ML20096D325

Amend 78 to License DPR-68,changing Tech Specs to Reflect Reload & Plant Mods Made During Fuel Cycle 6 Operation

ML20096D325
Person / Time
Site:	Browns Ferry
Issue date:	08/27/1984
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Tennessee Valley Authority
Shared Package
ML20096D328	List: ML20096D325
References
DPR-68-A-078 NUDOCS 8409060025
Download: ML20096D325 (41)

Text

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[fCtcyIo UNITED STATES q

{j) y[ g NUCLEAR REGULATORY COMMISSION

'5 jfp p WASHING TON. D. C. 20555

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TENNESSEE VALLE AUTHORITY DOCKET NO.40-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 78 License No. DPR-68 1.

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

~

A.

The applications for amendment by Tennessee Valley Authority (the licensee) dated January 23 and June 6,1984, comply 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 Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with tne Comission's regulations; O.

The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.

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-68 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

78, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

8409060025 840827 DR ADOCK p

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This license amendment is effective as of the date of issuance.-

FOR THE NUCLEAR REGULATORY COMMISSION-i-m,,,.

Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to-the Technical Specifications Date.of Issuance: August 27,1984 h

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ATTACHMENT TO LICENSE AMENDMENT NO. 78 FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Revise Appendix A as follows:

1.

Remove the following pages and replace with identically numbered

pages, iii, iv, v, vii, 3, 4, 21, 32, 33, 34, 36, 38, 39, 40, 43, 69, 76, 81, 82, 83, 94, 99, 102, 102a, 129, 149, 242, 264, 264a, 268, 279, 285, 289, 290, 353, 354, 355 2.

The marginal lines on these pages denote.the area being changed.

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Section Pag e No.

B.

Coolant Chemistry 187 C..

Coolant Leakage 191 D.

Relief Valves 192 E.

Jet Pumps 193 F.

Recirculation Pumn Operation 195 G.

Structural Integrity 196 H.

Seismic Restraints, Supports, a.1 Snubbers 198 3.7/4.7 Containment Systems 231 A.

Primary Containment 231 B.

Standby Gas Treatment System 247 C.

Secondary Containment 251 D.

Primary Containment Isolation Valves 254 E.

Control-Room Emergency Ventilation 256 F.

Primary Containment Purge System 258 G.

Containment Atmosphere Dilution System (CAD) 260 H.

Containment Atmosphere Monitoring (CAM)

System H, Analyzer 261

3. 8/ 4. 8 Radioactive Materials 299 A.

Liquid Effluents 299 B.

Airborne Effluents 302 C.

Mechanical vacuum Pump 307 D.

Miscellaneous Radioac ive Materials Sources 308

3. 9/ 4. 9, Auxiliary Electrical System 316 A.

Auxiliary Electrical Eggipment 316 B.

Operation with Inoperable Equipment 321 iii AmendmentNo.[,78

sag 3190 Page_ Eo s

C.

operation in Cold shutdown 326 3.10/4.10-Core Alterations 331 A.

Refueling Interlucks 331 B.

Core Monitoring 336 C.

Spent Fuel Pool Water 337 D.

Reactor Building Crane 338 E.

Spent Fuel Cask 339 F.

Spent Fuel Cask Handling-Refueling Floor 339 3.11/4.11 Fire Protection Systems 347 A.

High Pressure Fire Protection System 347 B.

CO Fire Protection System 351 2

C.

Fire Detectors 352 D.

Roving Fire Watch 353 E.

Fire Prottetion Systems Inspections 354 F.

Fire Protection Organization 354 G.

Air Masks and Cylinders 355 H.

Continuous Fire Watch 355 I.

Open Flames, Welding, and Burning in the Cable Spreading Room 355 5.0 Major Design Features 360 5.1 Site Features 360 5.2 Reactor 360 5.3 Reactor Vessel 360 5.4 containment 360 5.5 Fuel Storage 360 5.6 Seismic Design 361 6.0 Administrative controls 362 6.1 Organization 362 m

iv Amendment No. 78

seetion Page No.

6.2 Review and Audit 6.3 Procedures 368 l

6.4 Actions to be TdP.en in the Event of A Reportable Occurrence in Plant Operation 376 6.5 Action to be Taken in the Event a safety Limit is Exceeded 376 6.6 Station Operating Records i

376

)

6.7 Reporting Requirements 379 6.8 Minimiam Plant Staffing 388 h'

v Amendment No. 78

_m....

g-6.2.E Minimum Test and Calibration Frequency for Dryuell Le k Detection Instrumentation a

101 4.2.F Minimus Test and Calibration Frequency for Su.veillance Ins trumen ta tion 102 4.2.C Surve111ance Requirements for control Room Isolation Ins trumenta tion 103 4.2.M Hinimum Test and Calibration Frequency for Flood Protection Instrumentation 104 4.2.J Seismic Monitoring Instrument Surveillance Requfrements 3.5.-1 105 Minimus RRR$W and EECW Pump Assignment l

3.5.I -

156a MAPLHCR vs. Average Planar Exposure

" Reactor Coolant System inservice Inspecti M1. 182. 1824, 182b 4.6.A on Schedute 3.7.A 209 Primary Containment Isolation Valves 3.7.3 262 Testable Penetrations with Double 0-Ring Seals 3.7.C 268 Testable Penetrations with Testable Bellows 3.7.0 269 Air Tested Isolation valves 3.7.E 270 Primory Containment Isolation Valves vnich Termin t Suppression Pool Water Level a e Eclew the 279 3.J. T Primary Containment Seismic Class 1 LinesIsolation valves lucated in Water Sealed 280 3.7.C Deleted

- 3. 7.H Testable Electrical Penetrations 4.8.A 283 Radioactive Liquid Waste Sampling and Analysis i

4.8.8 310 Radioactive Caseous Waste Sampling and Analysis 4.9.A.4.c voltage 311 Relay Setpoints/ Diesel Cenerator start 3.11.A 126a Fire Protection Systen Hydraulic Requirements 6.8.A 35Sa Minimum Shitt Crew Requir ment:

390 o

vii AmendmentNo.[,g,g,78 n

i I. -Hot Standby Condit on - Hot standby condition means operation with coolant' temperature greater than 212 F, system pressure less than 0

1 1055 psig, the main steam' isolation valve closed and the mode switch in the Startup/ Hot Standby position.

-J.. Cold Condition - Reactor coolant temperature equal to or less than 2120F.

K.

Hot Shutdown -'The reactor is in the shutdown mode and the reactor coolant temperature greater than 2120F.

L.

Cold Shutdown Th'e reactor is in the shutdown modo and the reactor coolant temperature equal to or less than 2120F.

M.

Mode of Operation - A reactor mode switch selecta 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-system is energized with IRM neutron monitorinq system trip, the APRM 15 percent high flux trip and control rod withdrawal interlocks in service. This is often referred to as just Startup Mode. This is intended to imply the Startup/ Hot Standby position of the moda switeb.

2.

Run Mode - In this mode the reactor system pressure is at or above 825 psig and the reactor protection system is energized with APRM protection (excluding the 15 percent high flux trip) nnd RBM interlocks n service.

