of fuel assemblies l

4 in the reactor building which ensures adequate shielding, cooling, and l

reactivity control of irradiated fuel. An analysis has been performed l

which shows that a water level at or in excess of eight and one-half i

feet over the top of the stored assemblies will provide shielding such

)

i 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.

i 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 l

st such a height that their presence does not provide a possible i

drainage route that could lower the normal water level more than i

one-half foot.

The fuel pool cooling system is designed to maintain the pool water i

temperature less than 125'T during normal heat loads.

If the reactor core is completely unloaded when the pool contains two previout discharge batches, the temperature may increase to greater than 125'F.

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

i 3.10.n/4.10.n mAeR3 7

I taaetor Buildina crana 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 i

detection of signs of distress or wear so that corrections can be promptly made if needed.

i i

I

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

3.10.1/4,.10.E l'

Inant Fual Caak i

The spent fuel cask design incorporates removable lifting trunnions.

L 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 l

cask for lifting of the cask and the visual inspection assures correct installation.

k BFN 3.10/4.10-14 Unit 1 j

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UNITS 0 states

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NUCLEAR REGULATORY COMMIS$10N i

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WAS**lWO7088. D. C. 70946 '

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i TENNESSEE VALLEY AUTHORITY D0CKET N0. 50-260 BROWNS FERRY NUCLEAR PLANT. UNIT 2 AMEN 00ENT TO FACILITY OPERATING LICENSE j

i Amendeont No. 175 f

l License No. DPR-52 l

1.

The Nuclear Regulatory Comunission (the Comunission) has found that:

f i

A.

The application for amendment by Tennessee Valley Authority (the licensee)datedJune 20. 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comeission's rules and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comeission; C.

There is reasonable assurance (1) that the acti'#ities authorized by this amendment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in compliance with the Commission's rtgulations; D.

The issuance of this amendeent will not be inim' cal to the common i

defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applica)1e requirements have l

been satisfied.

l *

- - - - - -. - - - - - - - - - -. - - - - - - - - ~ ~ ~ - ~ - - - - - - - - - ~ ~ ~ - - - ~

,----_n..._----

r-4 2-2.

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 her6by amended to read as follows:

(2) Technical Specificatiog, The Technical Specifications contained in Appendices A and 6, as revis6d through Amendment No. 175, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specift:ations.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY Com ISSION Jhk b (ctor4et.

Euzanne Black, Assistant D

< TVA P for Projects rojects Division Office of Nuclear Reactor Regulation

Attachment:

l Changes to the Technical Specifications Date of Issuance: November 3, 1989 l

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e-ATTACHMEhi TO LICENSE AMENDMENT N0.175 l

FACILITY CPERATING LICENSE N0. DPR-52 i

DOCKET NO. 50-260 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 l

lines indicating the area of change.

EEEE I.EEB1 t

1.0-7 1.0-7 i

1.0-8 1.0-6*

l 3.3/4.3-11 3.3/4.3-11 3.3/4.3-12 3.3/4.3-12*

t 3.10/4.10-3 3.10/4.10-3*

i 3.10/4.10-4 3.10/4.10-4 l

l 3.10/4.10-5 3.10/4.10-5 3.10/4.10-6 3.10/4.10-6 l

3.10/4.10-13 3.10/4.10-13 l

3.10/4.10-14 3.10/4.10-14*

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1.4 DEFINIT 10ES (Cont'd) l Q.

operatina cvela - Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refuelins outase for the same unit.

l R.

Refuelina Outama - Refueling outage is the period of time between f

the shutdown of the unit prior to a refueling and the startup of the unit after that refuelins. For the purpose of designating frequency of testing and surveillance, a refuelins outage shall mean a i

regularly scheduled outage: however, where such outases occur within s months of the completion of the previous refueling outase, the I

required surveillance testins need not be performed until the r.azt

[

regularly scheduled outase.

l

(

S.

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

vessel. Normal movement of in-core instroentation and the l

traversing in-core probe is not defined as a Core Alteration.

{

Suspension of Core Alterations shall not preclude completion of the t

movement of a component to a safe conservative position.

[

Rameter v ssal Praamura - Unless otherwise indicated, reactor vessel T.

a pressures listed in the Technical Specifir.ations are thosa measured by the reactor vessel steam space detectors.

i i

U.

Thermal Paramatare 1.

