ML18283B750

Technical Specifications Changes to Add Containment Isolation Valves Associated with the Drywell to Torus Differential Pressure Control System. Corrected Pages to Amendments Nos. 27 and 24

ML18283B750
Person / Time
Site:	Browns Ferry
Issue date:	02/15/1977
From:	Schwencer A Office of Nuclear Reactor Regulation
To:	Gerald Williams Tennessee Valley Authority
References
Download: ML18283B750 (85)

Text

4 DISTRIBUTION Dockets(2)

NRC PDR(2 Loca1 PDR ORB81 Reading VSte11o FEB L G i'RGo1ler/TJCarter SMSheppard TNambach Tennessee Valley Authority OELD ATM:

Hr. GoA>in l<illiams, Jr.

OI &E(5)

Manager of Po>ier BJones(8)

N8 Power Building BScharf(15)

Chattanooga, Tennessee 372O1 JMcGough ACRS(16)

Gentlemen:

OPA(CMiles)

DRoss The Comission has issued the enclosed Alrendments Nos.QS and 4~to Facility Licenses t/os.

DPR-33 and DPR-52 for the Browns Ferry Nuclear Plant, Units 1 and 2.

These amendments consist of changes to the Technical Syecii'ications in response to your requests of September 1,

'ctober 1 pnd'ctober, 12, 1976.

BHarless TBAbernathy JRBuchanan Dockets Nos.

a 5O-26O Copies of the Safety Evaluation and the Federal Register Notice are also

enclosed, Sincerely, aoei>>i S'<""da The amendments change the Technical Specifications to add containment isolation valves associated with the drylcell to terus differential pressure control system to the valve listing (Table 3.7.0) for the limiting condition for operation and surveillance requirementa.of primary containm nt.

A clarification inl the wording of tho temperature surveillance requirement for the torus l~ater has also been made.. This latter change is different from what you had proposed in your October 1, 19T7 request (but your staff has agreed that this modification sufficiently clarifies the specification.

In addition, the allo<<able operating time with tin inoperable Automatic Oepressurization I+stem (ADS) valves has been reduced from thirty days to seven days to reflect, the fact that the KCS Appendix K analysis was perfoymd srith five of the siW ADS valves operable rather than four as stated previously.

We are a1so taking this opportunity to co} rect typo-graphical errors page misnua&ering.

and valve misnumbering that occurred v>hen the specifications <<ere reissued in their'ntirety on Angst 2Q, 1976.

A, Schwencer, Chief Operating Reactors Br nch 8 Division of Operating Reactors OFFICE x27433; fg,~

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Tennessee Yalley Authority 2

February 15, 1977

Enclosures:

l.

Amendment No.

28 to DPR-33 2.

Amendment No.

25 to DPR-52 3.

Corrected Pages to Amendments Nos.

27 8 24 4,

Safety Evaluation 5.

Federal Register Notice cc w/enclosures.

See next page

Tennessee Val 1 ey Authority

-3 February 15, 1977 cc:

H. S.

Sanger, Jr.,

Esquire General Counsel Tennessee Valley Authority 400 Commerce Avenue E

11B 33 C

Knoxville, Tennessee 37902 Mr. D. McCloud Tennessee Valley Authority 303 Power Building Chattanooga, Tennessee 37401 Mr. William E. Garner Route 4, Box 354 Scottsboro, Alabama 35768 Athens Public Library South and Forrest

Athens, Alabama 35611 Mr. Charles R. Christopher

Chairman, Limestone County Commission Post Office Box 188

Athens, Alabama 35611 Ira L. Myers, M.D.

State Health Officer State Department of Public Health State Office Building Montgomery, Alabama 36104 Mr. C.

S. Walker Tennessee Valley Authority 400 Commerce Avenue W 9D199 C

Knoxville, Tennessee 37902 Chief, Energy Systems Analyses Branch (AW 459)

Office of Radiation Programs U. S. Environmental Protection Agency Room 645, East Tower 401 M Street, S,

W, Washington, D.

C.

20460 U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN.i EjS COORDlNA3)R 345 Courtland Street, NE Atlanta Georaia

30308,

s

~

~Pg AEVI P0

+w*e+

UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROMNS FERRY NUCLEAR PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.'8 License No.

DPR-33 1.

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

A.

The applications for amendment by Tennessee Valley Authority

. (the licensee) dated September 1, October 1

and October 12,

1976, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

E.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulitions and all applicable requirements have been satisfied.

)

i 4

2.

Accordingly, the license" is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility License No.

DPR-33 is hereby amended to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

28, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the

.Technical Specifications.

3.

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

FOR THE hjUCLEAR REGULATORY COf'MISSION

Attachment:

Changes to the Technical Specifications j;-,:~cF/F~~

A. Schwencer, Chief Operating Reactors Branch P1 Division of Op rating Reactors Date of Issuance:

February 15, 1977

ATTACHMENT TO LICENSE AMENDMENTS ANEHDNENT HO, 28 TO FACILITY LICENSE NO.

DPR~33 AMENDMENT NO.

25 TO FACILITY LICENSE NO, DPR-.52 DOOKBS HOS.

50-.259 5 50-.260 Revise Appendix A as follows:

Remove pages l57, l58, 167, 227, 259, and 262 and replace with identically numbered pages.

(.)

I.IxITIr"r:cn:lnr rrn. >>

FDR nr FRAvl<w slrRVc.l i.l.ANCE Rl UIRF.'lF'.rTS

3. 5. F R<<act or Cnr<<

l so la tion Cooling 4.5.F Reactor Cora Isolation Cooling 2.

If thc'CICS is inoperab'c, the reactor may tcmain in operation for a period not to exceed 7 days if thc HPCIS is operable during such time.

2.

When it is determined that the RCICS is inoperabla, chc HPCES shall be demonstrated to be operable immediately and weekly theres~ter.

3. If specifications 3.5.F.3.

or 3.5.F.2 aza noc'ct, an orderly shutdown shall be initiated and tha reactor shall bc daprcssuriicd eo lass than 122 pc"Ig wi.chin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C.

Automatic Danrcss<<rization G.

Automatic Denressuriaation S stem (ADS) 1.

Five of the si,x valves of the Automatic Dapressuri-aation System shall be

. operable:

(1) prior to a itart<<p from a Cold Condition, or, (2) whenever there is i,rza-diatad fuel in the reac-tor vessel and the reaceor vessel press<<re ia greater than 105 psir except as speci fied in 3.5.C. 2 and 3.5.C.3 below.

1.

During each operacing cycle the following tests shall be

~ performed on the ADS:

A simulated automatic actuation test shall be perfoencd pzior eo scaztup after each rafuelinp out-age.

Nanual surveillance of the relief valves is covezed, in 4.6.D.2.

2. lf more than one ADS valve is known to be incapable of automatic operation, the reactor may remain in opera-tion for a period not to exceed 7 days, provided the.

HPCI system is operable.

(Note that the pressure relief function of these valves is assured by section 3.6.D of these specifications and that this specification only applies to the ADS function.)

2, When it is determined that more than one of the ADS valves are incapable of automatic operation, the HP'CIS shall be demonstrated to be operable immediately and daily thereafter as long as Specification 3.5.8,2

applies,

.1 57 Amendments Nos.

28

& 25

1.[hIZTINC CA.'II)I'I'TONS FOR OPEPAVIOH SURVKILLAHCF. RE U IRENENTS 5.G Autnmnt ic Ocnrcsnurlzntion

~Sstrin (AI)S)

'4.5.G Automntic Dc ressurization 3

If specifications 3.5.Gel a'nd 3.5.G. 2 cannot be met, an orderly ohutdoun vill be initia ted nnd the reactor vcoscl prcssure sliall be reduced to lO5 psig or less Mithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

fl.

Hnlnccnnnce of nffled ~nfochnc e

~Pl c

Mhcncvcr thc core spray oyotems, LPCI, IIPCI, or RCEC.are required to bc opcrnblc, thc discharge pipl>>g from thc pump discharge of these systems to thc 'last block valve shall bc filled.

H.

Nnintenancc of Filled Dischar c

~Pi e The folloving surveillance rcquircnn ments. shall bc adhered to to assure that thc discharge plplnp, of the core spray systems, LPCI, HPCI, and RCIC arc filled:

158 amendments

Nos, 28 5 25

1 C

3.5.C Automatic De ressurization S stem (Ans)

This specification ensures the operability of the ADS under all condi-tions Eor Mhich the depressurication of the nuclear system is an csscn" tial response.

to station abnormalities.

The nuclear system p'ressure relieE system provides automaeic nuclear system dcpressurisation for small breaks in the nuclear syseem so chat the loM-pressure coolant in)ection (LPCI) and the core spray subsystems can operate eo protect the fuel barrier.

Hate that this specification applien only to the nutamatic feature oE the pressure relief syseem.

Specification 3.6.D specifies the requirements for the prcssure relief function af the vnlvcs.

Zt is possible Eor any number of the valves assigned to the ADS to be incnpable of performing their ADS functions because of Lnstrdmentaeinn failures yet be fully cipable of performing their prcssure relief function.

Because the automneic dcpressurizneian system docs not provide makeup eo ehe reactor primnry vessel, no credit is taken for the steam cooling of the core caused by the sys'em actuation to provide further conservatism to the CGCS ~

With one ADS valve known to be incapable of automatic operation, five valves remain operable to perform their ADS function, 7he fCCS loss-of-coolant accident analyses for small line breaks assumed that five of the six ADS valves were operable.

