ML18025B761

Proposed Revisions to Tech Specs,Consisting of Minor Corrections & Clarifications to Reflect Actual Plant Configuration More Accurately

ML18025B761
Person / Time
Site:	Browns Ferry
Issue date:	04/30/1982
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18025B763	List: ML18025B761 ML20052F568
References
NUDOCS 8205130203
Download: ML18025B761 (97)

Text

UNIT 1

PROPOSED CHANGES REIlIII.UBRVMCKE'f FILE CgPy

,8205i30203 820430 PDR ADOCK 05000259 P

PDR

0

1.0 DEYINITIOHS. continued 2 ~

Q4<<n 4 syst<<m, subsystem,

train, component or d<<vice is dot<<rminod to be inoperable solely because its onsite power source is inoperable, or solelybecause its offsita power source is inoperable, it may be considered operable for the purpose of satisfying the requirements of its applicablc Limiting Condition For Operation, provided:

(1) its corresponding offsite or di<<eel power source is op<<rabl<<;

and (2) all of its redundant system(s),

subsystem(s),

train(s),

component(s) and device(s) are opersblc, or likewis<<satisfy these requirements,

. This provision describes what additional conditionsmust be satisfied to permit operation to continue consistent with the specifications for power sources, vhen an offsite or onsite power source is not operable.

It specifically prohibits operation when one division is inopcrablc because its offsite or diesel power source is inoperable snd a system, subsystem,

tzain, component or device in another division is inoperable for another reason.

This provision permits the requirements associated with individual systems, subsystems,

trains, components or devices to be consistent with the requirements of the associated electrical power source.

It allows operation to be governed by the time limit of the requirements associated with the Limiting Condition For Operation for the offsite or diesel power source, not the individual requirements foz each

system, subsystem, train, component or device that is determined to be inoperable solely because of the inoperability of its offsite or diesel power source.

D.

DELETED

1.0 DFFINITIA'HS (Cont'd) 2.

Run 11ndc - In this 'mode tlie reactor system prcssure is at or above 825 psig nnd thc reactor protection system is encrpksed with APR.'1 protection (excluding the 1'5% high fl>>x trip) and the RBM int'erlocks in service

~

3 0 Shutdown Mode

- Placinp the mode switch to the shutdown position initiates a reactor scram and power to the contrnl rod drives is removed.

After a short time period (about 10 sec),

the scram signal Xs removed allowinp a scram reset and restoring the normal valve lineup in the control rod drive hvdraulic system: also, the main steam line isolation scram and main condenser low vacuum scram are bypassed if reactor ve.,scl prcssure is below 1055 psig.

O.

4, Rcfiinl >!nde - l)ith th<. mode switch in the refuel po..itin>>

int<>>r1oc10 are cst,abl isticd so that one control rod only may bc withdrawn when ttie Soiircc Range Monitor 1ndicit<:.it 1<!asi 3 cp<<

I<n<l th<<refuel 1ng crane 1s n<it ov< r thc r<<nci.<ir; n].;n Lhc main stcam linc isolat.i<>n scram and mi1n condenser l<>w vnc>>>>m scram arc hypass<d if thc r<<actor vessel press>>rc is 1i< )ow 1055 psiR.

If the refuclinp crane is over the reactor, all rods must be fully inserte8 and none can be withdrawn.

Rated Power - Rated power refers to 1002 of the nominal value of the full, steady-state licensed power level of 3293 M<<t authorized by the operating license.

Small, non-steady state power excursions to above 3293 1Nt are permissible for brief periods of time provided that the average power level for an eight hour period does not exceed tho full, steady-state licensed power level of 3293 MMt. Rated steam flow, rated coolant flow, rated neutron flux, and rared nuclear system pressure refer to the values of those parameters when the reactor is at rated power.

Design power, thc power to which thc safety analysis applies, corresponds to 3,440 Wt.

Primary Contninmcnt Tnt<igritv - Primary containm<<nt integr1tv mrn>>!:

that the drywell and pressur<<suppression chnmb<<r are inta< t an<1 all of the following conditions arc satisfiedi 1.

All non-automatic containmcnt isolation valves on lines connected to the reactor coolant systems or containment which are not required to be open during accident cond1tions are closed.

These valves may be opened to perform necessary operational activities.

2.

At least one door in each airlock is closed and sealed 3.

