ML19246A640

Forwards Requested Amend to Licenses DPR-33,DPR-52 & DPR-68, Changing Tech Specs & Providing Justifications for Proposed Changes Re Administration,Clarification & New Organization

ML19246A640
Person / Time
Site:	Browns Ferry
Issue date:	06/29/1979
From:	Gilleland J TENNESSEE VALLEY AUTHORITY
To:	Harold Denton Office of Nuclear Reactor Regulation
References
TVA BFNP TS 123, TVA-BFNP-TS-123, NUDOCS 7907060188
Download: ML19246A640 (67)

Text

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TENNESSEE V ALLEY AUTHORITY

(.it AT f ANOOG A. TENN ESSEE 37401 500C Chestnut Street Tower II TVA BFNP TS 123 JUN D 91979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.',. Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Denton:

In the Matter of the ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 In accordance with the provisions of 10 CFR Part 50.59, we are enclosing 40 copies of a requested amendment tc licenses DPR-33, DPR-52, and DPR-68 to change the technical specification 1 of Browns Ferry Nuclear Plant units 1, 2, and 3 (Enclosure 1). Also enclosel are justifications for the proposed changes (Enclosure 2). The enclosed changes . nsist of administrative changes, clarification chan, es, NRC requested. changes, and changes to reflect new plant organization.

In accordance with the requirements of 1.0 CFR Part P 0.22, we have determined these proposed amendments to be Class III for unit 1 and Class I for units 2 and 3. Thcsa classifications are based on the fac s i ht the proposed amendment involves an issue which does not involve a significant hazard consideration for unit 1, and the proposed amend. Tents for uni' 2 sud 3 are duplicates of the unit 1 proposed n::endment submitted by this lecter. The remittance of $4800

($4000 for unit 1, $800 for unita 2 and 3) is being vired to the NRC, Attention:

Licensing Fee Management Branch.

Very truly yours,

) ,7 ,,

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' J. E. Gilleland Assistant Manager of Power S ubscribed and sworn to bafore ce this ,'[ / day of

/h aC 1979.

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' ' Notary.Public f[

My Commission Expires eh// 298 056 Enclosures 7 9 0 7 0 6 018E -

cc: See page 2 An Equal Oppor tur.ity Employrr

6 M r. Harold R. Denton JUN 0 91979 cc (Enclosures):

Mr. Charles R. Christopher Chairman, Limestone County Cocuission P.0 Box 188 Athens, Alabama 35611 Dr. Ira L. Mycra State Health Officer State Department of Public Health State Office Building Montgomery, Alabama 36104 8 057

ENCLOSURE 1 298 058

s PROPOSED CHANGES TO BROWNS FERRY UNIT 1 TECHNICAL SPECIF ALiTIONS 298 059

TABLE 3.1.A (Ccntir.ed)

M i r, . No.

cf 3

• Mo d e s i n '#r,i c h Fur-tion 7 I4"St 8' 00*'3DI*

Char.nels Per Trip Startup/ Hot Sp t en (1 ) Trip Function Trip Level Setting i Refuel (7) Standby Run Action (II X(3)(6) X(3)(6' X(6) 1. A or 1.C 4 Main Ster Line Isolation 1 10". Val ve Closure -

Valve Closure Turbine Cont. Valve rest upon trip of tne f ast x(4) x(4) x(4) 1. A or 1.0 2

Closure acting soler.oid valves ,

4 Turbine Stop Valve Closure < 101 Valve Closure X(4) X(4) X(4) 1.A or 1.0 2 Turbine Centrol Valve - 1 550 psig X(4) X(4) X(4) 1.A or 1.0 Loss of Control Oil 9 .. m are 2 Turbine First Stage 1 154 psig X(18) X(18) X(18) (19)

Pressure Permissive Turbine Ccodenser Low 1 23 In. Hg, Vacuu m X(3) X(3) X 1.A or 1.C ,

2 Vacuum Main Stear. Line Hign 1 6)( No nc l Ful l p c',,e r X(9) X(9) X(9) 1. A or ! .C 2

adiation (14) Background (20)

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CD CD ON CD (C (C (C

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l 1o. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.

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

12. The AFRM downscale trip is automatically bypassed when the v IRM instrumentation is operable and not high.

13. Less than 14 operable LPRM's will cause a trip system trip.

14 Channel shared by Reactor Protection System and Primary cont a irur.ent and Reactor vessel Isolation control System. A channel failure may be a channel failure in each system.

15. W e APRM 15% scram is bypaesed in the Run Mode.

16. Channel sha red by Heactor Protection System and Reactor Manual Cont rol Syst em (Bod Block Portion) . A channel failure ma y be a cf. ann e l failure in each system.

17. Nct required while performing icw pcuer physics tests at atmospheric pressure du-ing or after refueling at pcwer levels not to exceed 5 MW (t) .

18. Operability is required whe.n ncrmal first. stage pressure is below 30t (< 154 psig}.

19. Action 1.A or 1.0 shall be taken only if the permissive f alls in such a manner to prevent the affected RPS logic from O parf orming its intended function. Othervise, no action is raquired.

20. ?n alarm setting of 3D times norea' backgecund at rated power shall v t>e established to alert the operatcr to abnomal radiation levels in

. prima ry coolant.

21. The APPM High Flux and Inoperative Trips do not have to be operable in the Refuel Mode if the source Range Monitors are connected to give a non-coincidence, High Flux scram, at _< 5 x 105 cps. The SRM's shall be operable per Speci fication 3.10.B.l . The removal of eight (8) shorting links is required to provide non-coincidence high-flux scram protection from the Source Range Monitors.

22. T he three required IPM's per trip channel is not required in the Shutdown or Refuel Podes if at least four IRM's (one in each core quadrant) are connected to give a non-coincidence, High Flux scram.

The recoval of four (4) shorting links is required to provide non-coincidence high-flux scran protection from the IRM's.

O A

36 v

298 . .,

TABI.E 4.1.A RZACTOR FROTEC ICR STSTD( (SCRAM) INSTIQCGTICM FUNCTIONA1. TESTS HINIE"M FUNCTICRA1. TEST FREQUENCIES FOR SAFEU INSTR. A'm CONTROL CIRCUITS Croue (2) Functional Test Minimum Frequency (3)

Place Mode Switch in Shutdown Each Refueling Outage Mode Switch in Shutdown A Manual Scram A Trip Channel and Alaru Every 3 Months .

IRM C Trip Channel and Alarm (4) Once Per Week During Refueling High Flut and Before Each Startup not to exceed once per week C Trip Channel and Alarm (4) Once Per Week During Refueling Inoperative and Before Each Startup not to exceed once per week N

Trip Output Reisys (4) Before Each Startup and Weekl U Iligh Flux (ISI sc % C OvA When Required to be ')perable High Flux B Trip Output Relays (4) Once/ Week Inoperative o B Trip Output Relays (4) Once/ Week ON Downacale N B Trip Output Relays (4) Once/ Week g,q Flow Bias B (6) (6)

N6 Trip Channel and Alarm Once/ Month (1) mm High Peactor Pressure A

.I High Dryvell Pressure A Trip Channel and Alarm Once/ Month (1) czem Trip Channel and Alarm Once/ Month (1)

DJg Reactor Low Water Level (5) A W1:.Tua Every 3 Montha High Water Level in Scram Discharge Tank A Trip Channel and Alarm Cy;q CXJ ~~ Once/ Month (1)

Turbine Condenser Low Vacuum A Trip Channel and Alarm Main Steam Line Ifig5 Rsdiation B Trip Channel and Alarn (4) Once/Veek

"."_ J.N. f.'l1.AE d.Id dl

1. For the startup and run p *tiona of the 1(cac to r Mode Selector Switcle, there shall be two operat a s tripped trip syste== for each function. "

The $RH, IM. and arm ( .% t a r t u p re d e ) , blecka need not be operabler in

• bn" eso d e , and the APM (Flow biased) and RD4 rod blocks need riot be operable in "Startup" rode. If the first column cannot be taet for one of the two trip systems, this condition may exist for up to seven days provided that durina that time the operable system is functionally tested t e ndiately and daily thereafter; if this condition last lon2er than seven days, the system with the inoperable channel eMll t>e tripped.

If the fir s t colu2n c anno t be r.e t,f or both trip systes *, loth trip systen shall be t r ip p i.d .

2. W is the recirculation loop flow in percent of desir,n. Trip 1cvel settine is in percent of rated power (3293 in t). A ratio of FRP/CMFLPD <1.0 it, pemitted at reduced power. See Specific'atica 2.1 for APR.1 control rod block setpoint.

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1.

4 If refueling zone setoridary contatnment cannot be natntained J the following conditions shall be met:

a. Iland11ng of spent fuel and all operations over spent fuel pools and open reac-tor wells containing fuel shall be prohibited,

b. The standby 2as treatuent system suction to the re-fueling zone vill be blocked except for a con-trolled leakage area sized to assure the schieving of a vacuum of at least 1/4-inch of water and not over ,

3 inches of water in all three reactor zones.

Primary Conrainment Inolation Valves D. Primary Contair. ment Isolation Valves

1. During reactor power operation, 1. The prinary containment isola- ed all isolation valves listed in tion valves surveillance shall '

Table 1.7.A and all reactor be performed as follows:

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coolant system instrument line flow check valves shall be a. At least once per operating operable except as specified cycle the operable isols-in 3.7.D.2. tion valves t ha t are power operated and auto-matically initiated shall te teated for simulated automatic initiation and closure tires.

b. J.t least on'e per quarter:

(1) All normally open power operated isolation valves (except for the main sesam line power-operated isolation valves) shall be fully closed and recpened except where specific written relief from ASME Section XI requirements has been granted by NRC pursuant to 10 CFR 50, 242 Section 50.55a(g)(b)(1). .

