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) BHarless NRC PDR(2 TBAbernathy Loca1 PDR JRBuchanan Dockets Nos. ORB81 Reading a 5O-26O 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.

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.

Copies of the Safety Evaluation and the Federal Register Notice are also enclosed, Sincerely, aoei>>i S'<""da A, Schwencer, Chief Operating Reactors Br nch 8 Division of Operating Reactors OFFICE next page - '

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Tennessee Yalley Authori ty 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 U. S. Environmental Protection Agency General Counsel Federal Activities Branch Tennessee Valley Authority Region V Office 400 Commerce Avenue ATTN.i EjS COORDlNA3)R E 11B 33 C 345 Courtland Street, NE Knoxville, Tennessee 37902 Atlanta Georaia 30308, 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

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~ ~Pg AEVI P UNITED STATES 0

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

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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. 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 regulitions and all applicable requirements have been satisfied.

)

i4

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 j;-,:~cF/F~~

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

Attachment:

Changes to the Technical Specifications 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 t ion Cooling 4.5.F Reactor Cora Isolation Cooling

2. If thc'CICS is inoperab'c, 2. When it is determined that the the reactor may tcmain in RCICS is inoperabla, chc HPCES operation for a period not shall be demonstrated to be to exceed 7 days if thc operable immediately and weekly HPCIS is operable during theres~ter.

such time.

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 1. During each operacing cycle the Automatic Dapressuri- the following tests shall be aation System shall be ~ performed on the ADS:

. operable: '.

A simulated automatic (1) prior to a itart<<p actuation test shall be from a Cold Condition, perfoencd pzior eo scaztup or, after each rafuelinp out-age. Nanual surveillance (2) whenever there is i,rza- of the relief valves is diatad fuel in the reac- covezed, in 4.6.D.2.

tor vessel and the reaceor vessel press<<re ia greater than 105 psir except as speci f i ed in 3.5.C. 2 and 3.5.C.3 below.

2. lf more than one ADS known to be incapable valve is of 2, When it is determined that more than one of the ADS valves are incapable automatic operation, the of automatic operation, the HP'CIS reactor may remain in opera- shall be demonstrated to be operable tion for a period not to immediately and daily thereafter as exceed 7 days, provided the. long as Specification 3.5.8,2 HPCI system is operable. applies, (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.)

.1 57 Amendments Nos. 28 & 25

1.[hIZTINC CA.'II)I'I'TONS FOR OPEPAVIOH SURVKILLAHCF. RE U IRENENTS 5.G Autnmnt ic Ocnrcsnurlzntion '4.5.G Automntic Dc ressurization

~Sstrin (AI)S) 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 nf fled ~nfochnc e H. Nnintenancc of Filled Dischar c

~Pl c ~Pi e Mhcncvcr thc core spray oyotems, The folloving surveillance rcquircnn LPCI, IIPCI, or RCEC.are required ments. shall bc adhered to to assure to bc opcrnblc, thc discharge that thc discharge plplnp, of the pipl>>g from thc pump discharge core spray systems, LPCI, HPCI, and of these systems to thc 'last RCIC arc filled:

block valve shall bc 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

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. OO ENCLOSURF.

OO LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREi~mNTS

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

~cb 'ective 1 ~cb ective To assure the integrity of the primary and secondary To verify the integrity of the containment systems. primary and secondary containme nt.

Sneci fication Speci fication Primar'ontainment Primar Containment At any time. that the irradiated fuel is in Pressure Suaoressicn the reactor vessel, Chamber and the nuclear system is pressurized a. The suppression aLove atmospheric chamber water level pressure or work is be checked once per being done which has day, Whenever heat the potential to is added to the drain the vessel, the pressure suppression suppression pool by pool water volume and testing of the ECCS temperature shall be, or relief valves the maintained within the pool temperature shall following limits be continually monitored

~ except as .specified and shall be observed in 3.7.A.2. and logged every 5 a., Minimum wate" volume - 123, 000 minutes until the heat.

addition is terminated.

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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within 227 the following 30 Amendments Nos. 28 & 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.