Reactor operation with two ADS valves inoperable is only allowed to continue for seven days provided that the HPCI system is demonstrated to be operable.

Amendments Nos.

28 5 25

$ ~

~

. OO ENCLOSURF.

OO LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREi~mNTS

3. 7 COtP'QXNMENT SYSTEMS 0. 7 CONTAINMENT S YST EMS Ao licabilit Applies to the operating status of,the primary and seconda"y containment systems.

~cb 'ective 1

To assure the integrity of the primary and secondary containment systems.

Sneci fication A licabilit Applies to the primary and secondary containment integrity.

~cb ective To verify the integrity of the primary and secondary containme nt.

Speci fication Primar'ontainment At any time. that the irradiated fuel is in the reactor vessel, and the nuclear system is pressurized aLove atmospheric pressure or work is being done which has the potential to drain the vessel, the pressure suppression pool water volume and temperature shall be, maintained within the following limits

~except as.specified in 3.7.A.2.

a., Minimum wate" volume -

123, 000 ft>i b.

Maximum 'uater volume - 135,000 ft~

c.

With. the suppression pool water temperature 95'F initiate pool cooling and restore the temperature to 95'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least hot shutdown within the next 6

hours and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

227 Primar Containment Pressure Suaoressicn Chamber a.

The suppression chamber water level be checked once per

day, Whenever heat is added to the suppression pool by testing of the ECCS or relief valves the pool temperature shall be continually monitored and shall be observed and logged every 5

minutes until the heat.

addition is terminated.

Amendments Nos.

28

& 25

~

~

ENCLOSURE 1

TABLE 3,7.D (Continued)

Valves43-28B 43-29A 43-29B Valve Identification RHR Suppression Chamber Sample Lines RHR Suppression Chamber Sample Lines RHR Suppression Chamber Sample Lines Test Medium Water~

~

T.est Method Applied between 74-226 and-;43-28B Water 'pplied between 74-227 and 43-29A (2)

Water 'pplied between 74-227 and 43-29B (2) 64-17 64-18 Drywell and Suppression Chamber air purge in1et Drywell air purge inlet Air Applied between 64-.17, 64-18, 64-19, and 76-24 Applied between 64-17, 64-. 18, 64-19, and 76-24 64-19 Suppression Chamber air purge Air inlet Applied between 64;.17, 64-18, 64-.19, and 76-24 64 20 64-(ck)

Suppression Chamber vacuum relief Suppression Chamber vacuum relicf Ai (1) s Ai (1)

Applied between 64-20 and 64-(ck)

Applied between 64-20 and 64-(ck) 64-21 64-(ck) 64-29 Suppression Chamber vacuum relief Suppression Chamber vacuum relief Drywell main exhaust 64-30 Drywell main exhaust 64-.31 Drywell exhaust to Standby 64-32 64-33 64-34 Suppression Chamber Main Exhaust Suppression Chamber Main Exhaust Suppression Chamber to Standby Gas Treatment Applied between 64-21 and 64-(ck)

Applied between 64-21 and 64-(ck)

Air( )

Applied between 64-29, 64-30, 64-32 64-33 and 84-19 Air Applied between 64-29, 64-30, 64-32; (1) 64-33 and 84-19 Air( 'pplied between 64-31,64-141, 84-20 (1) and 64-140 Aiq Applied between 64-32, 64-..33, 64-29, (1) 64-30 and 84-19 Air(1)

Applied between 64-32, 64-33, 64-29, 64-30 and 84-19 Air( )

Applied between 64-34,64-141 and 64-139 259 Apepdments Nos.

28

& 25

~,

1

~0 v

TjUlLE 3.7.D (Coatinucd)

Valvaa 90>>257A 90-2575 84-8A Valve Identification Radiation Monitor Diaeharga Radiation Hoaitor Discharge Containment Atmospheric Dilution Teat Nadiim Air<'>

Ai ")

Air Teat Macha'pplied betveea 90-2571 aad 90-257:,

Applied be~en 90-237A sad 90-257'pplied between 84-8A and 84MOO 84 BB 84-8C 84-8D 84-19 Containment Atmospheri.c Dilution Air Containment Atmosphex'ic Dilution Ai'r Containment Atmospheric Dilution Containment Atmospheric Dilution Air Applied between 84MB and ~l...

Appli.ed between 84-8C and 84-603 Applied between 84~ and 84-60$ :

Applied between 64-32, 64-33, 64-29, 64-3A', and 84-19 I

(1)

(2)

Air/nitrogen test to be displacement flower.

Rater test to be injection loss or downstream collection.

y ~

~

Valve Identification Test Method Applied between 76-215 and 76-218 Applied between'7

-217 and 76-218 Applied between 76-220 and 76-223 Applied bctvreen 76-222 and 76-223 Applied Letween 7G-225 nc'G-227 Applied between 76-226 and 76-227 Applied between 76-229 and 76-231 Applied between 76-230 and 76-231 Applied between 76-237 and 76-240 Appl.ed between 76-239 and 76-240 Applied between 76-242 and 76-244 Applied between 76-243 and 76-244 Applied between 76-248 and 76-253 Applied between 76-250 and 76-251 Applied between 76-253 and 76-255 Applied between 76-254 and 76-255 Applied between 84-20,64-141, 64-140, and 64-31 Applied between 84-8A and 84-600 Applied between 84-88 and 84-601 Applied between 84-80 and 84-603 hppli.ed between 848D and 84-602 hpplied between 64-141,64-140, 64-30, and 84-20 Applied between 64-141,64-140, 64-31, and 84-20 Applied between 64-139.,64-141, and 64-34 Test Bed ium Vnlves 76-? 15 76-217 76-220 76-222 76~~25 76 22G 7G-229 76-230 76M37 76-239 76-242 76M43 76-248 76-250 76 253 76-254 Air<>>

Air Air Mr Air Air Air Air Air Air Air Air Ai.x Air Air t Air(>>

Containment Atmospheric tlonitor Containment Atmospheric llonitor Containment Atmospheric tlonitor Containment Atmospheric Honitor Containment AtmospiU.ric iionitor Containment Atmospheric Monitor Containment Atmospheric 1lonitor Containment Atmospheric Monitor Containment Atmospheric Honitox Containment Atmospheric Monitor Containment Atmospheric Monitor Containment'tmospheric Monitor Containment Atmosph ric Honitor Containment-Atmospheric Monitor Containment Atmospheric Monitor Containment Atmospheric Wi nitor I

tlain Exhaust to Standby Gas Txeatmen

~ 1 Main Exhaust to Standby Gas Treatment Main Exhaust to Standby Gas Treatment Main Exhaust to Standby Gas Treatment ttain I'.xhaust to Standby Gas Treatment Drywel'1 Vressurization, Comp.

Bypass 84-20 Nitrogen(1)84-600 84-601

,84-602

'4-603 F64-l4l tVtroi.en tt4trogen t$itx'oqen Air (>)

,64-140 Drywell Pressurization, Comp. Disc.

earn collection.

262 Amendments Nos.

28 8

25

'64-139 Dr~~ell Pressurization, Comp. Suction (1) Air/nitrogen test to'be displacement flow (2) Mater test to be injection loss or downstr

CORRECTED PAGES TO AMENDMENT NO.

27 TO DPR-33 ANENDNENT NO. 24 TO DPR-52 DATED AUGUST 20 1976 Revise Appendix A as follows:

Remove the following pages:

36 44 54 89 thru 95 123 124,'143'44

, 145i 146I 150I

151I

'154' 185I 187 252 267 thru 270 286 295 296 322 326 332 333 337 346 349 350 354 356 357 and replace with identically numbered pages.

10'.

Hot required to be ope able when the reactor pressure vessel head is not bolted to the vessel.

'11.

The APRM downscale trip function is only ac"-ive when t:he reactor mode switch is in run.

12.

The APBM downscale t ip is automa i,cally bypassed when the 1RM instmxmentation is operable and not high.

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

10. Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Isolate.on Control System.

A channel failure may be a channel fa'lure in each system.

15.

The APBM 15'X scram is bypassed in the Run Mode.

16. Channel shared by Reactor Protection System and R actor Manual Cont ol System (Rod Block Portion).

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

17. Not" re<paired while pe forming low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MM(t).

'l8.

Operability is required

>Aen normal first-stage pressure iz beIov

(

154 Psig)

~

19.

20.

Action 1.A or 1.D shall be taken only if the permissive fails in such a manner to pxevent the affected RPS logic from performing its intended function.

Otherwise, no action is required.

An alarm setting of 1.5 times normal background at rated power sha11 be established to alert the operator to abnormal radiation levels in primary coolant.

36 Amendments Nos.

27 8 24

d~SES modes.

In the power range the APRM system provides zequ'zed protection.

Ref. Section

7. 5.7 CESAR.

Thus, the IRM System is not required in the, Run ~de.

The APRM's aad the IRM's pzovide adequate coverage ~ the startup and intermediate range.

The hiph reactor pressure,.

high dzyvell pressure, reactor lov water level and scram discharge volume high level scrams aze required for Starcup and Run modes of plant operation.