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LIMITING C0tnIT10NS FOR OPERATION SURVEILLANCE REQUIREMENTS .

3.7 (ONTAINMENT SYSTDis 4.7 CONTAINMENT SYSTEMS .

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G. C on t a '.nm e n t Atmascnere G. Containment Atmosphere Dilution System (CAD) OTlution System (CAD)

1. The Containment Atmosphere Dilution (CAD) Systen shall be 1. sucem operability cperable ,ith:

a. At least acce per month. cycle each solenoid

a. Two indeperadent operated air / nitrogen valve with its hand syStens Capable switch (no containment isolation override) through at least one complete cycle of full Of Supplying travel and verify that each manual valve nitrogen to the in the flow path is open.

drywell and torus. b. Verify that the CAD System contains a minimum supply of 2500 gais, or liquid nitrogen twice

b. A rni n i num supply of 2500 gallon 3 c. At each cold shutdown verify that ex h solenoid of liquid operated air / nitrogen valve in ene flow patn can be opened with a containment isolation nitrogen per signal present (containment isalation override).

system.

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2. The C o n t a i n;n e n t Atmosphere Dilu ti on (CA D) Systen shall be ope ra bl e '.shenev er the reactor rode switch is in the "RUN" position.

3. If one systen is ino pe ra b l e , the reactor may remain in operation for a period of 30 days -

provided all active CoJ) p C n e n t s in the other systen are operable. ,

4 I f S.ceci f icati on

3. 7 . G .1 and 3.7.G.2, or 3.7.G.3 cannot be ret, an j orderly shutdcwi chall b- .nitiated l and tne r" actor snall .

be in the Cold .

Shu tf :wn condition within 2; neurs.

5. Primary containrent pressure 248 shall be lirited to a maxirum of 30 psig during 298 06/;

repressurization following a loss of coolant accident. ,

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TABLE 3,7. A (Continued)

Nurber of Power Maximum Action en Operated Valves Operating Nomal Initiating Group Valve Identi fication inboard Outboard Time (sec.) Position Signal 6 Suppression Charrber purge inlet 1 100 C SC (FCV-64-19) 6 Drywell/ Suppression Chamber nitro- -

gen purge inlet (FCV-76-17) l 10 C SC 6 Drywell Exhaust Valve Bypass to Standty Gas Treatrnent System 1 10 C SC (FCV-64-31) 6 Suppression Chamber Exhaust Valve Bypass to Standby Gas Treatment 10 C SC System (FCV-64-34) 1 m

7 RCIC Steamlir.a Drain (FCV-71-6A, 68) 2 5 0 GC 7 RCIC Condensate Pump Drain

( FCV-71-7A, 7B) 2 5 C SC E

] 7 HPCI Hotwell pump discharge isola-tion valves (FCV-73-17A,178) 2 5 C SC HPCI steamline drain (FCV-73-64,68) 0 GC fg 7 2 5 mm

(, 8 TIP Guide Tubes (5) 1 per guide NA C GC com 3,w;na; N db W Ep .:- O gDM;.,

f'.:LW L C3 Ch W s

( ) ( ) ( )

l.IMITING CON!)tTIONS FOR OPERATION SURVEllt.ANCE REQUIRDENTS 3.10 A Refueling Interincks 4.10.A Refueling Interlocks refueling interlocks shall be operable.

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b. A sufficient number of control rodo shall be operable so that the core can be made sub-critical with the strongest operable con-trol rod fully with-drawn and all other operable control rods fully inserted or all directional control valves for remaining control rods shall be disarwed electrically and aufficicnt margin to criticality shall be demonstrated.

c. If naintenance is to be perferned on two control rod drives they must be separated by more than two control cells in any J1rection, y d. An appropriate nte.ber of SiUt's are available as defined in specifi-cation 3.10.A.

6. Any number of control rods may be withdrawn or removed 3. With the mode selector switch in from the reactor core pro _ the refuel or shutdown mode, no viding the following condi. Control rod may be bypassed until tiono are satisfied: tw; licensed operators have confirred

s. The reactor mode switch that either all fuel has been removed is loc'acd in the "re-from around that rod or that all fuel" position. The Control rods in imediately 6dj3Cer; i refueling interlock cells have been fully inserted and which prevents tnore than electrically disarmed, one control rod f ro:a 298 070 r J v

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V s.o MA. ton nestcN n AtuREs 5.1 S iT F. f_L A Tlik l.S Browns Ferry unit 1 is located at Browns 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 on the north shore of tInecler Lake a; Tennessec River Mile 294 in Limestone County, Alabama. ihe minimum distance from the outside of the seccndary containment building to the boundary of the exclusion area as defined in 10 CFR 100.3 sh ll be 4.000 feet.

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'5.2 REACTOR A. The reactor core may contain 764 fuel assemblies consisting of 7x7 assemblies having 49 fuel rods each, 8x8 assemblies having 63 fuel rods each, and 8x8 R assemblies having 62 fuel rods each. The number cf each type in the core is given in the most recent reload amendment topical report.

B.

Tne reactor core shall contain 185 cruciform-shared control rods. The control material shall be boron carbide powder (B4 C) compacted to approximately 70 percent of theoretical density.

5.3 REACTOR VESSEL V

The reactor vessel shall be as described in Table 4.2-2 of the FSAR. The applicable design codes shall be as described in Table 4.2-1 of the FSAR.

5.4 CONTAIN9LNT A. The principal design parameters for the primary containment shall be as given in Table 5.2-1 of the FSAR. The applicable desi;;n codes shall be as described in Section 5.2 of the FSA.R.

B. The secondary containment shall be an described in Section 5.3 of the FSAR.

C. Penetr1tions to the primary containment and piping psJsing through such penetrations shall be designed in accordance with the standards set forth in Section 5.2.3.4 of the FSAR.

5.5 f tJ E L S Tn RAG E A. The arrangement of fuel in the new-fuel storaRe facility shall be such that k for dry conditions, is less than 0.90 and flooded i s T e s,7 s.

than 0.95 (Section 10.2 of FSAR).

298 071 b

'..'8 Al8til H l !!T!@'!. [ Y t. .CistJTJ,'f)l,y A summary of the more significant discussions and conclusions of the NSRB will be transmitted along with the final minutes.

9 7 Charter A writtan charter delineatir'q the es ta blis hme nt ,

composition, and mission of the NSHB and the dissemination of NSRB minutes and reports shall be maintained; this may be amended as required. The charter shall identif y the responsibility and authority of the NSBB in conducting reviews, including responsibility to identify problems and to recommend solutions to the Manager of Power.

B. Plant Operations Review Committee (PORC)

1. Mem be r s hip The PORC shall consist of the plant superintendent, electrical maintenance supervisor, mechanical maintenance supervisor, instrument maintenance supervisor, health physics supervisor, operations supervisor, results supervisor, and QA staff supervisor. An assistant plant supervisor may serve as an alternate cocmittee member when his supervisor is absent.

The plant superintendent will serve as chairman of the PORC. The assistant plant superintendent will serve as chairman in the absence of the plant V superintendent.

.: . Meetina Frequency Ti.e PORC shall meet at regular r.onthly intervals and for special meetings as called by the chairman or as requested by individual members.

3. Quorum Superintendent or assista.it superintendent, plus ffiveofthe seven other members, will constitute a quorun.

or their A memper alternate, will be considered present it he is in telephone communication with the committee.

335 298 072 P

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(, , q Argpt t3ISTPATIVE CONTROLS plant, the applicable codes required f atigue .

usage evaluation for the reactor pressure '~'

vessel only. The locations to be monitored shall be:

1. The feedwater nozzles

2. The shell at or near the waterline

3. The flange studs

b. Recording, Ev a lu a t i ng , and Re porting (1) Transients that occur during plant operations will he reviewed and a cumulative f atigue usage factor determined.

(2) For transients which are more severe than the transients evaluated in the stress report, code fatigue usage calculat ions will be made and tabulated s e pa r a t e ly.

i (3) In the monthly Operating Report, the fatigue usage factor determined for the transients defined in (1) an d (2) a bove shall be added and a cumulative fatigue usage f actor to date shall be listed.

When the cumulative usage factor reaches a va lue of 1.0, an inservice inspection shall be included for the specific '~'

location at the next scheduled inspection (3-1/3-year interval) period and 3-1/3-year intervals thereafter, and a subs equ ent evaluation performed in accordance with the rules of ASME Section XI Code if any flaw indications are detected. The results of the evaluation shall be submitted in a Special Report (Section 6. 7. 3) for review by the Commission.

B. Except where cevered by applicable 2 egulations, items 1 through 8 above shall be retained for a period of at least 5 years and items 9 through 17 shall be retained for the life of the pla nt. A cc.nplete inventory of radioactive materials in possession shall be maintained current at all times.

348 298 073 A Q 6%..,9 ka m- r - ,

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6.0 ADMINISTRATIVE CONTROLS (b). Annual Report (1) A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mren/yr and their associated man-rem exposure according to work and j ob functions, 3.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignment to various duty function:,

may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

(2) A report of facility changes, tests or experiments required pursuant to 10CFR50.59(b).

c. Monthly Operating Report. Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Office of Inspection and Enforce-ment, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to the appropriate Regional Of fice, to be submitted no later than the tenth of each month following the calendar month covered by the report. A narrative summary of operating experience shall be submitted in the above schedule.

2. Reportable Occurrences Reportable occurrences, including corrective actions and measures to prevent reoccurrence, shall be reported to the ITC. Supple-mental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.

351 298 074

6. 0 ADMINISTRATIVE CONTROLS B. Source Tests Results of required leak tests performed on sources v if the tests reveal the presence of 0.005 microcurie or more of removable contamination.