~ ~

ENCLOSURE 1 TABLE 3,7.D (Continued)

Valve Test T.est Valves Identification Medium Method 43-28B RHR Suppression Chamber Sample Water~ ~

Applied between 74-226 and-;43-28B Lines43-29A RHR Suppression Lines Chamber Sample Water 'pplied (2) between 74-227 and 43-29A 43-29B RHR Suppression Chamber Sample Water (2)'pplied between 74-227 and 43-29B Lines 64-17 Drywell and Suppression Chamber Applied between 64-.17, 64-18, 64-19, air purge in1et and 76-24 64-18 Drywell air purge inlet Air Applied between 64-17, 64-. 18, 64-19, and 76-24 64-19 Suppression Chamber air purge Air Applied between 64;.17, 64-18, 64-.19, inlet and 76-24 64 20 Suppression Chamber vacuum Ai (1) Applied between 64-20 and 64-(ck) relief s 64-(ck) Suppression Chamber vacuum Ai (1) Applied between 64-20 and 64-(ck) relic f 64-21 Suppression Chamber vacuum Applied between 64-21 and 64-(ck) relief 64-(ck) Suppression Chamber vacuum Applied between 64-21 and 64-(ck) relief 64-29 Drywell main exhaust Air( )

Applied between 64-29, 64-30, 64-32 64-33 and 84-19 64-30 Drywell main exhaust Air(1) Applied between 64-29, 64-30, 64-32; 64-33 and 84-19 64-.31 Drywell exhaust to Standby Air((1)'pplied between 64-31,64-141, 84-20 and 64-140 64-32 Suppression Chamber Main (1) Applied between 64-32, 64-..33, 64-29, Aiq Exhaust 64-30 and 84-19 64-33 Suppression Chamber Main Air(1) Applied between 64-32, 64-33, 64-29, Exhaust 64-30 and 84-19 Air( )

64-34 Suppression Chamber to Standby Applied between 64-34,64-141 and Gas Treatment 64-139 259 Apepdments Nos. 28 & 25

v

~, 1

~0 TjUlLE 3.7.D (Coatinucd)

Valve Teat Teat Valvaa Identification Nadiim Macha'pplied 90>>257A Radiation Monitor Diaeharga Air<'> betveea 90-2571 aad 90-257:,

90-2575 Radiation Hoaitor Discharge Ai ") Applied be~en 90-237A sad 90-257'pplied 84-8A Containment Atmospheric Dilution Air between 84-8A and 84MOO 84 BB Containment Atmospheri.c Dilution Air Applied between 84MB and ~l...

84-8C Containment Atmospheric Dilution Air Appli.ed between 84-8C and 84-603 84-8D Containment Atmosphex'ic Dilution Ai'r Applied between 84~ and 84-60$ :

84-19 Containment Atmospheric Dilution Applied between 64-32, 64-33, 64-29, 64-3A', and 84-19

I (1) Air/nitrogen test to be displacement flower.

(2) Rater test to be injection loss or downstream collection. y ~ ~

Valve Test Test Vnlves Identification Bed ium Method 76-? 15 Containment Atmospheric tlonitor Air<>> Applied between 76-215 and 76-218 76-217 Containment Atmospheric llonitor Air Applied between'7 -217 and 76-218 76-220 Containment Atmospheric tlonitor Air Applied between 76-220 and 76-223 76-222 Containment Atmospheric Honitor Mr Applied bctvreen 76-222 and 76-223 76~~25 Containment AtmospiU.ric iionitor Air Applied Letween 7G-225 nc'G-227 76 22G Containment Atmospheric Monitor Air Applied between 76-226 and 76-227 7G-229 Containment Atmospheric 1lonitor Air Applied between 76-229 and 76-231 76-230 Containment Atmospheric Monitor Air Applied between 76-230 and 76-231 76M37 Containment Atmospheric Honitox Air Applied between 76-237 and 76-240 76-239 Containment Atmospheric Monitor Appl.ed between 76-239 and 76-240 .76-242 Containment Atmospheric Monitor Air Applied between 76-242 and 76-244 76M43 Containment'tmospheric Monitor Air Applied between 76-243 and 76-244 76-248 Containment Atmosph ric Honitor Air Applied between 76-248 and 76-253 76-250 Containment-Atmospheric Monitor Ai.x Applied between 76-250 and 76-251 76 253 Containment Atmospheric Monitor Air Applied between 76-253 and 76-255 76-254 I