They are, therefore, required to be opera-tional for these modes of reactoz operation.

The requirement to have the scram functions as indicated in Table 3.1.1 operable ia'the Refuel mode is to assure that shifting to the Refuel mde during reactor povcz operation does not dimiaish the need for the reactor pzotection system.

The turbine condenser Iov vacuum scram is only zequired during povez operacion and must be bypassed co scarc up the unit.

Belov 154 psig tur-bine first stage presauze (30X of rated),

the sczam signal due to turbine stop valve closuze, turbine'ontzol valve fast closure, and tuzbiae coa-trol valve'oss of contzol oil pressure, is bypassed because flux sad pressuze scram are adequate to protect the reactor.

Because of the APRM downscale limit of

> 3Z vhea in the Run mode and high level limit of

<15X vhen in the Startup Mode, the transition bet"eea the Startup and Run Modes gust be made vith the APRN instzumcntatioa ind'cat'ag becveen 3X and 15X of rated power or a control zod scram vill occur.

In

addition, the IRM system must be indicating belov the High Plux setting (120/125 of scale) or a sczam vill occur vhen ia the Scaztup Mode.

Poz normal operating conditioas, these limits provide assuzaace of over3.ap betveen

the, IRM system and APRM system so that theze aze ao "gaps" a the pover level indications (i.e.,

the povez level is contiau'oualy monitored from beginning of scaztup to full paver aad from full pover to shutdovn),

Vhen power is being reduced, if a tzansfer to the Staztup mode is made and che IRM's have noc been fully inserted (a malopezatioaal but aot mpossibla condition) a control rod block i edistaly occurs so that react vity iassz" tion by control rod vithdraval caaaot occur.

Amendments Hos.

27

& 24

I

~ I 1

LXseeTXHG COhDXTEO';iS eOR OPKRAZTO~I SUB.'.T

'"tC:.. ".qb:"..-.

3.2.J'eiamc Mani prinz Tnst=men ation I

Set ~t.e

)fats Qted eeg t 1 1.

The seism=c run'tari ~ inst ments 1isted in table 3.2.Z sh~~

De

. Qaemb1e.at ~ times.

2.

Viith the neer ox seisw'c ~"'tar'~

~~s n erts 1ess tmn the neer 1is ed in tsb1e 3.2.Z, estore the Xnoaereb1e instru en (s) to onemb1e ststns d~ 30 ~s.

3.

Nth one or na e of the inst. ments usted 'n tsb1 3.2.Z inapemM.e o

mre then 30 ~+s, s~~

2, SaeciM Resort ta the Ccrr "ss'on pu~s~t to scecdddcedce 6.7.3. C uduhdc ""e c"+

10 ~s desc"d'b~

~".'.".e causa c';.e A@3.fu,ction a& p2.sns for r s o~~

the 'nst ments

".o oae"eble sts,.ns.

2.

72 Ch Q

toto So'd cn't ee <<~et'at 't see instr nts s'g~t De d

,Q ~

i

'ageable Dy nerfo~~-c o

tes s

e.t the f eonenc's

~'sted in table 4.2.Z.

R3 se s

c inst~ 't s 8.ctn te d~~

2 seis~~ c ev&~ 2~8. 2~a Jze tQ determines

+~1ts t A c d ~Ass Q.

P. aaw sh~" ~ be sub='e'o the CQM~~ ss~ Qn QU sl'"n Q ~~~c'Rt Qn 6.7.3.D 'edeeds 10 days de-"- 'ud.~

C'le c" s '*'2" "ct'e c'J'cec"M veStgs tered est 'teC> nt~~q p'7 tt tt+

fes tUZ'es ~~~no~~t~t to M~ctj e Amendments Nos.

27 5 24

PAGES DELETED 89 thru 95 Amendments Nos.

27 IN 24

LIHITIHG CONDITIONS FOR OPERATION SURV. ILLAHCZ RZ. VIREOS

".3. 8 Control Rods be During the shutdown procedure no rod movement is permitted between the testing performed above 20'o power and the rein; statement of the RSCS re-straints at or above 20~a power.

Alignment of rod groups shall be accomplished prior to performing the tests.

c.

Whenever the zeactor is in the startup oz run nodes below 20% rated paver the Rod North Hin~~izer'hall be operable oz a second licensed operator shall verify that the operator at the reactor console is fallowing the control rod program.

4.3 '

Control Rods The capability of the RSCS ta pro-perly fulfillits function shall be verified by the rollow'rg tests:

Sequence portion - Select a sequence and attempt to withdraw a rod in tha remaining sequences.

Move one rod in a sequence and select the z~~ in-ing sequences ard atte pt to move a rad in eacn.

Repeat for a11 sequences.

Group notch ao "'on Zor each or the six comparator circuits go through test initiate; compazator irhioit; verify; reset.

On seventh atte=pt test is allowed to continue until completion is indicated by ill~~tion of t st cauplete. light.

b.

The capability of the Rod North Hiniai"er ('X~A) shall hs sac< 44 sA Lee

~ 0 C'

~ iiai svgiorg checks:

The correctness o" the contzol rod vithdzawal sequence input to tha'kH computer shall be verified berore reactor startup oz shutdown.

If Specifications 3.3.3.3.a thraugh.c cannot be net the reactor shall nat be started,

'oz if the reactor is in the run or startup" mades at less than 20% zatad power, it shall be brought to a shut-down condition ir=edixtaly.

123 2.

The RHH computer on line diagnostic test sha'1 ae successfully perf armed.

3.

Prior to startup, proper annunciation or the selec-tion er.ar of at least ane aut-of-sequerce contra'ad shall be verified.

4.

Prior to star up, the "cd block function of the P.'~~i shall oe ve i~ied by =ov"ng an aut-or-sequence cart"o zad.

5.

Prior to obtaining 20% rated power during rad inse tian at shutdown, ve 'fy latch""g ar the proper rac group and proper a nunciatian sitar irsert e

ors.

Amendments Hos.

27 5 24

L" "TktlG CO.tO'BUTTONS POR OPERATION SURVEELLAhCE RE UTRE!~'.lTS 3'.3.B'ontrol Rods 4.3.B Control Rods 4.

Control rods shall not be withdrawn for startup or refueling unless at least tan source range channels have an observed count rate equal to or greater than

~three counts per econd..

a.

Both RBN channels shall be operable:

or b.

Control rod withdrawe3.

shall be blocked:

5..Durin"operation with limiting control.rod pat-

terns, as determined by the designated qualified person-nel, either:

4 5.

When requir d, th presence of e second l'censed operator to verify th foll~ing of the correct rod program s~l3.

be verified.

Prior to contro3. rod withdrawal for startup or duri..g refueling, verify that at least two source range channels have an observed count rate of at least three counts per second.

%hen a limiting control rod pattern exists, an inst~ ent functional test of the P~A shall oe pe fo~ d p.ior to withdrawal of the designa ed rod(s) and at least; once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

~g C.

Scram Tnsertion Times C.

Sc an Tnsertfon T'mes

<<nscr>>>>d c<<om T

~ <<J Pu3."d--'m Avg. Scram Tnser-

-'n T < -...es (sec) 5 20 50 90 0.375 0.90 2.0 5.0 The average scram insertion time, based on the deenergi-zation of the scram pilot valve soleno'ds as ti=

zero, of all operable control rods 'n th reactor power operas'on condi-tion shall be no greater than:

After each refueling ou age all operable rods shall be scram ti=,e tested from the fully w'thdra -..

position with the n c3.ear syste..

pressure above 950 ps'g ('wth saturat" on taupe acute)

I T. is test g

shall oe, comp eted prior to ezc 'ding and B34) wh ch wer>>>> ' y withdrawn in the region from lOOZ

>rod dens ty'to 505 rod density sha be scram t'."..e testee.

During all sc"am t.-e t st-'-.

below 20.';:o-e" the 3,~~ shall be operabl Amendments

Nos, 27

& 24

7-> fITZHG CONDITIONS FOR OP RATTON SURVETLLANCZ RKQUTR HKNTS 3.5 CORK AND CONTAKls..""NT COOLING SYST HS 4.5 CORE A?H) CORRAL'?Pi{""NT COOL""iiG SYSTEMS Aoolicabili.t Aoolicabilit Applies to the operational status of the core and contain-ment cooling systems.

Applies to the surveillance requirements of the core and containment cooling systems when the corresponding Limiting condi-tion for operation ia in affect.

~Ob ca r Sv e OSiecaive To assure the operability of the core and containment cooling systems under all conditions for which this cooling capability is

.an essential response to plant abnormalities.

To verify the operabili y of the core and contain nt cooling systems under all conditions for which this cooling capability is an esschtial response to plant abnormal'ti se Soecification S ecification A.

Core Sore-System (CSS)

.1.

The CSS shall be opera-ble:

(1) prior to reactor startup from a cold condition, or (2) when there is irra-diated fuel in the vessel and when the rcac or vessel. pres-sure is greater than atmospheric

prcssure, except as specified in specifications 3.5.A.2, 3.5.3.2, or 3.9, B. 3.

Item a.

Simulated Automatic Actuation teat

~aauaaa Once/

Opera ting Cycle b.

Pump Opera-.

Onc /

'oility anth c.

Hotor Onc /

Operat d

month Valve Operability d.