C. Soecial R e co rt s (in writing to the Director of Regional of fice of Inspection and En forcement) .

1. Reports on the folicwing areas shall be submitted as noted:

a. Socondary Containment 4.7.C Within 90 Leak It a t e Testing (5) days of completion of each test.

b. Tatique Usage 6.6 monthly Evaluaticn Operating Report

c. Seismic Instrumentation 3.2.J.3 Within 10 days i n ope ra bi li ty after 30 days of inoperability dv 356 d"

298 075

6.0 ApMINISTRATIVE CONTROLS f00TNOTES b A single submittal may be made for a multiple unit station.

1.

The subnittal should combine those sections that are common to all units at the station.

2. The tern " forced reduction in power" is normally defined in the elect ric power industry as the occurrence of a component f ailure or other condition which requires that the load on the unit be reduced for cor rective action immediately or up to and including the very next weekend. Note that routine preventive maintenance, s ur ve il la nce, and calibration activities requiring power reductions are not covered by this section.

3. The term " forced outage" is normally defined in the ele ctric power industry as the occurrence of a component failure or other condition which requires that the unit be removed from service f or corrective action immediately or up to and including the very next weekend.

4 This tabulation supplements the requirements of 520.407 of 10 CFR Part 20.

5. Each integrated leak rate test of the secondary containment shall be the subject of a sunmary technical report. This report should include data on the wind speed, wind direction, p concurrent reactor building pressure, and emergency ventilation ficw rate. The report shall also include analyses and

%' interpretations of those data which demonstrate compliance with the specified leak rate limits.

357 o

298 076 v

PLANT QUALITY ASSURANCE SUPERINTENDENT MANAGER

{

SUPERVISOR QUALITY ASSURANCE f '

DIVISION OF STAFF MIROMM DIVISION OF ASSISTANT MEDICAL SERVICES PLANT QUALITY PLAN'iING ASSURANCE SUPERINTENDE'fr -

COORDINATOR I QUALITY ASSURANCE 8

6 g ENGINEERS 8 8 l l

- I HEALTH t ELECTRICAL MECHANICAL INSTRUMENT RF.SULTS OMATIONS PHYSICIST NURS F*

SUPERVISOR MAINTENANCE MAINTENANCE MAINTENANCE SUPERVISOR SUPERVISOR SUPERVISOR SUPERVISOR i I i ASSISTANT , ASSISTANT j HEALTH / ASSISTANT ASSISTANT ASSISTANT ASSISTANT

' INSTRUMENT INSTRUMENT PHYSICS RESl'LTS OPERATIONS ELECTRICAL MECHA'iICAL MAINTENANCE MAINTENANCE TECHNICIANS SUPERVISOR SUPERVISOP, MAINTENANCE MAINTENANCE

UPERVISOR SUPERVLSOR SUPERVISOR SUPERVISOR (INST) (COMP)

SHIFT TECHNICAL TECHNICAL TECHNICAL TECHNICAL TECHNICAL ENGINEERS ENGINEERS ENGINEERS OPERATING ENGINEERS ENGINEERS A'!D CRAFT AND CRAFT AND CRAFT b g , PERSONNEL FOREMAN FOREMAN voRFuAN AND CRAFT FOREMAN 8

w CO CD 3ROWN5 FERRY HUCLEAR PLANT wJ IIN AL 5AFETY ANALY315 REPORT N

RTNCTIONAL ORGANIZATION FICURE 6.1-2  ;

l l

l

h PROPOSED CHASGES TO BROWSS I W ustT 2 TECHSICAL SPECIFICATIONS 298 078

TAC:1 3.1.- (Cent n -: j )

M i r. . i;o .

Of C m able h' des in Wr i cn Fur: t t ar.

I r.$ t Cha,[ls Must Be Coerable Per Trip Startup/ Hot 5 vite" (l )_ _

Trip function Trip Leve' Sottirc Pafcel(7) _Stardby Run Action (l) 4 Mc;n St: 3 Line Isolation < 10". Val.s closurc X(1)(6) X(3)(6' X(5) 1.A or 1.C Valve Clot;re -

2 Turbine C mt. Valve r:st Upon tric of the fest X(4) X ( -: ) x(4)  !.A or 1.D Closure acting soler.oic valses 4 Turbine Stop Valve Clorure _ 10 Valse Closure

< X(4) X(4) X(4) 1.A or 1.0 2 Turbine Centrol Valse - > 550 psig X(4) )(4) X(4) 1.A or 1.D Loss of Controi 011

'g T. ,> re 2 Turbine First Stage i 154 psig X(18) X(IS) X(18) (19)

Pressure Permissive

.' Turbioc C; eenser Lcw  ; 23 In. H;, Vac s X(3) X(2) X 1.A or 1.C ,

Vacuu"I 2 sain Stear. Line Hign 1 M Nor-a! Fu'l F:;.er X(9) X(9) X(9) 1. A or ! .C

a11ation (14) Background /^0) h

?

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- w /

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hpie';&l%l i>

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tr .y

10. Not required to be operable when the reactor pressure vessel head is not bolted to th e vessel. O 11 The APRM downscale trip function is only active when the reactor node switch is in run.

12. The AFFli downscale trip is automatically tfpassed when the v IBM instrurienta tion is operable and not high.

13. Less than 14 oper able LPPM's will cause .a trip system trip.

14 Channel sha re d by Reactor Protection Systen and P rima ry Containment and Reactor Vessel Isolation Control System. A channel failure may be a channel failure in e ach sy s t em.

15. 'Di e AP R M 15% scram is bypassed in the Run Mode.

16. Channel shared by IMactor Protection Systen and Reactor Ma nu al Ocntrol Syoten (Rod Block Portion). A channel failure may be a channel failure in each system.

17 Uct required while perforning low power physics tests at atmcapheric press"re during cr after refueling at power levels not to exceed 5 Mw (t) .

18. C erability is required when normal first-stage pressure is below 3Ct (< 154 psig}.

19. Action 1.A or 1.0 shall be taken only if the pernissive f itil s in such a manner to prevent the affected PPS logic frcm performing its intended function. Otherwise, no action is raquired.

An a l a r:n setting of 3.0 ti es nor a', background at ratea power shall I 20. t' astablished to alert the c;er3tcr to abnomal radiation lenis in v primary coolant.

21. The APPM High Flux and Incperative Trips do not have to be operable in the Refuel "oje if the source Range Monitors are connected to g've a non-coincidenc e, High Fl ux scram, at < 5 x 105 cos. The SRM's shall be operable per Spmification 3 .10 . B .1. The removal of eight (8) shorting links is rc';uired to provide non-coincidence high-flux scram protection from the Source Range Monitors.

22. llo tFree required IF"'s per trip channel is not required in the Shu tda.in or Pe f uel 'edes i f a t l eas t four IR"'s (one in each core qwjra n t ) a re c onnec t % te give a ron-coincidence, High Flux scram.

The re nal of four (4) shorting links is req; ired to provide non-cuincidente high-flux scra protection frcm the IRM's.

298 080 m

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1. For t he e t a r t..p and run positiona of the henctor Mende Selector Switch, there plull be two operable or tripped trip systeme for each functinn. '

The SRN, IM, and APR?t (Startup vede), blocks need not be op e r a ble, in

• • Run" pod e , a nd t h e APM (l'l ow b i a s e d ) and REM rod blocks need not be operable i n "S t a r t up" arid e . If the first column cannot be m t for one of the two trip systema, this condition mey exist for un to savan days provided that during that time the operable system is functionally tested i mediately and daily thereafter; if this condition laat lonxer than seven days, the system with the inoperable channal shall be tripped.

If the first colu n cannot be rw t,f or both trip syst ens, loth trip sy s t e rve sh.all be tripp=d.

2. W is the recirculation loop flow in percent of desir,n. Trip Icvel settine is in percent of rated power (3293 1R;t). A ratio of FRP/C?!FLPD <l.01:, peniitted at reduced power. See Specific'ation 2.1 f or APM c ont rol rod block set point.

3. . IRM down s cal e i s typa s s +d wh en i t i s en it s I cu e s t r an g e .

4. SRS1's A and C downscale function is bypass when IR>f's A C , E, and G are above range 2. SPJi's B and D downscale function is by-passd when IPli's B, D, F, and II are above range 2.

SFlf detector are not in startup position is bypassed when the count rate is ;e.100 CPS or the above condition is satisfied.

5. One i n s t ru:w n t chonnel; i.e., one APRM or 1RX c.- RSM, per trip systn may be bypa s sed ex cept only one of four SAN rey be bypasand.

6. IR.H cha nnels A, E , C , C a ll in ra n g e 8 1 f7 s u e s SM chic c o lo A f C functions.

I R.H c ha n n e l s B , F. D, H all in range 8 bypasses SM channolo 3 6 D functions. ,

7. The following operational restraint s apply to the RBt cnly.

n.  !!oth FBI channels ~ are bypassel when reactor power is f 30%.

b. The RD1 need not te operable in the "startup" position of the reacter nede selector switch.

c. ".v o R ht channels are provided and cnly one of these may be typassed from the console. An Rh'1 channel nay be out of service for testing and/or maintennnce provided this conditien does not last icnger thsn 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty day period.

d. 11 minimu.:n cend;tions for Table 3.2.C are not net, administrative controls, shall be imediately imposed to prevent centrol rod ]h 08u3 withdrawal.

i t' :h.

L U Vd OffiUiDhdl

( ( ()

IA3tE 3.2.E INSTRDIhTATION THA"" F.0:iLT nS LEAYA0" INTO DRTJLL System (2) Setcointa Action Re= arks ,

Equipment Drain (1) 1. Used to determine identifiable reactor Flow Integrator N/A Coolant leakage.

Sump Fill Rate 2. Conside. red part of sump system.

Ti=er 120.1 min.