All automatic containment isolation valves are opcrabl" or deactivated 1n the isolated position.

4<.

All blind flanges anil manwnys arc closed.

P.

Secondary Containmcnt I<it<:gritv - Secondary conta1nm<;nt int<i<<r1tv means that tlie reactor huildlng is intact and the fol]o'wing conditions aro met:

2.1 BASES

LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed throughout the spectrum of planned operating conditions up to the design thermal power condition of 3440 MWt.

The analyses were based upon plant operation in accordance with the operating map given in Figure 3.7-1 of the FSAR. 'ue to t)lc statistical nature of the nrocess points to the )teat balance calculation, a certain number of )teat balances will in-dicate a core themal power above t:he full, steady-state licensed power level of 3293 Yhlt when, in

) act, the unit is operating at rated power.

This is eqpecially t:rue following perturbat:iona to steady-state when operating close to <<ate<) power.

Smai't, on-steady state power excursions to above 3293 lit are permissible for brief. periods of time provided that the average power level for an eight hour period does not exceed the full, steady-state licensed power level of 3293 )St.

Xn addition, these excursions are further limited to 100 1/2% of rated power for one hour, 101% of rated power for 1/2 hour, or 302% of rated power for 15 minutes.

But, under no condition

'hould power <<xceed 102% of rated po~er.

Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes.

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

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

Results obtained from a General Electric boiling water reactor have been compared with predictions made by the model.

The comparisions and results are summarized in References 1, 2, and 3.

.The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to be about 25% greater than the nominal maximum value expected to occur during the core lifetime.

The scram worth used has been derated to be equivalent to approximately 80% of the total scram worth of the control rods.

The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay. and slowest insertion rate acceptable by Technical Specifications as further described in reference 4..

The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion.

The rapid insertion of negative reactivity is assured by the time requirements for 5% and 20t); insertion.

By the time the rods are 60% inserted, approximately four dollars of negative reactivity has been inserted which strongly turns the transient, and accomplishes the desired effect.

The times for 50% and 90% insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.

For analyses of the thermal consequences of the transients a

MCPR > limits specified in specification 3.5.K is conservatively assumed to exist prior to initiation of the transients.

This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

In Summary The full, steady-state licensed power level is 3293 MWt and is the maximum time-averaged (8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) power level.

B.

Non-steady-: tate power excur.,ions are further limited to:

a.

< 100 1/?

% of rated power excursions for not more than one hour.

b.

101%

of. rated power for not more than 1/2 hour.

c.

102% of rated power for not more than 15 minutes.

19a

0

2.

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

3.

The abnormal operational transients were analyzed to a power 1cvel of 3,440 MWt.

4.

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

The bases for individual set points are discussed below:

A.

Neutron Flux Scram 1.

APRM High Flux Scram Trip Setting (Run Mode)

The average power range monitori'ng (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated power (3,293 Wt).

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

Duiing transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel.

Therefore, during transients induced by disturbances, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses reported in Section 14 of the Final Safety Analysis Report demonstrated that with a 120 percent scram trip 'setting, none of the abnormal operational transients analyzed violates the fuel safety limit and there is a substantial margin from fuel damage.

Therefore, use of a flow-biased scram provides even additional margin.

Figure 2.1.2 shows the flow biased scram ns a function of core flow.

An increase in the APRM scram setting would decrease the margin pre-sent before the tuel cladding integrity safety limit is reached.

The APRM scram setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation.

Reducing this operating margin would increase the frequency of spurious

scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses.

Thus, the APRM setting was selected because it provides adequate margIn for the fuel cladding integrity safety limit yet allows opcratIng margin bhat reduces the possibility of unnecessary scrams.

20

~

s h

TABLE 3.2.A (Continued)

F ainur K>>" 'on Trio Level Settin Actioa (1 )

Remarks Ins runent Channel-Y~in Stean Line Tunael High Temperature

< 200oF l.

Above.rip setting initiates Yogin Stem Line'!solation 2 (ll)

nstrur ent Chaanel-Reactor Meter Cles:.up Sys e~ Floor Drain High Te.=perature 160 180oF l.

Above trip setting initiates Isolation of Reactor Mater Cleanup Line fro".. Reactor and Reactor Rater Return Lire.