System flow ratae Each loop shall deliver at lease 6250'pm against a system head corr s-ponding to a Onc /3 mon ths 1.

Core Spray System Testinge 143 Amendments Nos.

27

& 24

2 ~

3 ~

'i( Uns CSB loop ia inopera-

ble, the reactor may remain in operation for a period.

not to exceed 7 days provi-ding all active components in the other CSB loop and the RHR system (LPCX mode) and the diesel generators are operable.

Lt'pecification 3.5.A.l or specification 3.5.A.2 cannot:

be met, the reactor shall be shutdown in the Cold Condition within 24 ho'urs

~

When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop

. with one operable pump and associated diesel generator shall be operable, except vith the reactor vessel head, removed as specified in 3.5.A.5 or prior'o reactor startup as speci ied in 3.5.A.1.

TZMZTZNG CONDZTZONS FOR OP~TZON 3.5.h Coro Borav 8 stem CSS)

SUR~rZILANCZ RZQUZ~~PXS 4,5,A Core Sara S stem CSS) 105 psi dif-ferential pressure between the reactor ves-seI., end the primary con-tainment.

e.

Check. Valve Once/

Operating Cycle 2.

When it is determined that one core spray 1oop is inoperabl'e, at a time when operability is

reqoired, the other core spray loop, the RHRS (LPCl mode),

and the diesel generators shall be demonstrated to be operable immediately.

The operable core spray loop shall be demonstrated to be operable daily thereafter.

5 ~

'Men irradiated. fuel is in the reactor vessel and. the reactor vessel head. is removed, core spray is not recused

provided, vora's not in prog.ess vhich has the potential to drain the

vessel, provided, the fuel pool gates a

e open and the fuel pool is maintained, above the lov level alarm point, and. provided one HERSE pump and associated.

valves supplying the standby coolant supply are operable.

144 Amendments Hos.

27 8 24

LDKTIIfGCOt(DZT'ZOOS POR OPEcUTZON

-. 3. 5. 8 Res i(junl Heat Removal Svo tern

~RHRS (LPCI and Contain~at Coal ititt)

SUHVEILLN(CZ BZQUZHZ~PZS 4.5PB Restdual Heat Removal S stem

~RHRS (LPCI and Contain=ant Caaling)

The RHRS shall be opexable:

{1) prior to a reactor scarcup fran a Cold Condition; or (2) when there is irra diated fuel in the x'eactor vessel and when the reactor vesseL pres-sure is greater than atmospheric, except as ape'cified in specifica-tions 3.5.B.2, through 3.5.8.7 and 3.9.B.3.

1.

a, Simulated Automatic Actuation Test b.

Pump Opera<<

bility c.

Hotor Opera-ted vaLve operability d.

Pump PLow Race once/

Opera ting Cycle Once/.

r(anth Once/

month Once/3 esontha 2.

With the reactor vessel pr s-surc less chan l05

psig, the PHRS msy be removed from ser-vice {except that twa RHR pumps-containment, cooling made and associated heat.

cxchangers must remain aperablc) for a pex'iod nat to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while being dranned of suporession chamber quality water and

~ filled with primary coolant quality water provided that duxing cooldo~m two loops N th one pump per loop or one loop with cwo pumps, and associated diesel generators, in the core spray syste are operable.

~

3.

Lf one RHK pump (LPCI node) is inoperahLe.'the reactor may remain in operation for a period not co exceed 7

days provided the remaining RHR pumps (LPCI mode) and both access pacha of the RHRS (LPCI made) and the CSS and the diesel generators remain operable.

145 e.

Test Check Valve Once/

Operating Cycle Each LPCZ pump shah'e3.iver 9,000 gpm against an indicated system pressme of ~&5 psig.

Two LOCI PumPs in the same looP shaly deliver 15,000

<~m against an indicate(i system pressu e of 200 psig.

2.

An air test on the dryweLL and torus hcade."8 and no=ales shaLL be conducted once/5 years.

A water test may be performed on the torus heade.

in lieu of the air test.

3.

When it is determined that one PHR pump (LPCI mode) is inoperable ac time when operability is rcqu'-.e'-.

che remaining RM pumps (LPCI node'nd active components

'n both aces" paths of the RHRS

{LPCI mode) and the CSS and the diesel generators shall be de=onstrated to be opera-ble i~ediately.

The operab'e RHR:

pumps

{LPCI node) shall be do. on-strated to bc operable every 10 da thax'eafter until che inoperable pump is returned to nax se v c Amendments Nos.

27 8 24

PAGE DELETED amendments Nos.

27

& 24

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS inoperability, pipe break, etc), the reactor may remain in operati on for a peri od not to exceed 30 days provided the remaining RHR pump and associated diesel generator are operable.

13. If RHR cross-connection flow or heat removal capability is lost, the unit may remain in operation for a period not to exceed 10 days unless such capability is res tored.

14.

All recirculation pump discharge valves shall be operable prior to reactor startup (or closed if permitted elsewhere in these specifications).

a time when operability is required, the re-maining RHR pump and associated heat exchanger on the unit cross-connec-tion and the associated diesel generator shall be demonstrated to be oper-able immediately and every 15 days thereafter until the inoperable pump and associated heat exchanger are returned to normal service.

12.

All recirculation pump discharge valves shall be tested for operability during any period of reactor cold shutdown exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if operability tests have not been performed during the preceding 31 days, 150 Amendments

Nos, 27 8 24

1 I

LI!<ITINC CONDITIONS FOR OPERATION SURVEILLANCE REOUIREii~i TS 3.5.C RHR Service Water and Emeraenc E ui ment Coolin Water S stems 1.

Prior to reactor startup from a cold condition,,9 RHRSW pumps must be operable, with 7 pumps (includ-ing pump Dl or D2 for unit 1 and one of pumps Dl, D2, Bl, or B2 for unit 2) assigned to RHRSW service and 2 automatically starting pumps assigned to EECW service.

4.5.C RHR Service Water and E er encv E ui ment Coolin Water S stems (EECVS >

1.

a.

Each of the RHRSW pumps normally assigned to automatic service on the EECW headers will be tested automatically each time the diesel generators are tested.

~ Each of the RHRSW pumps and all associated essential control valves for the EECW headers and RHR heat evchanger headers shall be demon-strated to be operable once every three months.

b; Annually each RHRSW pump shall be ilow-rate, tested.

To be considered

operable, each pump shall pump at least 4500 gpm through its nozmally assigned flow path.

151 gmengments

Nos, 27 5 24

LIHITT,'lC Cni~~ITTONS POR OPERATION

/'.5.5 E ui ment Area Coolers SURVEILLANCE REOUTBMi.iTS 4.5iD E uf.a cnt Azcc Coolers 1.

The equipment area cooler associated vith each RK~

pomp and the equip=cr.t area coo lez assoc ia tcd vith ecch set of core sprav pumps

{.<.

and C or 8 and D) must be opezable at all times vhen the pump oz'umps se~ed by that specific cooler is considered to be operable.

l.

Each equ'pment area cooler is operated in con)uncticn vith the equipment served by that particular cooler; therefore, the equip-nt area coolers aze tested at t¹ same frequency as the pumps vhich they serve.'

2.

linen an equip=ant a

e cooler is not operable, the pump(s) served by that cool r must be considered inoperable for Technical Specification pur-poses.

E.

Hizh Pressure Coolant Intection S stem H?CIS)

Z.

High Prcssure Coolant Tniection 1.

The HPCI system shall be operable:

e 1.

HPCX Subsystem testing shall be per'a~ad as iollovs:

(1) prior to staztup from z Cold Condition; or (2) vhenever there is irra-diated fuel in'he reac-tor vessel and the reactoz vessel pressure is greater than 122. pslg, except aa specified in specifica-tion 3.5.E.2.

a.

SimuLated

~

Auto-atic

'ctuation Test C ~

.totor Operated Valve Opeza-b'lity be Pu.0 Opera bility Once/

operating cyc'e Once/

=anth Once/

month

d. - Plov Rate at normal reactoz vesseL opera-ting pre..sure Once/3 no=the e.

PLov Rate at 150 paid Onc /

cycle The HPCi pu=p shall de' rer at least 5000 Spy cuzir~

each =lov

". te t st.

154 Amendments Nos.

27 In, 24.

~

~

~

LIPXTTNG CONDXTXONS FOR'PKPATXON SURV" XLLANCZ RZQUXRZ+"NTS 3

6 PRIMA,RY SYST H

BOUNDARY 6

PRX >WRY SYST "~!:"OUND~Y Shock Suooressors Snubbers H.

Shock Suppressors C'Snub5ers During all modes of

,operation except. Cold Shutdown and Refuel, all safety-related snubbers sha3.1 be operabl.e except as noted in 3.6.H.2 through 3.6.H. 5 below.

The following surve'llance requirements apply to all hydraulic snubbers listed in 3. 6.H. 2.

All hydraulic snubbers whose seal material has been demons ated by ope ating experience, l.ab testing or analysis to be comoa t'le witn the operat'ng environment shall be visually inspected.

This inspec tion sha13.

include, but not necessarily be limited to, inspection o= the hydraulic fluid reservoir, f3.u' connec ions, and 1'kage connect'ons to the pip'g and ancho to ver' their operability in accordance wi"h the fol.lowing schedule:

Number of Snubbers Pound Inoper-able During Inspection or During Inspec-tion Interval Ne:c Requirec Xnsp ction Interva3.