Containment Atmospheric Wi nitor Air Applied between 76-254 and 76-255 84-20 tlain Exhaust to Standby Gas Txeatmen

~1 t Air(>> Applied between 84-20,64-141, 64-140, and 64-31 84-600 Main Exhaust to Standby Gas Treatment Nitrogen(1) Applied between 84-8A and 84-600 84-601 Main Exhaust to Standby Gas Treatment tV troi.en Applied between 84-88 and 84-601

,84-602 Main Exhaust to Standby Gas Treatment tt4trogen Applied between 84-80 and 84-603

'4-603 ttain I'.xhaust to Standby Gas Treatment t$ itx'oqen hppli.ed between 848D and 84-602 F64-l4l Drywel'1 Vressurization, Comp. Bypass Air (>) hpplied between 64-141,64-140, 64-30, and 84-20

,64-140 Drywell Pressurization, Comp. Disc. Applied between 64-141,64-140, 64-31, and 84-20 Dr~~ell Pressurization, Comp. Suction Applied between 64-139.,64-141, and 64-34 '64-139 (1) Air/nitrogen test to'be displacement flow (2) Mater test to be injection loss or downstr earn collection.

262 Amendments Nos. 28 8 25

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:

36 252 357 44 267 thru 270 54 286 89 thru 95 295 123 296 322 124,'143'44 326 332

, 145i 333 146I 337 150I 346 151I 349

'154' 350 185I 354 187 356 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) .

'l

8. Operability is required >Aen normal first-stage pressure iz beIov

( 154 Psig) ~

19. 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 r equired.

20. 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.

Amendments Nos. 27 8 24 36

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 1

~ I 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 72 Ch Q toto So'd cn't ee <<~et'at 't see 1isted in table 3.2.Z instr ~

. Qaemb1e.at ~ times. sh~~ De nts s'g~t

'ageable Dy nerfo~~-c o tes e.t the f eonenc's ~'sted in De d ,Q i s

2. Viith the neer ox seisw'c ~"'tar'~ table 4.2.Z.

~~s n erts 1ess tmn the neer 1is ed in tsb1e 3.2.Z, estore the 2.

inst~

ststns d~

Xnoaereb1e instru 30 en

~s.

(s) to onemb1e R3 d~~ 2 se s tQ determines c

seis~~

't s 8.ctn te c ev&~ 2~8. 2~a Jze

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

3. Nth one or na e of the inst. ments usted 'n tsb1 3.2.Z inapemM.e o P. aaw sh~" ~ be sub='e'o the mre then 30 ~+s, s~~ 2, SaeciM CQM~~ ss~ Qn QU sl'"n Q ~~~c'Rt Qn Resort ta the Ccrr "ss'on pu~s~t to 6.7.3.D 'edeeds 10 days de-"- 'ud.~

scecdddcedce 6.7.3. C uduhdc ""e c"+ C'le c" s '*'2" "ct'e c'J'cec"M 10 ~s desc"d'b~ ~".'.".e causa c';.e veStgs tered est 'teC> nt~~q p'7 tt tt+

A@3.fu,ction a& p2.sns for r s o~~ fes tUZ'es ~~~no~~t~t to M ~ ctj e the 'nst ments ".o oae"eble sts,.ns.

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 4.3 ' Control Rods be During the shutdown procedure no rod movement is permitted The capability of the RSCS ta pro-between the testing performed perly fulfillits function shall be above 20'o power and the rein; verified by the rollow'rg tests:

statement of the RSCS re- -

straints at or above 20~a Sequence portion Select a sequence and attempt to withdraw a rod in tha power. Alignment of rod Move one rod groups shall be accomplished remaining sequences.

prior to performing the tests. in a sequence and select the z~~ in-ing sequences ard atte pt to move a rad in eacn. Repeat for a11 zeactor is sequences.

c. Whenever the in the startup oz run nodes Group notch ao "'on Zor each or the below 20% rated paver the six comparator circuits go through Rod North Hin~~izer'hall be test initiate; compazator irhioit; operable oz a second licensed verify; reset. On seventh atte=pt operator shall verify that test is allowed to continue until the operator at the reactor completion is indicated by console is fallowing the ill~~tion of t st cauplete. light.

control rod program.