3 Shutdown Mode - Placing the mode switch to the shutdown position initiates a reactor scram and power to the control red drives is removed. After a short time period (ahout 10 seconds), the scram signal is removed allowing a scram reset and restoring the nornal valve lineup in the control rod drive hydraulic system.

4 Refuel Mode - With the mode switch in the refuel position, interlocks are established so that one control rod only may be withdrawn when the Source Range Monitor indicates at least 3 cps and the refueling crane is not over the reacter except as specified by TS 3.10.B.1.b.2.

If the refueling crane is over the reactor, all rods must be fully inserted and none can be withdrawn.

N.

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

Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system press'ure refer to the values of these parameters when the reactor is at rated power.'. Design power, the power to which the safety analysis applies, corresponds to 3,440 MWt.

3 Amendment No.

, 78

O.

Primary Containment Integrity - Primary containment integrity means-that the drywell and pressure suppression chamber are intact and all of the following conditions are satisifed:

1.

All non-automatic containment isolation valves on lines connected to the reactor coolant system or contairment which are not required to be open during accident conditions are closed. These valves may be opened to perform necessary operational activities.

2.

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

3 All automatic containment isolation valves are operable or deactivated in the isolated position.

4.

All blind flanges and manways are closed.

1 4

l Amendment No.,56', 78

5-decade, instrument which covers the ; range of. power level between that

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covered byithe'SRM and the APRM.

The 5 decades are covered by.the IRM by means of 'a-range. switch,;and. the 5 decades are broken down into 10 ranges,=each beingfone-half of a_ decade in size.

of;120 divisions is active.in each range of the IRM.The IRM scram setting

~

For example, if'

' the instrument was ~ on. range.1, the scram. setting would be 120 divisions-for t. that range; :likewise, 'if, the instrument was on range 5, the scram.

' setting would,be'120 divisions on~that range.

Thus, as the IRM is

. ranged'up,to accommodate the increase in power level, the' scram setting

~1s also: ranged up.

A scram at 120. divisions on the IRM instruments remains in ef fect as'long as the reactor is in the startup mode.

The

APRM 15-percent scram will prevent higher power operation without being in.the run mode. ' The IRM scram provides protection for changes which

. occur both locally and over the entire core.

The most significant sources of reactivity change during _ the power increase are duc ~ to control rod withdrawal. ' For insequence control rod withdrawal, the rate of. change.of, power is slow enough, due to the physical limitation of withdrawing control rods, that heat flux _is in equilibrium with the

. neutron flux and an IRM scram would result in a reactor shutdown well before any safety limit is exceeded. _For the case of a single control ~

rod withdrawal error, a range of rod withdrawal accidents was analyzed.

This analysis included starting the accident at various power levels.

The most severe case involves an initial condition in which the reactor.

is just subcritical and the 'IRM s:' stem. is not yet on scale.

This

.cendition exists at quarter rod density. Quarter rod density is illustrated

.in paragraph 7.5.5.4 of the FSAR.

Additional conservatism was taken in this analysis by. assuming that the IRM channel closest to the withdrawn rod Is bypassed. The results of this analysis show that the reactor is scrammed'and peak power limited to one percent of' rated power, thus maintaining MCPR above 1.07.

Based on the above analysis, the IRM provides protection against local-control rod withdrawal errors and -

continuous withdrawal of control rods in sequence.

4. - Fixed High Neutron Flux Scram Trip The average power range monitoring (APRM) system,.which'is calibrated using heat balance data caten during steady-state conditions, reads in percent of rated power (3.93 MWt).

The APRM system responds directly to neutron flux.

Licensing nalyses have demonstrated that with a neutron flux scram of 120% of rated power, none of the abnormal 6perational transients analyzed violate the fuel safety limit and there is a sub-scantial margin from fuel damage.

B.

APRM Control Rod Block Reactor power level may be varied by moving control-rods or by varying the recirculation flow rate.

The APRM system provides a control rod

. block to prevent rod withdrawal beyond 21 Amendment No.)d, 78 J

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TABLE 3.1.A REACTOR PROTECTION SYSTEH (SCRAM) INSTRUMENTATION REQUIREMENT Hinimua Number of Operable Modes in Which Function Instrument Must be Ooerable Channels Per Startup/Hst' Trip Systes (1)(21)

Trip Funetton Trip Level Setting Shutdown Refuel (7)

Stsedby Run Action (1) 1 Hode Switch in Shutdown X

X X

X 1.1 1

Hanual Scram X

X X

X.

1.A IRM (16) 3 H(gh Flux

$ 120/125 indicated on X(22)

X(22)

I (5) 1.A scale 3

Inoperative X

X (5) 1.A' APRH (16)(24)(25) 2 High Flux (Fixes Trip) $ 120 percent X

1.4 or 1.3 w"

2 Hagh Flux (Flow Blased) See Spec. 2.1.A.1 X

1.A or 1.3 2

High Flux

$ 15 percent rated power X(21)

X(17)

(15) 1.A or 1.B 2

Inoperative (13)

X(21)

X(17)

X 1.A or 1.8 2

Dcwnscale 1 3 indicated on scale (11)

(11)

X(12)

1. A or 1.B 2

High Reactor Pressure

$ 1055 psig X(10)

X X

1.A p

2 High Drywell 1 2.5 psig, X(8)

X(8)

X 1.A

,q

?ressure (14) 5 h

2 Reactor Low Water 1 538 inch above vessel zero X

X X

1.A

-(

Level (14) g 2

High Water Level in Vest Scram Discharge Tank 1 50 gallons X

X(2)

X X

l.A (LS-SS-45A-0)

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TABLE 3.1.A (cont'd)

REACTOR PROTECTION SYSTEN (SCatM) INSTRUMEdTATION REQUIREMENT 11nimus Number or Operable 1

Instru=ent Modes in Which Function hannels For Most be Operable frip Systes (1) (23) Trio Function Startup/ Hot Trio Levet settine shutdown peruel (7)

Standby Run

• Aetton(1) 2 High Water Level in j[ $Q gallons X

I(2)

I I

1.A East Scraa Discharge Tank (LS-85-45E-N) 4 Main Stena Line f( 10 percent valve closure Isolation Valve X(6)

-1.A or 1.C.

Closure I

2 Turbine Control Talve Fast closure or

2. 550 psig Turbine Trip 2(4) 1.A or 1.D us 4

Turbine Stop Valve j[ 10% Valve Closure

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Closure X(4) 1.A or I.D.

2 Turbine First Stage not 1,154 psig Pres'sure Permissive X(18)

X(18)

X(18) (19) 2

• adP Turbine Condenser 3,23 In. Hg. Vacuum I,ow vacuum X

l.A or 1.C i

2 Main Steam Line High 3X Nornal Full Power X(9)

X(9)

X(9)

I.A or I.C Radiat ion (14)

Background (20) 3 CL

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A.

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's.e

i NOTES FOR TABLE 3 1 A 1here shall be two opeable or tripped trip systems for each 1.

function.

If the minimum number of operable instrument channels per trip system cannot be met for both trip systems, the appropriate actions listed below shall be taken.

A.

' Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

In refueling mode, suspend all operations involving core alterations and fully insert all operable control rods within one hour.

B.

Reduce power level to IRM range and place mode ewitch in i

the startup/ Hot standby position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C.

Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

D.

Reduce power to less than 30% of rated.

2.

scram discharge volume high bypass may be used in shutdown or refuel to bypass scram discharge volume scram with control rod block for reactor protection system reset.

3.

Deleted.

4 4

Bypassed when turbine first stage pressure is less than 154 psig.

i 5.

IRM's are bypassed when APRM's are onscale and the reactor mode switch is in the run position.

a The design permits closure of any two lines without a scram u.

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being initiated.

7.

When the reactor is suberitical and the readtor water i

temperature is less than 212*F, only the following trip functions need to be operable 4

A.

Mode switch in shutdown B.

Manual scram i

C.

High flust IRM D.

Scram discharge volume high level E.

APRM 155-scram 8.

Not required to be operable when primary containstant integrity is not required.

9.

Not requi' red if all main steamlines are isolated. :

10. Not required to be operable when the reactor pressure vessel head is not bolted to ths vessel.

11. The APRM downscale trip function is only active when the reactor mode switch is in run'.

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k TABLE 4.1.A ADCToa FFCTICT8038 S T 5itas 15C AA rel IsdsrauMrstTATI0ta FU 4CT BoraAL T ESTS selsspeu1 rusactICisAt 1Z5r F At;uceCIES toa SAFETY 1985 T m. Af;.*t Co!8FAOL C!?-CUjT5 s

Croup (2)

Functional Test Mangnum troquency (3) seode switch in Ihutdown A

Place MJde $ witch in Shutdown EaCh SeIueI!D9 outage Manual screa A

Trip Channel and Alara gwery 3 peonthe Isse sigh rius C

Trip Channet and Alara ge)

Once Per n8eek taa ri ng 5e f u el a n*J

\\

and sefore tach startup Inoperat&we C

Trip Cnannel and Atara (el once Fer Week thasing aefueling g

and sefore tach startup Ata n' migh rius (150 acrang c

Trip sput melare gel aefore rach startup and weekly liigh Flux (Flow Riased)

B Trip Output Relays (4)

Once/ weck When pequi red to t,e Cseerable sigh rius (plxed Trip) a Trip astput pelays tel once/ Week u.

Inoperative a

Trip output pelare (*l cncrNeek e

Downscale s

Trip output melays (e3 once/ week Flow Blas 8

16)

(6) s{gh peactor Pressure A,

Trip Channel and Alara once/Manth (Il sigh Drywell Pressur*

A Trip Channet and Alara once/nanth g li meactor 14w watet Level (5)

A Trip criannel and Alara once/ Month (y

, migh water g4ve) in scram otscharge Tank l

g Float. Swi tches (I.5-85-45C-l')

A Trip Ch.innel and Alarn Once/'tonth

<o Elect ronic Level Swi tches B

irip (:liannel anal AI.stm (7)

Once /:'on t ti E

(LS-85-45A, B, C, 11) a en 3

'+

Turbine Condenser Low Vacuum A

Trip Channel and Alarm Once/ Month (1) i

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1. 7 IcitUlly the minimum frecuency for the indicated tests shall s

be once per month.'

2.

A descri0 tion of the, three groups ie included in the Bases of this specification.'i 3.

Functional ~ tests are not required when the systems are not required to be operable or are operating (i. e., already tripped).

If tests are, missed, they shall be performed prior to returning the systems to an operable status.

4.

This instrumentatibn is exempted from the instrument channel test definition.

This instrument channel functional test will consist of injecting'a simulated electrical signal into the measurement channels.

5.

The water level in the reactor v'essel will be perturbed and the corresponding level indicator changes will be monitored.

This perturbation test will be performed every month after completion of the monthlf functional test program.

6.

The functional test of the flow bias network is performed in accordance with Table 4.2.C.

7.

Functional test consis:s of the injec:1on of a si=ulated signal into the electronic 'crip circui:ry in place of the sensor signal to veri:7

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operabif.1ty of the trip end alarm functions.

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4, 38 Amendment No. 78 s

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TABLE 4.1.8 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Minimum Frequency (2).

Instrument Channel Group (1)

Calibration Note (86)

C Comparison to APRM on Control-IRM High Flux led startups (6)

Once every 7 days APRM High Flux B

Heat Balance output Signal Calibrate Flow Bias Signal (7)

Once/ operating cycle B

Flow Bias Signal B

TIP System Traverse (8)

Every 1000 Effective LPRM Signal Full Power Hours A

Standard Pressure Source Every 3 Months High Reactor Pressure A

standard Pressure Source Every 3 Months High Drywr.ll Pressure A

Pressure Standard Every 3 Months Reactor Iow Water Level High Water Level in Scras Disch'arge y

to Volume Float Switches A

Calibrated Water Column (5)

Note (5)

(LS-85 8*5C-F)

Electronic Level Switches B

Calibrated Vater Column Ducs/ Operating Cycle (9)

(LS-85 t:51. B. C. H)

Turbine Condenser Low Vacuum A

Standard Vacuum Source Every 3 Nonths it 69 Main Steam Line Isolation valve closure A

Note (5)

Note (5) m Qg Main Steam Line High Radiation B

Standard Current Source (3)

Every 3 Months a

c+

Turbine.First Stage Pressure Permissive A

Standard Pressure Source Every 6 Months g

lurbine Cont. Vilve Fast Closure or A

Standard Pressure Source Once/opegating cycle Turbine Trip surbine Stop valve Closure A

Note (5)

Note (5)

W

NOTES FOR TABLE 4.1.D 1.

A description of three groups is included in the bases of this specification.

2.

Calibrations are not required when.the systems are not required to be operable or are tripped.

If calibrations are missed, they shall be performed prior to returning the system to an operable status.

3.

The current source proNides an instrument channel alignment.

~

Calibration using a radiation source shall be made each refueling outage.

4 Required frequency is initial startup following each refueling outage.

5.

Physical inspection and actuation of these position switches will be performed once per operating cycle.

6.

On controlled startups overlap between the IRM's and APRM's will be verified.

7.

The Flow Bias Signal Cal ~ibration will consist of calibrating the sensors, flow converters, and signal offset networks during each operating cycle.

The instrumentation is an analog type with redundant flow signals that can be compared.

The flow comparator trip and upscale will be functionally tested according to Table 4.2.C to ensure the proper operating during the operating cycle.

Refer to 4.1 Bases for

• f urther.txplanation of calibration frequency.

8.

A complete tip system ;raverse calibrates the L*RM signals to the process computer. The individual LPRM meter readines will be adjusted as a ninimum at the beginning of each operating cycle before reaching 1007. power.

O Calibration consists of the adjustment of the primary sensor and associated components 'so th'at they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.

~

i 40 l

d AmendmentNo,f,p,78

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which a scram would be required but not be able to perform its function adequately.

A sourco range maitor. (SRN) system is also provided to supply additional neutron level information during startup but has no scran functions.

Ref. Section 7.5.4 FSAR.

Thus, the IRM is required in the Refuel and startup modes.

In the power range the APRM system provides required protection.