Instrument Channel-R a"tor L'eter Cleanup S; ste.= Spac HiG>>

Te. ccra'tu c

160 180oF C

1.

Sa. e as above Instruaen Channel-Beactcr Bu'dirg Venti-Re"c or Zone

< 100 nr/hr or dovnscale G

1 upscale or 2 dovnscale vill a.

Ini. a'c SOTS b.

Isol c re cto.

son anc.

re."Me.'ng floor.

c.

Close s-.=osphere c

a rol syst =.

Instruren Channel -

< 100 a./hr or dovnscale F

B ac or Building V

'ation High Badia.'cn Re.uleing Zoae l.

1 upscale or 2 dovnscele vill b.

Isolate re.uc'ng f'loor.

c.

Close.a

.".csp..erc control sys c.l, 2 (7) (<a,'. Instr'-.erat Chanael SG.S Flov - T. ain A

Heaters 2 (7)(o) Instruaeat Channe SG;S Flov Train B

Eeaters 2 (7)'.8) Instruaaent Channel SGTS Flov T".sin C

Hcs crs Charcoal Hea ers< 2000 cfh R.

H.

Hc crs< 2000 cf H aad (A or F) 2.

Charcoal Heaters<

2000 cfa R.H. Hca.: s< 2000 cf H and 1.

(A or F) 2.

l.

F)

C ~

Chcrccs Heaters<

200" "".- H cad R.H. Heaters<

2000 cfn (A or Belov 2000 cfa, trip se ting heaters vill tu:n on.

Belo'- 2000 cf...,

rip setting heaters v 1'hut off.>

Belov 2000 cfa, trip sct irg heaters vill u"n:cn.

Belov 2"00 c~, trip setting heaters v 11 shut off.

Below'000 c...-., trip sett ng heaters vi'1. 'rn oa.

Belo: 2000 c~, trip setting heaters vill shu o.f.

charcoal cherco 1

caarcoa R.H.

6.

Channel shared by RPS and Primary Containment G Reactor Vessel Isolation Contr'ol System.

A channel failure may be a

channel failuxe in each syscorn.

7.

A train is mnsidered a trip system.

8.

Two out of three SCTS trains required.

A failure of more than one will require action A and F.

9.

There is only one trip system with auto transfe to two power sources.

l0 Refer'o Tabla 3.7.A and its notes for a listing of Isolacidh Valve Groups and chair initiatinrr, signals.

ll.

Requires two independent channels from each physical location, there axe two locations.

61

TABLE 3.2.6 (Cont (nues)

'jin ta <n

~~o, Opera'olc Fcr

~Ss l)

Funct ian Trf Level Setting Action Renerks Instruncnt Chsaael Resctor lnv Prcssure (FS-68-93 d 9<<',

Su ll) 100 pafR

+ 15

~efav trip setrfng in con)unction ufth contafis<e<<t isolation signal and both suctian valves open Mill close qHR (LP~f) adr<issian valves, Core Sprsr Auto Sequencing o< t < S secs.

Tiners (5)

LPCl Auto Sequencing Tf~rs (5) 0< t <1 dec

~

QQSM A3.,

63, C3.,

end 0 3 l 3 < t < 15 sec

~

7 inc rs i.

41th diesel paver 2 ~

One per Rotor Mith diesel povet 2 ~

On ~ pcr eatat 4fth diesel paver Ooe per puup Core Spr4Z end LPC1 Auto Sequent lot Tlfoers (6) 0 < r. < 1 sec.

6 < t < 8 sec.

12

< t < 16 sec.

18 t < er sec.

1,

'Hfth navaal pover 2.

One per CSS notor 3.

Tva per RhR avatar RDRSM A3.,

5 3, Cl, end D 3 Timers 27< t <29 sec.

1.

Vith noraal pover 2.

Cne per punp

.ABLE 3.?.b (Continued)

'Njniaua 'vo.

Operable Per Function Trio Level Sectin Action Reearks Core Spra> Trip System bus pover oonicor H/A monitors availability of pover to logl,c sysceas.

ADS Trip System bus pover oonitor N/A

'Monitors availability of pover to'ogic systens and valves.

HFCI Trip System bua pouer noni ter H/A C

Non 1cors availability of pover to logic systens.

RCIC Trip System bus pover nonl tor H/A C

Honitors availability oi paw>>r to logic systeas.

