ML19318A631

Proposed Tech Spec Revisions for safety-related Sys.Change Justifications Encl

ML19318A631
Person / Time
Site:	Browns Ferry
Issue date:	06/13/1980
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML19318A626	List: ML19318A625 ML19318A631
References
NUDOCS 8006230520
Download: ML19318A631 (22)

Text

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ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION REVISIONS BROWNS FERRY NUCLEAR PIANT l

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1

1.0 DEFINITIONS

'The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.

A.

Safety Limit - The safety limits are limits below which the reason-able maintenance of the cladding and primary systems are assured.

Exceeding such a limit requires unit shutdown and review by the Atomic Energy Commission before resumption of unit operation.

Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory-review.

E.-

Limiting Safety System Setting (LSSS) - The limiting safety system setting are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represent margin with normal osaration lying below these settings. The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.

C.

Limiting Conditions for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.

1.

In the event a Limiting Condition for Operation and/or associated requirements cannot be satisifed because of circumstances in excess of those addressed in the specifi-cation, the unit shall be placed in at least Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown with the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are completed that permit operation under - the permissible discovery or until the. reactor is placed in an operational condition in which the specification is not applicable. Exceptions to these requirements shall be stated in the individual specifications. This provides actions to be taken for circumstances not directly provided for in the specifications and where occurrence would violate the intent of the specification. For example, if a specification calls for two systems "or subsystems) to be operable and provides for explicit cequirements if one system (or subsystem) is inoperable, then if both systems (or subsystems) are inoperable the unit is to be in at least Hot Standby in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> if the inoperable condition is not corrected. _

1~.0. DEFINITIONS (continued) 2.-

When a system, subsystem, train, component or device is determined to be inoperable solely because its onsite power source is inoperable, or solelybecause its offsite power source is inoperable, it may be considered operable for the purpose of satisfying the i

requirements of its applicable Limiting Condition For Operction, provided:

(1) its: corresponding offsite or diesel power source is operable; and (2) all of. its redundant system (s), subsystem (s), train (s),

1 component (s) and device (s) are operable, or likewise satisfy these requirements unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least Cold Shutdown within the following

-30 hours. This is not applicable if the unit is already in Cold Shutdown.or Refueling. This provision describes what additional conditions must be satisfied to permit operation to continue

. consistent with the specifications for power sources, when an F

.offsite or onsite power source is not operable. It specifically prohibits operation when one division is inoperable because its offsite or diesel power source is inoperable and a system, i :j' subsystem, train, component or device in another division is inoperable for another reason. This provision permits the i

requirements associated with individual systems, subsystems, trains, components or devices to be consistent with the requirements of.the associated electrical power source. It allows operation to be governed by the time limit of the requirements associated with the Limiting Condition For Operation for the offsite or diesel pcwer source, not the individual requirements for each l

system,. subsystem, train, component or device that is determined to be ' inoperable solely because of the inoperability of its offsite or diesel power source.

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1.0 DEFINITIONS (cont'd)

E.

Operable - Operability

.. system, subsystem, train, component, or device shall be Operable or have operability when it is

. capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary 3-attendant instrumentation, controls, normal and emergency electrical power sources, cooling or. seal water, iubrication or l'

other, auxiliary equipment that are required for the system,

' subsystem.. train,. component or device to perform its function (s) are also capable of' performing their'related support function (s).

F.-

Operating - Operating means that a. system or component is performing

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its intended functionsiin its required manner.

s

G.

. Immed'inte - Immediate. means that the required action will be

-initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.

H.

' Reactor Power Operation - Reactor power operation is any operation with the mode switch in the "Startup" or "Run" position with the reactor critical and above 1% rated power.

I.

Hot Standby Condition - Hot standby condition means operation with coolant temperature greater than 212*F, system pressure less than 1055 psig, the main steam isolation valves closed and the node switch in the Startup/ Hot Standby position.

J.'

Cold Condition - Reactor coolant temperature equal to or less than 212*F.

K.

Hot Shutdown - The reactor is in the shutdown mode and the reactor coolant temperature greater than 212*F.

L.

Cold Shutdown - The reactor is in the shutdown mode, the reactor coolant temperature equal to or less than 212*F, and the reactor vessel-is vented to atmosphere.

M.

Mode of Operation - A reactor mode switch selects the proper interlocks for the operational status of the unit. The following are the modes and interlocks provided:

1.