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

~

~ 0 C' checks:

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

If Specifications 3.3.3.3.a 2. The RHH computer on line thraugh .c cannot be net the diagnostic test sha'1 ae reactor shall nat be started, successfully perf armed.

'oz if the reactor is in the 3. Prior to startup, proper run or startup" mades at less annunciation or the selec-than 20% zatad power, it tion er.ar of at least ane shall be brought to a shut- aut-of-sequerce contra'ad down condition ir=edixtaly. 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 123 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 When requir d, th presence withdrawn for startup or of e second l'censed operator refueling unless at least to verify th foll~ing of tan source range channels the correct rod program s~l3.

have an observed count rate be verified.

equal to or greater than

~

three counts per econd.. 4 Prior to contro3. rod withdrawal for startup or duri..g refueling, 5..Durin"operation with verify that at least two source limiting control .rod pat- range channels have an observed terns, as determined by the count rate of at least three designated qualified person- counts per second.

nel, either:

a. Both RBN channels shall 5. %hen a limiting control rod be operable: pattern exists, an inst~ ent or functional test of the P~A shall oe pe fo~ d p.ior to

b. Control rod withdrawe3. withdrawal of the designa ed shall be blocked: 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 The average scram insertion After each refueling ou age all time, based on the deenergi- operable rods shall be scram ti=,e zation of the scram pilot valve tested from the fully w'thdra -..

soleno'ds as ti= zero, of all position with the n c3.ear syste..

operable control rods 'n th pressure above 950 ps'g ('wth reactor power operas'on condi- saturat" on taupe acute) I T. is test g tion shall be no greater than: shall oe, comp eted prior to ezc 'ding

<<nscr>>>>d T ~ <<J c<<om Avg. Scram Tnser-Pu3."d--'m -'n T < -...es (sec) and B34) wh ch wer>>>> ' y withdrawn in the region from lOOZ 5 0.375 >rod dens ty'to 505 rod density sha 20 0.90 be scram t'."..e testee. During all 50 2.0 sc"am t.-e t st-'-. below 20.';:o-e" 90 5.0 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 4.5 CORE A?H) CORRAL'?Pi{""NT COOL""iiG SYST HS SYSTEMS Aoolicabili.t Aoolicabilit Applies to the operational Applies to the surveillance status of the core and contain- requirements of the core and ment cooling systems. 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 To verify the operabili y of the the core and containment cooling core and contain nt cooling systems under all conditions for systems under all conditions for which this cooling capability is which this cooling capability is

.an essential response to plant an esschtial response to plant abnormalities. abnormal'ti se Soecification S ecification A. Core Sore- System (CSS)

.1. The CSS shall be opera- 1. Core Spray System Testinge ble:

Item ~aauaaa (1) prior to reactor startup from a a. Simulated Once/

cold condition, or Automatic Opera ting Actuation Cycle (2) when there is irra- teat diated fuel in the vessel and when the b. Pump Opera-. Onc /

rcac or vessel. pres- 'oility anth sure is greater than atmospheric prcssure, c. Hotor Onc /

except as specified Operat d month in specifications Valve 3.5.A.2, 3.5.3.2, or Operability 3.9, B. 3.

d. System flow Onc /3 ratae Each mon ths loop shall deliver at lease 6250'pm against a system head corr s-ponding to a 143 Amendments Nos. 27 & 24

TZMZTZNG CONDZTZONS FOR OP~TZON SUR~rZILANCZ RZQUZ~~PXS 3.5.h Coro Borav 8 stem CSS) 4,5,A Core Sara S stem CSS) 2 ~

'i( Uns CSB loop ia inopera- 105 psi dif-ble, the reactor may remain ferential in operation for a period. pressure not to exceed 7 days provi- between the ding all active components reactor ves-in the other CSB loop and the seI., end the RHR system (LPCX mode) and the primary con-diesel generators are operable. tainment.