Nint - Critical Power Ratio (McPR) - Mihinum Critical Power Ratio (MCPR) is the value of the critical power ratio associated

{

with the most limiting assembly in the reactor core. Critical

+

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

which is calculated to cause some point in the assembly to experience boiling transition, to the actual assembly operating l

power.

2.

Transition Railina - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur l

intermittently with neither type being completely stable.

l 3.

Cara==' - -- Fraction af Lf=(ttna Power Danalty (CNFLPD) - The highest ratto, for all fuel types in the core, of the maximua l

fuel ro4 power density (W/ft) for a siven fuel type to the limiting fuel rod power density (W/ft) for that fuel type.

4.

Avarmaa Flamar Linear Raat nanaration Rata (AF'") - The Average Planar Heat Generation Rate is applicable to a specific planar height and is equal to the sum of the linear heat i

seneration rates for all the fuel rods in the specified badle at the specified height divided by the number of fuel rods in the fuel 'Jandle.

1 i

l l

Brg 1.0-7 Amendment No. 154. 175 l

Unit 2 1

1

i

)

1.0 DEFINIT 10BS (Cent'd)

(

y.

Instrumentation t

1.

Instreant calibration - An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range, and accuracy, to a known l

value(s) of the parameter which the instrument monitors, j

2.

Chpanal - A channel is an arrangement of the sensor (s) and I

associated components used to evaluate plant variables and l

produce discrete outputs used in logic. A channel terminates l

and loses its identity where. individual channel outputs are combined in logie.

3.

Instreant Functional Tant - An instrument functional test means l

the injection of a simulated signal into the instrument primary sensor te verify the proper instrument channel response, alarm and/or initiating action.

l 4

instrument check - An instrument check is qualitative deteksination of acceptable operability by observation of instruneat behavior during operation. This detetuination shall include, where possible, comparison of the instrinsent with other l

independent instruments measuring the saae varisole.

5.

Lemie Sveten Functional Tant - A logic system functional test l

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 j

started and valves operated.

l 6.

Trin System - A trip systen 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 l

signals related to one or more plant parameters in order to l

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.

Protectiva,,Ag11gn - An action initiated by the protection system wk N a limit is reached. A protective action can be at a channel or systen level.

i g.

Protactiva Function - A systen protective action which results from the protective action of the channels monitoring a i

particular plant condition.

(

9.

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

l

\\

i t'

l I

\\

BFN 1.0-3 Amendment No. 165 l

I l

Unit 2

.. _. _. _.. _. _ _.~..._._ _ _..-- _ _ _ _ _

l 1.1/a.1 REACTIVITT coura0L

[

I LIMITING CORDITIONS Fot OPERATION SURVI!LLANCE REQUIREMENTS i

3.3.C.

scram Inmartion Timan 4.3.C. scram innartion Tim.m I

2.

The average of the scram inser-

2. At 16-week intervals, 10%

i tion times for tL taa;? festost of the OPERABLE control OPERASLE control rods of all rod drives shall be scram-stoups of four control rode in timed above 800 pois.

l a two-by-two array shall be no Whenever such scram time i

greater thant asasurements are made, an l

evaluation shall be made X Inserted From Avs. Scram Inser-to provide reasonable Fully Withdrawn tion Ti=== (ame) assurance that proper control rod drive l

5 0.398 perfornance is beins 20 0.954 maintained.

50 2.120 90 3.500 l

3.

The nazimum screa insertion time for 905 insertion of any OPERABLE control rod shall not j

exceed 7.00 seconds.

D.

Raaetivity Anomalian D.

taaetivity Anamaliaa The reactivity equivalent of During the STARTUF test the difference between the program and STARTUP followins actual critir 1 rod refueling outages, the configuration-4nd the expected critical rod configurations l

configuration during power will be compared ti the operation shall not exceed 1X Ak.

arpacted configurations at If this limit is exceeded, the selected operating conditions, reactor will be placed in These comparisons will be l

SEUTDOWN CONDITION until the used as base data for cause has been determined and reactivity monitorins during corrective actions have been subsequent power operation taken as appropriate, throughout the fuel cycle.

At specific power operating conditions, the eritical rod l

configuration will be compared to the configuration expected based upon t

appropriately corrected past date. This comoarison will L

be made at least every full l

power month.

i 1

5 l

l BFN 3.3/4.3-11 Amendment No. 129, 175 Unit 2 t

-r.-

-,..,.,.-,,...r,,~......

m e _,,,_

n-,-a n...-.

l 6

3.3/a.i enAeffv1TT confmot LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3.5.