Sump Puop Out Rate Ticer 113 4 Cin -

r# Ploor Drain (1) 1. Used to detersine unidentifiable

+

m4Q y Flou Integrator N/A reactor coolant leakage.

[(. 75 ,,)

Sunp Fill Rate 2. Considered part of su=p systes.

D- Timer 180.4 min.

E;;..* :7q Z Sucp Pucp Out EI4 28 Rate Timer N

<8.9 min.

A W (C'

.J CD Drywell Air Sampling Gas and .i x Average (3) f;' ;] ';

Particulate Background F; o h ,/

i CD

!) u b  ; HOTES:

i (1) 1."Senever a syctem is required te s e operable, there shall be one operable systc= either autosstic or canui, l or the action required in Section 3.6.C.2 shall be taken.

(2) An alternate system t o d e t e r.a i n a the leakige flou is a unual system ubereby the time between cu=p pu=p starts la =onitored. The time interval vill determine the leakage flov because the vole =a of the surp vill be knoun.

(3) '& x. r e. e t;

. ci alar , imeli st e c_tf- u 11 be taken to confirn the alat, and a sess the possibility of increas.J 1 r n a v, e . Refer to Tech - ' Sections 3.6.C .2 and 3.6.C . 3 f o r d rvwel l ,>ir sampling systen out of service.

T Ip.r. rownIT INNS FGit n P L ft.AT ION .4U AV M1 LL.* NC hA' E ;U I A .0 i2M i5

7. Secondsty Coa.afnment . 4.7.C Secondary Containesnt 4 If reiueling zone secondary con t a inmen t cannot be maintained the following conditions shall

be met:

a. llandline, of apent fuel and all operations over spent fuel pools and open reat-ter wells containing fuel shall be prohibited.

b. The standby ass treatuent system suction to the re-fvelin- zone will be blocked except for a con-trolled leakage area sized to assure the achieving of a v4cuum of at least 1/4-inch of water and not over ,

3 inches of water in all three reactor zones.

Prienry Cnnrainment Inolation Valves D. Primary Containment Isolation V31ves

1. Durine, reactor power operation, 1. The primary containment isols- -

all isolation valves listed in tion volves surveillance shall Table 1.7.A and all reactor be performed as follows: '

coolant eystem instrument line

%/

flow check valves shall be a. At lerat once per opersting operable except as apecified cycle the operable isola-in 3.7,0.2.

tion valvas t ha t are power operated and auto-metically initiated shall Ic tested (nr simulated automatic initiation and closure cines.

b. 1. t least once per quarter:

(1) All no rmally ope, power opereted isolation valves (except for the main stsam line power-operated isolation valves) shall be fullf closed and reopenen except where specific written relief from ASME Section XI require-ments has been granted by NRC

]hk 242 pursuant to 10 CFR 50, Section 50.55a(g)(b)(1) .

x

(

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[

LIMITING CONDITIONS FOR OPLRATION SURVEILLANCE REQUIREMENTS .

3.7 (O NTA ltN E".T S IST DIS 4.7 cot.TAItalEriT SYSTEMS '

j G. C on t a nm en t _ A : rra s pn a re G.

Diluticn System (CA D)

Containment Atrosenere 6Tlution System (CAOi 1 The Containrent Atmosphere Dilution (CAD) System shall be

1. system e r ,rability cperable with:

a. At least once per aanth, cycle each solenoid

a. Two independent operated atr/nttrogea valve with its hand systems capable =viten (r.o contatnment tseiation e erride)

'h' "4h l' "* C *Pl CYCI* ' f"11 of supplying travel and verify that eacn manual valve ni t r og en to the in the flow path is open.

. drvsell and torus, b. Verify that the CAD System contains a u.inimum supply of 2500 gals. of liquid nitrogen twice

b. A mininum supply of 2500 gallon 3 c. At eacn cold shu h ve r H y t.u t eA a solenoid Of liquid operated air / nitrogen valva ta Os 'cw path nitroqen per can be opened with a containmer.t is i tion sySt m. 81&n81 Pr'5ent (c ctainment isolata a override).

2. The Co n t a i nm en t -

Atmosphere Dilu ti on *

(CA D) System shall be o pe r a t:1 e 'sh ene v er the reactor rode switch is in the "RUN" posi tion.

3. If one system is inope ra b le, the reactor may remain in operation for a period of 30 days -

provided all active components in the other sy9 tem a re operable.

14 . I f Speci ficat. ion

3. 7 . G.1 and 3.7.G.2, or 3.7.G.3 cannot te ret, an orderly shutdevn l sh 11 t; . initiated and tna r"icter snall

• be in the Cold .

Shutfewn cardition within 2; neurs.

5. Prinary contairrent pressure 248 snall be li 1ted to a raximum of 30 psig during repressurization following .

a loss of coolant accident. {,

, , .* n ..,f...

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v b O ',;J .) G S W Vg , - , , . - - -

_, - --,-.e--

TA31.E 3. 7. A P 7. NJ.2Y CO W A. BET I S 01.C IO N V.al.*T ES Nu=ber of Power v.axi t:2 Action en

')

_ era:cd Valves 0;tra:ing Nc: sal lattisting Croup , valve identistc3 .33 yng rd cu t u = r d Tir e (sec.) Position Siznal 1

%in steanline isoletten valves 4 4 3<;<5 0 GC (FCV-1-14,26,37,A5s ,1-15, 27, 3 8, & 52) 1 Main sten 11ae drain isolation 1 1 15 C SC valves TCV-1-55 & 1-56 1

Reactor Vater saeple lin e isole- 1 1 5 C SC

, tion valvem

~,

o 2

~*

R' IRS shutdown cooling supply isolstion valves FCV-74-48 & 47 bN 1 1 40 C SC 2 PJIXS - LPC1 to reactor FCV-74-53, 67 C"%' "' 2 30 .C SC gG "i 2 Reactor vessel nead spray isola- .

tion valves PC7-74-77, 78

'M/

1 1 30 C SC N 2 P.liRS flush and drain vent to e suppreastot. ch2d er 4 20 C SC g Q TCV-74-102, 103, 119, & 120 n

2 Supperaaion Cit.mber Drain 2 15 C sc gg FCV-76-57, 58 n 22);,.,, CD 2 Drysall equip.:.eot drain disch.arge W

1:r2:q; isolstion valves FCV-77-15A, & 153

a. ,

C w;*.1m w 2 15 0  %

L. u w 2 Dry.; ell floor drain discharge U ' ac 28 lualation valves FCV-77-2A & 2B 2 15 0 CC

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0 9

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cl ao I

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TABLE 3.7. A (Continued)

NuMer of Pcwer Maxtrum Action en Opera ted Valves Operating Nomal Initiating Groua Valve Identi fic3 tion Inboard Ou tb oa rd Tine (sec.) Position Sicnal 6 Suppression Chanter purge inlet (FCV-64-19) 1 100 C SC 6 Drywell/ Suppression Chanter nitro- -

gen purge inlet ( FCV 17) 1 10 C SC 6 Drwell Exhaust Valve Bypass to Standby Gas Treatment System (FCV-64-31) 1 10 C SC 6 Suppression Chanter Exhaust Valve Bypass to Standby Gas Treatment g Systen(FCV-64-34) 1 10 C SC C- W CN #

2 7 RCIC Steamline Drain (FCV-71-6A, 68) 2 5 0 GC

~ b"

^

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

% 7 HPCI Hotwell pump discharge isola-OW tion valves (FCV-73-17A,170) 2 5 C SC gbh 7 HPCI steamline drain (FCV-734A,68) 2 5 0 GC gec- ni 8 TIP Guide Tubes (5) I per guide NA C GC EEI?IJ tube L"-my

-N Q d2iD  %

72a

( ) ( ) ( ) .

1.IHitING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRIM NTS 3.10.A Refuelt_ng Interl.eks 4.10.A R_efueling Interlocks refueling interlocks shall be operabla,

b. A sufficient number of control rodo shall be operable so that the core can be made sub-critical with the strongest operable con-trol rod fully with-drawn and all other operable control rods fully inserted, or all directional control valves for remaining control roda nhall be disarmed ricctrically and aufficicnt mergia to criticality shall be d e sion s t r a t ed .

c. 'f naintenance is to be perforned on two control rod drives they must be separated by more than two control cells in any direction.

d.

d y An app ropr iate nuruer of Sal's a r e available as definct in specifi-cation 3.10.A.

6. Any number of control roie ru y b e w i t h<t r nvn or removed 3. With the mode selector switch in from the reactor core pro- the refuel or shutdown n. ode, no viding the following condi. control rod nay be bypassed unti!

tiono are sati fled: two licensed operators have confir. rd that either all fuel has been removeJ

s. The reacto. code switch from around that rod or that all is leckes in the "re-control rods in ircediately ac,act u fuel' ronition. The cells have been fully insertec'ar.d refueling interloc" Which prevents ro r e than electrically disarmed.

one control red frca 304 298 0g9 p

v J h

u,lig g( gg  !

v Ju ..

A

m 5.0 KAJOR DES!G Q LAlURES 5.1 SiTF. F LATilRLS Browns Ferry tmit 2 is located at Brovns Ferry Nuclear Plant site on property ovned by the United States and in custody of the TVA. The site sha ll consist of approximately 840 acres on the north shore of Wheeler Lake at Tennessee River Mile 294 in Limestone County. Alabama. The minimum distance from the outside of the secondary containment building to the boundary of the exclusion ares as defined in 10 CFR 100.3 shall be 4,000 feet.

5.2 REACTOR A. The reactor core may contain 764 fuel assemblies consisting of 7x7 assemblies having 49 fuel rods each, 8x8 assemblies having 63 fuel rods each, and 8x8 R assemblies having 62 fuel rods each. The number of each type in the core is given in the most recent reload amendment topical report.

