ML20072K148

Amend 85 to License DPR-52,revising Tech Specs to Incorporate Limiting Conditions for Operation During Cycle 5 & Reflecting Design,Equipment & Procedural Mods Made During Refueling Outage

ML20072K148
Person / Time
Site:	Browns Ferry
Issue date:	03/11/1983
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Tennessee Valley Authority
Shared Package
ML20072K153	List: ML20072K148 ML20072K159 ML20072K173
References
DPR-52-A-085 NUDOCS 8303300396
Download: ML20072K148 (77)

Text

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UNITED STATES 8

NUCLEAR REGULATORY COMMISSION o

E, wasmworow, p.c. asses TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDENT TO FACILITY OPERATING LICENSE Amendment No. 85 License No. DPR-52 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated October 15, 1982, as supplemented by letters dated November 17, 1982 December 10, 1982 and January 7,1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

The issuance of this amendment will not be inimical to the common defense and securit.y or to the health and safety of the public; and E.

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

2.

Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility License No. DPR-52 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 85, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

8303300396 83o311 PDR ADOCK 05000260 P

PDR o

r 6

2 3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specificctions Date of Issuance: March 11, 1983 6

m e

ATTACHMENT TO LICENSE AMENDMENT NO. 85 FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Review Appendix A as follows:

1.

Remove the following pages and replace with identically numbered pages:

ii 27' 159 234 266 iii 30 160 235a 267 iv 33

.168 250 268 y

37 169 251 269 vii 39 s 171 252 273 viii 40 172 253 293a 256 294a 8

66 172a 9

98 181 258 297b 10 99 182 259 298 16 126 219 260 300 19 1 31 221 261 321 262 322 20 143 222 22 145 227 263 323 146 233 264 330 25 2.

Marginal lines on the above pages indicate the area being revised.

3.

Add the following new pages:

36a 160a 168a 253a 9

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t

,Section Page No.

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D.

Reactivity Anomalies..............

125 E.

Reactivity Control...............

126 F.

Scram Discharge Volume 126 3.4/4.4 Standhy Liquid Control System 135 I

A.

Normal System Availability...........

135 B.

Operation with Inoperable Components......

136' C.

Sodium Pentaborate Solution 137 3.5/4.5 Core and Containment Cooling Systems........

143 A.

Core Spray System (CSS) 143 D.

Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling).........

145 C.

RHR Service Water System and Emergency Equipment Cooling Water System (EECWS).................

151 D.

Equipment Area Coolers.............

154 E.

High Pressure Coolant Injection System (HPCIS)......

.154 F.

Reactor Core Isolation Cooling System (RCICS)...................

156 G.

Automatic Depressurization System (ADS) 157 H.

Maintenance of Filled Discharge Pipe......

158 I.

Average Planar Linear Heat Generation Rate...

159 l

J.

Linear Heat Generation Rate (LHGR)....

159 K.

Minimum Critical Power Ratio (MCPR) 160-I L.

Reporting Requi rements.............

160a 3.6/4.6 Primary System Boundary 174 l

A.

Therral and Pressurization Limitations.....

174 B.

Coolant Chemistry 176 l

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A endment No. 85

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y w-

Section Page No.

C.

Coolant Leakage 1 60 o.

Re11e1 v.1 es.

18i I

181 E.

Jet Pumps F.

Recircylation Pu=p Operation 182 G.

Structural Integrity..............

182 H.

Seismic Restraints, Suppor'ts, and Snubbers..

185 l

3.7/4.7 Containment Systems 227 A.

Primary Containment 227 B.

Standby Gas Treatment System..........

236 C.

Secondary Containment 240 D.

Primary Containment Isolation Valves......

242 9

E.

Control Room Emergency Ventilation.......

244 F.

Primary Containment Purge System..,......

246 G.

Containment Atmosphere Dilution System (CAD)..

248 H.

Containment Atmosphere Monitoring (CAM) System 249 l

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H2 Analyzer 3.8/4.8 Radioactive Materials 281 A.

. Liquid Effl uents................

281 B.

Ai rbo rne Ef fl uent s...............

282 C.

Mechanical Vacuum Pump.............

286 D.

Miscellaneous Radioactive Materials Sources 286 3.9/4.9 Auxiliary Electrical System 292 A.

Auxiliary El,ectrical Equipment.........

292 B.

Operation with Inoperable Equipment 295 C.

Operation in Cold Shutdown...........

298 3.10/4.10 Core Alterations..................

302 1

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A.

Refueling Interlocks..............

302 i ii l

Amendment No. 85 e

---w

,g 9-

Secticn Pace No.

B.

Core Monitoring 305

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C.

Spent Fuel Pool Water 305 D.

Reactor Building Crane......... 307 E.

Spent Fuel Cask

. ~.

307 F.

Spent Fuel Cask Handling-Refueling Floor 308 3.11/4.11 Fire Protection Systems 315 A.

High Pressure Fire Protection System.

315 l

B.

00 Fire Protection System...

319 2

C.

Fire Detectors......

320 321 D.

Roving Fire Watch E.

Fire Protection Systems Inspection... 322 F.

Fire Protection Organization...... 322 G.

Air Masks and Cylinders 323 H.

Continuous Fire Watch 323 I.

Open Flames, Welding, and Burning in the Cable Spreading Room.

323 J

330 5.0 Major Design Features 5.1 Site Features............. 330 5.2 Reactor............

330 53 Reactor Vessel 330 330 5.4 Containment.

330 5.5 Fuel Storage 331 5.6 Seismic Design 6.0 Administrative Controls 332 6.1 Organization 332 6.2 Review and Audit 333 iv l

4 endment i:o. 85

Section Pare fio.

.63 Procedures 33B 6.t Actions to be Taken in the Event of a Reportable Occurrer.ce in Plant Operatien............... 346 6.5 Actions to be Taken in the Event a 346 Safety Limit is Exceeded 6.6 Station Operating Records........ 346 6.7 Reporting Requirements 349 6.8 Minimum Plant Staffing 358 6.9 Environmental Qualification..

358 6.10 Integrity of Systems outside Containment.

359 6.11 Iodine Monitoring.

359 I

i s

V Amendment No. 85

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LIST Of TRLES (Cent'd) i

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Title Pace No.

Table 4.2.F Hinimum Test and Calibration Frequency for 105 Surveillance Ir.strumentation 4.2.G

' 5'urvei11ance Requirements for Control 106 Room Isolation Instrumentation-4.2.H Minimum Test and Calibration Frecuency 107 for Flood Protection Instrumentation 108

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4.2.J.

Seismic Monitorino Instrument Surveillance 3.5-1

'liinihu:n RHRSW and EECW Pu=p Assignment 152a '

-3.5.1 MAPLHGR Versus Average Planar F.xnosure.

. 171, 172 Reactor Coolant System inservice Inspection 4.6.A 209 Schedule....................

250 3.7.A Primary Containment isolation Yalves 3.7.8 Testable Penetrations with Doubic 0-Ring 256 Seals......................

257 Testable Penetrations with Testable Dellows....

3.7.C 258 3.7.D Primary Containrent Testabic Isolation Valves...

Suppression Chanber influent Lines Stop-Check

.3.7.E 263 Globe Valve Leakage Rates............

Check valves on Suppression Chamber influent 3.7.F 263

'~

Lines 265 3.7.H Testable Electrical Penetrations 287 Radioactive Liquid Waste Sampling and Analysis 4.0.A 288 Radioactive Gascous Waste Sampling and Analysis..

4.8.8 4.9.A.4.c Voltage Relay Setpoints/ Diesel Generator 298a Start 3.11.A Fire Protection Syste:n Hydraulic - ~~

j Require =er.ts 324 360 6.S.A Minimum Shif t Crew Requirements..........

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1 s,...

Amendment No. 85-~

1

0 LIST or 1ttu5;RATICN5_

Title Pete No.

.Finure 2.1.1 APFr. Flow Reference Scrae. and APRM Red Block 13 Settings....................

2.1-2 APRM Flow Bias $ cram Vs. Reacter Core Flow 26 4.1-1 Craphic Aid in the Selection of an Adecuate 49 Interval Between Tests 119 4.?-1 System Unavailability...............

3. 4 -1 Sodium Pentaborate Solution Volume Contentration 130 Requirements..................

3.4-2 Sodium Pentaborate Solution Temperature 139 Recuirements..................

172a l

3.5.K-1 MCPR Limits 173 3.5.2 K Factor.....................

7 3. f.- 1 Minimum Temperature 'F Aoove Change in' Transient

191, Temperature...................

3.6-2 Change in Charpy V Transition Teeperature vs.

195 lieutron Exposure................

'G.1-1 TYA Office of Power Organization for Operation 361 of Nuclear Power Plants.............

362 6.1-2 Functional Organization..............

363 6.7-1 Review and Audit Function...

364 6.3-1 In-Plant Fire Program Organization i

l viii Amencment No. 85

0 LIMITING SAFETY SYSTEM. SETTING

-- SAFCTY LIMIT 2.1 FUEL CLADDING INTEGRITY 1.1 FUEL CLADDING INTEGRITY Appli ca bili ty Applicability Applies to trip settings of the Applies to the interrelated instruments and devices which variables associated with fuel are provided to prevent the thermal behavior, reactor system safety. limits from being exceeded.