If Specificatieas 3.3.C and.D 4.3.E.

Surveillance requirements arc l

above canaat be met, an orderly as specified in 4.3.C and.D shutdown shall be initiated and above.

f the reacter shall be in the i

SNUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

F.

scram nia.hmena vaia-- (EDV)

P.

scram Discharme volume (sDV) 1.

Tae scram discharse volume 1.a. The scram discharge l

drain and vent valves shall volume drain and vent I

be OPERABLE any time that valves shall be verified the reactor protection open PRIOR 70 i

system is required to be STARTUP and monthly

[

OPERASLI ext.ept as thereafter. The valves i

specified in 3.3.F.2.

may be closed I

intermittently for testinS not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in any 24-hour period during operation.

1.b. The scram discharge I

volume drain and vent valves shall be j

demonstrated OPERABLE in accordance with Specification 1.0.MM.

t 2.

In the event any SDV drain 2.

When it is determined i

or vent valve becomes that any SDV drain or IN0PERABLE, REACTOR POWER vent valve is INOPERABLE, OPERATION say continue the redundant drain or provided the redundant vent valve shall be i

drain or vent valve is demonstrated OPERABLE OPERA 312.

immediately and weekly thereafter.

3.

If redundant drain or vent 3.

No additional valves become INOPERABLE, surveillance required.

the reactor shall be in NOT STANDBY CONDITION within 24 hosts.

i BFN 3.3/4.3-12 Amendment No. 129, 155 Unit 2 i

.n

I e

a i

e t

3,tafa,in ener At.eematrama t!M!T!pG CONDIT!0R8 FOR OPERATION

$URYt!L1.A303 REQUIRDENTS 3.10.A.

Rafna1 Lam Interlocka 4.10.A. marumlina interiech.

6.

A maximum of two non-

6. Prior to performing adjacent control rods may control rod or control simultaneously be withdrawn rod drive maintenance from the core for the purpose on two control cells of perforetas control rod simultaneously without and/or control rod drive removing the fuel from j

saintenance without removing the cells, two Stos l

the fuel from the cells shall verify that the provided the following requirements of conditions are satisfiedt Specification 3.10.A.6

)

are satisfied.

j 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 byp6 seed j

i for one of the control rods on which maintenance 3

is belas performed. All other refueling interlocks l

shall be OPERABLE.

7 b.

All directional control valves for remaining control rods shall be i

disarmed electrically ascept as specified in 3.10.A.7 and sufficient i

margin to criticality i

L shall be demonstrated.

l I

c.

The two asiatenance cells must be separated by more than two control cells in I

any direction.

l i

d.

As appropriate number of SENs are available as defined in Specification 3.10.5.

i BrN 3.10/4.10-3 Unit 2

_________.____g

'i

..s

\\

I 1.1o/4.10 CORE ALTERATIONS

_ 11ElfiEG_00EI.f{0RS Pot 0FEllAT!0N SURVEILLANCE REQUlkEMENTS 3.10.A.

Rafnaling Intarlocka 4.10.A.

Refuelina interlocka

}i 7.

Any number of control rods 7.

With the mode nelection

(

may be withdrawn or removed switch in the REFUEL or i

from the reactor core

$NUTDOWN mode, no note l

providing the following than one control rod conditions are satisfied may be withdrawn j

without first removing 1

a.

The reactor mode switch fuel from the cell f

is locked in the except as specified in REFUEL position. The 4.10.A.6.

Any number I

refueltag interlock which of rode may be f

prevents more than one withdrawn once i

control rod from being verified by two

{

withdrawn may be bypassed licensed operators on a withdrawn control that the fuel has been rod after the fuel removed from each cell.

assemblies in the cell i

containing (controlled by) that control rod

[

have been removed from 4

the reactor core. All l

other refueling interlocks

{

t shall be OPERABLE.

i l

i B.

Core Monitorina B.

Cora Monitorina 1.

During core alterations, Prior to making any except as specified in alterations to the 3.10.8.2, two SRMs (FLCs) core, the SRMs (FLCs) l shall be OPERABLE, one in shall be functionally l

and one adjacent to any tested and checked for quadrant where fuel of neutron response.

control rods are beine Thereafter, while l

moved. For an SIM (FLC) required to be to be considered CPERABLE, OPERABLE, the SRMs will the following shall be be checked daily for f

satisfied response.

f i

a.