Startup/ Hot Standby Mode - In this mode the reactor protection scram trips initiated by condenser low vacuum and main steam line isolation valve colsure, are bypassed when reactor L

pressure is less than 1055 psig, the reactor protection system is energized with IRM neutron monitoring system trip, the APRM 15% high flux trip, and control rod withdrawal interlocks in service.- This is of ten referred to as just Startup Mode.- This is intended to imply the startup/ Hot Standby position of the mode switch.

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1

2. only)

(onit

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l.1 BASES Decause the boiling transition correlation is based on a large quantity of rull scale data there is a very high confidence that operation nf a fuel nasembly at the condition of HCPR =1.07.would not produce. boiling tran-sition. Thus, although it is not required to establish the safety limit additional margin. exists between the safety limit and the actual occurence 1

of loss of cladding integrity.

)

Novever, if boiling transition were to occur, clad perforation would not be erpected. Cladding temperatures would increase to approximately 1100 T vnteh is below the perforation temperature of the cladding 0

material. This has been verified by tests in the General Electric Test Reactor (CETR) where fuel similar in design to BFNP operated above' the critical heat flux for a significant period of time (30 stautes) without clad perforation.

If reactor p essure should ever exceed IbOO psia during normal power operating (the limit of applicability of the boiling transition corre-lation) it would be assumed that the fuel cladding integrity Safety Limit' has been violated.

In addition to the boiling transition' limit (MCPR = 1.07 a,. cation is constrained to a maxinum LHCR of 18.5 kv/f t for 7x7 fuel and 13.4 kv/f t f r.

ggg 8x8 fuels. This linit is reached when the Core Haximun Fraction of Limiting Power Density equals 1.0 (CHTLPD = 1.0).

For the case where Core Maximus Traction of Limiting Power Density exceeds the Traction of Rated Thermal Power, operation is permitted only at less than 100% of rated power and only with reduced APRM scram settings as required by specification 2.1.A.1.

At pressures belov 800 psia, the core eieration pressure drpp (0 pove'r, O flow) is greater than h.56 psi.

At low powers and flows this pressure l

differential is maintained in the bypass region of the core. Since the pressure drop in the bypass regl' n is essentially all elevation head, o

the core pressure drop at lov powers and flov vill alvsys be greater than b.56 psi. Analyses show that with a flow of 28X10J 1bs/hr bundle flov, bundle pressure drop is nearly independent of bundle power and has i

a value of 3.5 pai. Thus,3the bundle flov with a b.56 psi driving head vill be greater than 28x10 lbs/hr. Full scale ATLAS test data taken at pressures from Ib.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3 35 MWt. With the design peaking factors this corresponds to a core thermal power of more than 50%. Thus, a core themal power limit of 255 for reactor pressures' belov 800 psia is conservative.

For the fuel in the core during periods when the reactor is shut down, con-sideration must also be given to water level requirements due to the effect of decay heat. If water level should drop below the top of the ruel during this time, the ability to remove decay. heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and elad perforation. As long as the fuel remains covered with vnter, sufficient cooling is available to prevent fuel clad perforation.

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s TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCRAM) INSTR *JMENTATION REQUIREMINT Min. No.

of Operable Inst.

Modes in Which Function Channels Must Se Operable Per Tr1P Shut-Startup/ Mot System (1)

Trio Function Trip Level Setting down Refuel (7)

Standby M

g,,g,nggy 1

Mode Switch in Shutdown y

1 Manual scram x

x x

x g,g IRM (16) 3

!!igh Flux

• 12J/ g ndicated X(22) xM X

W U

3 Inoperative g

w APRM (16) 2 High Flux See Spec. 2.1.A.1 2

Htqh Flux

< 15% rated power x

1.A or 1.3 X(21) x(17)

(15)

1. A or 1.3 2

Inoperative (13) 2 counseale

> 3 Indicated on Scale X(21)

X(17)

X l A #f 1.3 (11)

(1 )

1(12) 1.'A or 1.5 2

High Reactor Pressure < 1055 peig g({0)

X X

l'A 2

Hir.h Orvve11

• 2*5 psig Pressure (14)

X(8)

X(8)

X 1*^

i 2

Reactor Lov Water

> 539" above vessel 5

Level (14)

K X

X 1A

~

4 1

2 Hieh Water Level in Scram Discharge Tank

-< 50 Callons x

X(2)

X X

l'A j

.TC LE 3.2.0

~

INSTILMD." TAT 10N TT.Af INITIATLS OR COCCLS THE CORI AA CCNTA110ENT COOLING SYSTIMS

.tiatAna No.