Obiective Obiective To define the level of the process variables at which To establish limits which automatic protective action is ensure the integrity of the initiated to prevent the fuel fuel cladding.

cladding integrity safety limit from being exceeded.

Specification Speci fica tions The limiting safety system settings shall be as specified A.

Thermal Power Limits

)

below:

1. Reactor Pressure > 800 A. Neutron Flux Trip Settings psia and Core Flow > 10%

~

of Rated.

APRM Flux Scram Trip 1.

Setting (Run Mode)

When the reactor pressure is greater than 800 psia, (Flow Biased) Mode Switch a.

When the the existence of a minimum is in 'the RUN critical power ratio position, the APRM (MCPR) less than 1.07 flux scram trip shall constitute violation setting shall be:

of the fuel cladding integrity safety limit.

ss(0.66W + S 4 %)

where:

S= Setting in per-cent of rated thermal power (3293 MWt)

W= Loop recircu-lation flow rate in Fer-cent of rated (rated loop recirculation flow rate equals

.3462x106 lb/hr) 8 Amend No. 25,' /6, E2, 85

1IMITING Cf.Tr.TY SYST :t sc;;Inc

.t n:TY

.1 t:17 1.1 fur!, CL*,0 !NG INTEGRITY 2.1 FUEL CLADDING 1:TEGRITY

b., in the event of c;cration with the core maximun f raction of lic.iting Nyer deosity (c'.TI.TD) greater than fraction of rated therust power (Fr#)

the setting chall be modified as follows:

s i (0.66W + $4%) FRP aouc For no combination of loop recir'cu-lation flow rate and core thermal c.

• Power shall the APRM flux scras trip setting be allowed to exceed 120:

of rated therrsal power.

(Note:'These settings assume operation within the basic thermal hydraulic design criteria. These criteria areLEGR 1113.4 kw/f t for 8x8 8x8R. and P2x8R, and MCPRwithin limits of k If is determincJ that 't ither of these tt asesign criter:s is be'ing v! ole ed during operation, action siis11 be

~

initiated within 15 cir.utes to restore operation within prescribe:d limits 4

Surve111acce requirenents f or APRM scram setpoint are given in specifitation 4.1.B.

The APRM Rod block trip

• d.

setting shall be:

S 5 (0. 6 6W + 42%)

g where:

g = Hod block settingin percent of' rated S

thermal power (3293 MWt)

= Loop recirculation W

flow rate in percent of rated (rated loop recirculation flow.

rate equals

34. 2 x 10* lb/hr) 9 Amendment No. 32, 35, 46, EB, E2, 85

's

. c

shr:Th. U MIT LIMITIII0 SAIETY SYSTEM EE IIIIC

\\

1. t rus:, cLArt:::c It:TEGRITY 2.1 FUEL CLADDING It:TEGRITY In the event of operation with the core maximum fraction of limiting power density (CMTLPD) greater than fraction of rated thermal power (TMP) the setting shall be modified

~

as follows:

S S (0.66W + 442% ) TRP RB CMTLPD Eixed High Neutron Flux Scram Trip e.

Setting--When the mode switch is in the RUN position, the APRM fixed high flux scram trip setting shall

• ~ '

be: Sil20% power.

2.

APRM and IRM Trip settings 2.

Reactor Pressure f800 PstA (Startup and 11ot Standby Modes).

or Core Flow 110% of rated.

k' hen the reactor prest.ure a..APRM--When the is $800 PSIA or core flow reactor mode switch is 610% of rated, the core is in the STARTUP thermal power shall not position, the APRM exceed E23 int (^'25% of scram shall be set at rated thernal power)'.

less than or equal to 15% of rated power.

b.

IRM--The IRM scram shall be set at less than or equal to 120/125 of full scale.

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i Amendment'Id. 22, 82, 85

o

.1.1 BASES r,ecause the boiling transition correintien is based on a Itzge quantity ot' rull scale data there is a very high confidence' that operation of a fuel assembly at the condition of HOPR = 1.07, vould not produce boiling tran-sition.. Thus, although it is not required to establish the esfeiy limit additional margin exists between the safety limit and the cetusi occurence of loss of cladding integrity.

However, if boilinE transition vere to occur ~ clad perforation vould not be expected. Cladding te=peratures vould increase to cpproxtuately 0

1100 T vhich is below ti perforation temperature of the cladding _

materi al.

This has been verified by tests in the General E3 ectric Test Reactor (CETR) vbere fuel sitilar in design to ETNP operated above the critical heat flux for a significant period of ti=e (30 minutes) vithout clad perforation.

If reactor pressure should ever exceed Ih00 psia during nor=sl power operating (the limit of applicability of the boilins transition corre-lation) it vould be assu=ed that the fuel cladding ii.tegrity Safety Limit has been violated.

In addition to the boiling transition limit.(MCPR - 1.07) operatior is l

constrained to a maximum LHGR of 13.5 kv/f t for 8x8, 8x8R,and F8x8R.This limit is reached when the Core Maximum Fracticn of Limiting Pouer Density equals 1.0 (CMFLPD - 1.0).

For the este where Core Maximun Traction of Limiting Power Density exceeds the Trcction.of Rated Therns1 Power, operation is permitted only at less than 2001 of rated power and only with reduced APRM scram settings as required by specification 2.1.A.1.

At pressuren belov 800 psia, the core elevstion pressure drop (0 power, l

0 flow) is greater than L.56 pai.

At lov powers and flers this pressure differential lo maintained in the bypass region of the core. Sicce the i

pressure drop in the bypass region is essentially all elevation haad, the core pressure drop at lov powers and flov vill alvags te greater than h.56 pei. Analyses show thet with a flev of 2EX10 lbs/hr bundle flov, bundle pressure drop is nearly independent of bundle pover and has a value of 3.5 psi. Thus,3the bundle flov vith a 4.56 psi driving head vill be greater than 28x.10 lbs/hr. Full cesle ATLAS test data taken at presettres frc= 1k.7 psia to 800 psia indicate that the, fuel nace=bly critien1 power nt this flov is approximately 3 35 Kn6 Vith the design t

peaking factors thie corresponds to a core ther=al povrr cf core than i

50f.

Thus, a core thermal powcr it=1t of 25% for rea: tor pressures belov 000 psia is conservative, r

~

for the fuel in the core during periods when the reactor is shut dovn, -con-sideration must also be given to water level requirer.ents due to the effect nr decay heat.

If vster level should drop below the top of the fuel during Liais time, the ability to remove decoy heat is reduced. ' Thin reduction in cooling capability could lead to elevated cladding temperatures and clad perforation.

As long as the fuel ramains covered with vater, sufficient cooling is available to prevent fuel clad perforation.

16 l

/

Amendment No. 32, EE, AE, Es, 85 t

2.1 BASES:

LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CLADDING INTEGRITY The abnormal operations 1' transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed' throughout the spectrum of planned operating conditions up to'the design thermal power condition of 3440 MW t.

The

^

analy se s were based upon plant o;sration in accordance w i th -

the operating map given in Figure 3.7-1 of the FSAR.

In addition, 3293 MWt is the licensed maximum power level of Browns Ferry Nuclea'r Plant, and this represents the maximum steady-state pow er which shall not kn o.w i n g l y be exceeded.

Conservatism is' incorporated i n' 'th e* t r a n s i e n t analyses in estimating the controlling factors, such as void reactivity coefficient, control rod s cram worth, scram delay time, peaking.{ actors, and axial power shapes.

These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current a naly si s model.

This transient model, evolved over many years, has been substantiated in operation.as a conservative tool for evaluating reactor dynamic performance.

Results.

s obtained from a General El e c t r i c boiling water reactor have been compared with predictions made by the model.

The

- compa risons 'a nd re sul ts are summarized in Referencg 1, 2

and 3.

~

~The absointe value of the void reactivity coefficient used

'in the analysis is conservatively estimated to be about 25%

than the norma 1' maximum value expected to occur greater ~ the core during 1,i f e t im e.

The s cr am w or th use d ha s been derated'to be equivalent to approximately 80% of the total scram worth of the control rods.

The scram delay time and rate of rod insertion allowed by the analyses are conseryatively set equal to the longest delay and slowest insertion rate acceptable by Technical Specifications as g

further-described in Reference 4.

The effect of screm I

w or the scram de(ay time and rod insertion rate, all c on s e rv a t iv ely applied, are of greatest significance in the early portion of the negative reactivity insertion.

The 4

. rapid insertion of ne g a t iv e reactivity is assured by the time requirements for 5% and 20% insertion.

By the time the rods are 60% inserted, approximately four dollars of negative reactivity has been inserted which strongly turns the t r a n s i e n't, and accouplishes the desired effect.

The times for 50% 'and 90% insertion are given to assure proper completion of the expe c te d pe rf ormanc e in the earlier portion of the transient, and to establish the ultimate

~~'

fully sh ut d ow n steady-state condition.

~

For analyses of the thermal-consequences of the transients a MCPR > limits specified in specification 3.5.k is co'n s e r v a t'sv e ly -assumed to exist prior to initiation of the transients.

This choice of using conservative values of controlling parameters and initiating transients at the design power level produces more pessimistic answers than would result by using expected valuee of control parameters 7

and analyzing at higher power levels.