1 2

3,4 5,6,7 Ooera ting

.12 months 6 months 124 days 62 days

+25~

+25~

+25%

Amendments Nos.

27 8 24 185

>Q 31 days

+25%

The required inspection intervai shall not be lengthened more than one step at a time.

~

~

e

~ 4 (MlTIl4%BY SYSTEM BOUNDARY 6

PRIMARy SYSTEM'1 BOUNDARY If tne requirements of 3. 6. H. i anci 3.6.H.3 cannot be

met, an orderly shutdown shall be initiated and the reactor shall be in a cold shutd~

condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

lf a snuo6er ts determined to be in-operable while the reactor is in tlie sfiutdown or refuel

mode, the snubber shall be made operable or replaced prior to reactor startup.

Snubbers may be added to safety related systems without. orior license amendment to Table 3.6.H provided that a revision to Table,3.6.H is included with a subsequent license amendment request.

Once each refueling cycle, a representative sample of 10 snubbers or approximately 10~ of the snubbers, whichever is less, shall be functionally tested for operability including verification of proper piston movement, lock up and bleed.

For each unit and subsequent unit found inoperable, an additional 10 or ten snubbers shall be so tested until no more failures are found or all units have been tested.

Snubbers of rated capacity greater than 50,000 lb need not be functionally tested.

187 Amendments Nos.

27 8 24

t

~Grou Valve Identification TABLE 3,7,A (Continued}

(umber of Power 0 crated Valves lnboar Outboard Maximum Action on Operating Hormal Initiating Time sec.

Position

~Si nal Suppression Chamber purge inlet

( FCV-64-19)

Drywel1/Suppressi on Chamber nitro-gen purge inlet (FCV-76-17)

Drywell Exhaust Valve Bypass to Standby Gas Treatment System (FCV-64-31)

Suppression Chamber Exhaust Valve Bypass to Standby Gas Treatment Sys tern (FCV-64-34) 100 10 10 10 SC SC SC SC S

I C+

7 O

cia v

8 RCIC Steamlind'rain (FCV-71-6A, 6B)

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

HPCI Hotwell pump discharge isola-tion valves

( FCV-73-17A, 178)

HPCI steamline drain (FCV-75-57, 58)

TIP Guide Tubes (5) 1 per guide NA tube 0

GC GC SC GC GC

PAGE DELETED 267 Amendments Hos'7

& 24

t

~ ~

W BASES

. 3.7.A 8 4. 7.A Primar Containment The integrity of the primary containment and operation of the core standby cooling system in combination, limit the off-site doses to values less than those sQggested in 10 CFR 100 in the event of a break in the primary system piping.

Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists.

Concern about such a violation exists when-ever the reactor is critical and above atmospheric pressure.

An exception is made to this requ'.'rement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required.

There will be no pressure on the 'system at this time, thus greatly reducing the chances of a pipe break.

The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occur ring.

Procedures and the Rod 'r(orth Minimizer would limit control worth such that a

rod drop would not result in any fuel damage.

In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment

system, which shall be operational during this time,- offer a sufficient barrier to keep offsite doses well below 10 CFR 100 limits.

The pressure suppression pool water proviges the heat sink for the reactor primary system energy release following a postulated rupture o

the system.

The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1,035 psig.

Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the Iiquid must not exceed 62 psig, the suppression chamber maximum pressure.

The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water volumes given in the specification, containment pressure during the design basis accident is approximately 49 psig which is below 'the maximum of 62 psig.

Maximum water volume of 135,000 ft3 results in a downcomer submergence of 5'2-3/32" and the minimum volume of 123,000 ft3 results in submergence approx imtely 12 inches less.

The majority of the Bodega tests were run with a submerged length of 4 feet and with complete co'ndensation, Thus, with respect to downcomer submergence, this specj ication is adequate, The maximum temperature at the end of blowdown 'tested during the Humoolt Say and Bodega Bay tes s was 170'F and'his is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170'F.

288 Amendments Hos.

27 5 24

BASES Should it be necessary to drain the suppression

chamber, this should onl'y be done when there is no requirement for core standby cooling systems operatibility.

JJnder full power operation conditions, blowdown from an initial suppression chamber water temperature of 95'F results in a peak long term water temperature of 170'F which is sufficient, for complete condensation.

At this temperature and atmospheric

pressure, the available iVPSH exceeds that required by both the RHR and core spray

pumps, thus there is not dependency on containment overpressure.

Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160 F during any period of relief valve operation with sonic conditions at the discharge exit.

Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressuirzed in a timely manner to avoid the regime of potentially high suppression chamber loadings.

Limiting suppression pool temperature to 105'F during RCIC, HPCI, or relief valve oper ation when decay heat and stored energy is removed from the primary system by discharging reactor steam directly to the suppression chamber assures adequate margin for controlled blowdown anytime during RCIC operation and assures margin for complete condensation of steam from the design basis loss-of-coolant accident.

In addition to the limits on temperature of'he suppression 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:

(I) use of all available means to close the valve, (2) initiate suporession pool water cooling heat exchangers (3) initiate reactor

shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.

If a loss-of-coolant accident were to occur when the reactor water temperature is below approximately 330'F, the containment pressure will not exceed the 62 psig code permissible pressures even if no condensation were to occur.

The maximum allowable pool temperature, whenever the reactor is above 212'F, shall be governed by this specification.

Thus, specifying water volume-temperature requirements applicable for reactor-water temperature above 212'F provides additional margin above that available at 330'F.

~Inerti n The relatively small containment volume inherent in the GE-BMR pressure suppression containment and the large amount of zirconium in the co"e are such that the occurrence of a very limited (a percent or so) reaction o.

the zirconium and steam during a loss-of-coolant accident could lead to the liberation of hydrogen combined wi th an air atmosphere to result in a flarrmable concentration in the containment.

If a sufficient amount of hydrogen is generated and oxygen is available in stoichicme.ric quantities the subsequent ignition of the hydrogen in rapid recombination rate could lead to failure of the containment to maintain a low leakage ntegrity.

Tl e 4% oxygen concentration minimizes the possibility of hydrogen combust.'on following a loss-of-coolant accident.

269 Amendments

Nos, 27 8 24

BASES The occurrence of primary system leakage following a majo~ refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-of=coolant accident upon which the specified oxygen concentration limit is based.

Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety.

Thus, to preclude the possibility of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled during startup periods, when the primary system is at or near rated operating temperature and pressure.

The 24-hour period to provide inerting is judged to be sufficient to perform the leak inspection and establish the required oxygen concentration.'o ensure that the oxygen concentration does not exceed 4'~ following an

'ccident, liquid nitrogen is maintained on-site for containment atmosphere dilution.

About 2260 gallons would be sufficient as a 7-day supply, and replenishment facilities can deliver liquid nitrogen to the site within one day; therefore, a requirement of 2500 gallons is conservative.

Following a loss of coolant accident the Containment Air B'fonitoring (CAN) System continuously monitors the oxygen and hydrogen concentration of the containment volume.

Two independent systems

(

a system consists of one oxygen and one hydrogen sensing circuit) are installed in the drywell and one system is installed in the torus.

Each sensor and associated circuit is periodically checked by a calibration gas to verify operation.

Failure of a drywell system does not reduce the ability to monitor system atmosphere as a second independent and redundant system will still be operable.

Failure of the torus system would require a reactor shutdown as no means would be available under accident conditions to monitor torus atmosphere.

Until a redundant system becomes available in the torus, the monitoring requirements of either a hydrogen or oxygen sensing circuit will be utilized.

While this reduces the offered protection slightly, one sensor can be used to prevent a combustible atmosphere In additition the torus atmosphere will be mixed with the drywell atmosphere through the drywell to torus check valves and any increase in the torus hydrogen or oxygen concentration would proportionally change the drywell atmosphere.

270 Amendments Nos.

27 8 24

1

T i'll'I(NC CONO

. TONS cOR OPK?ATIONS SUPVc.ILL;AC:: ? KC J~ ~F'TS C.

Hechanical Vacuum Fu=a Q.

Meehan'czl'Vacuu=

Pv-a The nechanica! vacu~ p~p shall be capable of being automatically isalated and secured on a signal aE 'high radioactivity in,the stea=

lines Mhcnever the cain stean isalatian valves are open.

IE th liMts af 3. 8. C. 1 are nat net, the vacuu.- pu=p shall be isolated.

~

~

D.

Lfisccllanenus %dig=et va Yiaterials Sacr "as.

Source I, azace Test I

I Each sealed source-containing I

radioactive material in exccssl'f those 'quantitics of byprod-uct mate ial 'isted in ~ 10 CFR 30.71 Schecule B and all other

sources, including alpha

emitters, in excess of O.l microcurie, shall be free of

> 0.005 microcu.i.e o

remov-able contarinati,on.

"=ach sealed sou=c with removable contamination in excess of the above limit shall be immediately withdrawn from use and (a) eith " decon-taminated and repai"ed, or (b) disposed of in accordance with Commission regulations.

At 'mesc ance during each operating cycle vcr'ify auto"atic securing and isolation aE the mechanical vacu n pucp i O.