B. The reactor core shall contain 185 cruciform-shaped control rods. The control material shall be boron carbide powder (B4 C) compacted to approximately 70 percent of theoretical d e r.s i ty .  %

5.3 REACTOR VF.SSEL The reactor vessel shill be as described in Table 4.2-2 of the FSAR. The applicable design codes shall be as described in Table 4.2-1 of the FSAR.

5.4 ,CO NT A !:.N ENT A. The principal design parameters for the prinary containment shall be as given in Table 5.2-1 of the FSAR. The applicable design codes shall be as described in Sec tion 5.2 of th FSAR.

B. The secondary containment shall be as described in Section 5.3 of the FSAR.

C. Penetritions to the primary contain ent and pipine, passing t hrough such pene t ra t for.s shall be designed in accordance with the standards set forth in Section 5.2.3.4 of the FSAR.

5.5 FUEL STou]

A. The arranRement of fuel in the new-fuel storage facilit/

chall be such that k for dry conditions, is less than 0.90 and flooded is TNs, than 0.95 (Section 10.2 of FSAR).

298 090

8. 4 Alstil H l :;TP A'l, j Yh CUilfl'Ol/>

p A summary of the more significant discussions and conclusions of the NSRD will be transn.itted along with the fanal minutes.

7 C h a r t_e r A writton charter delineating the es ta bl is hme nt ,

composition, and mission of the NSRB and the dissemination of NSRB minute

• and reports sha}l be maintained; this may be amended as required. The charter s ha ll identif y the responsibility and authority or the NSRB in conducting leviews, including responsibility to identify problems and to recommend solutions to the Manager of Power.

B. Plant Operations Review Committee (PORC)

1. Membership The PORC shall consist of the plant superintendent, electrical maintenance supervisor, mechanical maintenance supervisor, instrument maintenance supervisor, health physics supervisor, operations supervisor, results supervisor, and QA staff supervisor. An assistant plant supervisor may serve as an alternate co=ittee tember when his supervisor is absent.

The plant superintendent will serve as chairman of tr.e PORC. The assistant plant superintendent will serve as chairman in the absence of the plant V superintendent.

.: . Meeting Frequency The PORC shall meet at regular monthly i n te rv al s and for special meet ings as called by the cha irma n or an requested by individual members.

3. Quorun Superintendent or assistant superintendent, plus five of the seven other members, or their alternate, will constitute a quorum. A memoer will be considered present it he is in telephone communicition with the conci t t ee.

335 b

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(, 0 6Dm NISTRATIVE CO MROL_g plant, the applicable codes r equ i r ed f at igue usage evaluation for the reactor pressure

vessel only. The locations to be monitored sh,ill be:

1. The feedwater nozzles

2. The shell at or near the waterline

3. The flange studs

b. Recording, Ev a lu a t i ng , and Reporting (1) Transients that occur during plant operations will be reviewed and -

cumulative fatigue usage tactor determined.

(2) For transients which are more severe than the transients cvaluated in the stress report, code fatigue usage calculut ions will be made and tabulated s epa ra t e ly .

(3) In the monthly 0;erating Report, the fatigu._ usage factor determined for the transients defined in (1) and (2) above shall be added and a cumulative fatique usage factor to date shall be listed.

When the cumulative usage f actor reaches a va lue of 1.0, an inservice inspection shall be included for the specific

location at the next scheduled inspection (3-1/3-year interval) period and 3-1/3-year intervals thereafter, and a subsequent evaluation performed in accordance with the rules of ASME Sec*. ion XI Code if any flaw indications are detected. The results of the evaluation shall be submitted in a Special Report (Section 6. 7. 3) for review by the Commission.

B. Except where covered by applicable 2 egulations, items 1 through 8 above shall be retained for a period of at least 5 years and items 9 through 17 shall be retained for the life of the pla nt. A complete inventory of radioactive materials in possession shall be maintained currer t at all times.

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6.0 ADMINISTRATIVE CONTROLS (b). Annual Report (1) A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-Tem exposure according to work and j ob functions, 3.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or fila badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major

.ork functions.

(2) A report of facility changes, tests or experiments required pursuant to 10CFR50.59(b).

c. Monthly Operating Report. Routine reports of operating statistics and shutdown experience shall be submitted on a nonthly basis to the Office of Inspection and Enforce-ment, U.S. Nuclear Regulatory Commission, b'ashington, D.C.

20555, with a copy to the appropriatc Regional Office, to be submitted no later than the tenth of each month following the calendar month covered by the report. A narrative summary of operating experience shall be submitted in the above schedule.

2. Reportable Occurrences Reportable occurrences, including corrective actions and measures to prevent reoccurrence, shall be reported to the NRC. Supple-mental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.

298 093 WRvRn tiMinw. '  ;

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6.C sDMItJISTRATIVE cot 3TROLS D. Source Tests Results of required leak tests performed on sources v if the tests reveal the presence of 0.005 microcurie or more of removable contamination.

C. Soecial R a co rt s (in writing to the Director of Regiona l Of fice of Inspection and Enforcement) .

1. R<> ports on the following areas shall be sutnitted as noted:

a. S.condary Containment 4.7.C Within 90 Leak la t e Tes ting (5) days of completion of each test.

b. Fatigue Usage 6.6 monthly Evaluati on Operating Report

c. Seismic Instrumentation 3.2.J.3 Within 10 days inope rabi li ty af ter 30 days of inoperability Jv

,rmu,<m 298 094 M

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6.0 ApMINISTR ATIVE CotlTROLS h FOOTh0TFS v 1. A single submittal may be made for a multiple unit station.

The submittal should combine those sections that are common to all units at the station.

2. The tern " forced reduction in power" is normally defined in the elect ric power industry as the occurrence of a component f ailure or other condition which requires that the load on the unit be reduced for corrective action innediately or up to and including the very next weekend. Note that routine prevantive maintenance, s ur ve illa nc e , and calibration activities requiring power reductions are not covered by this secticn.

3. The term " forced outage" is nornally defined in the electric power industry as the occurrence of a component failure or other condition which requires that the unit be removed from service f or corrective action immediately or up to and including the very next weekend.

4 This tabulation supplements t' c requirements of $20.407 of 10 CFR Part 20.

5. Each integrated leak rate test of the secondary containment shall be the subject of a summary technical report. This report should include data on the wind speed, wind direction, concurrent

(*5 reactor building p r e s s ur e , and emergency ventilation ficw rate. The report shall also include analyses and

%- in terpretaticns of those dat a whica descastrate compliance with the specified leak rate limits.

298 095 357

[ t v

pg l QUALITY S 'JPERINTENDENT l ASSURANCE MANACER

- SUPERVISOR QUALITY ASSULNCE DIVISION OF STAFF DIVISION OF ASSISTANT OME PLANT QUALITY MEDICAL SERVICES PLMNING ASSURANCE SUPERINTENDENT COORDINATOR QUALITY ASSURANCE l l ENGINEERS HEALTH ELECTRICAL MECHANICAL PHYSICIST F. ATIONS INSTRUMENT ICRSE MAINTENANCE MAINTENANCE SUPERVISOR SUPERVISOR MAINTENANCE SUPERVISOR SUPERVISOR SUPERVIS0h 1 - t ASSISTANT , ASSISTANT HEALTH / ASSISTANT ASSISTANT ASSISTANT ASSISTANT INSTRU:!ENT INSTRUMENT RESULTS OPERATIONS ELECTRICAL MECHANICAL PHYSICS

CE MINTENANCE TECHNICIANS SUPERVISOR SUPERVISOR MAINTENANCE ltAINTENANCE SUPERVISOR SUPERVISOR SUPERVISCR SUPERVISOR (INST) (COMP) l SHIFT TECHNICAL TECHNICAL TECHNICAL TECHNICAL W N l TECHNICAL l

ENGINEERS ENGINEERS OPERATING ENGINEERS ENGINEERS W ENGINEERS l PERSONNEL AND CRAFT AND CRAFT AND CRAFT AND CRAFT

, CD j FOREMAN FOREMA'i FORF"AN FOREMAN CD 4

• SROWH1 FERRY HUCLE AR PLANT IINAL SAFETY ANALY115 REPOR T 00  ! O i 'U f' a h m.G E FM b : G L2 L di'quL L n1NCTIOm ORcmurION W

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PROPOSED CHANGES TO BROWNS FERRY UNIT 3 TECHNICAL SPECIFICATIONS 298 097

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TABLE 3.1.A REACTOR I POTirTION SYOTEM (GCRAM) IN97 RUM EWTATION P EQU T P D4LNT Min. No.

of operable Inst, redes in which Function Channels Must Pa operable Per Trip Shut- S ta r t up/ Hot System (1) Trip runction Tr i p Tevel Settinq down Pefuel (7) Standby F Act ion (1) 4 Turbine Stop valve Closure 5 101 Valva closure X(4) X (4) X(4) ..A or 1.D 2 Turbine con trol valve - Loss ot Control Pressure 2 550 psi ; X(4) X (e) X (4) 1.A or 1.0 2 Turbine Fi r s t Stage Pressure Dermissive 5 154 psig X ( 18) X (18) X (18) (19) 2 Turbine Condenser Iow vacuum 2 23 In. Eg, Vacuum X ( 3) Z ( .1) I 1.' or 1.C w

2 Main Steam Liae High 5 bX Norm l Full Power X(9) X (9) X (9) 1.A or 1.C Radiation (14) Bac k gr ound (20)

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12. The APRM downscale trip is automatically bypassed when the IRM instru:nentation ie operable and not high.

13. Less than 14 operable LPRM'o uill cause a trip system trip.

14. Channel shared by Reactor Protection System and Primary A Containment anc Reaccor Veccel Isolation Control Syste.m.

channel failure may be. a channel failure in each system.