Ie> 1 1

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Amendment ik 32. 35. W 8_E

2.1 BASES In su= mary 1.

The licensed maximum power level is 3,293 KWt.

2.

Analyses of transients employ adequately conservative values of the controlling reactor parameters.

3 The abnormal operational transients were analyzed to a power '1evel of 3440 MWt.

4.

The analytical procedures now used result in a more logical answer than the alternative method of assuming a higher starting power in conjunction with the expected values for the parameters.

The bases for individual set points are discussed below:

A.

Neutron Flux Scram 1.

APRM Flow-Biased High Flux Scram Trip Setting (Run Mode)

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3293 MWt).

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

During transients, the instantaneous fuel surface heat flux is less than the instantaneous neutron flux by an amount depending upon the duration of the transient and the fuel time constant. For this reason, the flow-biased scram APRM flux signal is passed through a filtering network with a time constant which is representative of the fuel time constant.

As a result of this filtering, APRM flow-biased scram will occur only if the neutron flux signal is in excess of the setpoint and of sufficient time duration to overcome the fuel time constant and result in an average fuel surface heat flux which is equivalent to the neutron flux trip setpoint. This setpoint is variable up to 120% of rated power based on recirculation drive flow according to the equations given in section 2.1.A.1 and the graph in figure 2.1.2.

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

biased provides additional margin to the thermal limits for l

> slow transients such as loss of feedwater heating. No safety credit is taken for flow-biased scrams.

I 20 Amendment No. 85 1.

O 2.1 BASES IRM Flux Scram Trip Settine (Continued) example, if the instrument were on range 1, the scram setting would be at 120 divisions for 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 acco=modate the increase in power level, the scram setting is also ranged up.

A scram at 120 divisions on the IRM instruments.

remains in effect as long as the reactor is in the startup mode.

In addition, the APRM 155 scram ~ prevents 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 due 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 suberitical and the IRM system is not yet on scale. This condition exists at quarter rod density.

Quarter rod density is illustrated in paragraph 7.5.5 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 l

calibrated using heat balance data taken during steady-state l

conditions, reads in percent of rated power (3293 MWt). The l

APRM system responds directly to neutron flux. Licensing analyses have demonstrated that with a neutron flux scram of 120% of rated power, none of the abnormal operational transients analyzed violate the fuel safety limit and there is a substantial margin from fuel damage.

l 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 a given point i

at constant recirculation flow rate and thus to protect against the condition of a MCPR less than 1.07.

This rod block trip setting, which is automatically varied with recirculation loop

~

flow rate, prevents an increase in the reactor power level to excess values due to control rod withdrawal. The flow variable I

trip sett$ng provides substantial margin Amendment No. 45, f,6, 85

e a 1.I 8,AS,EJ I

J. & K.

Reactor low vater level set point for initiation of HPCI and isolation valves, and starting LFCI KCIC, closann main s t e.1m and core spray pumps.

These systens maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures.

The design of these systems to adequat ely per f orm the intended func-tion is based on the specified lov level scram set point and initia-tion set points. Transient analyses reported in Section 1/. of the TSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the systen pressure.

L.

Re f e r enc e_s, Linf ord, R.

B., " Analytical Methods of Plant Transient Evaluations f or 1.

the Ceneral Electric Boiling Water Reactor," b'EDo-10802, Teb., 1973. _.

2.

Generic Reload Fuel Application, Licensing Topical Report NEDE-20411-P-A, and Addenda.

3. " Qualification of the one-Dimensional Core Transient Model for Boiling Water Reactors," NEDO-25154, NEDE-24154-P, October 1978.

~

4. Letter from R. H. Buchholz (GE) to P. S. Check (NRC), " Response to NRC Request f or-Information on ODYN Computer Model," September 5, 1980 25 Amendment tto. 85 e

e.

I,f MITING. SArt. 'T f;Y:'T!R St.7ING

.gytTY Lilt!"'

_.............._.---.g a.2 RIACT0R C001 ANT SYSTD4 I NT ECR_1_T_T 2.2 REACTOR Com. ANT SYSTD( INTECRITT_

Applicsbility gpiscability Ap' plies to trip settings of the A, plies to limits on reactor coolant and devices'which are in,struments s; sten pressure provided to prevent the reactor systen safety limits from being exceeded.

Oge.Q!e, 0)1ective, fu estaolish a limit belew which To defior the level of the process

~

variables at which automatic pro-the integrity of the reactor coolant tective action is initiated to system is not threatened due to en the pressure safety liasit prevent overpsessure condition, from being exceeded, s

Specification.

5 ecifiestion The limiting asfsty system settings A.

The pressure at the lovest point shall be as specified below:

of the reactor vessel shall not exceed 1,375 psig whenever Limiting Safety irradiated fuel is in the re'ac-Fratective Action Syetem Settini, tot vassal.

-m

-ee+-e,4 A.

Nuclear systen 1105 pois +

relief valves 11 psi (4 open--nuclear valves )

system pressure 1115 pois +.

11 psi (4 valves) 1125 p s i t +.

11 psi ( 5 I

valves)

B.

Scrae--nuclear f l.05) peig system high pressure 27 Amendment f'0. M, 85

i 2.2 EASES REACTOR COOLA!!T SYSTEM INTEGRITY To ceet the safety basis, thirteen relief valves have been installed' en the unit with a total capacity of 84.1% of nuclear boiler rated steam flow. The analysis of the worst. overpressure transient (3-second closure'of all cain steam line isolation valves) neglecting.

the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves ' operable, results in adeq'uate margin to the. code allowable overpressure limit of 1375 psig.

To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well.below the allowed vessel overpressure of 1375 l

psig.

1 l

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l 30 Amendment No. 25, f6, ES, 85 L

se i

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TAf3LE 3 1. A IlEACT0ft PROTECTION SYSTEr1 (SCR Af t) It!STRiit1EllTATION llE0tlIREr1EllT

1.30

,/.

.:=

.s-2 ';. -,..:

?. 4=.=

1.29

=

'..,L '

oc 1.26

/.,

c

+=

u

C 1.27-

/

_2 :-

=

v,.-

1 1.26-

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f M-

.'f

. ~. - -

-. 'n.

_ _ _. M.

,-f

__4,

^1 N:

1*25 -7

=

,y

.w

. - =

e=.=.__

1.24 --

etr

.,. c.

J.

1.23 i~~'

,,y. q _,._

a

~

- _~

1.22

~~-

I n

i i

i i

4 6

.l 0.0 0.1 0.2 0.3 0.4 0.5~ <0.6 0.7 0.8 0.9-1.0 Y

Figure 3.5.K-1 MCPR Limits

172e,

' l' Amendment flo. H, EB, 77, 85 l

n

i

.N e

e-

~

> l-1.1P.*T!SO' CC:CITICSS TCF OPEFJ. TICS STxygII,1;J:ct pzg;py,gd;$

3.4,C coelaat Leaken 4.6.C Ceelant tests _ee 3.

If' the condition in 1 or 2

~

above te nnot be met. an orderly shute m shall be initiated and the reactor shall be smut.

D. Relief Valves i

dove la the Cold Conditten vithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1.

' A pproxiestely one.hslf of att t

D.

Relief Valves relief valves shall be benen.

checked or replaced with a i

1 When more than or.r relief bench-checked valve each opert..

r

r. int, cyc1t. All 13 valves j

j valves are known to be vill

'r. ave failede an ' orderly shutdown been checked or re;1 aced upon shall be initiatad and the th* to=81stion of every setend cycle.

reactor depressurised to

~

less than 105 psig within y

21. hours.

2.

Once during each operating cycle, each relief valve

[

shall be canually opened until thermocouples and g

acoustic aonitors dounstrean of the valve indicate stes:n is flowing from the valve.

3.

The integrity of the relief /

safety valve believs shall be continuowsly sonitored.

r 4.

At least ene relief valve shall be disessembled and inspected

/ ~,

each operating cycle.

I j',*.

1 g.

Jet Pumps i

t.

Jet rumes s'_

1.

Whenever the resetor le in the 1.

Vhenever there is recirculation startur or run emes. all jet flov vith the reacter in the

,o pumps shall le operable. If startup er run c.oits with both

~

.it is deterwined that a jet recirculation pu=;s running.

punt le inoperette or if cvo jet pusp operability shall be or more jet pwn; flow instrv-checked' daily by verifying that

~-

e.ent failures occur ar.d can-the following cenditicas ce not 3..

c.

i

-?

not be corrected within 11 occur striultaneously hovrm. en ordi:rty shutdo e i

ehall be initiatet and the a.

The two recirculation toeps l

' ~

reacter sh.11 be shucJe w in have a flow te.halance a:

3 l

l the Cold Condition within 24 15t or mere when the ;w u s heure.

are operated at the s ne t

I speed.

~,

1 161 t

t s ~

l

[.

G

/

l Ame!idment No. E, 3 6 f t.x85 5

w

}.;

- r e

^f t

1 I,IHTTING CONDITIONS FOR OPERATION S U R VE I!,l. A N C E RFOUIREMENT a

4.6.E Let Punes 3.6.F Recirculation Pump Operation b.

The indicated value of core flow rate varies 1.