3 ~ Lt'pecification 3.5.A.l e. Check. Val ve Once/

or specification 3.5.A.2 Operating cannot: be met, the reactor Cycle shall be shutdown in the Cold Condition within 24 2. When it is determined that one core spray 1oop is inoperabl'e, ho'urs ~

at a time when operability is When the reactor vessel reqoired, the other core spray pressure is atmospheric loop, the RHRS (LPCl mode), and and irradiated fuel is in the diesel generators shall be the reactor vessel at demonstrated to be operable least one core spray loop immediately. The operable core

. with one operable pump and spray loop shall be demonstrated associated diesel generator to be operable daily thereafter.

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.

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

LDKTIIfG COt(DZT'ZOOS tern POR OPEcUTZON SUHVEILLN(CZ BZQUZHZ~PZS

-. 3. 5. 8 Res i(junl Heat Removal Svo 4.5PB Restdual Heat Removal S stem

~RHRS (LPCI and Contain~at ~RHRS (LPCI and Contain=ant Coal ititt) Caaling)

The RHRS shall be opexable: 1. a, Simulated once/

Automatic Opera ting

{1) prior to a reactor Actuation Cycle scarcup fran a Cold Test Condition; or (2) when there is irra b. Pump Opera<< Once/.

diated fuel in the bility r(anth x'eactor vessel and when the reactor vesseL pres- c. Hotor Opera- Once/

sure is greater than ted vaLve month atmospheric, except as operability ape'cified in specifica-tions 3.5.B.2, through d. Pump PLow Race Once/3 3.5.8.7 and 3.9.B.3. esontha

2. With the reactor vessel pr s- e. Test Check Valve Once/

surc less chan l05 psig, the Operating PHRS msy be removed from ser- Cycle vice {except that twa RHR pumps-containment, cooling made and Each LPCZ pump shah'e3.iver associated heat. cxchangers must 9,000 gpm against an indicated remain aperablc) for a pex'iod system pressme of ~&5 psig. Two .

nat to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while LOCI PumPs in the same looP shaly being dranned of suporession deliver 15,000 <~m against an chamber quality water and indicate(i system pressu e of

~

filled with primary coolant 200 psig.

quality water provided that 2. An air test on the dryweLL and duxing cooldo~m two loops N th torus hcade."8 and no=ales shaLL one pump per loop or one loop with cwo pumps, and associated diesel be conducted once/5 years. A generators, in the core spray syste water test may be performed on are operable. ~

the torus heade. in lieu of the air test.

3. Lf one RHK pump (LPCI node) is inoperahLe.'the reactor 3. When it is determined that one PHR pump (LPCI mode) is inoperable ac may remain in operation for a time when operability is rcqu'-.e'-.

period not co exceed 7 days che remaining RM pumps (LPCI provided the remaining RHR active components 'n both aces" node'nd pumps (LPCI mode) and both access pacha of the RHRS paths of the RHRS {LPCI mode) and the CSS and the diesel generators (LPCI made) and the CSS and shall be de=onstrated to be opera-the diesel generators remain ble i~ediately. The operab'e RHR:

operable. 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 145 Amendments Nos. 27 8 24

PAGE DELETED amendments Nos. 27 & 24

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREMENTS a time when operability inoperability, pipe break, is required, the re-etc), the reactor may remain maining RHR pump and in operati on for a peri od not associated heat exchanger to exceed 30 days provided on the unit cross-connec-the remaining RHR pump and tion and the associated associated diesel generator diesel generator shall be are operable. demonstrated to be oper-able immediately and every 15 days thereafter until the inoperable pump and associated heat exchanger

13. If RHR cross-connection flow or are returned to normal heat removal capability is lost, service.

the unit may remain in operation for a period not to exceed 10 days unless such capability is res to red.

12. All recirculation pump discharge valves shall be tested for operability

14. All recirculation pump during any period of discharge valves shall reactor cold shutdown be operable prior to exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, operability tests if have reactor startup (or closed if permitted elsewhere in these not been performed during the preceding specifications). 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 4.5.C RHR Service Water and E er encv E ui ment Coolin Water S stems E ui ment Coolin Water S stems (EECVS >

1. Prior to reactor startup from a 1. a. Each of the RHRSW pumps cold condition,,9 RHRSW pumps must normally assigned to be operable, with 7 pumps (includ- automatic service on ing pump Dl or D2 for unit 1 and the EECW headers will one of pumps Dl, D2, Bl, or B2 for be tested automatically unit 2) assigned to RHRSW service each time the diesel and 2 automatically starting pumps generators are tested.