CZs Oa 1(2)

Iostrunen hanuel Elev.

551'ondensate Storage Tank Lou Level (LS-73-55A &. B) h-Belou trip setting vill open E?

suction valves to the suppression chamber.

l(2)

Inscruaent Channel-Suppressioo Chaaber High Leva)

< 7" above instrument zero A

l.

Above trip setting vill open HPCI auction valves to the suppression chamber.

2(2)

Ins c runen c Channel Reactor High Mater Level,

I Instrument Channel RC)C Turbine Stean Line High Flov

< 583" above vessel zero.

< 450" H 0 (7)

1. 'bove trip setting trips 8

IC turbine.

1.

Above trip setting isolates R IC system and trips RCIC turbine.

NOTES fOR TABLE 3.2.8 Mhenevor any CSCS Sy tem io required by section 3.5 to be operable, yo If' thoro shall ba two operable trip systems excopt ao noted.

'a requirement of tha first coluzan is reduced by one, the indicated action shall bo taken.

If tho oamo function is inoparablo in caoro than one trip aystem or tho first column reduced by ~re than one, action B shall be taken.

Action4 A.

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

If the function ia not operable in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, take action B.

B.

Declare the system or component inaperablo.

C.

Ioaaodiataly take action B until power ia verified on the trip systems D,

No action required, indicators are conoidored redundant.

2.

In only one trip ayoten.

3.

Hot considered in a trip system.

4.

Requires one channel from each physical location (there are 4 loca<<

tiona) in the steam line space.

5.

Mith diesel power, each RHRS pump ia achoduled to start immediately and each CSS pump io sequenced to start about 7 sec later.

6.

Mith normal power, one CSS and one RllRS pump ia scheduled to otart-inatantaneously, one CSS and one RHRS pump is sequenced to start after about 7 sec with similar pumps starting after about 14 aoc and 21 ooc, at which time tho full complement of CSS and RES pumps would be oporatinW.

7.

The RCIC and HPCI steam line high flow trip level settings are ivan in terms of differential pressure.

The RCICS setting of 450" of water corresponds to at least 150% above maximum steady state steam flow to assure that spurious isolation does not occur while ensuring the initiation of isolation following a postulated steam line break.

Similarly, the HPCIS setting of 90 psi corresponds to at least 150% above maximum steady state flow while also ensuring the initiation of isolation

~following a postulated break.

8.

Note 1 does not apply to this item.'.

The head tank is designed to assure that the discharge piping from the CS and RHR pumps are fu11.

The pressure sha11 be maintained at or above the values listed in 3,5.H, which ensures water in the discharge piping and up to the head tank.

7l

TABLE a 1

A SmVRZLLAmd RROaiaVCmTS tOR IRuuay COmhimCeRT ARD RBACTOR BDZLDZRG ZSDLATim ZVSTROXurTATZ(RI Function Group 6 Logic Group d (Initaating)

Logic Reactor Buiiding Isolation (refuel ing floor) Logic Reactor Building Isolation (reactor xone)

Logic SGTS Train A Logic SGTS Train B Logic sGTs Train c Logic static pressure Control co (refueling floor) Logic Static pressure Control (reactor xone)

Logic functional Test once/operating cycle (10)

Checked during channel functional test.

Ho further test required.

once/6 saonths

( Id) once/6 months (18]

once/6 oonths (19) once/6 months (19) once/6 sonths (19) once/oper a' ing cycl e (1d) once/opera ting cycle (1 d)

Calibration fr uenc 5/A (6)

(6)

II/h (6)

(6)

Instrument Check 8/A 8/A M/A Instruaent Channel Reactor Cleanup Systen-tloor Drain High Tenperatur>>

lnstrusent Channel-Reactor Cleanup Systea space High Temperature once/gperating cycle 8/h once/operating cycle'/A

Tel.E 4.2.h ((.oat~nved).