Ref. Section 7.5.7 FSM4.

M us, the IRM System is not required in the Run mode.

The APRPc and the IRM's provide adequate coverage in the startup and interwediate range.

The high reactor pressure, high drywell pressure, reactor low water level and scram discharge volume high level scrams are required for Startup and Run modes of plant operation.

Wey are, therefore, required to be operational for these modes of reactor operation.

The requirement to have the scram functions as indicated in Table 3.1.1 operable in the Refuel mode is to assure that shifting to the Refuel mode during reactor PC,ser operation does not diminish the need for the reactor protection system.

The turbine condenser low vacutzt scram is only required in the,run mode.

Below 154 peig turbine first stage pressure (305 of rated), the scram signal due to turbine stop valve closure, and turbine control valve rase closure, or turbine crip is bypassed because flux and pressure scram are adequate to protect the reactor.

.Because of the APRM downscale limit of it 35-when in the Rui mode and high level limit of 5 155 when in the Startup Mode, the transition between the Startup and Run Modes must be made with the APRM inst 1mtation indicating between 35 and 155 of rated power or a control rod scram will occur.

In addition, the IRN system must be indicating below the High Fir,uc setting (120/125 of scale) or a scram will occur when in the Startup Mode.

For normal' operating conditions, these limits provide assurance of overlap between the IRM system and APRM system so that there are no " gape in the power level indications (i.e., the power level a

is continuously monitored from beginning of startup to full power and from full power to shutdown).

Sehen power is being reduced, if a transfer to t'1e Startup mode is made and the IRM's have not been fully inserted (a maloperational but not impossible condition) a control rod block immediately occurs so that reactivity insertion by control rod withdrawal cannot occur.

43

, f, 78 Amendment No.

4 a

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Table 3. 2.3 I.-

2CTREMDsTATJoM THAT IutTIATes on COWrROLS 782 CORE ANo COVrAINNENT COOLING SYSTE r.inimua No.

operaMe Per II11.HI? J R hneti23 Trio Level settine Action Rebarks 1

Core Spray Trip N/A i

tysten bus C

1 Manitors availability of power psver w. iter to logic systems.

1 1.DS *Irip System bus N/A pcwer mer.itor C

1.

Monitors availability of power to logic systems and valves.

I NPCI trip systes, bus WA power monitor C

1.

Monitors availability of power to logic systems.

1 BCIC Trip system bus power monitor N/A C

1.

Monitors availability of power to logic systems.

1(2)

Instrument, Channel -

2 Elev. 5518 condensate Header A

1.

selow trip setting will open i

14 vel (LS-73;56A & 33 MPCI suction valves to the suppreselon chamber.

2 (2)

Inscrument Channel -

57* above normal water Suppressico chamber sigh level A

1 Above trip settimig erill open 14,el 3

EPCI suction valves to the supprescion chamber.

{

2 (2)

Instrument channel -

5 593* above vessel seco meactor sigh Water Level a

1.

Above trip setting trips BCIC turbine.

1 Instrument Channel -

1 tS48 54(1)

DCIC Turbine Steam Line A

1.

Above trip setting inolates Nigh Flow AC'C system and trips TCIC tL;bine.

4 (t)

Inst;rument channel -

5200* r.

ACIO Steam Line Space A

1.

Above trip setting' isolates Bigh Tesperature BCIC system and trips RCIC tur bine.

F

<o o

C1.

a 3

i (D

'3 2:

.O Co I

i

)

i

[ -

~

.- y TabtE 3.2.C l

IN5thUtitNI ATION lHAT INiiTAlt3140D hlACKS I

Hinleue Operable Channels Per Trip Funetton (5 runction Trip 1.evel Setterg

')

~.

4(1)

APiet Upsw - (Flai plas) 10.664 e' 42? (2) 4(1) april Upscale (Startup Honte) (8) 1925 4(l) april Deanscale (9) 2 31 4(1)

APRtlInopeIntive (10h) j 2(7)

Risi Upscale (Flere Blas) 1 0.669 401 (2)(11) 2(7)

RHP: Dounscale (9) 234 2(7)

RIIH Inoperative (10c) 6(l)

INH Upscale (8) 1 108/125 or rull scale 6(l)

Inti Imnscale (3) (8) 15/125 or ruil scale 6(1)

Isti netector not in startup Position (a)

(ll),

6(I)

IR'l Inope-ative (8)

(loa) cm S

3(1) (6) salt lipscale (2) glNIO counts /sec.

3(1) (6) skrs furenscale (4) (8)

A 3 counts /seo.

3(4) (6) stdl Detector not in Startup Posillon (4)(8)

(11) 3(1) (6)

.,H?s Inoperative (R)

(10 1 2(l) rim stag owyv ator s lof ilfrc~;nce to rectreulation rina, F

1 o

2(1)

Flou plean tipscale 19 4 5% rec treutetin, ricre

_s 1

Pat RiocP Lne,1c.

11/ 4 tD 3

2 ( 13 ascs asetseint 147 peig terbias (ps-85-6 th and first-stage pressere po-es-slat 1(12) liigh Water Level in West d25 gal.

Scram Discharge Tank

'S' (LS-85-45L) 1(12) liigh Water Level in East

$25 gal.

5 Scram Discharge Tank (1.S-85-4 5M) s

c;

+-

.~-

- ~ ~-

, ? *-[$.M, t,

s s

1

=

TABLE., 3.2.r 50fNEILIMCE !!GTFffME.VTATIOli 9

Minimue $ of Type Indication Ope r able In st s omer.t Cria r.rsri s Instrument e Instrument and Range Not es 2

LI-3-46 A Reactor Water Level Indicator' - 8 55" so.

. (1) ( 2) (3)

• 60" LI-3-46 5 2

PI-3-54 Peactor Pressure Isid a cat os o-1500 psig (1) (2) (3)

PI-3 -61 2

PR-6 4-50 Drywell Pressure Recorder 0-80 psia (1) (2) (3)

Indicator 0-80 psia PI-6e-67 2

TI-64-52 E,rywell Temperature Recorder, Indicator (1) (2) (3) 0-e00*r TR-64-52 1

1 TR-64-52 Suppression chamber Air Recorder 0-400*r (1) (2) (3)

Temperature 3

1 N/A Control Rod Position 6V Indicating

)

Lights i

1 N/A Neutron Monitoring SRM, IR N, LPRM.

)

(1) (2) (3).. g 4 )

O to 1001 power )

1 PS-64-67 Drywell Fressure Alarm at 35 psig )'

N TR-64-52 and

'Drywell Temperature and Alarm if temp.

)

1 Ps-64-58 B and Pressure and Timer

> 281*r and

).