assigned to EECW service. ~

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 SURVEILLANCE REOUTB Mi.iTS

/'.5.5 E ui ment Area Coolers 4.5iD E uf.a cnt Azcc Coolers

1. The equipment area cooler l. Each equ'pment area cooler associated vith each RK~ is operated in con)uncticn pomp and the equip=cr.t area vith the equipment served by coo lez assoc ia tcd v i th ecch that particular cooler; set of core sprav pumps {.<. therefore, the equip- nt area and C or 8 and D) must be opezable at all times vhen coolers aze tested at same frequency as the pumps t¹ the pump oz'umps se~ed by vhich they serve.'

that specific cooler is considered to be operable.

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 Z. High Prcssure Coolant Tniection S stem H?CIS)

1. The HPCI system shall be 1. HPCX Subsystem testing shall operable: be per'a~ad as iollovs:

e (1) prior to staztup from z a. SimuLated Once/

Cold Condition; or ~

Auto-atic operating

'ctuation cyc'e (2) vhenever there is irra- Test diated fuel in'he reac-tor vessel and the reactoz be Pu .0 Opera Once/

vessel pressure is greater bility =anth than 122. pslg, except aa specified in specifica- C~ .totor Operated Once/

tion 3.5.E.2. Valve Opeza- month b'lity

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

e. PLov Rate at Onc /

150 paid 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 The following surve'llance

,operation except. Cold Shutdown and Refuel, requirements apply to hydraulic snubbers listed all all safety-related in 3. 6.H. 2.

snubbers sha3.1 be operabl.e except as noted in 3.6.H.2 All hydraulic snubbers whose seal through 3.6.H. 5 material has been below. 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 in spec 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 Ne:c Requirec Snubbers Xnsp ction Pound Inoper- Interva3.

able During Inspection or During Inspec-tion Interval Ooera ting 1 .12 months +25~

2 6 months 3,4 124 days +25~

5,6,7 62 days +25%

185

>Q 31 days +25%

Amendments Nos. 27 8 24 The required inspection intervai shall not be lengthened more than one step at a time.

~ ~

e

~

4 (MlTIl4%BY SYSTEM BOUNDARY 6 PRIMARy BOUN DARY If tne requirements of 3. 6. H. i anci Once each a

refueling cycle, representative sample 3.6.H.3 cannot be of 10 snubbers or met, an orderly approximately 10~ of shutdown shall be the snubbers, whichever initiated and the is less, shall be reactor shall be in a functionally tested for cold shutd~ operability including condition within 36 verification of proper hours. piston movement, lock up lf a snuo6er ts and bleed.

and subsequent For each unit unit found determined to be in-operable while the inoperable, an additional reactor is in tlie 10 or ten snubbers shall sfiutdown or refuel be so tested until no mode, the snubber more failures are found shall be made or all units have been operable or replaced tested. Snubbers of prior to reactor rated capacity greater startup. than 50,000 lb need not be functionally tested.

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.

187 Amendments Nos. 27 8 24

t TABLE 3,7,A (Continued}

(umber of Power Maxi mum Action on .

0 crated Valves Operating Hormal Initiating

~Grou Valve Identification lnboar Outboard Time sec. Position ~Si nal Suppression Chamber purge inlet

( FCV-64-19) 100 SC Drywel 1/Suppressi on Chamber ni tro-gen purge inlet (FCV-76-17) 10 SC Drywell Exhaust Valve Bypass to Standby Gas Treatment System (FCV-64-31) 10 SC Suppression Chamber Exhaust Valve Bypass to Standby Gas Treatment Sys tern (FCV-64-34) 10 SC RCIC Steamlind'rain (FCV-71-6A, 6B) 0 GC RCIC Condensate Pump Drain (FCV-71-7A, 7B) GC S

I C+

HPCI Hotwell tion valves pump discharge isola-( FCV-73-17A, 178) SC 7 HPCI steamline drain (FCV-75-57, 58) GC O

cia v 8 TIP Guide Tubes (5) 1 per guide NA GC tube

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 At 'mesc ance during each operating shall be capable of being cycle vcr'ify auto"atic securing and automatically isalated and isolation aE the mechanical vacu n secured on a signal aE 'high pucp i radioactivity in,the stea= O. Miscellaneous 3 c.cnctive l!a eri 1s Sources lines Mhcnever the cain stean ccu'c.-.cn isalatian valves are open. Surv~il'ance  ?