I'ua:tion Inst~at Channel Laactot Lou tressure (FS-68-93 4 94)

Functional Test Calibre tioa once/3 <<oaths Insttmaent

&~el<.

nons Cote Spray Auto Sequnacing Ttaers

(ÃareLal t~er)

(4) once/cpernting cfcle Corn Spray Auto Sequencing Tiinrn QCesnl Paver) fPCI Auto Sequencing Tieet's

-(Lo~i l'~r) oncn/0~rating cp Xs once/operating c~e LK[ Auto Sequencing Theetn Qiasel Parer)

(4) once/opcrntiilg ~Is KQKV Al. g3 Cl, 03 Tiaori 45otce1 2'~r)

(4)

EMn/0$4rnt JJtg cfcle

~ Al, R.3 cl, 93 TiNEsrs el Penner)

(4) men/operating eycla (4) oncelepmt ing ej-In

HOTtS fOll TASLES 4. 2. A THROUCH 4. 2.H Conc tnued 14.

Upscal ~ trip ts (unctionally tested during (unct tonal test liras aa required by rection 4.7.b.l,a and 4.7.C.l.c.

15,,

The ( lou bias cocaparator vttl ba tested by putting one (lou uaLt Teat" (producing 1/2 scram) and ad)us( tng the test fnpu t to sbts fn cosparator rod bloc'k.

The (lov bt ~ ~ upscale vill be verfffcd by observing a local upscale trip light durtng operation and vert fied that Lt vill produce a rod block during thc operating cycl ~.

1&.

Performed during operating cycle.

Portions o( the logic Ls checked wre frequently during fun'ctional taste o( the (unctions that produce a rod block.

17.

Thta calibration const its o( removtng the (unctLon (ron service and per(oaafng an electronic calibration of the channel.

LS.

Punctional tc ~ t ts ttratted to the condition vhcre

~ econdary contafnmcnt Lntegrtty ts not required as spccf tied tn sections 3.7.C.2 and ],7

~ C,),

29, Puncttonat test ts ttsst ted to thc t tee uhere the SOTS Ls required to laeet thc requtrenents o( section 4.7.C.l.c.

20, Cattbrst ton o( thc cosparator require ~ the inputs (rory both recirculation loops to be tnterruptad, thereby removing the (lou bta ~

~ ignal to the APRlt and IUD aid scramming the reactor.

Thfs calibration can oaly be per(armed durtng an outage.

21.

Logic tent ts limited 'to the ttms vhere actual opcratLon of tha equipment Ls permtssibt

~

22'nc channe 1 o( ctthcr the reactor tone ur re(ueting tone Reactor guttdtng Vcntilatton Radiation Honitortng System oay be adntnistrattvety bypassed for a period not to exceed 2'ours (or (unctional testtng and cattbratipn.

23.

Deleted

'4 This instrument check consists of comparing thc thermocouplc readings (or all va1vas for consistence and. for nominal expected values (not required during refueling outages)

During each refueling outaRc, all acoustic sonitoring channels shall be calibrated.

This calibration includes vcrificatfon of accelerometer response dua to mechanical cxciration in tha vicinity of thc sensor

~

26.

This tnstrument check consists of comparing the backRround signal levels (nr all valvas for consistency and for nominal expected values (not required durinR rafuelLng outapes).

110

~ ~

ll'r IHF: CnnnlTtnutt Vdu( ilPYHhTlnH gunvEtu.met:

Hu ~~xy ~H:Hr I,S.S non Id oi llollt II n o'S'.o.

~nnnSI ICFCI ond Contoinnont Cooling) 4 ~ 5 ~ b Residual Heat R Fdnval g

atedc

~IIKIS (I,FCI nnd Contodnr ont Cooling)

4. If any 2 RHR puffs become inoperable, shall be placed Ln shutdown condition 24 bourse'LPCI mode) th reactor the cold within 4.

No additional surveillance required.

5.

1i one RHR puFFp (con:ain>>

~ant cooling IF >dc) o'a-nocidsted heat exthanxer Ls inoperable, the reactor uay renaln in operation (or a period not tu exceed 30 days pre vldtod thc reualnlng glib. puops (conteLneent cooling node) and aaao-ciatcd heat exchengera snd d'.eccl groeratnrs and all access paths u( th.

RHRS (cont~invent cooling node) are operable.

5, Mnen it ia determined that one "othe pu'.ap

{Cuntainsaent cooling esode) or r.saociat d heat exchanger io inoperable nt o

tine uhen operability lo'e-

quired, the reaalninst RHR pumps (containrsent cooling node).

the aosociatcd heat cxchsng"rs erd diesel generators, and all active conponenta

'n t'e access pacha on the lUBS (contalntaent cooling node) shsLL bc ole.nn-atrated to bc operabl ~ icsiediatcly anti use'd:ly'thereanter unn ll t'e inoperable RHR pucFp (contslntttent cooling fade) and aaaos'strd hest.

exchanger ie returned to normal service.