The SRM shall be inserted to the normal operating level.

(Use

(

of special stoveable, l

dunking type detectors during initial fuel loading and major core i

I I

BFA 3.10/4.10-4 Amendment No. 175 i

I Unit 2 1

l l

I i

e' u

J. '

l e

t h; '

3.1o/4.10 cott ALT 11ArtoNs i.rurrima cammiffons FOR OPenATION SURVIfft nCE REQUIREMENTS

{

3.10.5.

Cara ahmilaring i

3.10.5.1.a.

(Cont'd) f f

alterations in place of normal detectors is permissible as long as the detector is connected to the i

normal SRM circuit.)

b.

When one or more fuel

[

assemblies are in the core, sacept as specified in 3.10.5.2, the SRM (FLC) shall have a minimum indicated l

reading of 3 eps while monitoring the i

loaded assembly (assemblies) with all rods fully inserted

{

in the core.

t

~

i i

L i

1

?

l BrN 3.10/4.10-5 Amendment No.175 Unit 2 l

i l

[

3.10/4.10 CQRE ALTERATIONS

[

a LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.10.5.

Care Manitorina 4.10.8 Core Manitorina 2.

During a complete core removal, the SIMs l

shall have an initial minimum count rate of i

3 cpe prior to fuel removal. With all rods fully inserted t

and rendered l

alectrically disarmed and inoperable, once the SEM count ratt decrsases below 3 ups, the SRMs will no i

longer be required to be OP1 TABLE during fuel removal.

Individual control rods outside the periphery of the then existing fuel matrix may be electrically anned and moved for maintenance after all fuel in the cell contatning (controlled l

by) that control rod have been removed from i

the reactor core.

l I

)

BrN 3.10/4.10-6 Amendment No. 175 Unit 2 m-y

-,y-,

,,,-._,s-..,

.,---,,,----y,....----.._%,

,,.-...,_,,-.-c,,,-.--.-,

s s

r.

1 J.1'0 RAAE1 (Cont'd) 5.

cara Manitorina The SENs 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 l

adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations. The requirement of three counts per second provides assurance that neutron flus is being monitored and ensures that startup j

is conducted only if the source range flus level is above the miniaun assumed in the control rod drop accident. During a full core reload, l

)

the fuel will be loaded in control cells that are contiguous to previously loaded control cells. This provides coupling of the loaded fuel matrix which is being monitored by the SENs (FLCs).

Under the special condition of removing the full core with all control rode inserted and electrically disarmed, it is permissible to allow SRM i

count rate to decrease below three counts per second. All fuel moves

]

during core unloading vill reduce reactivity.

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

unloaded. Since there will be no reactivity additions during this period, the low number of counts will not present a hasard. When sufficient fuel has been removed to the spent fuel storage pool to drop l

l the SRM count rate below 3 epe, SRMs will no longer be required to be i

operable. Requirinb the SRMs to be fisketionally tested prior to fuel

(

removal adeures that the SRMs will be operable at the start of fuel removal. The daily response check of the SRMs ensures their continued operability until the count rate diminishes due to fuel removal.

l 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 l

electrically and moved for maintenance purposes during full core removal, provided all rode that control fuel are fully inserted and i

electrically disarmed.

REFEREBCES l

1.

Neutron Monitoring System (BFNP F8AR Subscetion 7.5)

+

f 2.

Morgan, W.

R., "In-Core Neutron Monitoring System for General Riectric Boiling Water Reactors," General Electric Company, Atomic Power Reuipment Department, November 1968, revised April 1969 i

(APRD-5706) l l

l

?

L BFN 3.10/4.10-13 Amendment No. 175 Unit 2 1

1 1,. _... -.. -. - _,...,,.. - _ _ _ _., _ _, _ _.,.. _ _ _. -.., _.. _. _. _ _. _. _ _ _ _. _ _ _. _ _. _ _ _ _ _ _. _. _ _ _ _ _ _. _. _ _

e n

t

• . O I

t 3.10 &&ggi (Cont'd)

(

i C.

Saant Fual Peel Water i

1 The desiga of the spent fuel storage pool provides a storage location f

for approximately 140 percent of the full core load of fuel assemblies in the reacter building which ensures adequate ahtelding, 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 naziam calculated radiolegical doses do not exceed the limits of 10 CFR 20. The normal water level provides 14-1/2 feet of l

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 i

storage tanks. All penetrations of the fuel pool have been installed at such a height that th61r presence does not provide a possible drainage route that could lower the normal water level more than one-half foot.

l 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 mioaded when the pool contains two previous

[

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

The RER system supplemental fuel pool cooling mode will be used under i

these conditions to maintain the pool temperature to less than 125'F.