15. The APRM ,155 scram is byprssed in the Run Mode.

16. Channel shared ay Peacter Protection System and Reactorfailure A channel Manual Control yctem (Rod Llock Portion).

may be a channel f ailure in each system.

17. Not required while performing lcw power physi"s tests at atmospheric pressure during or af ter refueling at power levels not to exceed 5 Mu tt) .

18. Operability is required when recctor thermal power is belev 30% (high-pressure turbine first-stage pressure (5 154 psig) .

19. Action 1. A or 1.D shall be take, only if the permissive f ails in cuch a manner to prevent t!c Gffected RPS logic frca Otherwise, no action is performing its intended funct.on.

required.

20. In alam setting of 3.0 times r.omal background at rated power shall O be established to alert the operator to abnomal radiation icvels in the primary ccolant.

21 . The APRM High Flux and Inoperative Trips do not have to be operable in the Refuel Mode if the Source Range Monitors are connected to give a non-coincidence, High Flux scram, at < 5 x 105 cps. The SRM's shall be operable per Specification 3.10TB.l . The removal of eight (8) shorting links is required to provide non-coincidence high-flux scram protection from the Source Range Monitors.

22. The three required IRM's per trip channel is not required in the Shutdown or Refuel Modes if at least four IRM's (one in each core quadrant) are connected to give a non-coincidence, High Flux scram.

The removal of four (4) shorting links is required to provide non-coincidence high-flux scram protection from the IRM's.

298 099 35 O

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TABLE 4.1. A REACTOR PROTECTION SYSTEM (SCRAM) I N STR tJHENT ATIO N FUNCTIONAL TESTS MIN IMUM FUNCTIONAL TTST FREQUENCIES FOR SAFETY INSTR. AND CONTROL CIRCUITS Minimufs Frequency (3)

Grou p (2) Functional Test Each Refueling Outage Mode Switch in Shutdown A Place Mode Switch in Shutdown Trip Channel and Alarm Every 3 Months Mnual Scram A IRN Trip Channel and Alarm (4) Once Per Week During Ref ueling High Flux C and Before Each Starttp not to exceed Once per week Trip Channel and Alarm (4) Once Per Week During Ref ueling Inoperative C and Before Each Startup not to exceed once per week APRM Trip output Relays (4)

Before Lach Gtartup and Weekly High Flux (15% scram) c When Rcquired to ue 09erable Trip Output Relays (4) Once/ Week High Flux B B Trip output Relays (4) Once/ Week g Inoperative B Trip Output Relays (4) Once/ Week Downscale B (6) (6)

Flow Dias A Trip Channel and Alarm Once/ Month (1)

g% High Reactor Pressure Ap / A Trip Channel and Alain Once/ Month (1)

High Drywell Pressure A Trip Channel and Alarm Once/ Month ( 1)

Reactor Low Water Level (5)

Trip Channel and Alarm Every 3 conths j High Water level in Scram Discharge Tank A A Trip Channel and Alarm Once/MontA (1)

Turbine Condenser Low vacuum g, e a

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(FrT[Uyf, rABLE QC

1. For t hs

• t or t up and r un pos it ions of the Reactor Mode Selec t or Switch, there shall be two operable or trapped trip systems for each ronction. The SRN, IM, and APRM (Startup mode) , bl oc k s n e ~1 not be operable in " R una mode, ar d the APRM (Flow bi as e d) and R BM rod blocks need not be operable in "Startup" mode.

If the first column cannot be met for one of the t wo t ri p syst ems, this condition nay exist for up to seven days provide 1 that derang tnat time the operable system as functionally testei immediately and daily thereaf ter; if this condatton last lonaer than seven days, the system with the inoperable channel shall be tripped. If the first column cannot be tripped.

tw met for both trip systems, Doth trip s, stems shall

2. W is the recirculaticn loop flow in percent of design. Trip level setting is in parcent of ra ted power ( 3 29 3 Mwt ) .

A rat ic of FM /.'."F1.P? < l . C i s permit t e a. reJucea pcwe r.

so t po i n t .

S?* Specification 2.1 f or APRM cont rol rod block 1.

IPM downscale is bypassed when it is on its lowest range.

4. SRM's A and C downscale function is bypassed when IRM's A, C, E, and C are above range 2. SRM's B and D downscale function is by- '

passed when IRM's B, D, F, and H are above range 2.

SRM detector not in startup position is bypassed when the count rate is b 100 CPS or the above condition is satisfied.

5 One in str ument cha rin el ; i.e., one APRM or IRM or RBit, per trip syst em may be bypassed except only one of f our SRM may be bypassed.

6 IRM chann els A, E, C, G all in range 8 bypasses SFN channels A6C f unct tons.

IRM channeis B, F, D, 11 all in range 8 bypasses SRM channels D6 D functions.

T. The follcvin6 ;peratiencl restraints appl:' te the PL" enly:

A.

30th P3.'; channe;c are typassed when resctor power it f 3M.

t. The RE" need nct be c;erarle ir, the "rtartup" position of the react r n:de se?tetcr svitch.

c.

Tvc PE!' che_nnels are provideo and cnly one cf there ray be ty;assed frc. the ccr.s le. en PiM channel r.uy be ct.: cf ser'tice fcr testing ana/t- r. air.tenance provided this cor.diticn does not last lcreer than 2t tours ir. any thirty day periv1.

d. If .ini ce. censiticnr. fer L ble 3.2.0 are net r.et, ad.iinistrctive centr:ls shall te irr.ediately i.cycsed to preven. centrc; red withdrawal.

298 101 f96%MG H Ogury B UJh MingSykp

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TABLE 3.2.E INSTRUMENTATION TIIAT KNITORS LEAKAGE Itfr0 DRYWELL Action Remar ks System (?) S et points

( 1) 1. Usel to determine aden t a-DTuipment Dr La WA fiable reactor coolant Flow In t eg r at or leakage.

2. Considered part of surp -

Sump Fill Rate system.

Timer 2 20.1 min.

Sump Pump Out Rate Timer 5 13.4 min.

( 1) 1. Used to determine unidenti-Floor Drain fiable reactor coolant Flow Integrator WA leakage.

2. Considered part of surp Sump Fill Rate system.

Tim er 2 80.4 min.

Sump pump Out Rate Timer 5 8.9 min.

6 Gas and 3 x Average (3)

Drywell Air Sampling Particulate Background "h

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64

• Whenever a rystem is required to be operable, there shall be one operable system either automatic or m.anual, or the action required in Section 3.6.C.2 shall be taken.

(2) An alternate system to determine the leakage flow is a manual system whereby the time tetween sump pump starts is monitored. The time interval will determine the leakage flow because the valua.e of the sump will be known.

p (3) Upon receipt of alarm, isonediate action will be taken to confirm the alarm and assess the possibility of increased PTM leakage. Refer to Tech Spec Sections 3.6.C.2 and 3.6.C.3 for drywell air sampling systen Out of service.

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 totGArnsenT sYsress 4.7 cottrAINMENT SYSTEMS

-D. Primary Con t a intnent Isolation Valves D. Primary Containment

_ Isolation Valves _

1. During reactor power 1. The primary operation, all isolation valves containment i rola t ion listed in Table 3. 7. A valves surveillance shall be performed as and all reactor coolant system follows:

instrument line flow a.

check valves shall be At least once per operable except as operating cycle the specified in 3.7.D.2. operable isolation val ves that are power oporated and autcmatically initiated shall be tested for sitrulated automatic initiation and closure times.

b. At least onecper quarter:

(1) All normally open power b operated isolation valves (except for the tain steem line power-operated icolation valves) shall be fully closed and reopened except where specific written relief from ASME Section XI has been granted by NRC pursuant to 10 CFR 50, Sec tion 50.55a (g) (b) (1) .

(2) With the reactor power less than 75% trip nain steam isolation valves individually and verify closure time.

298 103 P 2sa f @  !.

LIMITING CONDITIONS FOR OPERATION BURVEILLANCE REQUIREMENTS

,b

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3. 7 (OJTA INMEffr S YST_F2iS 4.7 CONTAINMENT SYSTEMS G. Containment Atronphere G. Containment Atmosohere DiluM on System (CA D}_ Dilution System (CAD)

1. The Containnent Atmosphere Dilution l' E"*" * # '?

(CAD) System shall be operable with: a. At least once per manth, cycle each salenoid operated air / nitrogen valve with its hand switch (na cantainment isolatteu override)

a. Wo independent * ' "'" l" "* '

• l*** "7'1" "# I"Il systems capable travel anj verity that each manual valve of supplying in the flaw path is open.

nitrogen to the

b. verity that the CAD System contains a minimu:n drywell and supply t 2500 gals. of liqui.i nitrogen twice torus. per weet.

A minimum supply c. At each cold shutdown verif y that each salenoid b.

oP"ated air /nttragen valve in the ficw path of 2500 gallons can be opened with a containment isolation of liquid signal present ( c on t aic::.e n t isolation override).

nitrogen per system.

( *

2. The Containment Atmosphere Dilution (CAD) System shall be operable whenever the reactor mode switch is in tha "RUN" position.

3. If one cystem is inoperable, the reactor may remain in operation for a period of 30 days provided all active components in the other system are operable.

4 If Specification

3. 7 . G .1 and 3.7.G.2, .

or 3.7.G.3 290 lgg UT cannot be eet, an orderly shutdown shall be initiated and tho reacter shall ~i ;;j be in the Cold DU mC'o 4 -

Shutdown condition .

b '1'! \d L.

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 9 d Q 2e ( , .

TABLE 3. 7. A PRIMARY CONTAINPtDTT ISOLATION VALVES Nomber of Power Ma11 mum Acti on on Operated Valves Opera ting Normal Initia tiry Valve Identifica^. ion Inboard Out tx>a rd Time (sec.) Position Sig nal Group 3 Reactor w.ter cleanup system supply 30 0 GC t oolation valves (FCV 1 & 23 1 1 4 EPCIS steamline isolation valves 1 1 20 0 .