The reactor shall not be from the value derived ope re te d with one from loop flow recirculation Icop out of measurenents by more service for more than 24 th a n 10%.

hours.

With the reactor operating, if one c.

The diffuser to lower recirculation loop.is out plenan differentisi of service, the plant shall pressure reading on an be placed in a hot shutdown individual jet pump

' condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> varies from the mean of uniesa the loop is sooner all jet pump returned to service.

differential pressures by more than 10%.

2.

F o l,l o w i n g one pump operation, the discharge

2. Whenever there is valve of the low speed pump recirculation flow with the may not be opened unless reactor in the Startup or the speed of the faster Run Mode and one pump is less than 50% of recirculation pump is its rated speed.

operating with the equalizer valve closed. the

3. Steady state operation with diffuser to lower plenum both recirculation pumps differential pressure shall out of service for up to 12 be_ checked daily and the hours is permitted.

During differential pressure of an such interval restart. of individual jet pump in a the recirculation pumps is loop shall not vary from permitted, provided the the mean of all je t pump loop discharge temperature differential pressures in is within 75*F of the that loop by more than saturation temperature of 10%.

the reactor vessel water as determined by dcme F.

Rjcirculation Purr.p Operation pressure.

The total elapsed time in natural f l.

Recirculation pump speeds circulation and onc pump shall be checked and logged operation must be no at least once per day.

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

2. No additf onal surveillance required.

3.

Before startinn either recirculation pump during steady state operation, check and lor the loop discharge temperature and dome saturation temperature.

C.

Structural Interrity G.

Structural Inteerity 1..The structural integrity of 1.

Table 4.6.A together with the primary sy s t e m shall be supplementary notes.

L specifies the 182 Amendment No. 5/, E2, 85

3.6/4.6 BASES

~

detected reasonably in a catter of few hours stilizing the ava'ilable leakage detection schemes, and if the origin cannot be determined in a reasonably short tice the unit should be shut down to allow further investigation and corrective action.

The total leakage rate consists of all leakage, identified and unidentified, which flows to the dr ywell floor drain and equipment drain sumps.

The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 sp=.

Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.

REFERENCE Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10) 3.6.D/4.6.D Relief Valves To meet the safety basis, thirteen relief valves have been installed on the unit with a total capacity of 84.1% of nuclear boiler rated steam flow. The analysis of the worst overpressure transient (3-second closure of all main steam line isolatien i

valves) neglecting the direct scram (valve position scram) results in a maximme vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate-cargin to the code allowable overpressure limit of 1375 psig.

To meet operational design, the analysis of the plant isolation I

0 transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is.well below the allowed vessel overpressure of 1375 psig.

i 219 Anendment No. 35, fE, EE,, 85 4

m,

3. t. / t. f.

af.s.rd:

differ by 10 percent or cere. the core flew tata nessured by the.

he puan s'.f fuser dif f erential ytessure sys tem cust be checked against tr If ther de jet I'f the dif f erence between pescured and dertved core flow rate core flow (with the derived value higher) diffuser measurepents correlatic..

10 per:ent er retet aken to ' define the location within the vessel of f ailed jet pursp is nottle (or etter) and the unit shut devn for repairs. If the potential vt11 be resistance to the recirculation increased, the system blowdown flow area is the af f ected drive pwsp vill "run out" to a puae is also reduced; hence,her flow rate (epstoximately lif perceht to 120 percent e nha t ant t.4117 hit the If ths two loops are balanced in flov at for e ate le not:1e failure). speed, the resistance charactettetics cannot have changed.

Atty tabalance between drive loop flow rates would be indicated by the plant esme pump procese instrurentation. In addition. the affected jet pusy would provide a the core thus reducing the core flow rate. The reverse 4

le ska r,e pa th pa st flow tht ourh the inactive j et pv=p would s till be indicated by a po sitivethe net e f fe dif f erential press.ure but cent to 6 percent) in the total core flow measured. This decrease, together with the loop flov increase, would result in a lack of correlation between Finally, the affected jet pv=p diffuser meneured.snd derived core flow rate.

dit f erential pressur's signal vould be reduced because the backflov vould be less than the r.ormal f orvard flow.

f ailure could also generate 'th$ coincident f ailure of N

A nar.tle-r t eer erotem n )ct pu-p diffu er body: however, the converse is not true. The lack of impussible any sdatantial stress in the jet pump dif fuser body makes f ailuss an initial nottle rtser system fatture.

a without i

i 3.6.F/4.6.7' Recirculation pumn Operation _

4 Steady-state operation without forced recirculation vill not be permitted And the start of a recirculation pump from the for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

natural circulation condition will not be permitted unless the temperature dif ference between the loop to be started and the core coolant temperature This reduces the posit 1~ e reactivity inscrtion to an v

is less than 750F.

acceptably low value.

Requiring the discharge valve of the lower speed loop to remain closed w 50% of its rated speed until the speed of the faster pu=p is belo.

provides assurance when going frc.n one to two pump operation that excessive vibration of the jet pump risers vill not occur.

8 i

221 Amendment No. 55, 85

3.6/4.6 BASES:

3.6.Gic.6.G Structural inteority The requirements for the reactor coolant systems inservice inspection program have been identified by' evaluating the need for a sampling examination of areas of. high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.

The program reflects the built-in limitations of access to the reactor coolant systems.

It is intended that the required examinations and inspection be completed during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.

Only proven nondestructive testing techniques will be used.

More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to These welds provide additional protection against pipe whip.

were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems Selection was based on judgement from actual or control systems.

plant obsrevation of hanger and support 1ccations and review of Inspection of all these welds during each 10-year drawings.

Inspection interval will result in there additional examinations above the requirements of Section XI of ASHE Code.

An aegmented inservice survelliance program is required to determine whether any stress corrosion has occurred in any stainless steel piping, stainless ccmponents, and highly stressed alloy steel such d

l as hanger springs, as a result of environmental ' conditions i

associated with the March 22,1975 fire.

l 222 Amendment No. 85 1

1

L:M:T;NC CONDITICNS TCR CPIRATION SURVEZLI.ANCE REO*.*!RIMT. HTS

3. { CCf.ChlN"CN7.SJOI-E Ma7 CU f#~ A I "ENT SYS7E"S U

A App 12cabality y1 t ea bi li t y Applies to.the operating status Applies to the primary and of the primary and secondary secondary containment containment. systems.

integrity.

Obiective Obieetive To assure ths integrity c! the primary and secondary To verify the int yrity of.the containment systems.

primary and secondary containment.

E2ts1(1 Sat. lod Scecification A.

Primar,,y Con H 1_n, 3 A.

Primary Containment 1.

As any time that the irradiated fuel is in 1.

Pressure Sueeressiet.

the.* actor vessel, Chamoer and the nuclear system is pres surized a.

The suppression above at.mospheric charter water level pressure or work is be checked once per being done wnica has day. Whenever heat the potential to

• is added to the drain the vessel, the

~

suppression D001 by pressure suppression pool water level and testing of the ECCS temperature shall be or relief valves the maintained within the pool temperature shall felacwing lim:ts be continually rnonitored eceept as specified and shall be observed in 3.7.A.2.

and logged every 5 minutes until the heat a.

Minimum water level =

addition is terminated.

l

-6.25" (Differential pressure control

>0 psid) l

-7.25" (O PSID Differen -

tial pressure control) b.

Maxi =u:n water level =

-1" 227 Amendment No. #2, 85

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A Prirery Centefreent 4.7.A Prirery Containnent within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> f ollowing detection of excessive local leakage, the reactor shall be shut down and depressurized until repairs are ef fected and the local leakage meets the acceptance criterion as demonstrated by re te st.

i. The main steamline isolation valves shall be tested at a pressure of 25 psig for leak-age during each refueling outage.

If the leakage rate of 11.5 scf/hr for any one main steamline isolation valve is exceeded, repairs and retest shall ie performed to correct the condition.

J.

Continuous Leak _ Rate Monitor When the primary containment i's inerted the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements. This monitoring system may be taken out of service for maintenance but shall be returned to service as soon as practicable.

l k.

Drvwell end Torus _ Surfaces The interior surfaces of the drywell cnd torus above the level one f oot below the normal water line and outside surfaces of the

~

torus bel ow the water line shall be visually inspected each operating

(

cycle f or deterioration and any signs of structural damage with particular attention to piping connections and supports and f or signs of distress or displacement.

233 Amendment No. 85

"^

1 IMIT tHC Ct1NatT 10M FOR CPFA\\ TION SURVEIL 1J NCE REQ;'IT Ewy,NTS Primary Ce n t.s inm e n t_

4.7. A Prima rv containment J.7.A _

3.

Freenure Suppreeston Chamber -

Reactor Buildine vacuum Dreakers Except as specified in a.

3.7.A.3.b belov, two 3.

Pressure Suppression Chamber-Reactor i

pressure suppression chamber-reactor building Building Vacuum Breakers vacuum breakers shall be a'.

The pressure suppression chamber-operable at all times when primary containment inte-reactor building vacuum breakers I

shall be exercised and the associ-trity is required. The f

ated instrumentation including set point of the differen-setpoint shall be functionally l

tial pressure instrumenta-tested for proper operation each

. tion which actuates the f

three months.

pressure suppression cham-4

-ber-rcoct.or building b.