C;crable Per Trs, Sys (1)

Function Trip level Setting Action Remarks 2

lustrument Channel -

1970[at,ove vessel sero.

A

1. Below trip setting initiated RFC1.

Resctor Im Water Level 2

Instruzent Channel -

~> &70*above vessel zero.

A

1. Multiplier relays. initiate RCIC.

EAact")'r lov Vater Lev 41 2

Instrument Channel -

3,378" above vessel seto.

A

1. Below trip se$ttag initiates Css..

' deactor Lov Vater Level Moltiplier relays inittete LFCI.

(L15-3-53A-0. su #1)

2. Multiplier. relay from CSS initiates 5'j accidest signal (15).

9 2(16)

Ica rursent Chel -

3, 378" ahora vessel sero.

A

1. telow trip settings in conjunction Ecactor Im Uotar Level with dryvell high pressure. low (LIS-3-58A-D. SV #2) water level peratesive. 110 sec. del tia:er and CSS or 112 pm running.

isitiates ADS.

1(16)

Instru:.ent Channel -

1 SM" above vessel zero.

A

1. Eclow trip setting permissive for reactor low Water Level initiating signals on ADS.

Pers1.sive (L15-3-1M &

f 16). Su #1) 1 Instrument Channel -

1 312 5/16" abcvc vessel zero.

A

1. Below trip sat' tog prevents inadver-Reac:or Lov uater te el (2/3 core height) tant operation of contaicaent spray (LITS-3-52 & 62. SU #1) during accident conditica.

2 Instruz. cat Channel -

11 p12.5 psig A

1. Eclow trip setting prevents inadver-Drywelt High Pressore tent operation of conteicnent ePray

~

fr0-64-58 E-u) during accident conditions.

Amendment No. 40

TA3LE 3.2.5 (Centinued)

Min!=ue No.

dyerable Per

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Trin sys (1) runction Trip Level setting Action Renarks' 2

Instrument Channel -

• 2.5 psig A

1. Above trip settins. in conjunction wt Drywell High Pressure low reactor pressure initistes CSS.

(PS-64-58 A-D, SV 12) hultiotter relays Lnitiate HPCI.

2. ?tultiplier relay f rom CSS initiates acc14ent signal.(15).

A

1. Belo s trip setting trips recirculed 2

Instrument Channel -

y,4/70"above vessel zero Reactot Low Water Level tion pumps (LS-3-56A. B, C, D) 2 Instrument Channel

c1120 psig A

1. Above trip setting trips rectrcula.

Reactor High Pressure tian pwups e,-

(PS-J-404 A,

5. C, D) 2 Inst rument Channel -

1 2.5 psig A

1. Above trip setting in conjunction v1 Drywell High Pressure low reactor pressure initiates LPCI.

(PS-64-saA-D, SV #1) 2(16)

Instrument Chennel -

12.5 psig A

1. Above trip setting in conjunctica vi Dryvell High Pressure low reactor water level, dr7vell hig (TS-66-57A-D) pressure,120 sec. delay timer and C or RER pump running, ' initiates ADS.

f, l

.2 lastrument Channel -

L30 psig f; 15 A

1.

elow tri; setting pernissive far :penirt Reactor Lov Pressure CSS and l?CI admissi r. valles.

(PS-3-74-A & B, SV #2)

(PS-63-95, SV #2)

(PS-43-96, SV # 2) 2 Instrument channel -

230 peig 3; 15 A

1. Re:irculation discharge valve Rea c t o r Low Pressure ac tu ation.

(PS-3-74A & B, SV #1)

(PS-68-95, su 11)

(PS-65-96, SV #1)

Amendment f:o. 40 I

TABLE 3.2.F SURVEILLANCE INSTRUMENTATION Minimum i of Operable Instrument Type Indication Chsnacis Instrument #

Instrument and Range Notes 2

LI-3-46 A Reactor % ter Level Indiestor -107.5" to (1) (2) (3)

LI-3-46 a

+107.5" 2

FI-3-54 Reactor Pressure Indicator 0-1200 psig (1) (2) (3)

PI-3-61 2

PR-64-50 Dryvell Pressure Recorder 0-80 psia (1) (2) (3)