(rCV-73-2 & 3) 5 RCICS steamline isolation valves 1 1 15 0 GC (PCV-71-2 & 3) 6 Drywell nitrogen purge inlet isola-10 C SC tion valves (FCV-76-18) 1 gQ 6 Suppression chamber nitrogen purge 10 C SC

. inlet isolation valves (FCV-76-19) 1 p- 6 Drywell Main Exhaust isolation

• • valves (PCV-64-29 and 30) 2 90 C SC 5 M r 6 S::ppression chamber main exhaust u isolatloa valves (frV-64-32 and 33) 2 90 C SC 47 6 Drywell/ Suppression Chamber purge 90 C SC inlet ( FCV-6 4-17) 1 6 Drywell Atmosphere parge inlet

- af ( frV-6 4-18) 1 90 C SC wn ~

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I TABLE 3.7.A FBIAAF Y CORIAINMENT ISOLATION V ALV ES 54axtrum Actten on Numter of P r>w e r Operated Valves Operattni Normal Insttatsnq Position Signal Inboard out tx>a r d Time (sec.)

Group valve Ident2fication 6 Suppression Chamber purge inlet 1 100 C SC

( FC V- 64-19) 6 Drywell/ Suppression Chamber nitro- 10 C SC 1

gen purge inlet (FCV-76-17) 6 Drywell Exhaust Valve Bypass to Sta ndby Gas Treatment System C SC 1 10 (TCV-64-31) 6 Suppression Chamber Exhaust Valve Bypass to Standby Gas Treat 2nent 1 10 C SC System (FCV- 6 4-3 4) 5 0 GC 2

7 RCIC Steamline Drain (FCV-71-6 A & 6B) 7 RCIC Condensate Pump Drain C SC 2 5 M. . - (FCV-71-7A & 7B) 7 HPCI Hotwell pump discharge isola- 5 C SC 2

n tion valves (FCV-73-17A & 17B) 5 0 GC 2

HN1 steamline dratn (FCV-73-6A & 6B)

E[h 7 1 per guide NA C GC 8 TIP Guide Tubes (5) tube G71 gtwJ cmr s

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LIMITING CONDITIONS FOR OPEPATION SURVEILLANCE REQUIREMENTS 3.10 CORE ALTEPATIOtJ S

6. Any number of control 3. With the mode selector rods may be withdrawn switch in the refuel or or removed from the shutdown mode, no control reactor core rod may be bypassed until providino the two licensed operators following :onditions have confirmed that either are satisfied: . all fuel has been removed from around that rod or that

a. The reactor mode all control rods in switch is locked immediately adjacer.t cells in the " refuel" have been fully inserted and

' Th electrically disarmed.

P[ffe interlock which prevents more than one control rod from being withdrawn may be bypassed on a withdrawn control rod after the fuel assemblies in Q

the cell containing (controlled by) that control rod have been removed from the reactor core.

All other refueling interlocks shall be operable.

107

. 298

.', M A.Jo e L, L 'i t '.N f f A T 'h E ',

V 5.1 f, l T L F EATUI t.S Drowns Ferry units 1, 2, and 3 are located at Browns Ferry Nuclear I' l a n t site on property owned by the United St.ates and in c u s t od y o f 't n e TVA. The site f. hall consist of approximately 84 0 acres on the north shore of Wheeler Lake at Tennessee River Mlle 294 in Limestone County, A l a ba m a. The minimum distance from the outside of th e s econ d a ry cont a i nma nt building to the boundary or the exclusion area ac defined in 10 CFR 100.3 chall be 4,0 00 f e et.

5.2 REACTOR A. The reactor core may contain 764 fuel assemblies consisting of 8x8 assemblies having 63 fuel rods each, and 8x8 R assemblies having 62 fuel rods each. The number of each type in the core is given in the most recent reload amendment topical report.

B. The reactor core shall contain 185 cruciform-shaped control rods. The control material shall be boron carbide powder (B4 C). compacted to approximately 70 percent of theoretical density.

5.3 R EACTOP VESSEL The reactor vessel shall be as described in Table 4. 2-2 of tne FSAR. The applica ]le design codes shall be as described s- in Ta ble 4. 2- 1 of the FSAR.

5.4 CONTAINMENT A. The Trincipal desion parameters for the prinary contatnmen; shall be given in Table 5.2-1 of the PSAR.

The applicable desion codes shall be ao described in Section 5.2 of the FSAR.

B. The sacondary containment chall be as described in Section 5.3 of the FSAR.

C. Penetrations to the primary containment and piping pa:sinq enrough such penetro t!cns shall be designed in accor4ince with the standards set f ort h in section

5. 2. 3. 4 of the FSAR.

5.5 FUEL STOFAGE A. The arrangement of the fuel in the new-fuel storage facilitity shall be such that k,gg , f or dry conditions, 298 108

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A summary of the more significant discussions and conclusions of the NSRB will be transmitted along with the final minutes.

7. Charter A written charter delineating the establishment, composition, and mission of the NSRI and the dissemination of NSRB minutes and reports shall be maintained; this may te amended as required. The charter shall identif y the responsibility and authority of the NSRB in conducting reviews, including responsibility to identify problems and to recommend solutions to the Manager of Power.

B. Plant Operations Review Committee (PORC)

1. Membership The PORC shall consist of the plant superintendent, electrical maintenance supervisor, mechanical maintenance supervisor, instrument maintenance supervisor, health physics supervisot, cperations supervisor, results supervisor, and QA staff supervisor. An assistant plant supervirer may nerve as an alternate committee member when his supervisor is absent.

The plant superintendent will serve as chairman of the PORC. The assistant plant superintendent will serve as chairman in the absence of the plant superintendent.

2. Meeting Frequency The PORC shall neet at regular monthly intervals dnd for special meetings as Called by the chairman or as requested by individual members.

3. Quorun Superintendent or assistant superintendent, plus

\fiveofthesevenothermembers, or their alternate, will conutitute a quorum. A member will be considered p1.esent if he is in telephone communication with the committee.

- 298 109

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6.0 ADMINISTPATIVE CONTPOLH TW plant , the appl a c.s tile code., a equ i re>l tat ique usage evaluatiori for the reactor pt<tsnure vessel only. The locatioria to be monitored shall he:

1. The feedwater nozzles

2. The shell at or near the waterline

3. The flange studs

b. Recording, Ev a lu at i ng , and Reporting (1) Trensients that occur during plant operatioas will te reviewed and a cumulative fatique usage factor determined.

(2) For transients which are more severa than the transients evaluated in the stress re por t , code f atique usage calculations will be made and tabulated separately.

( (3) In the monthly Operating Report, the fatique usage factor determined for the

([>- transients def ined it. (1) and (2) above shall be added and a cumulative fatique usage factor to date shall be listed.

When the cunulative usage factor reaches a va lue of 1.0, an inservice inspection shall be included for the specific location at the next scheduled inspection (3-1/3-year interval) period and 3-1/3-year intervals thereafter, and a subs equ ent evaluation performed in accordance with the rules of ASME Section XI Code if any flaw indications are detected. The results of the evaluation shall be submitted in a Special Report (Section 6.7.3) for review by the Commission.

D. Except where covered by applicable regulations, items 1 through 8 above shall be retained for a period of at least S years and items 9 through 17 shall be retained for the life of the pla nt . A complete inventory of radioactive materials in possession shall be maintained current at all times.

I10 7

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6.0 ADM?.NISTRATIVE CONTROLS (b)) Annual Report (1) A tabulation on an annual basis of the number of station, utility and other per.onnel (in-c.'uding contractors) receiving exposures greeter than 100 mrem /yr and their associated man-reu exposure according to work and job functions, e.g., reactor operations and sur-veillance, inservice inspection, routine rain-tenance, special maintenance (describe main-tenance.f, vaste processing, and refueling.

The dose assignment to various duty functions may be estimates based on pocket dosiccter, TLD, or film badge measurements. Small exposures totaling less than 20*. of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received frca external sources shall be assigned to specific major work functions.

(2) A report of facility changes, tests or experi-ments required pursuant to 10CFR50.59(b).

c. Monthlv Operating Report: Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Office of Inspection and Enforcement, U. S. Nuclear Regulatory Co=ission, Washington, D.C. 20555, rith a copy to the appropriate Regional Office, to be sub-mitted no later thaa the tenth of each conth following the calendar conth covered by the report. A narrative summary of operating experience shall be submitted in the above schedule.

2. Reportable Occurrences Reportable occurrences, including corrective actions and ceasures to prevent reoccurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.

298 111 381

f 6.0 CDt4INISTR ATIVE CONTROLS

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B. Source Tests Results of required leak tests performed on sources if the tests reveal the presence of 0.005 microcurie or more of removable contamination.

C. Special Repog s (in writing to the Director of Regional Office of Inspection and Enforcement) .

1. Reports on the following areas shall be submitted as noted:

a. Secondary Centainment 4.7.C within 90 Leak Rate Testing (5) days of completion of each test.

b. Fatigue Usage 6.6 monthly Evaluation Operating Report b

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FOOTNOTES

1. A single submittal may be made for a multiple unit station.

The submittal should combine those sections that are common to all units at the station.

2. The term " forced reduction in power" is normally defined in the electric power industry as the occurrence of a component f ailure or other condition vehich requires that the load on the unit be ' reduced for corrective action immediately or up to and including ti.e very next weekend. Note that routine perventive maintenance, surveillance, and calibration activities requiring power reductions are not Covered by this section.

3. The term aforced outage n is normally defined in the electric power industry as the occurrence of a component failure or other condition which requires that the unit be removed from service f or corrective action immediately or un to and including the very next weekend.