D.

Raatter Buildina crana The reactor building crane and 125-ton hoist are required to be I

operable for handling of the opent fuel in the reactor building. The I

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

l 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 interlocks assures their proper operation when the crane is used.

E.

Saant Pual Caak l

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 i

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

j i

l BFN 3.10/4.10-14 Unit 2

l l

,e h

A,

/

1

• g UNITED STATES 1

NUCLEAR REGULATORY COMMIS$10N

{

wAsmovorow. o. c. rosos TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 143 License No..OPR-68 1.

The Nuclear Regulatory Commission (the Comission) has found that:

i l

A.

The application for amendment by Tennessee Valley Authority (the 11censee) dated June 20. 1989, complies with the standards and l

requirements of the Atomic Energy Act of 1954, as amended (the Act).

l and the Consission's rules and regulations set forth in 10 CFR i

j_

Chapter It j

L B.

The facility will operate in conformity with the application, the l

l provisions of the Act, and the rules and regulations of the t

Commission; l

C.

There is reasonable assurance (1) that the activities authorized by I

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; D.

The issuance of this amendment will not be inimical to the common l

defense and security or to the health and safety of the public; and E.

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.

i a

i 1

-.. ~. -.

4

[

4 i

l

. l l

2.

Accordingly, the license is amended by thanges to the Technical Specifications as indicated in the attachment to this lin nse amendment and paragraph 2.C.(2) of Facility Operating License No. OPR-68 is hereby amended to read as follows:

l t

(2) Technical Soecifications I

The Technical Specifications contained in Appendices A and B as l

revised through Amendment No.143, are hereoy incorporated in the t

license. The licensee shall operate the fact 11ty in accordance with the Technical Specifications.

)

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

l FOR THE' NUCLEAR REGULATORY C0pg415$10N c.Sr#dw fvt I

)

anne lack. Assistant D tor

\\

for Projects TVA Projects Division Office of Nuclear Reactor Regulation i

l L

Attachment:

)

Changes to the Technical l

Specifications l

Date of Issuance: November 3. 1989

)

I J

l

)

l 0

\\

1.

l 1

l l

l L-i l

i

,2,

, -. - - - - ~ -., -, -,.. - - - -,, - - -., - -. - - - -, -,,. -,.,... -, - -.. - - - - -. - - _ _. - - _. - - _ -.. - _. - -,. -.. - -. -. - - -. -

l

@f.f 4...

. ~,,

1 1.,. '.

i ATTACHMENT TO LICENSE AMENDMENT NO.143 FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 1

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 change.

Rggg INSERT 1.0-7 1.0-7 1.0-8 1.0-8*

3.10/4.10-3 3.10/4.10-3*

3.10/4.10-4 3.10/4.10-4 3.10/4.10-5 3.10/4.10-5 3.10/4.10-6 3.10/4.10-6 3.10/4.10-11 3.10/4.10-11*

3.10/4.10-12 3.10/4.10-12 l

e l

1

)

?

O

. 8 l

2

\\

'~,

1 4,DKFIBITICMS (Cont'd) l q.

Omaratina cvela - 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.

Refuelina Outama - Refueling outage'is.the period of time between j

the shutdown of the unit prior to a refueling and the startup of the l

unit after that refueling.

For the purpose of designating frequency i

of testing and surveillance, a refueling outage shall mean a regularly scheduled outage; however, where such outages occur within l

3 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage.

l 3.

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

. vessel. Normal movement of in-core instrumentation and the p

l tsaversing in-core probe is not defined as a Core Alteration.

I l

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

T.

Raneter Vanaal Praamura - Unless otherwise indicated, reactor vessel pressures listi 9 the Technical Specifications are there measured l-by the reactor 11 steam space detectors.

I' l

U.

Tharsal Paramatara l

1.

Miniana critical Power Ratio (MCPti - Miniaan Critical Power Ratio (MCPR) is the value of the critical power ratio associated L

with the most limiting assembly in the reactor core. Critical L

Power Ratio (CPA) is the ratio of that power in a fuel assembly, l

which is calculated to cause some point in the assembly to

?

l experience boiling transition, to the actual assembly operating h

power.