A visual examination and determina-vacuum breakers shall be tion that the force required to O.5 paid.

open each vacuum bre ker (check valve) does not exceed 0.5 psid b.

From and after the date

. Will bs made each refueling outage.

that one of the pressure i

suppression chamber-reactor building vacuus breakers is i

made or found to be inopera-ble for any reanon. reactor operation is permiscible only during the succeeding 4

Drywell-Pressurc Suooression s-seven days, provided that Chamber Vacuum Breakers the repair procedure does j

not violate prir.ary contain-i a.

F.ach drywell-suppression ment int e r,rit y.

chamber vacuum breaker j

shall be exercised through 4.

Drywell-Preo ure Suppression an opening-closing cycle Chsmber Vacuum Breakers every month, When primary containment a.

is required. all drywell-

.oupprenoion chamber vacuum brer. hero shall be operable

.and positioned in the fully b.

When it is deter =ined that closed position (except two vacuum breakers are during testing) except as inoperable for. opening at a cpecified in 3.7.A.4.b and time when operability is requtr c, below.

all other. Vacuum breaker b.

One drywell-suppression chamber vacuum breaker may be non-fully closed so long as it ic determined to be not more than 3*

~.

open as indicated by the L-position lights.

734 v

s Amendment ?!o. 35

LIEITING CONDITICNS FOR OPEPATION SURVEILLANCE REQUIREMEh'25 3.7 Cot?"AINMENT SYSTE.MS t.7 CONTAINKENT SYSTEMS 6.

Dryvell-Suppression Chamber 6.

Dryvell-Suppression Chamber Diff erential Pressure Diff erential Pressure -

a.

Diff erential pressure a.

The pressure diff er-between the dryvell and ential between the suppression chamber shall dryvell and suppression be maintained at equal chamber shall be recorded l

to or greater thsn 1.1 at'least once each shif t.

psid except as specified in (1) and (2) below:

(1) This differential shall be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of achieving operating te=perature and pressure. The differential pressure nay be reduced to j

less than 1.1 psid l

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scheduled shutdown.

(2) This differential t

(

nay be decreased to l

less than 1.1psid for a maximum of four l

hours during required l

operability testing l

of the EPCI systen, RCIC system and the l

dryvell-pressure

~

suppression chamber vacuum breakers.

l b.

If the differential l

pressure of specifica-tien 3.7.A.6.a cannot be maintained and the differential pressure cannot be restored within

'~'

the subsequent six (6) hour period, an orderly shutdown shall be init-iated and the reactor shall be in the Cold

~

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

235a l

A.re n dm e r.t. !'O. 42, 85

4 8 5

o TABLE 3.7.A D

PRIMART (X)lffAINHENT ID0tATION VALVES J

' (D Number of Power Maximum Action on

' :s Operated Valves operating Normal Initiating O

Croup Valve Idmtification Inboard outboard Time (sec.)

Position Signal 1

Main steamline isolation valves 4'

4 3<T<5 0

GC (FCV-1-14, 26, 37, s 51; 1-15, 27, 3a & 52) 0 CC 1

Main steamline drain toolation 1

1 15 valves (FCV-1-55 5 1-56) 1*

Reactor water sample line isola-1 1

5 C

SC tion valves 2

RRRS shutdown cooling supply isolation valves (rcv-74-48 s 47) 1 1

40 C

SC 2

30 C

SC 2

RHRS - LPCI to reactor (FCV-74-53 5 67) 2 Reactor vessel hiemt neray inola-tion valves (FCV-74-77 & 78) 1 1

Je c

SC w

2 RERS flush and drain vent to

=

Sc 4

20 c

y suppression chamber (FCV-74-102, 103, 119, s 120) 2 15 c

SC l

2 suppression Chamber Drain I

(FCV-75-57 & 58) 2 Drywell equipment drain discharge Isolation valves (ftv-77-15A 6 158) 2 15 0

GC 2

Drywell floor drain discharge isolation valves (rcV-77-2A & 28) 2 15 0

GC I

l 5

\\

linc l

• These valves isolate only on., reactor vessel low low water level.,"(470',') and maln steam high radiation of. Croup 1 iso,1ations.

0 e

a

s,

~

~

ca

=

m e

2 TABLE 3.7.A (Continued) e Action os n

Maximum Initiating Normal Necber of Power Operating 51 pg, 7

Operated Valves _

Time (sec.)_

Position rc Inboard Outboard yelve Identification CC e

gs, 0

m 30 Reactor water clunup system supply 1

1 1 solation valves FCV-69-1, & 2 3

l CC o

10 1

CC 0

FCV 73-31 (Bypass around FCV 73-3)-

1 1

20 4

-UFCts steauline isolation valves

~'

~

CC ;

4 0

TCV-73-2-& 3 15 1

1 RCICS stesoline trolation valves 5

FCV-71-2 & 3 y

SC g

C l

Dryvell nitrogen purge intet isola-1 5

6 tion v4lves (FCV-7 6-18)

SC C

5 suppression chanber nitrogen purse 1

isolation valves (TCV-76-19) 6 talet SC C

2 2.5 Dryvell Main Exhaust isolation

, SC 6

valves (rcy-6f.-29 and 30)

C 2

2.5 Suppraesion chamber main exhaust isolatten valvve (rcy-64-32 and 33) 6' 3C C

Dryve11/Suppreaston Chamber porge 1

2.5 6

inlet (ICf-54-17)

SC C

0 Drywell At.noephere purge inlet -

1 2.5 (IEV-64-18) r I

e

.-...,.m.+e

...s.

..mm e--.

e.

Ti

}{!Y TABLE 3.7.A (Continued) f(unber of Power Maximum Action on s

Operated Valves Operating Ilonnal Initiating R

Group Valve Identification _

Inboard Outboard Time (sec.) Position Signal g

6 Suppression Chanter purge inlet 1

2. 5' C,

SC (FCV-64-19) 6-Drywell/Suppmssion Chamber nitro-1 5

C SC gen purge inlet (FCV-76-17) 6 Drywell Exhaust Valve Bypass to Standby Gas Treatment System 1

5' o

cc (FCV-64-31) 6 Suppression Chanber Exhaust Yalve Bypass to Standby Gas Treatment j