,,, 4. This tabulation supplements the requirements of s20.407 of 10 ,_s CFR Part 20.

5. Each integrated leak rate test of the secondary containment shall be the subject of a summary technical report. This report should include data on the wind speed, wind direction, concurrent reactor building pres s ur e, and emergency ventilation flow rate. The report shall also include analyses and interpretations of those data which demonstrate compliance with the specified leak rate limits.

298 113

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PLAh7 QUALITY ASSURANCE SUPERINTENDENT MANACER SUPERVISOR QUALIIT ASSURANCE DIVISION OF STAFF DIVISION OF ENVIRONMENTAL MEDICAL SERVICES PLANT SUAL U PLANNING ASSURANCE l SUPERINTENDENT -

COORDINATOR QUALITY ASSURANCE i i ENGINEER 3 e

HF I ELECTRICAL MECllANICAL p p* RESULTS OPERATIONS INSTRUMENT Sb?ERVISOR MAINTENANCE MAINTENANCE MAINTENANCE SUPERVISOR SUPERVISOR SUPERVISOR SUPERVISOR I __ i _

ASSISTANT ASSISTANT HEALTH g' ASSISTANT RESULTS ASSISTANT ASSISTANT ELECTRICAL ASSISTANT MECilANICAL INSTRUMENT INSTR 12ENT PHYSICS OPERATIONS MAINTENANCE MAINTENANCE TECHNICIANS SUPERVISOR SUPERVISOR MAINTENANCE MAINTENANCE SUPERVISOR g . SUPERVISOR SUPERVISOR SUPERVISOR

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SHIFT TECIINICAL TECHNICAL TECHNIC /.L TECHNICAL

' TECHNICAL ENGINEERS ENGINEERS OPERATING ENGINEERS ENGINEERS (A) ENGINEERS ' AND CRAFT AND CRAFT AND CRAFT AND CRAFT PERSONNEL

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_ gq( c IINAL 5AFETY ANALY3f 5 REPORT de a te L1:~py d2h k;hy FUNCTIONAL ORCANIZATION FIGl'RE 6.1-2 c5sa ECZe~I

ENCLOSURE 2 298 115 th d_i3L I,

REASONS AND JUSTIFICATIONS FOR PROPOSED CHANGES TO BROWNS PERRY NUCLEAR PLANT TECHNICAL SPECIFICATIONS FOR UNITS 1, 2, AND 3 UNIT 1 Page 34 (Table 3.1. A) and ---- It is proposed to change the Main Steam page 36 (Note 20) Line Radiation Monitor (MSLRM) setpoints from 1.5X and 3X background to higher valuan, name, 1X and 6X background.

The rece= mended setpoint in GEK 779, Vol. IV is 3X background for alarm and 6X background for iscal: ten trip. The accuracy of the MSLRM is i 25 percent for the 100- 1000 MR/ER range with a 13 percent per Vech drif t (Reference GEK 32426A). The current setpoint of 1.5% add 3X background is very hard to maintain without exceeding technical specifications when the instrument drifts down or having high alarms when the instrument drifts up (see Attachment 1).

The MSLRM is installed to detect and respond to increases in main steam line radiation that might indicate gross fuel cladding ruptures (NEDO-10174 Oct. 1977). Neither precision nor accuracy is required to measure gross failures as indicated by CE specif1-cations. Therefore, raising the ser-points will not degrade instrument tasponse to gross failure, but will allow instru-ment accuracy to be taken into effect when setpoints are calculated.

Page 37 - - - - - - - - - - - - - - - It is proposed to change the IRM Functional Test Minimum Frequency for the high flux and inoperative modes to read: "Once per week during refueling and bef ore each startup not to exceed once per week."

This change is proposed in order to provide agreement with the functional test requirements of Table 4m 2.C.

298 ii6 138 0 "jdP?oggav J U E@Mn U f Aba d

Page 74 -------------- It is proposed to change Note 4 to read as follows:

"SRM s A and C downscale function is 8

bypassed when IRM's A, C, E, and C are above range 2. SRM detector not in startup position is bypassed when the count rate is & 100 CPS or the above condition is satisfied." This change vill make the specification for instrumenta-tion that initiate rod blocks agree with the as installed equipment.

Page 77 - - - - - - - - - - - - - - It is proposed to add the following state-ment to Note 3 of Table 3.2.E " Refer to Technical Specification Sections 3.6.C.2 and 3.6.C 3 for drywell air sampling system out of service." This addition should clarify the requirements of Section 3.2.E.

Page 242 ------------- It is proposed to change specification 4.7.D.l.b(1) to read : ". . . . closed and reopened except where specific written relief from ASME Section XI requirements has been granted by NRC pursuant to 10 CFR 50, Section

50. 55a (g) (b) (1) . " The NRC has identified certain valves which are not to be cycled except at cold shutdowns.

HPCI and RCIC steaa supply valves are examples of these valves.

Page 248 -------------

It is prop, d to change Section 4.7.G.1 to read as shovn. This proposal reflects discussions held with NRC on June 1, 1979.

The proposed change prevents the possibility of overriding a containment isolation signal, except during testing while in the cold shutdown condition.

Page 250 ------------- It is proposed to change "FCV 74-57, 58" to "FCV 75-57,58" - the present valve n ubers as quoted are incorrect.

Page 251 -------------- It is proposed to delete valve FCV 69-12 from Tabic 3.7.A. This valve is not a containment iso 1& tion valve. Isolation is provided by check valves 69-57,9 agd 3 3-572. 9 \\l m

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Page 252 - - - - - - - - - ----- It is proposed to change "FCV 75-57, 58" to "FCV 73-6A, 6B" and to change the normal operation of FCV 71-7A, 7B from "0" to "C" and change Action on Initiating signal from "GC" to "SC."

The present valve numbers as quoted are

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incorrect. The normal valve position of FCV 71-7A, 7B is closed.

Page 304 - - - - - - - - - --- -- It is proposed to delete the word "withdravn" in tech spec 4.10 A.3 and insert " bypassed." This would remove an otherwise ambiguous statement.

Page 330 - - - - - - - - - -- --- It is proposed to revise paragraph 5.2.A to delete the number of assemblies of each type of fuel in the core and add the following: "The number of each type in the core is given in the most recent reloca amendment tepical report."

This proposed change will serve to reduce the number of cycle dependent page changes to future reload technical specifications.

We consider this proposed change admini-strative in nature.

Page 335 - - - - - - - - - ----- It is proposed to change the membership and quorum of the Plant Operations Review Committee (PORC) to reflect new plant organization.

Pages 348, 351, and 356 r r- ---- It is proposed to change paragraph 6.6.A.17.b(3) and 6.7.3.C.1.b from a frequency of " annual operating report" to " monthly operating report" to further comply with NRC's letter from A. Schwencer to G. Williams dated September 16, 1977.

It is also propesed to add paragraph 6.7.1.b(2) in order to repert cumulative fatigue usage factors for the reactor vessels menthly in the monthly operating report.

Page 357 --------------- It is proposed to delete the words "inside and outside temperatures during the test" from footnote 5 of Section 6.0.

These temperatures are not pertinent to the resulta of the integrated leak rate test of the secondary containment.

Page 362 -------------- It is proposed to revise Figure 6.1-2 to reficct new plant organization.

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UNIT 2 Page 34 (Table 3.1.A) and - -- - - - Same as for pages 34 and 36 for page 36 (Note 20)

Page 37 ------------- - - Same as page 37 for unit 1 above Page 74 - - - - - - - - - - - - - - Game as page 74 fcr unit 1 above Page 77 - - - - - - - - - - - - - - - Same as page 77 for unit 1 above Page 242 - - - - - - - - - - - e Same as page 242 for unit 1 above Page 248 - - - - - - - - - - - - - - - Same as page 248 for unit 1 above Page 250 - - - - - - - - - - - - - - Same as page 250 f or unit 1 above Page 251 - - - - - - - - - - - - - - Same as page 251 for unit 1 above Page 252 - - - - - - - - - - - - - - - Same as page 252 for unit 1 above Page 304 - - - - - - - - - - - - - - Same as page 304 f or unit 1 above Page 330 - - - - - - - - - - - - - - - Same as page 330 for unit 1 above Page 335 - - - - - - - - - - - - - - - Same as page 335 for unit 1 above Pages 348, 351, and 356 -- - - --- Same as pages 348, 351, and 356 for unit 1 above Page 357 e - - - - - - - - - - - - - Same as page 357 for unit 1 above Page. 362 e e - -- - - - - - - - - - Same as page 362 for unit 1 above 9 9 hj u

bNIT 3 Page 33 (Table 3.1.A) and - - - - - Same as for pages 34 and 36 for page 35 (Note 20) unit 1 above Page 36 --------- - -- Same as page 37 for unit 1 above Page 77 ------ ------- Same as page 74 for unit 1 above Page 80 - -- ---- ---- Sane as page 77 for unit 1 above Page 248 -------------- Same ao page 248 for unit 1 above Page 254 -------------- Same as page 242 for unit 1 above Page 263 -------------- Same as page 251 for unit I above Page 264 ---- --------- It is proposed to change the normal position of FCV 71-7A, 7B frna "O" to "C" and change Action on Initiating Signal from "CC" to "SC". The normal position of FCV 71-7A, /B is closed.

Page 335 -- ----------- Same as page 304 for unit 1 above Page 360 -------------- Same as page 330 for unit 1 above Page 365 -------------- Same as page 335 for unit 1 above Page 378, 381, and 386 -e - - - -- Same as pages 148, 351, and 356 for unit 1 above .

Page 387 - - - - - - --- - - - - - - Same as page 357 for unit 1 above Page 392 -------------- Same as page 362 for unit 1 above

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