Miscellaneous 3 c.cnctive l!a eri 1s Sources Surv~il'ance

? ccu'c.-.cn Tes=s;or 3.caka;c an"/ar ccn'-a.".wast-n shaM~

be per"ar.-.ed

'oy t'ao lican ee or by o he.

per sans saeciE'ica lv authari=e=

by ha Co-.~issicn or a,. agreement State,,

as fo ~ lo"s:

ch saa 'd sou ca) axcett sta sources s'jcct to core flux, cant" wing radioactivo Tcatarial, athar than Hydra;en 3,::it., a h3 liSa greater than thirty d=':s and in =ny 'rm other than gas s.".all be tasted for lee.ka~~ and/ar contamination a" intervals ro ta exceed six months.

Thc lc kage

-.est shall be capab'e of detecting

.he pr senc o

0.005 microcuri of radioactive materia'n the tost sample.

b.

The periodic lea< test reouired

'oes not apply ta sea d sour"cs that sre stored and not be'~~

used, The sourc s

xceptec this tes s"..al be =ested

=a.

leakace p-..icr o:.nv "se a" rans.-.c-.

to another user unl ss they have baetf leak test d th'". s~- ;.cnt..s "a the absence ci a

ert icaticn

!rom a trans."arar

='-.dicat

".g th a

est has cee..

made w

..'". six months pr'or ta the tr=-",sf ".,

scale" sou"ces s~a'~ nct be put Startua sources she'e leak tested prior ta ~a~

a a'>

.g an) repair or r".tenance anc - 'ore be ".g subgec ed :o core fl"-.

286 Ci pen)!TIentq Nqs.,

27 8 24

LIMITIHg CONDITIONS FOR OPERATION SURVEILLANCE REOUIR~ZNTS 3.9.8 Operation vith Inoperable Eccukoa~en t Whenever a reactor is ia Staz'tup mode or Run mode and noc in a cold condition, the avaiiabilfty of electric pover shall be as speci-fied in 3.9.A, except as specified herein.

4.9.B Operation with Inoperable Eauiamenr.

1.

Prom and after the date that oae 161-kV line'or one common station transformer and its parallel cooling tower trans-former or one*scarc bus becomes inoperable, reactor operation is.

permissible under ehis condition for seven days.

2.

When one of the units 1 aad 2

diesel generacor is inoperable, continued reactor operacion is permissible during the succeeding 7 days, provided thac both off-site 161-kV transmission lines and boch common station trans-formers or one common transformer and oae cooling tover transformer (noe parallel vich the energized common transformer) are avail-

able, and all of che CS,. KK'(LPCI and Containment Cooling) Systems, and the remainiag three units 1

and 2 diesel generators are operable.

If this requirement cannot be met, an order'y shuc-dovn shall be initiated and both zeactors shall be shutdown and in ehe cold condition vithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

When one 161-kV line oz'ne common station transformer and its parallel cooling tower transformer or one scarc bus is found to be inoperable, all units 1 and 2 diesel generacors and associaced boards must be demonstrated co be operable immediately. and daily there after.';

When one oz the units 1 and 2

.diesel generator'is found to be inoperable, all of the CS, RHR (LPCI and Containment Cooling)

.Systems ahd the remaining diesel

~ generators and associated boards shall be demonstrated to be operable immediately and daily thereazter.

Amendments Nos.

27 8 24

Sl)RVZIl.LAVCF; REnu IR'~EWTS

). h 0~eral, ion Mtth lnoacrable

~Eel mene 4.9.8 Oacration with Inoperable Eccutoment

'Men ane un'ts 1 and 2 4-kV shutdown ooard is inoperable, continued reactor operation is permissiale for.a period not to exceed 5 days,

provided, that bath off-site 161-xV tr'ansmission lines nd both common stat'an transformers or ane common ransformer and one cooling tower transfor...er (not po"elle ath the energized common transformer) are availaa3.e and the, remain-'ng 4-RV shutdown boards an" associated d'esel gen ra ors, CS, RHR (I CZ and Contair ent Ccoling) Systems, a>>d eM~

460 V emergency powe boards a"

ape ab'.

his recuirement cannot be ret, an orderly shut-down sh&3. be in't'ated and both reactors sha'~

ce s>u da~z and in the cold condition WtBin 24 hau"s.

From and af"er the date that ane of the three 250>>Vol" unit ba-tories and/or its associated battery board is found ta be inopcraale for anv reason, continued reactar operaticn is permissible curing the succeeding seven days.

Except for routine surve llnnce testing tne HRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the s 'uation, the precautions to be taken during this period and the plans to return the failed component to an operable state.

3.

%hen one 4-kv shutdown board is found to be inoperable, all remaining 4-kv shutdown banters and associated d'ese'enera-

tors, CS and RPZ (LPCI and Containment Coo1.ing)

Systems supplies by the remaining 4-kv shutdown boards shall be demon-s tra ted to be operable, i...e-diatcly and daily thereafter.

5.

From and after the date that one of the faur 250-volt shutdown 296 Amendmeqts los.

27 8I 2C

LIHlTING CONDITIONS FOR OPEPATION SURVEILLANCE REQUIRE."ANTS 3-11 FIRE PROTECTION SYSTEMS Q. 1 1 FIRE PROTECTZON SYSTEMS 1.

An independent fire protection and loss prevention inspection and audit shall be

., performed annually utilizing either qualified TVA personnel o-an outside fire protection f'm.

2.

An inspection and audit by an outside qualified fire consultant will be performed at intervals no greater than 3 years.

(The first inspection and audit will be during the period of g~ne September 1977. )

If it becomes necessary'to breach a fire stop, an attendant shall be posted on each aide of the open penetrativn until work is completed and the penetration is resealed.

E.

Fire Protection S stem Ins ection E

Fire Protection Svs ems Insnection Any xnspecta.on or audit will review and evaluate the effectiveness of fi e prevention and, protection by physical inspect'on of plant. facilities, sys

ems, and equipment as related to fire safety.

Evaluations will bo made of, but not -nece ssarily limited to, the following:

Administrative control documentation, maintenance of fire related. records, physical plant inspection, relat d histor'cal research and application, and management interviews.

Gi The minimum in-plant fire protection oraanization and duties shall be as depicted in Figure 6.3-1.

322 amendments

Nos, 27

& 24

3.11 BASES The High Pressure Fire and CO> Fire Protection sp'ecifications are provided in order to meet the preestablished levels of operability during a fire in either or all of the three units.

Requiring a patrolling fire watch with portable fire equipment if the automatic initiation is lost will provide

{as does the automatic system) for early reporting and immediate fire fighting capability in the event, of a fire occurrence.

The High pressure Fire Protection System is supplied by three pumps aligned to the high pressure fire header.

The reactors may remain in operation for a period not to exceed 7 days if two pumps are out of service.

):f at least two pumps are not made ope able in seven days or if all pumps are lost during his seven day period, the reactors will be placed in the cold shutdown condition within 20 nours.

For the areas of applicability, the, fire protection water dis ribution system minimum capacity of 2664 gpm at 250~

head at the fire pump discnarge consists of the following design loads:

1 0 2 ~

3.

Sprinkler System

{0.30 gpm/ft~/4440 fthm ar ea) 1 1/2" Hand Hose Lines Raw Service Water Load TOTAL 1332 gpm 200 cpm 1132 anm 2660 apm The CO~ Fire Protection System is consid red op rable with a minimum of 8 1/2 tons (0.5 tank)

COq in storage for units 1 and 2;

and a minimum of 3 tons

{0.5 tank)

CO> in storage for unit 3-An immediate and continuous ir watch in the cable spreading room or any diesel generator building area will be established if CO< fire protection is lost in this room and will continue until CO~, fire protection is restored.

To assure close su.ervision of fire protection syst m activities, the removal from service of any component in either the H'ah

'Pressure Fire System or the CO~ Fire Protection System or any reason other than testing or emergency operations will require Plant Superintendent approval.

Early reporting and immediate fire fighting capability in the event of a fire occurrence will b~~ provided (as with the automatic system) by reauiring a patrolling fire watch '

more than one detector for a given protected zone is inoperable.

A roving ire watch for areas in which automatic fire suppression systems are to be installed will provide additional interim fire protection for areas that have been determined to need ada'ional pro ection.

326 Amendments Hos.

27 8 24

e 6. 0 ADMINISTRATIVE CONTROLS 6.

1 Or anizatian B ~

C.

D.

The plant superintendent has on-site responsibi 1ity for the safe operation of the facility and shall report to the Chief, Nuclear Generation Branch.

In the absence of the plant superintendent, the assistant superintendnet will assume his responsibilities.

The portion of TVA management which r 1ates to the operation of the plant is shown in Figure 6. 1-1 ~

The functiona 1 organizati on for the operation of the station shall be as. shown in Figure 6. 1-2.

Shift manning requirements

shall, as a minimum, be as described in section 6-8.

E G

H.

Qualifications of the Browns Ferry Nuclear Plant management and operating staff shall meet the minimum acceptable levels as described in ANSI - N 1 8. 1, Selection and Training of Nuclear Power Plant Personne l, dated Harch 8, 1 97 1 Retraining and replacement training of station personnel shall be in accordance with ANSI -

N1 8. 1, Selection and Training of Nuclear Power Plant Personnel, dated March 8,

1 97 1.

The minimum frequency of the retraining program shall be every two years.

An Industrial Security Program shall be maintained for the life of the plant.