1 o

GC System (FCV-64-34) 5j i

i6 System Suction Isolation Valves to Air Compressors "A" and "B" 2

15 o

cc j

(FCV-32-62, 63) y

!7 RCIC Steamline Drain (FCV-71-6A, 68) 2 5

0 GC 7

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

e sc 7

HPCI Hotwell pump discharge isola-tion valves (FCV-73-17A,178) 2 5

C SC 7

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

0 GC 8

TIP Guide Tubes (5)

~~~

l per guide NA C

GC tube k

~s r

, ~,., ~.....

$ s l{

.a s..

.3.e f.

' U.

In11.E 3.7.A (Continuail)

Action on Maximum operated valves opera ting Normal Initi a tirvi Number of rcwr Inboard outton.d Time (sec.)

Position _

signal Valve Identification

$g check valves (cv 63-526 & $25) 1 1

HA C

Process standby 11guld control systm 2

2 tiA 0

Process Feedwater check valves (CV-3-558, 572, 554 & 568) 1 1

MA 0

Process control rod hydraulic return check valves (CV-85-576 s 573) i g

RHRS - IECT to reactor check MA c

Process 2

valves (CV-74-54 5 68) t 7.

6 CAD' System Torus /Drywell Exhaust 2

10 C

SC to Standby Cas Treatment (FCV-84-19,20) 6 Drywell/ Suppression Chamber Nitrogen 1

5 C

SC Purge Inlet (FCV-76-24)

Core Spray Discharge to Reactor 2

NA C

Process Check Valves FCV-75-26,54 6

~~

+.. _..

• hhJh

&em#

aa

+

S R

TABLE 3 7.A (Continued) a

[

tiumber or Power Maximum Actton on 6

Operated Valves Operating flormal Initiatinr.

Group Valve Identification Inboard Outboard Time (sec.)

Position Signal 6

Drywell oP air compressor suction 1

10 0*

GC valve (FCV-64-139) 6 Drywell AP air. compressor discharge 1

10 0

GC-valve (FCV-6ff-1110) y 6

Drywell CAM suction valves 2

10 0

CC y,'

(FCV-90-2581A and 254D) 6 Drywell CAM discharge valves 2

10 0

CC

-(FCV-90-257A and 257B) 6 Drywell CAM suction valve 1

10 0

GC (FCV-90-255)

• This valve cycles open and closed during normal operation.

r, o

I e

TABLE 3 7.B TESTABLE PENETRATIONS WITH DOUBLE O-RING SEALS X-1A Equipment Hatch X-1B X-4 DW Head Access Hatch X-6 CRD Removal Hatch X-35A T.I.P. Drives X-35B X-35C X-35D i

X-35E

[

l X-35F X-35G X-47 Power Operations Test j

i X-200A Supp. Chamber Access Hatch X-2003 X-213A Suppression Chamber Drain f

l X-223 Supp. Chamber Access Hatch DW Flange-Top Head Shear Lug Inspection Cover #1 i

I Hatch #2 n

n n

n g3 n

n n

n og

1. 5

1. 6 a

n a

n p7 s

n n

.n n

p3 i

t

[

-256-i e

i.- n e a nn+

• M OR

TABLE 3.7.D AIR TESTED ISOLATICN VALVES Velve Valve Identification 1-14 Main Steam 1-15 Main Steam.

1-26 Main Steam 1-27 Main Steam 1-37 Main Steam 1-38 Main Steam 1-51 Main Steam 1-52 Main Steam l

1-55 Main Steam Drain 1-56 thin Steam Drain 2-1192 Service Water 2-1383 Service Water e

3-554 Feedwater j

3-558 Feedwater 3-568 Feedvater 3-572 Feedwater 32-62 Drywell Compressor Suction 32-63 trywell Compressor Suction 32-336 Drywell Compressor Return 32-2163 Drywell Compressor Return 33-1070 Service Air 33-785 Service Air 43-13 Reactor Water Sample Lines 43-14 Reactor Water Sampla Lines63-525 Standby Liquid Control Discharge 63-526 Standby Liquid Control Discharge 64-17 Drywell and Suppression Chamber Air Purge Inlet 64-18 Drywell Air Purge Inlet 64-19 Suppression Chamber Air Purge Inlet 64-20 Suppression Chamber vacuum Relief 64-c.v.

Suppression Chamber Vacuum Relief 64-21 Suppression Chamber Vacuum Relief 64-c.v.

Suppression Chamber Vacuum Relief 64-29 Drywell Main Exhaust 6t.-30 Drywell Main Exhaust 64-32 Suppression Chamber Main Exhaust 64-33 Suppression Chamber Main Exhaust 64-31 Drywell exhau:,t to Standby Cas Treatment 64-34

. Suppression Chamber to Standby Cas Treatment 64-139 Drywell pressurization, Compressor Suction 64-140 Drywell pressurization,. Compressor Discharge 68-508 CRD to RC Pump Seals 6S-523 CRD to RC Pump Seals68-550 CRD to RC Pump Seals68-555 CRD to RC Pump Seals 258 Amendment l'o. 85

TAB' E 3.7.D (Continued)

• c, v e Valve Identification e...

RWCU Supply ei_t RWCU Supply p,_h g RWCU Return 7._;

RCIC Steam Supply

-._3 RCIC Steam Supply l

7._

RCIC Pump Discharge 7. _3 RCIC Pump Discharge 4?_?

RCIC Steam Supply hl_[:

RCIC Steam Supply HPCI Pump Discharge i_g H_d HPCI Pump Discharge HPCI Steam Supply Bypass f, _,-

RHR Shutduwn Suction 7.: _ t 3 RHR Shutdown Suction

.. 633 RHR Shutdown Suction

. ;.; _3 52 RHR Shutriown Suction 3, _., -

Drywell/ Suppression Chamber Nitrogen Purge 76-16 Drywell Nitrogen Purge Inlet 7&_39 Suppression Chamber Purge Inlet 75 7; Drywell/ Suppression Chamber Hitrogen Purge 76-49 Containment Atmospheric Monitor 73_59 Containment Atmospheric Monitor 76 53 Containment Atmospheric Monitor

-' 6-5 2 containment Atmospheric Monitor.

76-53 Containment Atmospheric Monitor 7 6-5:;

Containment Atmospheric Monitor 76 g Containment Atmospheric Monitor 6-[6 Containment Atmospheric Monitor 76-57 Containment Atmospheric Monitor 76-55 Containment Atmospheric' Monitor 76-59 Containment Atmospheric Monitor 76-53 Containment Atmospheric Monitor 76-61 Containment Atmospheric Monitor 76-62 Containment Atmospheric Monitor 75-33 Containment Atmospheric Monitor -

u_na Containment Atmospheric Monitor

,. g _ g Containment Atmospheric Monitor 3 36 Containment Atmospneric Monitor

. _i7 Containment Atmospheric Monitor

- 3_..s Containment Atmospheric Monitor I

7_;A Drywell Floordrain Sump

~7_;3 Drywell Floordrain Sump 77 35A Drywell Equipment Drain Sump 77-153 Drywell Equipment Drain Sump-

- p _3 A Containment Atmospheric Dilution p_c3 Containment Atmospheric Dilutien

.p _g e Containment Atmospheric Dilution g _3 -)

Centainment Atmospheric Dilution l

c :;_39 Containment Atmospheric Dilution

g. 23 Main Exhaust to Standby Gas Treatment 9 5p Main Exhaust to Standby Gas Treatment p_gg)

Main Exhaust to Standby Cas Treatment 3,_ g 3 Main Exhaust to Standby Cas Treatment 9,g3 Main Exhaust to Standby Cas Treatment e, c -.. -

CHD Hydraulic Return y'_}IA

• d

• 259 e

i be n e.e ~ ' '.p. J 2. P.5

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TABLE 3 7.D (Continued)

".1ve Valve Identification 90 2343 Radiation Monitor, Discharge 90.:25 Radiation Monitor Discharge 90 257 A 3.diation Monitor Discharge 00 2573 hadiation Monitor Discharge i -

i:

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260 Ar t -te nt *:0, 27.52.85

This Page Intentionally Lef t Blank 4

i 261 I

1 Y

I TAIII 3 7.E l.

I?.D'ARY CO:CAI;tE:C ISOI.ATIO:I VAIL.'ES WHICH TETCC; ATE D2LCl TIE SU/I?ISSIO:I POOL WATER IrJEL Vcive Identificction Valve 12-738 Au::iliary Boiler to RCIC 12-Th1 Auxilicry Boiler to RCIC h3-20A Rim suppression Chan5er Sc= pie Lines h3 2?.)

R:m Suppression Choter Sa aple Lines h3-2?A R*iR Cupprestion Cha.ber Utmple Lines h3-29D RiiR Suppression Cht:.ber Sa:::ple Lines 2-11h3 Dominerclized Water 71-lh RCIC Turbine Exhaust 71.32 RCIC Vacuu.2 P.:=p Eischorge 71-530 RCIC Turbine Exhaust 71-592 RCIC Vacuum Pu=p Discharge 73-23 HPCI Turbine Exhaust 73-2h HPCI Turbine Exhaust Drein 73-603 HIcI Turbine Exhaust 73-609 HICI Exhaust Drain 7 -722 RHR h

75-57 Core Spray to Auxiliary Boiler 75 Core Spraf to Auxiliary Boiler Core Spray to Auxiliary Boiler j

I L

l l

l 262 l

Amendment No. 85 i

l l

l t

D 4

1 T/GE 3.~.F I?r'JiY CC:7'A2:3:C ISOLf,TIO:7 V/JNES L% ATE III J

TTA3 SEAED SEIS:dC CLASS 1 LI*ES Velve Identification Valve RI:R LTCI Discharge 7h-53 R!m 74-5h rim Suppression Chamber Sprey L

7 -57 74 58 RIE Suppression Chanber Spray 7 -60 RER Drywell Spray i

4 7 -61 RER Drywell Spray l'

h 7h-67 RIS LICI Discharge Th-68 RER LICI Discharge RER Suppression Chamber Spray 74-71 RHR Suppression Chamber Spray "4-72 Th -h R!!R Drywell Spray 7h 75 RHR Drywell spray F

-h-77 RIIR Head Spray

[

i Th-78 PJG Head Spray i

Core Sprav Discharge 75-25

\\

75-2$

Core Spray Discherge Core Spray Discherce 75-53 75 54 Core Spray Discharge j

i l

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Amendment.No. 85 i

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TAELE 3.7.C i

(This table intentionally Icft blank) l t

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264 Amendment l'o. 85

j.

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i 4

U 1

TABLE 3.7.H (Cont'inued)

J X-1073 Spare (testable)

X-108A Power v

X-1083.

CRD Red Pesition Indie.

X-109 9

X-110A Power t

l X-110h CRD Rod Position Indic.

j' X-230' Containment Air Monitoring System 1.

e t:

l X-200A-SC S/RV Test Instrumention (Temporary)

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266 w

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Amendment-No. 85.

A

nst.s

• 3.7.A & 4.7.A P rit-a ry centainmant The integrity of the prinary containment and operation of the core standby cooling j

systen in combination, ensure that the release of radioactive materials from the r

'centainment atmcsphere will be restricted to these leakage paths.and associated leak '

rates accu.ed in the accident analyses. Thts restriction, in conjunction with the leakt:;e rate limitation, vill limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

}

l Durir.; initial core leading 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 i

system thus greatly reducing the changes of a pipe break. The reactor may be taken I

critical during this period; however, restrictive operating procedures will be in t

effect to minimize the probability of an acciden'ti occuring.

The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the italue assumed in the accident analyses j

at the peak accident pressure of 49.6 psig, P. As an added conservatism, the j

a i

measured overall integrated leakage rate is further limited to 0.75 L during a

perfcrmance 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 l

requirements of Appendix J of.10 CFR Part 50 (type A, B, and C tests).

The

• essure su::ression Dool water prevides the heat sink for the react:r primary system energy release following a costulated rupture of the system. The pressure sucpressien chamber water volume must absoro the associated decay and structural sensible heat release; during primary sys te-blowdo-n f rom 1.035 psig. Since all of the gases in the drywell are purged into the pressure suppression chamoer air space during a loss of c::lant at:ident, the pressure resulting from isothermal compression plus the vacer Dressure of the licutd must not eiceed 62 psig, the suportssion chamoer maximum pressure. The design volume of the suppression chaeber (water and air) was ottained by considerinc that the total volume of reactor coolant to be concensed is discharged to the suppression chambe* and that the drywell vclume is purged to the suppression chamber.

Using the mini =um or maximum water levels given in the specification con-tainnent pressure during the design basis accident is approxicately 49 psig, which is below the max 1=u: of 62 psig. ne maxi =u=