( ll (2) (3) (4) u:s Q

IS-6e-67 pressur e > 2.5 psig i after 30 minute

)

e delay

)

3 c+

2 1

LI-84-2A CAD Tank " A" Level

- Indicator 0 to 1004 (1) g 1

.LI-84-13A CAD Tank *B" Level Isadicator 0 to 1006 (1) y

~

CD f*4

9

\\'

TABLE 3.2.F

'A Surveillance Instruner.tation Minimum # of 4

Operable Instru:nent Channels Instru:sent #

fnstrumen_t Type Indication 2

_ and Renee H H S k Notes _

2 Drywell and Torus 0.1 - 20%,

(1)

M H lob 2

Hydrogen Concentration 2

PdI-6b-137 Drywell to Indicator o to 2 psid (1) (2)-(3)

PdI-6%-138 Suppression Chamber Differential Pressure 1/ Valve Rettet Yalve Ts11 pipe (5)

Thermocouple Temperature or Acoustic Mcnitor on Relief Vedie Ta11 pipe a

o e

3 Q

to

<+

l

.i

.. =

2 2

LI-6%-139A O

Suppression ' *

• Indicator.

(1) (2) (3)

IR-6%-139 thaaber Water Recorder 0-2%0

Level-Vide Range k

2 PI 16M rs-6h-159 cryvc11 Trzerure Indicator. Recothr) (1) (q) '(1)

%J Vide Esase 0-300 p585 I

71-56-161

~

Tn.6L-161 Suppresstor. Puol Indicator. Accert.er) (1) (2) (3) H)M)

Du14 71-64-162 i

)

Tit-68.-162 Temperature 30* - 230* r I

)

v

, /~

-/

e NOP S TUR "" Ult t 3. 2. 7_

(1) Trom and af' ter the date that one of these parameters i

• is sooner made opera 2:le.during the succeeding s

entation (2) indicated in the centrol room,Trem and af ter the date 'that e s not continued operatica is instrumentatien is sooner made operable. permissible du (3)

If the requirasants of notes (1) and (2) cannot be eet of the indications cannot be restored in (6) hours, an, orderly and if one shutdown shAll be initiated and the reactor shall be in a c condicion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(el nose surveillance instruments are considered to to each other.

(3) From and af ter the date that both the acoustic mon control room, continued operation is permissi or and 'the e

succeeding thirty days, unless one of the two monitoring is soonce cade available.

channals on any SRV tailpipe la inoperable,If both the primary and secondary indica incressa which night be indicative of an open SR tored at mperature i

(6) A channel consists of 8 censors bay.

Seven sensors must be operable for the chan, one from each operable.

nel to be i

i

-e O

e 4

Amendment No, e h 78 4

TABLE 4.2.8 SURVE!!n' ICE REQUIRE.4INTS FCR INSTRUMCfTATION THAT INITIATE OR CONTROL THE CSCS Punction Functional Test Calibration inst rumen t check I

ADS Timer (4) once/ operating cycle none 3

~

Instru.nent channel (1) once/3 months none RSR Po p Discharge Pressure Instrument channel (1) once/3 months none Core Spray Pump Discharge Pre ssur e 1

Core Spray Sparger to RPV d/p (1) once/3 months once/ day j

Trip system aus Pcwer Monitor once/ operating cycle N/A none Instrument channel Condensate Header 14 vel (1) once/3 months none g

(IS-73-56A,B) 3 CL 3

64 3

ct 2O l

t j

N CD 4

TABLE 4.2.C SURVLII. LANCE REQOlaB6ENTS fos INSTSD4ENTATION THAT INITIATE 300 314CAS runction fitnctional Test Calibration (17)

Instrument Check APRM Opecale (Flow tiae)

(1)

(13) once/3 months once/ day (8)

APRA Upacale (Startup Mode)

(1)

(13) once/3 acoths once/ day (4)

Apan now. scale (t)

(s3) once/3 months once/ day (e)

Apam Inoperative (1)

(13)

WA once/ day (e) man Dpecale (Flow Blael (1)

(13) once/6 months sem Downecale once/ day (4)

(1)

(13) once/6 months once/ day (8) san Inoperative (s)

(13)

N/A once/ day (8)

Ian opacale (1) (2)

(13) once '3 months once/ day (a)

Jam Downscale (1) (2)

(13) once/3 months e

once/ day (8)

Jax Detector not in Startop (2) (once/operattag once/ operating cycle (12)

N/A Position cycle)

Jam 2aoperative (1) (2)

(13) avA N/A sam Upecale (1) (2)

(13) once/3 months once/ day (8) sam Downscale (1) (2)

(13) once/3 months once/ day (8) san Detector not in startup (2) (once/cperating once/ operating cycle (12)

N/A Poettion cycle) sam Inoperative (1) (2)

(13)

WA N/A 4

riov Blas Co'sparator (1) (15) once/ operating cycle 320)

VA j

riow Blas Upscale (1) (15) once/3 acothe M/A mod alock Logic (16)

N m/x w/A (D

BSCS Restraint 3

(1) once/3 acatha CL WA l

West Scram Discharge Tank Water Level liigh once/ quarter once/ operating cycle N/A 2

(LS-85-45L)

.o East Scram Discharge t

Tank Water Level liigh once/ quarter once/cperating cycle N/A (LS-85-45M) y

\\

\\

(

TABLE 4.2.F HINIMUM TEST AND CALIBRATION FREQUEllCY FOR SURVEIhLANCE IllSTRUMEtJTATION Instrument Channel Calibration Frequency Instrument Check

1) Reactor Water Level Once./6 months Fach Shift

2) Reactor Pressure Once/6 months Each Shift

3) Drywell Pressure Once/6 months Each Shift

4) Drywell Temperature Once/6 months Each Shift-

5) Suppression Chamber Air Temperature Once/6. months Each Shift l

5

8) Control Rod Position HA Each Shift

9) Neut;ron Monitoring (2)

Each Shift

10) Drywell Pressure (PS-64-67) once/6 months NA

11) Drywell Pressure (PS-64-58D)

Once/6 months 11A

12) Drywell Temperature (TR-64-52)

Once/6 months NA E'

13) Timer (IS-64-67)

Once/6 months NA

14) CAD Tank Level Once/6 months Once/ day co 5

15) Containment Atmosphere Monitors Once/6 months Once/ day 4

,16) Dryvell to Suppression Chamber Once /6 months Each Shift Differential Pressure u

__-r-i C.

~

- 'j s

}IlllIHUH TEST All0 CALIBRATIONTABLE 4.2.T PREOUCIICY ron sunVEILLANCE INSTRUllENTATION i

_ Instrument Channel Calibration Frequency

_ Instrument Check 37 RelJef valve Tallpipe

-l in Tliermoenuple Tem.erature s

IR Acouscle Honitor on Relief Valve Ta!! pipe Once/ cycle (25) i Once/nonth (26)

Bf

}

5 r 19. Suppression Chamber Water Level-Wide Range Once/ cycle Once/ month (LI-64-159A) (XR-64-159) a

20. Dryvell Pressure - Wide Range Once/ cycle (PI 160A)(XR-64-159),

Once/ shift

21. Suppression Pool Dulk Temperature Once/ cycle (T1-64-161) (TR-64-161)

Once/ shift (TI-64-162), (TR-64-162)

O.

e

{

2 P

m l

(

f

-j LINETING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT j

J.J REACTIVITY CONTROL 4.3 REACTIVITY CONTROL D.

Re ac tivi ty Anomalies D.

Reactivity Anomalies the reactivity equivalent During the startup test of the difference between program and startup the actual critical rod following refueling confiquiation and the outages,- the critical rod expected configuration configurations will be during power operation compared to the expected shall not exceed 1%ak.