PI-64-67 Indicator 0-80 psia 2

TI-64-52 Dryvell Temperature Recorder Indicator (1) (2) (3) 12-64-52 0-4CD'F 1

TR-64-52 Suppression Chamber Air Recorder 0-400*F (1) (2) (3)

Tes:perature y

cn 2

TI-44-55 scppression Chamber Vater Indicator. 0-400*F (1) (2) (3)

TIS-64-55 Tc peratura 2

LI-64-54 A suppression Chamber Vater Indicator -25" to (1) (2) (3)

LI-44-G6 Level

+25" 1

NA Control Rod Tosition 67 Indicating

)

Lights

)

1 rtA Neutron Monitoring SRM, IRM, LPRM

)

(1) (2) (3) (4' 0 to 1001 power) 1 PS-64-47 Dryvell Pressure Alara at 35 psig )

)

j 1

TR-44-52 and Dryve11 Temperature and Alarm if temp.

)

PS-64-58 B and Precoure and Timer

> 281*F and

)

(1) (2) (3) (4' I5-44-67 pressure > 2.5 psid j

after 30 minute

)

delay) l 1

f.1-84 ?A CAD tank "A" level Indicator 0 to 10M (1) 1 1.i-84-1 M CAD tank "C" level Inlicator O to 100(

(1)

FIGURE 3.6-1 CURVE #1 Minimum temperature for pressure tests such as required by Section XI I

2 3

1200

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Z '.' ~ ~

Il, CURVE #2 l'

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Minimum temperature L

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for mechanical heat up or cooldown l

i following nuclear i,

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shutdown 1000 I

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i CURVE #3

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~ j+-*--/-

Minimum temperature for core operation i

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g f_

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(criticality)

'i 800

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Includes additional

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margin req'd by

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10CFR50 Appendix G g

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400

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-- BOLT.UP.'I. E. M.R.(CT ~~ ~~

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O 100 200 300 400 MINIMUM TEMPERATURE ABOVE CHANGE IN TRANSITION TEMPERATURE

(*F)

(1) As measured on closure flanges, adjacent head, and shell material 188

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' UNIT 2 ONLY-Revised 1-31-80 C.

1.l!!1TitiC_C0!iDITIONM POR OPFRATION SURVEIffANCE REOUIRFEE!TTS 1.9.A Auxiliary Electrical Equipment.

4.9.A

, Auxiliary Electrical Equipment common transformer and the specified time sequence.

cooling tower trans-former capable of sup-

c. Once a month the quantity plying power to the of diesel fuel available shutdown boards.-

shall be logged, b.

A fourth operable units

d. Each diesel generator 1 and 2 diesel generator, shall be given an annual

4. Buses and Boards Available inspection in accordance with instructions based a.

Start buses lA and 1B on the manufacturer's are energized.

recommendations.

b.

The units 1 and 2 4-kV

e. Once a month a sample of shutdown boards are diesel fuel shall be energized.

checked for quality. The quality shall be within c.

The 480-kV shutdown the acceptable limits boards are associated specified in Table 1 with the unit are ener-of the latest revision gized to ASTM D975 and logged.

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d.

Undervoltage relays 2.

D.C Power System - Unit operable on start Barreries (250-volt) Diesel buses 1A and IB and 4-kV Cenerator Batteries (125-Volt' shutdown boards, A, B, and Shutdown Board Batteries C, and D.

(250-Volt)

5. The 250-Volt unit and shut-a.

Every week the specific dcwn board batteries and a gravity and the voltage o!

battery charger for each the pilot cell, and battery and associated temperature of an adjacen:

battery boards are operabic.

cell and overall battery voltage shall be measured

6. Logic Systems and logged.

a.

Common accident signal b.

Every three months the logic system is operable measurements shall he made of voltage of each b.

480-V load shedding cell to nearest 0.1 volt.

logic system is operable specific gravity of each cell, and temperature of

7. There shall be a minimum of every fifth cell. 1.se 103,300 gallons of diesel measurements shall be fuel in the standby diesel 1 gged.

. generator-fuel tanks.

c.

A battery rated discharge i

(capacity) test shall ta performed and the voltage,

('

time, and output currer.t measurements shall be logged at intervals not.

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to exceed 24 months'.'~ -

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1.0 DEFINITIONS

/

The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.

A.

Safety Limit The safety limits are limits below which the reasonable maintenance of the cladding and primary systems are assured. Exceeding such a limit requires unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation.

Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.

B.