Tes=s;or 3.caka;c an"/ar ccn'-a.".wast- n

~ ~ IE th liMts af 3. 8. C. 1 are shaM~ be per"ar.-.ed 'oy t'ao lican ee or nat net, the vacuu.- pu=p shall by o he. per sans saeciE'ica lv authari=e=

be isolated. by ha Co-.~issicn or a,. agreement State,, as fo lo"s: ~

Lfisccllanenus %dig=et va D.

Yiaterials Sacr "as. ch saa 'd sou ca) axcett sta sources s'jcct to core flux, Source I, azace Test cant" wing radioactivo Tcatarial, I

athar than Hydra;en 3,::it., a h3 liSa greater than thirty d=':s and I

Each sealed source- containing I radioactive material in in =ny 'rm other than gas s.".all be tasted for lee.ka~~ and/ar exccssl'f those 'quantitics of byprod-uct mate ial 'isted in 10 CFR ~ contamination a" intervals ro ta 30.71 Schecule B and all other exceed six months. Thc lc kage sources, including alpha -.est shall be capab'e of detecting emitters, in excess of O.l .he pr senc o 0.005 microcuri of microcurie, shall be free of radioactive materia'n the tost

> 0.005 microcu.i.e o remov- sample.

able contarinati,on. "=ach

b. The periodic lea< test reouired sealed sou=c with removable not apply ta sea d sour"cs 'oes contamination in excess of that sre stored and not be'~~

the above limit shall be immediately withdrawn from used, The sourc s xceptec use and (a) eith " decon- this tes s"..al be =ested =a.

taminated and repai"ed, or leakace p-..icr o:.nv "se a" rans.-.c-.

(b) disposed of in accordance to another user unl ss they have with Commission regulations. 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 ".,

286 scale" sou"ces s~a'~ nct be put Ci 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"-.

pen)!TIentq Nqs., 27 8 24

LIMITIHg CONDITIONS FOR OPERATION SURVEILLANCE REOUIR~ZNTS 3.9.8 Operation vith Inoperable 4.9.B Operation with Inoperable Eccukoa~en t Eauiamenr.

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.

1. Prom and after the date that When one 161-kV line oz'ne common oae 161-kV line'or one common station transformer and its parallel station transformer and its cooling tower transformer or one parallel cooling tower trans- scarc bus is found to be inoperable, former or one*scarc bus becomes all units 1 and 2 diesel generacors inoperable, reactor operation is. and associaced boards must be permissible under ehis condition demonstrated co be operable for seven days. immediately. and daily there after.';

2. When one of the units 1 aad 2 diesel generacor is inoperable, When one oz the units 1 and 2 continued reactor operacion is .diesel generator'is found to be permissible during the succeeding inoperable, all of the CS, RHR 7 days, provided thac both off- (LPCI and Containment Cooling) site 161-kV transmission lines .Systems ahd the remaining diesel and boch common station trans- ~

generators and associated boards formers or one common transformer shall be demonstrated to be operable and oae cooling tover transformer immediately and daily thereazter.

(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 />.

Amendments Nos. 27 8 24

Sl)RVZIl.LAVCF; REnu IR'~EWTS

). h 0~eral, ion Mt th lnoacrable 4.9.8 Oacration with Inoperable

~Eel mene 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 3. %hen one 4-kv shutdown board is energized common transformer) found to be inoperable, all are availaa3.e and the, remain-'ng remaining 4-kv shutdown banters 4-RV shutdown boards an" and associated d'ese'enera-associated d'esel gen ra ors, tors, CS and RPZ (LPCI and CS, RHR (I CZ and Contair ent Containment Coo1.ing) Systems Ccoling) Systems, a>>d eM~ 460 V supplies by the remaining 4-kv emergency powe boards a" shutdown boards shall be demon-ape ab'. his recuirement t to be operable, i...e-s tra ed cannot be ret, an orderly shut- diatcly and daily thereafter.