Responsibi 1ities of a post.-fire overall restoration coordinator will consist of duties as described in section 6. 9 I ~

The Safety Engineer shall have the following qualifications:

a.

Must have a sound understanding and thorough techni cal knowledge of safety and fire protection practices, procedures, standards, and other codes relating to electrical uti I ity operations Must be able to read and understand engi neeri ng drawings.

Must possess an analyti cal ability for problem solving and data analysis.

Must be able to communicate well both ora 1 ly and in writing and must be able to write investigative reports and prepare written procedures, ~ Must have the abi 1 ity to secure the cooperati on of management, employees and groups in the imp 1 ementati on of safety programs, Must be able to conduct sa fety pres entati ons for supervisors and employees b.

Should have experience in safety engineering work at this 1 evel or have 3 years experience in safety and/or fire protection engineering.

It is des irabl e that the i ncumbent be a graduate of an accredi ted college or uni versi ty wi th a degree in inductri al, mechani ca 1, electri ca 1, or safety engineering or fire protection engineering 332 amendments Nos.

27 5 24

~0 6

0 ADMINISTRATIVE CONTROLS 6.2 Review and Audit

. The Manager of Power is responsible for the safe operation of all TVA power plants,-including the Browns Ferry Nuclear Plant.

The functional organization for Review and Audit is shown in Figure 6.2-1.

Organizational units for the review of facility operation shall be constituted and have the responsibilities and authorities listed below.

Ai Nuclear Safet Review Board NSRB 1.

Membershi The NSRB shall consist of a chairman and at least five other members appointed or approved by the Manager of Power.

A m'ajority of the members shall be independent of the Division of Power Production.

The qualifications of members shall meet the xequirements of ANSI Standard'N18.7-1972.

Membership shall include at least.

one outside consultant and representatives of the following TVA organizations:

Office of Engineering Design and Construction; Division of Environmental Planning; Division of Power Production; Division of Power Resource Planning.

An alternate chairman may be designated by the chairman or, in his absence or inCapacity, may be selected by the NSRB.

The NSRB chairman shall appoint a secretary.

2.

Minimum Meetin Fre uen The NSRB shall meet at least quarterly and at more frequent intervals at the call of the chairman, as xequixedo 3 ~

uorum A quorum shall consist of four members, a minority of which shall be from the Division of Power Production..

4 ~

Responsibilities a.

Review proposed tests an'd experiments, and their results, when such tests or experiments may constitute an unrlviewed safety question as defined in Section 50.59, Part 50, Title 10, Code of Federal Regulations.

b.

Review proposed changes to equipment, systems or procedures, which are described in the Final Safety Analysis Report or which may involve an unreviewed safety question, as defined in Section 50.59, Part 50, Title 10, Code of Federal Regulations, or which are referred by the operating organization.

C ~

Review proposed changes to Technical Specifications or licenses.

333 A)gendrrlents Nos.

27 8 24 I

6 0

ADi4fIN'CSTRATIVE CONTROLS Review adequacy of employee training programs and recommend change.

5.

~Autharit The PORC shall be advisory to the plant superintendent.

6, Records Minutes shall be kept foz'll PORC meetings with copies sent to Director, Power Production; Chief, Nuclear Generation Branch;

Chairman, NSRB.

7.

Procedur es written administrative procedures for committee operation shall be prepared and maintained describing the method for submission and content of presentations to the committee, review and approval by members of committee actions, dissemination of

minutes, agenda and scheduling of meetings.

C.

alit Assurance and Audit Staf The Office of Power Quality Assurance and Audit Stazf (QAGAS) shall formally audit operation of the nuclear plant.

Audits of selected aspects of plant operations shall be conducted on a frequency commensurate with their safety significarlce and in such a manner as to assure that an audit of safety-related activities is completed within a period oz two years.

The audits shall be performed in accordance with appropriate wzitten instructicns or.procedures and should include verification oz compliance with internal

rules, proceduzes (zor example, normal off/normal, emergency, operating, maintenance, suzveillance,

test, security, and radiation control procedures and the emez'gency plan), regulations, and license provisions; t aining, qualification, and perform'ance of operating staf f; and corrective actions following pepo jtpQ]e occurrences.

337 Amendments Hos.

27 8 24

6 0

ADMIN ISTRATIVE CONTROLS

6. tI Actions to be Taken in the Event of a Reaortable Occurrence in Plant Ooeration Ref. Section 6.7 A.

Any reportable occurrence shall be promptly reported to the Chief, Nuclear Generation Branch and shall be promptly reviewed by PORC.

This committee shall prepare a separate report for each reportable occurrence.

This report shall include an evaluation of the cause of the occurrence and recommendations for appropriate action to prevent or reduce the probability of a repetition of the occurrence.

B.

Copies of all such reports shall be submitted to the Chief, Nucleai Generation Branch, the Manager of Power, the Division of Power Resource

Planning, and the Chairman of the NSRB for their review.

C The plant superintendent shall notify the NRC as specified in Specification 6.7 of the circumstances of any reportable occur ence.

6. 5 Action to be Taken in the>> vent a Sa fet Limit is, Exceeded Ef a safety limit is exceeded, the reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

A prcmpt report shall be made to the Chief, Nuclear Generation Branch and the Chairman of the NSRB.

A complete analysis of the circumstances leading up to and resulting from the situation, together with recommendations to prevent a recurrence, shall be prepared by the PORC.

This report shall be submitted to the Chief, Nuclear Generation Branch, the Manager of Power, the Division of Power Resource

Planning, and the NSRB.

Notification of such occurrences will be made to the NRC by the plant superintendent within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6.6 Station Ooeratin Records Ai Records and/or logs shall be kept in a manner convenient for review as indicated below:

'C..

1.

All normal plant operation including such items as power level, fuel exposure, and shutdowns 2.

Principal maintenance activities 3.

Reportable occurrences 34~

Amendments Nos.

27

& 24

I

'f 4

6 0

AOMEN ISTAATIVE CONTROLS 6.7 Re rtin Recuirements In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforcement unless otherwise noted.

Routine Reports a.

Startu Report..

A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier,= and (0) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test pxogram and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license conditions based on other commitments shall be included in this report.

b.

Startup reports shall be submitted witnin (1) 90 days following completion of the startup test

program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

Zf the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commexical power operation),

supplementary reports shall be submitted at least every three months until all three events have been completed.

Annual Oneratina Report.~

Routine operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to March 1 of each year.

The initial report shall be submitted prior to March 1 of the year following intial cxiticality.

Amendments Hos.

27 8 24

6 o 0 ADMINIST TIVE CONTROLS The annual operating reports made by licensees shall provide a comprehensive summary of the operating experience gained during the year, even though some repetition of previously reported information may be involved.

References in the annual operating report to previously submitted reports shall be clear.

Each annual operating report shall include:

(1)

A narrative summary of operating experience during the report period relating to safe operation of the facility, including safety-related maintanance not covered in item

1. b. {2) (e) below.

(2)

For each outage or forced reduction in powezm of ovei twenty pezcent of design power level where the reduction extends for greater than four hours:

(a) the proximate caUse and the system and major component involved (if the outage or forced reduction in power involved equipment malfunction);

(b)

A brief discussion of (or reference to reports of) any reportable occurrences pertaining to the outage of power reduction~

(c)

(d) corrective action taken to reduce the probability of recurr ence, if appropriate; operating time lost as a result of the outage or power reduction (for scheduled or fozced outages,~

use the generator off-line hours; for forced reductions in po~er, use the approximate duration of operation at reduced power);

(e) a description of major safety-related corrective maintenance performed during the outage or power reduction, including the system and component involved and identification of the critical path activity dictating the length of the outage or powe reduction; and gyp

. Amendments Hos.

27 5 24

6' ADMINISTRATIVE CONTROLS (9)

Performance of structures,

systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report or technical specifications that require remedial acticn or corrective measures to prevent the existence or development of an unsafe condition.

b.

Note:

This item is intended to provide for reporting of potentially generic problems.

Thirt -Da Written Re orts.

The reportable occurrences discussed below shall be tne subject of written reports to the. Director of the appropriate Regional Office within thirty days of occurrence of the event.

The written report shall include, as a

minimum,'

completed copy of a licensee event report form.

Information provided on the licensee event report form shall be supplemented, as needed,,

by additional narrative material to provide complete explanation of the circumstances surrounding the event.

(1)

Reactor protection system or engineered safety

~

feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.

(2)

Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.

Note:

Routine surveillance testing, instrument calibration, or preventative maintenance whic& require system configurations as described in items 2.b.(1) and 2.b.(2) need not he reported except where test resul'ts themselves reveal a degraded mode as described above.

(3)

Observed inadequacies in the implementation of administrative or procedural controls which 354 Amendments Hos.

27 8 24

r

~

6 0

ADMINISTRATIVECONTROLS 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.'aecial Re orts (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 Containment Leak Rate Testing(5) b.

Fatigue Usage Evaluation 4 ~ 7~C 6

6 Within 90 days of completion of each test.

Annual Operating Report co Seismic Instrumentation Inoperability 3.2.J. 3 Within 10 days after 30 days of inoperability 356 Amendments.Hos.

27 5 24

6 0

ADMXNXSTRATXVECONTROLS

'/

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 failure or other condition which requires that the load on the unit be reduced for corrective action immediately or up to and including the very next weekend.