vater level indi-cation of -1 inch corresponds to a downconer submergence of _3 feet i

7 in:hes and a water volu=e of 127,800 FT3 with or 128 700 FT3 vithout the drvvell-suporession cha=ber differential pressure control. The mini =um 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 a water volume of a.c.oroximate].y 123.,000 cubic feet. Maintainine the water level between these levels vill assure that the torus water volume and dovn-comer submergence are within the aforementioned limits during normal plant operation. Alarms, adjusted for instrument error, vill notify the operator when the li=its of the torus water level are approached.

The taxi =um per=1ssible bulk pool te:tperature is li=ited by the 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 above approximately 555 psig, the bulk pool te=perature shall not exceed 180oF. At pressures below approxi=ately 240 psig, the bulk te=perature may be as cuch as 1840F. At intemediate, pressures., linear

~

interpolation of the bulk te=perature is per=1tted.

267 j

A endrent "o.

22,12, 85

h 0

3Asts They also represent the bounding upper limits that are used in suppression pool temperature response analyses for safety relief valve discharge and LOCA cases. The actions required by specification 3 7.c-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 pump 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 is no requirement fur 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 condensatiun.

Limiting suppression pool temperature to 1050F during RCIC, HPCI, or relief valve operation when decay heat and stored energy is removed from the primary system by discharging reactor steam directly to the suppression chamber ensures adequate margin for controlled blowdown anytime during RCIC operation and ensures margin for complete condensation of steam from the design basis loss-of-coolant accident.

In addition to the limits on temperature of the suppression l

chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include:

(1) use of all available means to close the valve, (2) initiate suppression l

pool water cooling heat exchangers, (3) initiate reactor l

shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated l

(

form that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.

If a less-of-coolant accident were to occur when the reactor water temperature is below approximately 3300F, the containment pressure will.not exceed the 62 psig code permissible pressures even if no condensation were to occur.

l The maximum allowable pool temperature, whenever the reactor l

is above 212 F, shall be governed by this specification.

0 Thus, specifying water volume-temperature requirements 0

applicable for reactor-water temperature above 212 F i

provides additional margin above that available at 330 F.

268 l

l A endment No. f2, 85

)

In conjunction with the Mark I Centain=ent Short Term Progra=, a plant unique analysis was perfor=ed (" Torus Support System and Attached Piping Analysis for the Browns Terry Nuclear Plant Units 1, 2, and 3 " dated Septe=ber 9,1976 and supplemented October 12, 1976) which demonstrated a factor of safety of at i

least two for the weakest ele =ent in the suppression chamber support system s

I and attached piping. The caintenance of a dryvell-suppression chamber differen-tial pressure of 1.1psid and a suppression chamber water level corresponding to a downce=er submergence range of 3.06 feet to3.58 feet will assure the i

integrity of the suppression chamber when subjected to post-LOCA suppression l

pool hydrodynamic forces.

Inerting The relatively small containment volume inherent in the CE-Bk'R pressure suppres-i tion containment and the large amount of zirconium in the core are such that

[

the occurrence of a very limited (a percent or so) reaction of the zirconium l

and steam during a loss-of-coolant. accident could lead to the liberation of.

[

hydrogen combined with an air at=osphere to result in a firm =able concentracion t

in the containment. If a sufficient amount of hydrogen is generated and oxygen I

is available in stoichiometric quantitiew the subsequent ignition of the hydrogen in rapid recombination rate could lead to failure of the contain=ent to e,sintain a low leakage integrity. The <4% hydrogen concentration minimizes 4he possibilityi of hydrogen combustion following a loss-of-coolant accident.

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269 Amendment No. 26, 42, 85

t e

e BASFS The interior surfaces of the dryvell and suppressien chamber are coa'ted as necessary to provide corrosion protee. tion and to provide a r. ore easily decentaninable surface. The surveillance inspection of the internal surfaces each operating cycle assurcs ts=ely detection of corrosion. Dropping the torus ester level to one foot below the normal operating icvel enables an inspection of the suppressien chambcr where problems would first begin to show.

i j

ss 9

G The primary containment preoperational teat pressures are based upon the calculated primary containment pressure response in the event of a loss-of-coolant accident.

The pank dryw' ell pressure would be about 49 psig which would rapidly reduce to less than 30 psig within 20 seconds following the pipe break.

Following the pipe break, the suppression' chamber pressure rises to 27 psig within 25 seconds, equalizes with

,5 dryvell pressure, and d e c' a y s with the drywell pressure decay.

The design pressure of th6 drywell and suppression chamber is 56 psig.

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

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

Based on the calculated containment pressure response discussed above, the primary containment preoperational test I

pressures were chosen.

Also based on the primary containment pressure response and the fact.that the drywell and suppression chamber function-as a unit, the primary containnent will be tested as a unit-

~

rather than the individual components separately.

The calculated radiological doses given in Section 14.9 of the FSAR were based on an assumed leakage rate of 0.635 percent at the maximum l[

. calculated pressure of 49.6'psig.

The doses eniculated by the NRC using this bases are 0.14 rem, whole body passing cloud samma dose, j

and 15.0 ren, thyroid dose, which are respectively only 5x 10-s and 10-3 times the 10 CFR 100 reference d o s e s '.

Increasing the assumed leakage rate;at 49.6 psig to 2.0 percent as indicated in the i

a factor of 3.

specification's would increase these doses approximately s' t ! ! ! leaving a margin between the calculated dose and the 10 CFR 100 f

reference values.

c Establishing the test limit of 2.0%/ day provide: an adequate i

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

It is further considered that the allowable leak rate should not deviate sig'nificantly 273 Amencent No. 85 s

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0 I.lMi**'47 Cf.';DI*!O M FOk CPERATICN SUitVEI!.LWCE ftEOttifWMEt;TS 4.9 AtlXILIAPT E1.Er.rnICNI. 3L" TEM 3.0 f. n :..: ?, F Y Ft.EC RIC AL r,Y.", TEM vith ibn't.ritettons Inine d ori b.~The units 1 an! 2 4-kV the mariufacturer's

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renore.er.da tienu.

shutdown boards are h

energized,

e. Once r. month a marr.plc of

c. The 480-V nhutdown boards

, diesal fuel.: Mall Ac checked for qunlity..The ossociated with the unit

, '.quali ty shall, be, withi n are energized.

acecptable itnits specified

.d. The units 1 and 2 diesel in Table 1 of the latest revisich to ASTM D'J75 and' t

auxiliary boards are

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'l ogged.

energized.

f.oan of voltage and degraded 2.'O. C. Pnwer Sfotet: L tin i t.

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voltage relays operable on Generator 3.it.t e,ies (12r-Volt)..

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Datteries (O';*9(Volt) Dienel c.

e 4-kV shutdown boards A, D.

t.nd Shut 4ne noned Dat.t.crina C, and fi.

3 250-V olt.)

r. Shutdown busses 1 and 2 I(

energir.ed,

~s. Every week the specLtin

g. The 480V Rx. MOV Boar.is

~

gravi,ty and the soltacc' of the pilot. cell, and v.

D & E are energized with

< temper,ture or an odjacent 7 M-G sets 2DN,2DA,2EN, and cell and r.verall batt.cyy J~'

5 2EA in service.

voltage shall be mea:.9 red s

~ and loE804

5. The 250-volt unit and shutdown board batteries and

b. Every three monthstthe 2-1

• roeasurene.aty shall hed.ade 7 n battery charger for each

'or voltari or-eac1. cell to C

battery boards are operable.

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nenrent 0.1'yolt, spect r'c j

a

^ gravity of t ach, cell, ant >:,

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6. Logic Systems j.terp3raturn of yvery [1 f tb w

. l-cell. T.hese measureraenta==>~

- sha11 be") ogge 1., ' '

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a. Comon accident sigr.at y) lo; tic system 13 cperabic.

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c. A battery' rated dischar.pf[.,

(capacit'y) '*3st shall;De

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perfor#e{ and the volt.$e,). j g

b. 4SC-V load sheddirg lor,1c systen in cperabic.

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time, au:' output, curre..t. ~

. ceanortrw:nta sba11 be n,'

. logged At int.ce'vals not t o - '

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7. There shall be a minir:um of 103,300 gallons er diesel 7pxege.1 P'ipcachs.

fuel in the standby diesel g,

,f r,anerator fuel tanks.

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72, E2, 85 1,'

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y St!RVEILLANC:: PRCUITEMENTS a * < LIMITING' CdNDITICNS FOTOFFRATTON ti f6 7

x 4.9 AtlXILIt.'tY ELECTRICAI. SYSTEM

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T_,7,j.0KQ3 9 ' A'3XIf,I ART ELECTPiral SYSTElj

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demenntrate that the

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,., -. 4 associated diesel generator

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will start.

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c. The loss of voltage and de-(-..w

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.a graded voltage relays whit.h

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~ the diesel generators 7

statt from the 4-kV shutdown boards

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shall be calibrated annually I

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for trip and reset,and the

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measurements logged. These j