2.

Igaggition Boilina - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film bodling occur intermittently with neither type being completely stable.

3.

Cora Maxiana frantian af.. Lim

• lina Pavar Eensity. (CMFLPD) - The highest ratio, for all fuel types in the core, of the maxistan fuel rod power density (kW/ft) fer a given fuel type to the l

limiting fuel rod power density (kW/ft) for that fuel type.

4.

Averama Pinnar Liname Raat Caamention Rata (APT""") - The Average Planar Heat Generation Rate is applicable to a specific I

planar height and is equal to the sum of the linear heat generation rates for all the fuel rods in the specified bundle L

at the specified height divided by the number of fuel rods in the fuel bundle.

l..

I:

l BFN 1.0-7 Amendment No. 129, 143 l

Unit 3 l

l

~~r

-,--,.,n.

l

,O e

!1.0 D4FINITIOR4 (Cont't)

V.

Instrsmaantation i

1.

Instrument Calibration - An instrtasent calibration means the adjustaant 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.

Channel - 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.

Instrument Functional Tant - An instrument functional test means the injection of a simulated signal into the instrument primary sensor to verify the proper instrtasent channel response, alarm and/or initiating action.

4.

Instrument Check - An instrument check is qualitative l

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.

I.oale Svaten Functional Tant - A logic system functional test l.

means a test of all relays and contacts of a logic circuit to insure all camponents are operable per design intent. Where practicable, action will go to completion; i.e., pumps will be started and valves operated.

6.

Trin Svaten - 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 L

coincident tripping of two trip systems.

1 7.

Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.

8.

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

9.

Simulated Automatic Actuation - Simuisted automatic actuation means applying a simulated signal to the sensor to actuats the circuit in question.

l-t BFN 1.0-8 Unit 3 m+,----.

,,,,.,..,.,,...,.,-.,,.,,..,.~..m.,m.

.,,,,m,,,.-y,...

.._,_.-....,--.,..-.__...,___.m..

'O I

' J, 3.10/4.10 CORE AfTERAff0Ns LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.10.A.

Rafnalina Interlocks 4.10.A. Refuelina Interlocks

6. A maximus of two non-

6. Prior to performins adjacent control rods may be control rod or control simultaneously withdrawn rod drive maintenance from the core for the purpose on two control cells of performing control rod simultaneously without

)

and/or control rod drive removing the fuel from 1

maintenance without removing the calls, two SR0s the fuel from the cells shall verify that the provided the following requirements of

)

j conditions are satisfied:

Specification 3.10.A.6 l

are satisfied.

[

a.

The reactor mode switch shall be locked in the REFUEL position. The refueling interlock which i

prevents more than one L

control rod from being l.

withdrawn may be bypassed for one of the control i

l rods on which maintenance L

is being performed. All I

other refueling interlocks

(.

shall be OPERABLE.

l b.

All directional control I

valves for remaining control rods shall be disarmed electrically i

axcept as specified in 3.10.A.7 and sufficient margin to criticality shall be demonstrated.

c.

The twc maintenance cells must be separated by more than two control cells in l

any directism.

d.

An appropriate number of SENs are available as defined in Specification 3.10.8.

l' e

BFN 3.10/4.10-3 Unit 3 s

-v,.-

-,,.n-,--.n---...--,,-.,.

..--~---,------.--,-------------.-n----

- ~. - -. - - - - - - - - - - -. - - -

a.

4 i

~

W t

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3.10/4.10 CORE ALTERATI0Rs r

s LIMITING CONDIT!0N8'FOR OPERATION SURVEILLANCE REQUIREMENTS 3.10.A.

Refuelina Inggriocka 4.10.A.

Refuelina Interloeka 7.-

Any number of control rods 7.

With the mode selector A

~

may be withdrawn or removed switch in the RETUEL or from the reactor core SRUTDOWN mode, no more providing the following than one control rod conditions are satisfied:

may be withdrawn without first removing a.

The reactor mode switch fuel from the cell is locked in the except as specified in REFUIL position. The 4.10.A.6.

Any number refueling interlock which of rods may be-prevents more than one withdrawn once control rod from being.

verified by two withdrawn may be bypassed licensed operators on a withdrawn control that:the fuel has been rod after the fuel removed from each cell.

assemblies in the cell containing (controlled l

by) that control rod have been removed from

(

the reactor core. 'All other refueling interlocks shall be OPERABLE.