Note that routine preventive maintenance, surveillance, and calibration activities requiring power reductions are not covered by this, secti on.

3.

The term "forced outage" is normally defined in the electric power industry as the occurrence of a component failure or other condition which requixes that the unit be removed from service for'orrective action immediately ox up to and including the very next weekend.

This tabulation supplements the requi ements of 3 20. 007 of 10 CFR Part 20 ~

5 Each integrated leak rate test of the secondary containment shaU.

be the subject of a summary technical report.

This report should include data on the wind speed, wind direction, outside and inside temperatures during the test, concurrent reactor building pressure, and emergency ventilation flew rate.

The report shall also include analyses and interpr etations of those data which demonstrate compliance with the specified leak rate limits.

3/7 amendments Nos.

27

& 24

4

<g8 REGS4g

+

O~

4j A

Cl r+

~O

+**y4 UNITEDSTATES NUCLEAR REGULATORYCOMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROMNS FERRY NUCLEAR PLANT UNIT NO.

2 AMENDMENT TO fACILITYOPERATING LICENSE Amendment Ho. 25 License Ho.

DPR-52 1.

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

A.

The applications for amendment by Tennessee Valley Authority (the licensee) dated September 1, October 1

and October 12,

1976, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

I

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment'nd paragraph 2.C(2 ) of Facility License Ho.

DPR-52 is hereby amended to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B,

as revised through Amendment Ho ~ 25, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of i ts i ssuance.

FOR THE HUCLEAR REGULATORY COf~itlISSIOH

Attachment:

Changes to the Technical Specifications J

g~.g a~ rq/rgb'g~

A. Schwencer, Chief Operating Reactors Branch 81 Division of Operating Reactors Date of Issuance:

February 15, 1977

. ~ ~

~ 7

~S AE'gy~

IP

~>>*+~

UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT HO. 28 TO FACILITY LICENSE NO.

DPR-33 AND AMENDMENT NO.

25TO FACILITY LICENSE NO.

DPR-52 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS NOS.

1 AND 2 DOCKETS NOS.

50-259 AND 50-260 Introduction By application dated September 1, 1976, the Tennessee Valley Authority (TVA) requested amendments to the operating licenses for Browns Ferry Nuclear Plant, Unit Ho.

1 (DPR-33) and Unit No, 2

(DPR-,52) to change the Technical Specifications by adding the isolation valves for a new, drywell-torus differential pressure control system to the containment isolation valves listed in the Technical Specifications for containment.

By application dated October 1,

1976, TVA requested amendments to DPR-33 and DPR-52 to delete from the Technical Specifications the logging requirement for torus temperature when heat is being added to the torus.

By application dated October 12,

1976, TVA requested amendments to DPR-33 and DPR-52 to correct the basis in the Technical Specifications for the number of Automatic Depressurization System (ADS) valves required to be operable and to reduce the allowable time for reactor operation with two ADS valves inoperable from 30 days to 7 days.

Isolation Valves for Differential Pressure Control S stem Discussion As a result of recent structural analyses performed in conjunction with a generic review of pool dynamic loads for Mark I pressure

- suppression containments, it was determined that the margin of safety in the containment design for the Browns Ferry Nuclear" Plant as related to pool dynamic loads resulting from a postulated loss-of-coolant accident was less than originally thought to exist.

Consequently, TVA agreed to institute a

"differential pressure control system" to mitigate the pool dynamic loads and thereby restore the original margin of safety in the containment design.

The differential pressure control system establishes a positive pressure between the drywell and torus regions of the containment which reduces the height of the water leg in the downcomers and consequently reduces the hydrodynamic loads.

f'i ~

The differential pressure control system consists of a bypass installed in the containment purge line between the drywell and the torus.

A compressor is installed in the bypass line which takes suction from the torus and pressurizes the drywell until the appropriate differential pressure is established.

In conjunction with the piping modifications, three valves have been installed in the containment purge and bypass lines to serve as outboard containment isolation barriers and to provide proper system flow routing.

Evaluation The piping modifications associated with the inclusion of the differential pressure control system result in the addition of three containment isolation valves.

These valves serve as the redundant containment isolation valves and as such are designed to seismic Category I and Safety Class 2 criteria.

Automatic isolation occurs upon the receipt of a reactor vessel low water level, high drywell pressure, or high

.reactor building exhaust radiation signal.

These valves, their controls, actuation logic and installation meet all the requirements of the previously accepted criteria for Browns Ferry containment isolation valves.

Provisions have been made in the piping modifications to permit local leak testing of the isolation valves in accordance with Appendix J to CFR 50.

The differential pressure control system is designed such that its inclusion will not interfere with the safety related features incorporated in the existing plant design.

In addition, the system design is in conformance with the applicable regulations, regulatory guides, and staff positions.

Therefore, we find the proposed modifications together with the addition of Technical Specifications requirements for these valves to be acceptable.

Torus Tem erature Lo in The Technical Specifications include a requirement to log the torus water temperature every 5 minutes when heat is being added to the torus by the operation of relief valves.

TYA's application of October 1, 1976 requested deletion of this requirement since the torus water (suppression pool) temperature is continuously recorded on a strip chart recorder and the operator will receive an alarm if the suppression pool temperature exceeds 95oF.

TVA was concerned that the specification, as written, would require an operator to be logging temperatures during a period when abnormal conditions exist and safety priorities would require him to be doing other things in response to the abnormal conditions.

It was not our intent to require such logging during transient or accident conditions.

The limiting conditions for operation on suppression pool temperature include an allowance to exceed the normal 95oF limit up to 105oF during testing of ECCS and relief valves.

Therefore, during such testing the temperature alarm could annunciate at 95 F (its alarm point) but there would be no further alarm annunciation to attract the attention of the operator should

the water temperature exceed 105 F.

Consequently, the requirement to log the temperature at 5 minute intervals was specified.

We have, with this change, clarified the wording of the specification to more clearly indicate its intent.

Automatic De ressurization S stem ADS Th b sis for the limiting condition for operation for the ADS has stated that only four of the six valves are assumed operable for the small bre e

a k analysis of the ECCS evaluation.

On this basis operation with two inoperable valves was allowed for 30 days and operation with more than two inoperable valves was limited to seven days provided that the high pressure coolant injection system, which is a redundant alternate to ADS and low pressure coolant injection, is operable during that 7 days.

TVA's October 12, 1976 application for amendment indicates that the ECCS Appendix K analysis was performed with five of the six valves assumed to be operable.

Therefore, the time limit for continued operation must-

'e reduced to seven days whenever'ore than one valve is inoperable proviaed that the high pressure coolant injection system is demonstrated to be operable daily.

This change will maintain the reliability of the ECCS at a level commensurate with that previously evaluated and accepted and will maintain the margin of safety used as the basis for the Technical Specifications.

Environmental Considerations We have determined that the amendments do not'authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.

Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the amendments.

Conclusion We have concluded, based on the considerations discussed above, that:

{1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, and {3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.

Date:

February 15, 1977

UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKETS NOS.

50 259 AND 50 260 TENNESSEE VALLEY AUTHORITY NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES The U. S. Nuclear Regulatory Commission (the Coranission) has issued Amendment No.

28 to Facility Operating License No.

DPR-33 and Amendment No.

25 to Facility Operating License No.

DPR-52 issued to Tennessee Valley Authority (the licensee),

which revised Technical Specifications for operation of the Srowns Ferry Nuclear Plant, Units Nos.

1 and 2, (the facility) located in Limestone County, Alabama.

The amendments are effective as of the date of issuance.

/

The amendments change the Technical Specifications to add containment isolation valves associated with the drywell to torus differential pressure control system to the valve listing (Table 3.7,D) for the limiting condition for operation and surveillance requirements of primary containment.

A clarification in the wording of the temperature survei llance requirement for the torus water has also been made.

In addition, the allowable operating time with two inoperable Automatic Depress0rization System (ADS) valves has been reduced from thirty days to seven days to reflect the fact that the ECCS Appendix K analysis was performed with five of the six ADS

'alves operable rather than four aszstated previously.

The applications for the amendments comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations.

The Commission has made

appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license

'amendments.

Prior public notice of these amendmentsiwas not required since the amendments do not involve a significant hazards consideration.

The Commission has determined that the issuance of these amendments will not result in any significant environm ntal impact and that pursuant to 10 CFR 5 51.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with issuance of these amendments.

For further details with respect to this action, see

( 1) the applications for amendments dated September 1, October 1

and October 12,

1976, (2) Amendment No. 28 to License Ho.

DPR-33 and Amendment No. 25 to License Ho.

DPR-52, and (3) the Corrmission's related Safety Evaluation.

All of these items are available for public inspecti on at the Commission's Public Document

Room, 1717 H Street, H. W., Washington, D.

C.

and at the Athens Public Library, South and Forrest.,

Athens, Alabama 35611.

A copy of items (2) and

( 3) may be obtained upon request addressed to the U, S. Huclear Regulatory Commission, Washington, D.

C.

20555, Attention:

Director, Division of Operating Reactors.

Dated at Bethesda, Maryland, this 15tp day of february lg77, FOR THE NUCLEAR REGULATORY COMMISSION A. Schwencer, Chief Operating Reactors Branch 81 Division of Operating Reactors

'1