Limiting Safety System Setting (LSSS)- The limiting safety system setting are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits wil' not be exceeded.

The region between the safety limit and these settings represent margin with normal operation lying below these settings. The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.

C.

Limitina Conditions for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.

1.

In the event a Limiting Condition and/or associated requirements cannpt be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in at least Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are completed that permit operation under the' permissible discovery or until the reactor is placed in an operational condition in which the specification is not applicable. Exceptions to these requirements shall be stated in the individual specifications. This provides action to be taken for circumstances not directly provided for in the specifications and whose occurrence would violate the intent of the specification. For example, if a specifi-cation calls for two systems (or subsystems) to be operable and provides for explicit requirements if one syst.em (or sub-systems) is inoperable, then if both systems (or subsystemq are inoperable, the unit is to be in at least Hot Standby in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> if the operable condition is not corrected.

2

1.-

11.0 DEFINITIONS (cont'd)

^

2.=

When a system, subsystem, train, component or device is

-determined to be inoperable solely because its onsite apower source is inoperable, or solely because its offsite

. power source is inoperable, it may be considered opererable

.for the-purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:

(1) its corresponding offsite'or diesel power source is-operable and (2) all of its redundant system (s), subsystem (s),

train (s), component (s) and device (s) are operable, or likewise satisfy these requirements unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least Cold Shutdown within the'following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This is not applicable 4

if the-unit is already in Cold Shutdown or Refueling. This provision describes what additional conditions must be satisfied to permit operation to continue consistent with the specifications for power sources, when offsite or onsite power sources are not operable.. It specifically prohibits operation when one division is inoperable because its offsite or diesel power source is inoperable and a system, subsystem, train, component or. device in another division i

is inoperable for another reason..This provision permits the requirements associated with individual systems, subsystems, trains, components or devices to be consistent with the requirements of the associated electrical power source. It allows operation to be governed by the time limits of the requirements associated with the Limiting Condition for Operation for tbs offsite or diesel power source, not the individual. requirements for each system, subsystem, train, component or device that is determined to be inoperable solely because of the inoperability of its offsite or diesel power source.

D.

DELETED 4

E.

Operable - Operability - A system, subsystem, train, component or device shall be operable or have operability when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, i

controls, normal and emergency electrical power sources, cooling or 1

-seal water lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

F.

Operating - Operating means that a system or component is performing its intended functions in its required menner.

G.-

Immediate - Immediate means that the required action will be initiated as soon as pract'icable considering the safe operation of the unit and

'and importance of the required action.

H.

Reactor Power Operation - Reactor power operation is any operation with.the mode switch in-the "Startup" or "Run" position with the-

~

. reactor. critical and above 1% rated power.

2a

TABLE 3.1.A REACTOR F90fBCTIN ST3?De (SCRAM) EtsTRantwrRTIopt REQUIREMENT Mija. 300 of operable Peodes in adhich Function Inst.

tmst Be Operable Cunnele stat-startup/ sot Per Trip System f11 Trip Functices Trio level Settine dowrt p* fuel 01 Standby Engg Actiontti X

X X

X.

1.A 1

ptode Switets in Shetdown X

X X

X 1.A 1

Manual scram IRM (16) 3 algh rius 1 120/125 Indicated on scale X(22)

X (22)

X (5) 1.A X

X (5) 1.A 3

Inoperative APRn (16)

X 1.A.ar 1.3 La 2

Sigh rium See Spec. 2.1.A.1 N

2 sigh rius 5 ISS rated power I (21)

X(173 (15) 1.A or 1.3 If X(17)

X 1.A or 1.5 2

Inoperative (13) 2 Dcunuc.sle 2 3 Indicated on Scale (11}2N (11)

X(12) 1.A or 1.5 2

Eigh Re.nctor Pressure 5 1055 poig X(10)

X r

1.A 2

Righ Dryuell treneure (143 s 2.5 psig X (s)

X(eg I

1.A 2

Reactor Low Water Level (14) 4 53s= above vessel sero X

X X

1.A 2

sigh Water 14 vel in Scram Discharge Tank 1 S0 Gallone X

X(2)

X X

1.A 4

Main Steaes Lin's Isola ~

+1on valve closure 5 105 Valve closure Z(3) (6)

Z(3) (6)

Z(6) 1.A or f.C l

2

. Turbine cont. valve Opon trip of the feet

~X(4)

Z(4) 1.A or 1.0 tant closure acting oolenoid valwee X(4)

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i 1

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s Tabl e 3.2. D IV~TJ2NTAT!DN TEAT I:3171ATI3 Cr. NFeStr 72E a ;.I scD CatcAIri.%I r* CCCIE E Y 51 L".5 Ninitur. N2.