B.

Core Monitorina B.

Cora Monitorina 1.

During core alterations, Prior to making any l

L except as specified in alterations to the l

3.10.B.2, two SRMs (FLCs) core, the SENs (FLCs) l shall be OPERABLE, one in shall be functionally l

and one adjacent to any tested and checked for quadrant where fuel or neutron response.

l control rods are being moved.

Thereafter, while l

For an SRM (FLC) to be-required to be considered OPERABLE, the OPERABLE, the SRMs i

following shall be satisfied:

will be checked daily for response.

a.

The SEN shall be inserted to the normal operating level.

(Use of special moveable, dunking type detectors during initial fuel loading and major core BFN 3.10/4.10-4 Amendment No. 143 Unit 3

41 f

y.

p ^

l i

8' 2,1o/4.10 core At.Tf9ATIOgg e-LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS q

3.10.B.

Cara Manitaring l

3.10.8.1.a.

(Cont'd)'

j i

I alterations in place of normal detectors is permissible as 1

long as the detector i

is connected to the i

normal SRM circuit.)

b.

When one or more fuel assemblies are in the 1

i core, except as specified in 3.10.B.2, the SRM (FLC) shall.have a minimum indicated' reading of 3 cpm while monitoring the loaded assembly (assemblies) with all rods fully inserted in the core.

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BFN 3.10/4.10-5 Amendment No. 143 Unit 3 3

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',O 3.10/4.10 CORE ALTERAvf0N2 i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.10.8.

Cora._Manitorina 4.10.8 core Monitorin.

l 2.

Durind a complete core removal, the SRMs shall have an initial minim e count rate of 3 eps prior to fuel removal. With all rods fully' inserted and rendered electrically disarmed and inoperable, once the SEM 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 hava.been removed from the reactor core.

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BFN 3.10/4.10-6 Amendment No.143 Unit 3 Y

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e y

, '3.10 R&&E1 (Cont'd) 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 l

inadvertent criticality.

Fuel handling is normally conducted with the fuel grapple hoist. The total load on this' hoist when the interlock is required consicts of the weight of the fuel grapple and the fuel assembly. This total is approximately 1,500 lba, in comparison to the load-trip setting of i

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 t

is adequate to trip the interlock when one of the more than 600-lb fuel bundles is being handled.

l During certair periods, it is desirable to perform maintenance on two control rods and/or control ro4 drives at the same time without removing fuel from the cells. The n.aintenance 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 fron-being withdrawn at the same time. The requirement that an adequate E

shutdown margin be demonstrated and that all remaining control rods have their directional control valves electrienlly disarmed ensures that inadvertent criticality cannot occur during this maintenance.'

The adequacy of the shutdown margin is verified by demonstrating that i

at least 0.38 percent Ak 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 che 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 L

than one control rod, it is necessary to bypass the refueling l

L interlock on each withdrawn control rod which prevents more than one centrol rod from being withdrnwn 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

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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.

REFERENCES

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l 1.

Refueling interlocks (BFNP FSAR Subsection 7.6)

BFN 3.10/4.10-11 Unit 3 l~

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t 3.10 Rgg31 (Cont'd) i I

B.

cara Manitorina The SENs are provided to monitor the core during periods of station j

shutdown and to guide the operator during refueling operations and station startup. Requiring two OPERABLE SRMs (FLCs) one in and one l

. adjacent to any core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such i

alterations. 'The requirement of three counts per second provides i

assurance that neutron flux is being monitored and insures that startup is conducted only if the source range flux level is above the minimum assJaed in the control rod drop accident. During a full core reload, the fuel vill be loaded in control cells that are contiguous to previously loaded control cells. This provides coupling of the loaded i

fuel matrix which is being monitored by the SRMs (FLCs).

Under the special condition of removing the full core with all control h

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 l'

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 hasard. When

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sufficient fuel has been removed to the spent fuel storage pool to drop the SEM count rate below 3 cys SRMs will no longer be required to be operable. Requiring the SRMs to be functionally tested prior to fuel removal assures that the SIMs will be OPER&BLE at the start of fuel removal. The daily response check ef the SRMs ensures their continued operability until the count rate disintahes 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 rode that control fuel are fully inserted and electrically disarmed.

References 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 Departmenc, November 1963, revised April 1969 (APED-5706)

BFN 3.10/4.10-12 Amendment No. 143 Unit 3 n

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