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.

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 E. Fire Protection S stem E Fire Protection Svs ems Ins ection Insnection

1. An independent fire Any xnspecta.on or audit protection and loss will review and evaluate prevention inspection and audit shall be the effectiveness of prevention and, protection fi e

., performed annually by physical inspect'on of utilizing either plant. facilities, sys ems, qualified TVA and equipment as related personnel o- an to fire safety.

outside fire Evaluations will bo made protection f'm. o f, but not -nece ssarily

2. An inspection and limited to, the following:

audit by an outside Administrative control qualified fire documentation, maintenance consultant will be of fire related. records, performed at physical plant inspection, intervals no greater relat d histor'cal than 3 years. research and application, first inspection(The and and management interviews.

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.

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 the automatic initiation is lost will provide {as does the if 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 pumps are out of service. ):f at least two pumps are if not two 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 Sprinkler System {0.30 gpm/ft~/4440 fthm ar ea) 1332 gpm 2 ~ 1 1/2" Hand Hose Lines 200 cpm

3. Raw Service Water Load 1132 anm TOTAL 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 CO< fire protection is lost in this room and will be established will continue until if 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 I ADMIN STRATIVE CONTROLS

6. 1 Or anizatian The plant superintendent has on-site responsibi 1 ity for the safe operation of the facility and shall report to the Chief, Nuclear Generation Branch. In the absence o f the plant superintendent, the assistant superintendnet will assume his responsibilities.

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

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

D. Shift manning requirements shall, as a minimum, be as described in section 6- 8.

E Qua lif ications 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 ANS I - N1 8. 1, Selection and Training of Nucl ear Power Plant Personnel, dated March 8, 1 97 1 . The minimum frequency of the retraining program shall be every two years.

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

H. Responsibi 1 ities 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 i ty 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 ti wri ng and must be able to write investigative reports and prepare wri tten procedures, ~ Must have the abi 1 i ty 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 i rabl 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 .

amendments Nos . 27 5 24 332

~0 6 0 ADMINISTRATIVE CONTROLS 6.2 Review and Audit The Manager of Power is responsible for the Ferry safe operation of all TVA powerfunctional organization for Review andNuclear plants,-including the Browns Plant. The Audit is shown in Figure 6.2-1.

of facility operation Organizational units for the reviewresponsibilities shall be constituted and have the 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 I ADMIN ISTRAT VE 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 AOME N 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.

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.

b. Annual Oneratina Report.~ Routine operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to March of each year. The initial report shall 1

be submitted prior to March 1 of the year following intial cxiticality.

Amendments Hos. 27 8 24

6o0 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) corrective action taken to reduce the probability of recurr ence, appropriate; if (d) 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.

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

b. 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 ADMINISTRATIVE CONTROLS 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 Enforcement).

Director of Regional Office of Inspection and 1 ~ Reports on the following areas shall be submitted as noted:

a. Secondary Containment 4 ~ 7~C Within 90 Leak Rate Testing(5) days of completion of each test.

b. Fatigue Usage 6 6 Annual Evaluation Operating Report co Seismic Instrumentation 3.2.J. 3 Within 10 days Inoperability after 30 days of inoperability 356 Amendments .Hos. 27 5 24

6 0 ADMXNXSTRATXVE CONTROLS

'/

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 20. 007 of 10 3

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 REGS 4g UNITED STATES

+

4j O~

A NUCLEAR REGULATORY COMMISSION Cl WASHINGTON, D. C. 20555 r+ ~O

+**y4 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROMNS FERRY NUCLEAR PLANT UNIT NO. 2 AMENDMENT TO fACILITY OPERATING 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 J g~ .g a~ rq/rgb'g~

A. Schwencer, Chief Operating Reactors Branch 81 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: February 15, 1977

~ ~ ~ 7

~S AE'gy~

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 IP

~>>*+~

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

Automatic be

'k 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.

Th e b a sis operable.

De ressurization S stem Therefore, the time reduced to seven days whenever'ore ADS 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 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 limit for continued operation must-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

'e than one valve is inoperable proviaed 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 operable rather than four aszstated previously. 'alves 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 wi th 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