'Cperable Per

' Tri n Sys til Functio 9 Trio fevel tettino peylog Renarks 2

- Instrument Ch annel -

tips 2.5 psig A

1.

selow trip settini prevents Drywell High Pressure -

inadverter.t oFeration of (PS-64-58 E-up containnent spray durang accident conditions.

2 Lnstrument Channel -

s2.5 psig a

5.

Anove trip settins in -

Drywell High Pressure conjuncticn with low reactor (PS-64-5 4 A-D, SW 82) pressure initiates CSS.

Multiplier relays iniciate H7CI.

2.

Multiplier relay f rom CSS -

initiates accident signal. (15) 2 Instrument Channel -

2 44TO'ab2ve vessel' t rw A

1.

Below trip settirq trips Reactor Low Water Level recirculation tumps (LS 56 A, B, C, D) 2 Instrument Channel

$1120 psig A

1.

Above trip setting trips Ecactor High Pressure recirculation rumps -

(PS-3-204 A, B, C, 0) ue 2

Instrument Channel--

s 2.5 psig A

1.

Above trip settini in Drywell High Pressure conjunction with low rea: tor (PS-64-55A-D, SW f t) pressure initiates LPCI.

2(16)

Instrument Channel -

s 2.5 psig A

1.

Above tris settina in DrYvell High Pressure conjunction with low reactor (PS-64 -57A-D) water level, drywell high pressure, 120 rec, delay timer and CSS or PliA pump running, initiates ADS.

Aqf'

'ent No. 14

'G

• sAbt.1 3.2.9 LU2/EIs.LCCE IMikidENTAT3C.4 Mittesta e of type I.dic.atian Operable Instrumes.t and F.nne.

_Hat es mannels Inalrunent 8 I:. t in. nt 2

LI-3-4 6 A Raact.3r WLet Lev el

' Indicator -It,F.5" to (1) (2) (3)

+107.5" LI-3-46 a 2

FI-3-54 Remetor Prus.sure Indicator C-1200 poig (1) (2) (3)

PI-3-61 Recorder 0-00 psia til (2) (3) 2 PR-44-50 Dr pell Pressare Indicatot 0-00 psia PI-54-67 Recorder, Indicator (1) (2) (3) 2 TI-64-52 Dr pell Tesperature 3-400*F Tst-6 4-52 1

TR-64-52 suprzession Chamber Air Recorder 0-400*r til (2) (3)

Tc.mperature 2

TI-64-55 Suppression Chamber Water Indicator, 0-400*F (1) (2) (3) tis-64-55 Teaptrature 2

LI-64-54 A Coppres sion Chamber Water Indicator -25" to (1) (2) (3)

LI 66 Level

+25*

3 1

m/A Control Rod Position 6V Indicatin.;

)

Lights

)

1 N/A Meets on Mnitoring SRM, IR N,

'M

)

(1) (2) (3) (4)

O to 10.

,wer i 1

PS-64-67 Dryw 11 Pressure Alarm at 35 peig )

)

1 TR-64-52 and Drywell Temperature and Alarm if temp.

3 PS-64-58 B and Presaiure and Tfmer

> 281*F and

)

(1) (2) (3) (4)

IS-64-67 pressure >2.5 i

o I

O Psig after 30 minute delay h

1 LI-84-2A C'A Tank "A* Imvel Indicator 0 to 1944 (1) 1 LI-84-13A CAD Tank *8* 14W 1 Indicator 0 to 1004 (1)

U 8m4 m

Amendment No.19

.w--

w--r----

w e a y,

f TABLE 3.6.M SHOCK SUPPRESSORS (S.',11BBERS)

Snubbers Snubbers.in Sign Inaccessit:1e Snutters Radiatiet Area During Snubbers Especially During Normal Accessible During Snabber No.

System

-Elevaticts Shutdown Difficult to Remove Operation.

Norm 1 Operation.