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relays shall be.. calibrated as n.

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specified in Table 4.9 A.4.c.

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d. 4-kV shutdown board voltages shall be recorded

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once ever'/ 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4

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480V RMOV' boards D and E e,

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a. ~ Once per operating cycle the f

automatic transfer feature for

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480V RMOV boards D and E shall e

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be functionally tested to verify

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LIMITING CONDITIONS POR 0?ERATION SURYCILLANCE REQUIREMENTS 3.9 AUXILIARY ELECTRICAL SYSTEM 4,9 AUXILIARY ELECTRICAL SYSTEM shutdewn boards and undervoltage relays are operable. (Within the surveillance schedule of 4.9.A.4.b).

12. When ene 4BO-volt shutdown board is found to be inoperable, the reactor will be placed in hot standby within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold ahutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. -

13. Ir one 480-v'RMov board M-G.

' set is inoperable, the reactor may remain in operation fer a period not te exceed seven days, provided the remaining 480-V RMOV board m-g sets and their associated loads remain operable.

14. If any two 480-V RMOV board M-G sets become inoperable, the reactor shall be placed in the cold shutdown condition s.

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

15. If the requirements for operating in the conditions specified by 3 9.B.1 througn l

3.9.B.14 cannot be met, an orderly shutdown shall be initiated and the reactor shall be shutdown and in the cold conditi.on within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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Amendment 'o.

72, 85 l

LIMITING C0!;DITIC!;S FOR OPERATION SURVEILLANCE REQUIF2MENTS 4.9 AUXILIARY ELECTRICAL SYSTEM 3.9. AUXILI ARY ELECTRICAL SYSTEM C. Cperatien in Cold Shutdown Whenever the reactor is in cold shutdown condition with irradiated fuel in the reactor, the availability of electric power shall be as specified in Section 3 9.A except as specified herein.

1. At least two units 1 and 2

~

diesel generators and their associated 4-kV shutdown boards shall be operable.

2. An additional source of power consisting of at least one of the following:

a. The unit 1 or 2 unit station service transformers energized.

b. One 161-kV transmission line and its associated co=on station service transformer energized.

c. Either 161-kV line, one cooling tower transformer and the bus tie board energized and capable of supplyfng power to the units 1 and 2 shutdown boards energized.

I

d. A third operable diesel generator.

3. At least one 4BO-V shutdown board for each unit must be operable.

~

4. One 480-V RMOV board motor-generator (M-G) set is l

required for each RMOV board (D or E) required to support I

operation of the RHR system l

in accordance with 3.5.B.9 29 8 s.

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Amendment f:0. 72, S5

3.9 EASES (con't) control functions, operative power for unit cotor loads, and alternative drive power for a 115-volt a-c unit preferred motor-generator set.

One 250-volt d-c system provides power for coccon plant and transmission system control functions, drive power for a 115-volt a-o plant preferred motor-generator set, and energency drive power for certain unit large motor loads.

The four remaining systems deliver control power to the 4160-volt shutdown' boards.

~

Each 250-Volt d-c shutdown board control power supply can receive power from its own battery, battery charger, or from a spare charger.

The chargers are powered from normal plant auxiliary power or from the standby diesel-driven generator system. Zero resistance short circuits between the control power supply and the shutdown board are cleared by fuses located in the respective control power supply.. Each power supply is located in the reactor building near the shutdown board it supplies. Each battery is located in its own independently ventilated battery room.

The 250-volt d-c system is so arranged, and the batteries sized such, that the loss of any one unit battery will not prevent the safe shutdown and cooldown of all three units in the event of the loss of cffsite pcuer and a design basis accident in any one unit. Loss of control power to any engineered safeguards control circuit is annunciated in the main control room of the unit affected. The loss of one 250-Volt shutdown board battery affects normal control power only for the 4160-Volt shutdown board which is supplies. The station battery supplies loads that are not essential for safe shutdown and cooldown of the nuclear system. This battery was not considered in the accident load calculations.

, There are two 480-V ac Reactor Motor-Operated Valve (RMOV) Boards that contain motor-generator (M-G) sets in their feeder lines. These l

480-V ac RMOV boards have an automatic transfer from their normal to l

alternate power source (480-V ac shutdown boards). The M-G sets act as electrical isolators to prevent a fault from propagating between electrical divisions due to an automatic transfer. The 480-V ac Fy.0V boards involved provide active power to valves associated with the LPCI code of the RER system. Having an M-G set out of service reduces the assurance that full RER (LPCI) capacity will be available when required. Since sufficient equipment is available to maintain the minimum complement required for RHR (LPCI) operation, a 7-day servicing period is justified. Having two M-G sets out of service can considerably reduce equipment availability. Therefore, the affected unit shall be placed in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l k

300 s

k endrent No. 72, 85

e 1

,:tt:TINO COND: :CNS TOR OPLP.ATICN SURVEII, LANCE REQUIR1.MENTS

.p.

s.

+

u.11 TIRE PROTECTION SYSTEMS i

t1' FI RT. _P60T TCT T ON' SYSTEMS 3.

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

u.

The circuits'between

- t.h r loc.al panels in u.11.C. 3 and the main control room will be tested monthly.

5.

Smoke detector sensitivity will be checked in accordanca with manuf acturer 's instruction annually.

D.

Roving Tire Watch D.

Roving Fire Watch e

. ;. roving fire watch will A monthly walk-through by tour. each area in which the Saf ety Engineer will s.

aut<c.itic fire cuppression be made to visually systems are to be inspect the plant fire installed - (as described in protection system for the " Plan for Evaluation, signs of damage, Repair, and Return to deteriora tion, or abnormal

'Se:vice of. Browns Ferry conditions which could Units 1 a n d 2, Section X) jeopardize proper at i nt'<.rvale no granter operation of the syr, tem.

t i ps n 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

A keyclock recording type system chall'be used to monitor the rou.es'of the roving iira weteh.

Th+*

patrol will be di: con.tiriued as l'

the automatic suppression l-

r. s t ens a re installed end

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mhd* Operable for eSch

[.eci! icd arei..

l I-321 r

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I Atenecnt'ilo. 85 i

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e LI"IT.ING CONDIT! OMS FOR OPERATION SCRVEILLANCE REQUIREMENTS 3.11 FIRE PROTECTION SYSTEMS 4.11 FIRE PROTECTION SYSTEMS E.

~ Fire Protection Systems Inspection E.

Fire Protection Systems Inspections All fire barrier penetrations.

Each required fire barrier including 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 icnal at all times. With one following repairs or maintenance 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 watch -

on at least one side of the affected penetration or verify the OPERAEILITY of fire detect-ors on at least one side of the non-functional fire barrier and establish an hourly fire watch patrol until the work is com-pleted and the barrier is re-stored to functional status.

~

-F.

Fire Prorection Organization The minimum in-plant tire protection organization and duties shall be as depicted in Figure 6.3-1.

322 Amendment No. 63, 74, 85

LIMITIIG C0"DITIO"S FOR OPERATION SURVEILLANCE REQUIREMENTS 3 11 FIRE PPOTECTIO': SYSTEMS 4.11 FIRE PROTECTION SYSTEMS G..

Air Masks and Cylinders A minimun of fifteen air masks and thirty 500 cubic 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 A continuous fire watch shall be stationed in the immediate vicinity where work involving open flame welding, or burning is in progress.

I.

Ooen Flames, Welding, and Burnine in the Cable Spreadine Room There shall be no use of open flame, welding, or burning in the cable spreading room unless the reactor is in the cold shutdown condition.

323 Amendment No. 62, 85

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.v a,. o gAJnn ncstcs rrATUREs 5.1 51TEJLATuxts Brovns Terry unit 2 is located at Brevns Ferry Nuclear Plant site on property owned by the United States and in custody of the TVA.

The site shall consist.of approximately 840 acres the north shore of Wheelcr Lake at Tennessee River Mile on 294 in Limestone County, Alabama. The minimum distance from the outside of the secondary containment building to the boundary 'of the exclusion area as defined in 10 CFR 100.3 shall be 4,000 feet.

5.2 REACTOR A.

The reactor. core may contain 764 fuel assemblics consisting of 8x8 assemblies having 63 fuel rods each, and 8x8R -and P8x8R assemblies having 62 fuel rods each.

t B.

The reactor core shall contain 185 cruciform-shaped control rods. The control material shall be boron carbide powder (B C) compacted to approximately 70 percent of' theoretical 4

density.

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5.3 REACTOR VESSEL The rlactor vessel shall be as described in Table 4.2-2 of the FSAR. The applicable design codes shall be as dcscribed.in i

Table 4.2-1 of the FSAR.

5.4 CONTA1NMENT A.

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

B.

The secondary containment shall be an de;ceibed in Section 5.3 of the FSAR.

C.

Penetrations ~ to the primary containment and'pipinr. passing through such penetrations shall be designed in accordance I

with the standards set forth in Section 5.2.3.4 of the TSAR.

5.5 rutt STORAg l

A.

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

gg 4*

330 Amendment No. 35, #,2, 85

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