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

the reactor will be shut These comparisons will be down until the cause has used as base data for been determined and reactivity monitoring corrective actions have during subsequent power been taken as appropriate.

op er ation ' t hroughout the fuel cycle.

At specific power operating E.

If specifications 3.3.C and.D conditions, the critical above cannot be met, an orderly r d configuration will be co(([gua shutdown shall be initiated and on expected the reactor shall be in the based upon appropriately shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

corrected past data. This comparison will be made at F.

Scram Discharge Volume (SDV) least every f u) \\ power month.

E.

Surveillance requirements are 1.

The scram discharge volume drain and vent valves shall as specified in 4.3.C and.D be operable any time that the above.

reactor protection system is F.

Scram Discharge Volume (SDV) required to be operable except as specified in 3.3.F.2.

,1.a.

The scram discharge volume 2.

In the event any SDV drain or drain and vent valves shall vent valve becomes inoperable, be verified open prior to reactor operation may continue each starup and monthly thereafter. The valves may provided the redundant drain or vent valve is operable.

be closed intermittently for testing not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 3.

If redundant drain or vent in any 24-hour period durin6 valves become inoperable, the operation.

reactor shall be in hot stand-1.b.

The scram discharge volume by within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, drain and vent valves shall be demonstrated operable monthly.

2.

When it is determined that any SDV drain or vent valve is inoperable, the redundant dr'ir a

or vent valve shall be demon-strated operable immediately and weekly thereafter.

129 3.

No additional surveillance required.

Amendment No. 78

l f.tMITING CONDITIONS POR OPEPATION CURVCILLANCE ltt@TREM1;MTS I

1 3.5 S9BE_.6U9 99hTSIMMI!C 4.5 CORE AND CONTAINMENT COOLIPC SQQjellis_EXM-L4E SYSTEMS B.

'tasidual Heal Demoval B.

Residual Heat Removal Gys _ tem (EHRS) (LPCI and System (RHRS) (LPCI and Containment Cooling)

Containlient Cooling) 1 The BilRS shall be 1.

a. Simulated Once/

o pe rable:

Automatic Operating Actuation Cycle (1) prior to a Test reactor startup from a Cold

b. Pump Opera-Once/

Condition; or bility month (2) when there is

c. Motor Opera-Once/

irradiated fuel ted valve month in the reactor operability vessel and when the reactor vessel pressure

d. Pump Flow Once/3 is greater than Rate Months atmospheric,

e. Testable Once/

except as specified in check valve operatain specifications cycle 3.5.B.2, through 3.5.B.7

-Esch LPCI pump shall deliver 9,000 gpm against an indicated system pressure of 125 psig.

'No 2.

With the reactor LPCI pumps in the same loop shall vensel pressure less deliver 4 000 gpm against an than 105 psig, the indicated system pressure of HHH may be removed 250 psig.

from service (except that two RHR pumps.

2.

containment cooling Anairtestonthedhellandtorus mode and associated headers and nozzles shall be conducted once/5 years. A heat exchangers must water test may be perfor=ed on remain operable) for the torus header in lieu of the a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while air test.

being drained of 149 AmendmentNo.[,78

LIMITING CONDITIONS FOR OPERATION SURVEIT?TNCE REQUIREMENTS j

I7 GDl!IAllHfElfI IIIIEl$h 4.7 COMAIMMENT SYSTEMS v

\\

system may be taken out of service for maintenance but shall be returned to service as soon as practicable.

k.

The interior surfaces of the drywell and torus above the level one foot below the normal water line and outside surfaces of the torus below the water line shall be visually inspected each.

operating _^ cycle for deterioration and any signs of structural damage with particular attention to piping connections and supports and for signs of distress or displacement.

i f

i 242 l

Amendment No. 78 I

l 8

t 7

,c.

i TABLE 3.7.A (Continued)

Number of Power Maximum Action on Oper.ated Valves -

Operating Normal Initiating Croup

_ Valve Identification Inboard Outboard Time (Sec.)

Position Signal 6

Suppression Chamber purge inlet (FCV-64-19) 1 2.5 C

SC 6

Drywell/ Suppression Chamber nitro-1 5

C SC gen purge Jalet (FCV-76-17) 6 Drywell Exhaust Valve Bypass to Standby Cas Treatment System (FCV-64-31) 1 5

0 GC 6

Suppression Chamber Exhaust Valve Bypass to Standby Cas Treatment System (FCV-64-34) 1 5

0 CC 6

System Suction Isolation Valves to Air Compressors "A" and "B" (FCV-32-62, 63) 2 15 0

CC 6

Drywell/ Suppression Chamber Nitrogen Purge Inlet (FCV-76-24) 1 5

C SC 6

Torus Hydrogen Sample Line Valves Analyzer A (FSV-76-55, 56) 2 NA Note 1 SC 6

Torus oxygen Sample Line Valves

]

Analyzer A FSV-76-53, 54) 2 NA Note 1 SC' 6

Drywell Hydrogen Sample Line Valves Analyzer A (FSV-76-49. 50) 2 NA Note 1 SC i

E 6

Drywell Oxygen Sample Line Valves Analyzer A (FSV-76-51, 52) 2 NA Note 1 SC 5

j 6

Sample Return Valves - Analyzer A 2

j (FSV-76-57, 58) 2 NA 0

GC 6

Torus Hydrogen Sample Line Valves Analyzer B (FSV-76-65, 66) 2 NA Note 1 SC i

E

TABLE 3.7.A (Continued)

Number of Power thximum Action on Operated Valves Operating Normal Initiating-Croup Valve Identification Ir. boa rd

_ Outboard Time (Sec.) Position Signal 6

Torus Oxygen Sample Line Valves-Analyzer B (FSV-76;-63, 64) 2 NA Note 1 SC 6

Drywell Hydrogen Sample Line Valves-Analyzer B (FSV-76-59, 60) 2 NA Note 1 SC 6

Drywell Oxygen Sample Line Valves-Analyzer B (FSV-76-61, 62) 2 NA

. Note 1 SC 6

Sample Return Valves-y Analyzer B (FSV-76-67, 68) 2 NA 0

CC l

7 RCIC Steraline Drain (FSV j GA, 63) 2 5

C SC

]

7 RCIC Condensate Pump Drain (FCV-71-7A, 78) 2 5

C Se 7

HPCI Hotwell pump discharge isola-tion valves (FCV-73-17A, 17B) 2 5

C SC 7

HPCI steamline drain (FCV-73-6A, 6B) 2 5

0 GC 8

TIP Cuide Tubes (5) 1 Per g

guide tube NA C

CC i

[

t CL 5

NOTE: 1: Analyzers are such that one is sampling dryvell hydrogen and oxygen (valves from drywell open -

5 valves from torus closed) while the other is sampling torus hydrogen and oxygen (valves from torus 4

open - valves from drywell closed)

=

'I

=

b

(

)

t X

(,.

TABLE 3.7.B TESTABLE PENETRATIONS WITH DOUBLE 0-RING SEALS Penetratien No.