SS3-A (3 35 8)

Recirculation 564 x

SS 3-B ( 115 *)

Recirculation 564 x

SS3-B (154 8)

Recirculation 564 x

SSt-A Recirculation 570 x

SSt-B Recirculation 570 x

SS5-A (26 2 *)

Recirculation 581 x

SS5-A (3 258)

Recirculation 581 x

S55-B (3 58)

Recirculation 581 x

SS5-B (9 88)

Recirculation 581 x

If SS6-A Recirculation 568 x

SS6-8 Recirculation 568 x-I SS7 Recirculation 564 x

SS8 Recirculation 564 x

R-64 RHRSW 582 x

R-62 RHRSW 582 x

DRdmt.4

,\\*O.

3

7 FIGURE 3.6-1 CURVE #1 Minimum temperature for pressure tests such as required by Section XI I

2 3

(200 CURVE #2

i.

Minimum temperature

..Q g l' for mechanical heat up or cooldown

.i.i

• i'!!'*,

following nuclear i

f-

~-.Z _

shutdown

+

1000.

j._.

-t-t-Y '

CURVE #3 s

I Ii Minimum temperature for core operation

- 4' ; i (criticality) i, Includes additional 800 margin req'd by o

10CFR50 Appendix G e

i M

n i,,

1

.e m i

HO m 0; 600 Ym av I.,! -

t/)

m 1L y

n i

m

.o t

400 i,,.

.t__y y.,.

- ;}+-

~.

200

. DCLT.UP.. T. E. MP.9)..-._._ --.

H +-+-. - +.. - -

,.7, s.

.-.7.,.,...,...

..... r...

..__p..,.,_.

C 1

I I

I--

o 10 0 200 300 400 MINIMUM TEMPERATURE ABOVE CHANGE IN TRANSITION TEMPERATURE

(*F)

(1) As measured on closure flanges, adjacent head, and shell material 201

7

_ ~_

a ENCLOSURE 2 1

?

JUST1FICATION FOR PROPOSED

-TECHNICAL SPECIFTCATION REVISIONS 1.

Units 1 and 2, pages 2, 2a, and 3 Unit.3, pages 2 and 2a These proposed technical specification revisions are submitted

'in response to a letter from D. G. Eisenhut to All Power Reactor-10,.1980. These revisions are submitted to

-Licensees dated April make the -Browns Ferry technical specifications comply with the mo' el d

NRC specifications regarding operability of safety-related systems, provided as an enclosure to the April 10, 1980, letter.

2.

. Unit 1, page 16 This-revision corrects an administrative error. The safety limit MCPR was shown as 1.06.

It is being changed to 1.07, which is the correct and accepted value. This revision makes unit 1 consistent with units 2 and 3 specifications for this item.

3.

Units 1 and 2, pages 33, 62, 63, and 78 Unit 3, pages 32, 65, and 81 By letter from T. A. Ippolito to H. G. Parris dated November 9, 1979, the NRC issued Amendment Nos. 53, 49, and 26 to Facility License Nos. DPR-33, -52, and -68 for Browns Ferry units 1, 2, and 3.

The amendments were in response to TVA application dated August 27, 1979

.(TVA HFNP TS 129). The amendments changed the technical spectfication high drywell pressure' trip level setpoint from 2.0 psig to 2.5 psig.

These-changes are needed.to make the technical specifications consistent.

.TVA failed to include these proposed changes in application TS 129.

-4.

Unit 3, page 219, Table 3.6.H This revision is needed to add snubbers which were left off the list in the technical specifications.

5.

Units 1 and 2, page 188

. Unit 3, page 201 (Figure 3.6-1) 1Rie upper portion of the curves provides an additional 20*F shif t from the original curves for protection because of uncertainty of

-radiation. damage. 1The' lower portion of curves 2 and 3 reflect the

~11miting' conditions for protection of the feedwater nozzles from degradation. This lower portion includes the 40*F conservatism for nuclear heatup. These proposed curves are more conserative than those in the.present technical specifications and have been administrative 1y

- imposed.-

.1-4 m

m,.

w-..

-c.

4 s

I

.2

=-

t 6.- 1 Unit 2 page,293 Tliis reviaion corrects an administrative -error made in retyping of : this ; page for ' the ' revisions dated January 31,.1980. This revision changes'the surveillance =cequirement back to 24 months instead'of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as stated.

/

4 i

a t

t I

.s.

1 4

i e

5

+

f 9

a -

J

.. i i

'i l

n 5

4

.,', I e "

r e e

e i

e n w wee-Ja,

ne v a me

,m,~w~,

-, - + -- s-w, ~ ~, -

s-,~+=