_Identi fica tion X-1A Equipment Hatch X-13 Equipment Hatch X-4 Head Access, Drywell X-6 CRD Removal Hatch X-25 Flange on 64-18 X-25 Flange on 64-19 X-25 Flange on 84-8A X-25 Flange on 84-8D X-26 Flange on 64-31 X-26 Flange on 64-34 X-35a TIP Drive X-353 TIP Drive X-35c TIP Drive X-35d TIP Drive X-35e TIP Drive.1*

X-35r TIP Indexer Purge l X-47 X-35g Spare Power Operation. Test X-200A Suppression Chamber Access Hatch

, X-200B Suppression Chamber Access Hatch Drywell Head Shear Lug No.1 Shear Lug No. 2.

Shear Lug' No. 3 Shear Lug No. 4 Shear Lug No. 5 Shear Lug No. 6 Shear Lug No. 7 Shear Lug No. 8 X-205 Flange on 64-20 X-205 nange on 64-21 X-205 Flange on'84-8B X-205 Flange on 84-8C X-205 Flange on> 76219.

X-205 Flange on 76-18' X-219A Space (Unit 3 only)

X-223 Suppression Chamber Access Hatch X-231 Flange on 64-29 X-231 Flange on 64-32

~

e 8

0 268

{; -

lkmendmentNo

, 78

• 9 g-i r tj

• t f 1,

• * ~ *

• i I

h,ff $.i6 k ' '

TABLE 3.7.E 13tII*1RY CDi.*rAI;;"E!?? ISOLATIO:f VALVES Mi!IC1! TEKm; ATE DELCV TIIE SUFFRESSIO!! POOL UATER LEVEL Valve Valve Identificotto,12-733 Auxillery Boiler to RCIC 12-?lal Auxiliary Boiler to RCIC I3-28A RlR Suppression Charsber Staple Lines t

1+3-23D RIE Suppression Chaiaer Saaiple Lines 163-29A iUIR Ouppression Cho:nber Sample Lines 163-293 RER Suppression Chusber Sample Lines 71-14 RC'~

Turbine Exhaust 71-32 Re. Vacuum Pu.mp Discharge 71-530 RCic Turbine Exhaust

~1-592 nCIC Vacuu:n Pur:p Discharge 73-23 IIICI Turbine Exhcust "F3-2k

!!ICI Turbine Exhaust. Drain 73-o03

!ICI Turbine Exhaust 73- @

!!ICI Exhaust Drain

?!4 22 RHR 75-57 Core Spray to Auxiliary Boiler 75-58 Core Sprar to Auxiliary Boiler Core Spray to Auxilisry Boiler e

9

• c.

m 279 Amendment No" [ 78 n

,..--n.

y.-

r n.,

~

BASES,

3.7.A & 4.7.A Primary Containment The integrity of the primary containment and operation of the core standby cooling system in combination, ensure that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

During initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required, there will be no pressure on the system thus greatly reducing the changes of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect to minimize the probability of an accident occuring.

The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 49.6 psig, P,.

As an added conservatism, the measured overall integrated leakage rate is further limited to 0.75 La during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50 (type A, B, and C tests).

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure i

suppression chamber water volume must absorb the associated decay and structural sensible heat release during primary system blowdown from 1,035 psig. Since all of the gases in the drywell are purged into the pressure suppression chamber air space 4

during a loss of coolant accident, the pressure resulting form isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber,and that the drywell volume is purged to the suppression chamber.

3 Using the minimum or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 psig, which is below the maximum of 62 psig. The maximum water level indications of -1 inch corresponds I

to a downcomer suboergence of 3 feet 7 inches and a water volume of 127,800 cubic feet with or 128,700 cubic feet without the drywell-suppression chamber differential pressure control. The minimum water level indication of -6.25 inches with differential pressure control and -7.25 inches without differential pressure control corresponds to a downcomer submergence of approximately 3 feet and water volume of approximately 123,000 cubic feet. Maintaining the water level between these levels will ensure that the torus water volume and downcomer submergence are within the aforementioned limits during normal plant operation. Alarms, adjusted for instrument error, will notify the operator when the limits of the torus water level are approached. The maximum permissible bulk pool temperature is limited by the i

potential for stable and complete condensation of steam discharged from safety relief valves and adequate core spray pump net positive suction head. At reactor vessel pressures i

285 AmendmentNo.)I,78

44 above approximately 555 psig, the bulk pool temperature shall not exceed 180 F.

At pressures below approximately 240 psig, the bulk temperature may be as much at 1840F.

At intermediate pressures, linear interoolation of the bulk temperature is permitted.

i They also represent the bounding upper limits that are used B.\\SES in suppression pool temperature response analyses for sr.fety relief valve discharge and 1.0CA cases. The actions required by specification 3.7.e-f assure the reactor can be de-pressurized in a timely manner to avoid exceeding the maximum bulk suppression pool water limits. Furthermore, the 1840F limit provides that adequate RHR and core spray purap NPSH will be available without dependency on containment overpressure.

Should it be necessary to drain the suppression chamber, this should only be done when there la no requirement for 3

core standby cooling systems operability. Under full power operation conditions, blowdown from an initial suppression chamber water temperature of 950F results in a peak long term water temperature which is sufficient for complete condensation.

i N

!.imitino suppression pool temperature to 105'T during RCIC, MPCI, or relief valve operation when decay heat and stored energy is removed f rom the primary system by diccharging reactor etnam directly to the suppression chamber assures adequate margin for controlled blowdown anytime,during RCIC operation and assures margin f or complete condensation of steam from the design basis lose-of-coolant accident.

In additien to the litattr. on temper. cure of the cuppression chamber pool water, operating precedures define the action to be taken in the event a relief valve inadvertently opens or sticks'open.

Thin action would include:

(1) u:e of all available coins to close the valve, (2) initiate suoneesnan pnol water couling he.tr exchancers (3) initiate Tactor shutdwn,.n!

08) if other relief valves are used to depressuri:e the reactor, their diccharge shall bc Separated frtm that of the stuck-open relief valw to assure mixing and unifornity of energy insertien to the pool.

If a locs-of-coolant accident were to occur when the reactor water temperature is below approximately 3JO*r, the containment pressure will not even if no condensation were to occur. exceed the 62 peig code permissible pros

^

The maximum allowable pool tempe raturn, whenever the coactor is above 212*F, shall be governed by this specification.

Thus, specif ying water volume-temperature requirements applicable for reactor-water temperature e

i above 212*r prov,Ldes additional marginabove that available at 3 J0

• r.

In conjunction with the Mark 1 Centainment'$hort Term Program, a plant unique analysis was perfonced (" Torus Support Syntes and Attached piping Analysis for s

the Browns Terry Nuclear Plant Units 1, 2, and 3." dated SepteeJer 9,1976 and supplemented October 12, 1976) which demonstrated a facter of safety of at least two for the weakest element in the suppression chamber support systes and attached piping. The maintenance of a drywell-suppression chamber differen-i tial pressure of 11 paid and a suppression chamber water level corresponding to a downconer submergence range of 3,0f, feet to 3.58 feet will assure the integrity of the suppressies chamber when subjected to post-14CA suppreestes peel hydredynaste forces.

286 Amendment No. I, 78

5

\\

l The containment desian has been examined to determine that a leakage equivalent to one drywell vacuum breaker opened to no

^

more than a nominal 38 as confirmed by the red light is acceptable.

on this basis an indefinite allowable repair time for an inoperable red light circuit on any valve or an inoperable check and green or check light circuit alone or a malfunction of the operator or disc (if nearly closed) on one valve, or an inoperable green and red or green light circuit along on two valves is justified.

During each operating cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber.

The drywall pressure will be inrJeased by at least 1 pel with respect to the i

soppression chamoer psameure and held constant.

The 2 psig set point w111 not be exceeded.

The subsequent suppression chamber pressure transient (if any) will be monitored with a senaitive pressure gauge.

If the drywell pressure cannot be increased by 1 psi over the suppression thanhar pressure it would be because a significant leakage path exists; in this event the leakage source will be identified and eliminated before power operation is resumed.

siith a differential pressure of greater than 1 psig, i

the rate of change of the suppression chamber pressure must not exceed.25 inches of water per minute as measured over a 10-minute period, r

wtich corresponds to about 0.14 lb/sec of containment air.

In the event the rate of change exceeds this value then the source of leakage will be identified and eliminated before power operation is resumed.

The water in the suppression chamber is used for cooling in the event of an accident:

1.e., it is not used for nortnal operation; theref ore, a daily check of the temperature and volume is adequate to assure that adequate heat removal capability is 3

presen t.

The interior surfaces of the drywell and suppression chamber are coated as necessary to provide corrosion protection and to provide a more easily decontaminable surface.

The surveillance inspection of the internal surfaces each operating cycle assures e,imely detection of corrosion.

Dropping the torus water level to one foot below the normal operating level enables an l

inspection of the suppression chamber where problems would first begin to show.

i, 289 i

Amendment No. 78

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l The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the t

event of a loss-of-coolant accident.

The peak drywell pressure t

would be about 49 psig which would rapidly reduce to less than 30 poig within 20 seconds followin Following the pipe break,'the suppression cha'g the pipe break.

aber p essure rises to 27 psig within 25 seconds, equalizes with drywell pressure, and decays with the drywell pressure decay.

l The design pressure of the drywell and suppression chamber is 56 psiq.

The design leak rate is 0.5 percent per day at the pressure of 56 psiq.

As pointed out above, the pressure response di the drywell and suppression chamber following an accident would nun the same after about 25 seconds.

Based on the calculated containment pressure response discussed above, the I

primary containment preoperational test pressures were chosen.

Also based on the primary containment pressure response and the i

pact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than the 1

l individual components separately.

3 The calculated radiological doses given in Section 14.9 of the psAR were based on an assumed leakage rate of 0.635 percent at i

the maximum calculated pressure of 49.6 psig!

The doses calcukated by' the Npc using this bases are 0.14 rem, whole body l

passing cloud gamma dose, and 15.0 res, thyroid dose, which are respectively only 5 x 10-3 and 10-1 times the 10 CFR 100

)

reference doses.

Increasing the assumed leakage rate at 49.6 l

psig to 2.0 percent as indicated in the specifications would i

increase these doses approximately a factor of 3, still leaving a j

margin between the calculated dose and the 10 CFR 100 reference values.

l Establishing the test limit of 2.05/ day provides an adequhte i.

margin of safety to assure the health and safety of the general public.

It is further considered that the allowable leak rate j

should not deviate significantly from the containment design value to take advantage of the design leak-tightness capability l

of the structure over its service lifetime.

Additional margin to maintain the containment in the "as builta condition is achieved i

l by establishing the allowable operational leak rate.

The.

allowable operational leak rate is derived by multiplying the maximum allowable leak rate (49 poig Method) or the allowable j

test leak rate (25 psig Method) by 0.75 thereby providing a 25%

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Amendment No. 78

bIMITINC CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3*11 f f RE _Pl'_T.%7JON SYSTrN,3, 34.11 FIRF PR CT 'ECT ION SYSTEMS

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3.

The class A supervisad detector alarm circuits will be tested once each two months at the local panels.

u.

Th. ci rmiit.: between the local panels in u.11.C.3 and the main control room will be tested montaly.

5.

Smoke detector sensitivity will

'a c

checked in accordance with manufacturer's instruction annually.

D.

ROVING FIRE WATCH D.

ROVING FIRE WATCH A roving fire watch will tour each area in which A monthly walk-through by autom. tic fire cuppression the Safety Engineer will systems are to be be made to visually installed (as described in inspect the plant fire the " Plan for Evaluation, protection system for signs of damage, Repair, and Return to Service of Brnwns Ferry deteriora tion, or abnormal Onits 1 and 2,"Section X) conditions which could at intervals no greater jeopardire proper than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

A keyclock operation of the system.

recording type system shall be used to monitor the routes of the roving fire watch. The patrol will be discontinued as the automatic suppression systems are installed and made operable for each specified arec.

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353 1

Amendment No. 78

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t LIMITU!G CONDITI0tlS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.11 FIRE PROTECTION SYSTEMS 4.11 TIRE FROTECTION SYSTEMS E.

Fire Protection Systems ?.nspection E.

Fire Protection Systems Inspections All fire barrier penetrations.

Each required fire barrier includinn cable penetration penetration shall be verified barriers, fire doors and to be functional at least once fire dampers, in fire zone per 18 months by a visual. inspect-boundaries protecting safety ion, and prior to restoring a related areas shall be funct-fire barrier to functional status ional at all. times. With one following repairs or maintensnee or more of the required fire

- by performance of a visual in -

barrier penetrations non-spection of the affected fire functional within one hour es-barrier penetration.

tablish a continuoos fire vacch on at least one side of the affected penetration or verify the OPERABILITY of fire detect-ors on at least one side of the non-functional fire barrier and establich an hourly fire watch patrol until the work is com-plated and the barrier is re'-

stored to functional status.

F.

Fire Protection Organization F.

Fire Protection Organization l

The minimum in-plant fire i

protection organization and No additional surveillance duties shall be as depicted required.

l in Figure.6.3-1.

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354 AmendmentNo.[,f,78 f

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LIMITING CONDITIONS FOR OPERATION SURVEILLAtlCE REQUIREMENTS 3 11. FIRE PROTECTION SYSTEMS 4.11 FIRE PROTECTION SYSTEw.S G.

Air Masks and Cylinders C.

Air Masks and cylinders A minimum of fifteen air No additional surveillance masks and thirty 500 cubic required.

inch air cylinders shall be available at all times except that a time period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following emergency use is allowed to permit recharging or replacing.

H.

Continuous Fire Watch H.

Continuous Fire Watch A continuous fire watch No additional surveillance shall be stationed in th*

required.

immediate vicinity where work involving open flame welding, or burning is in progress.

I.

Open Flames, Velding, and I.

Open Flames, Welding, and Burning in the Cable Burning in the Cable Spreading Room Spreading Room There shall be no use of No' additional surveillance open flame, welding, or required.

burning in the cable spreading room unless the reactor is in the cold shutdown condition.

4 355 Amendment No. /I, 78 I

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