6, ll tvo RHR ruaps (Contain>ant coolins code) or associated hes exchangerv are Lnopers-bic, the

".cocci ~ r osy remain in apcrst lon t"r

~

pd riod not to exceed 7 days pro-vided the ree~>ndina kHR punps (contalndicnt cooling csctde) s associated heat exchangers, diesel generators, and a11 access paths of the PHRS (containment 'cooling mode) 6, N:en it is determined chat RHA puopa (conte incsent:ooL!ng etoffe) or aasoriat d heat ed:chsssd,are are inoperabLe at a

inc tdhen operab(lity Ls rediuired, Lhc refraining RHR ptnpa

(=ontalidtatcnt coo! ing mode),

thc associated heat

<<xchang rs, diesel gsneratorss and all oct."Ie cce" ponents in !nc access paths o(

tha RHRS {concainoent cool!n" 147

BASFS opening.

If the check snd green or check light circuit alone fo inoperable, the valve shall bc considered inoperable for full closure.

If the red and check light circuits arc inopcrablc the valve shall be considered inopera-ble and open greater than 3

For a light circuit to bc considered operable the light must go on and off in proper sequence during the opening-closing cycle. Ifnone of the lights change indication during the cycle, the valve shall be considered inoperable and open unleso the check light stays

'on and the red light stays off in which case the vaIve ohaII be considered inopera-ble for opening.

The twelve dryMell vac'uum breaker valves which connect the suppression, chamber snd.drywcll arc sized on the bsofo of the Bodcga pressure suppreo-efon system teoto.

Ten operablc to open vacuum breaker valves (18-inch) selected on this Lest basis and confirmed by the green lights are adequate to lfmit the preooure differential between the suppression chamber and dry" well during post-accident drywall cooling operationo to a value which is within ouppressfon system design values.

The containment design hao been examined to determine that a leakage equi-valent to one drywell vacuum breaker opened to no more'han a nominal 3's confirmed by the rcd light is acceptable.

On thio basis sn fndcffnite allowable repair time for sn inoperable red light circuit on any valve or an inoperable check and green or check light circuit alone or a malfunction of the operator or disc (if nearly closed) on one valve, or an inoperablc green and rcd or green light circuit alone on two valves is )ustified.

During each operating cycle, s leak rotc test shall be performed to verify that significant leakage flow paths do not exist between the dryweL and suppress'fon chamber.

The drywell preosure will be increased by at least I poi with respect to the suppression chamber pressure and held constant.

Thc 2 poig oet point will not be exceeded.

The subsequent suppression chamber prcssure transient (ff any) vill bc monftorcd with a sensitive pres-sure gouge.

If thc drywall prcssure cannot be increased by I psi over the suppression chcmber prcssure it would be because a significant leakage path exists; fn thin event the leakage source will be identified and eliminated before power operation fs resumed.

arith a differential prcssure of greater than I psig.

the rate of change of the suppression chamber presoure must not cxcevd 0.38 inches Of Water per minute 'as measured over a '10 minute period, which corresponds to about 0. 14 Ib/oec of containment sir.

Zn the event the rate of change exceeds this

'alue thon the source of leakage will be identified and eliminated before power operation fs resumed.

The water in the suppreosion chamber fs used for cooling in the event of an accident f.e., it is not used for normal operation; therefore,'"

a daily check of thc temperature and volume fs adequate to assure that adequate heat removal capability io present.

272

l.lK'ITIHO COHVlTIOHS FAR OPERATION SURVEZLLAHCE RE VIREKEHTS

3. 10 CORY. Al.TERATIOHS 4elO CORE ALTERATIONS Applies to the fuel handling and core reactivity liroitationo.

Applies to the periodic testing of those interlocks and instru-mntation used during refueling and core alterations.

O~t eettve

~Ot ective To ensure that core reactivity is Mithin thc capability of thc control rods and to prevent criticality during refueling.

To verify the operability of'nstruraentat ion and interlocks used in refueling and core alterations e

S ecification S ccification A.

Rcfuclin Interlocks h.

Refuel in Inter locks 1,

The reactor made ovitch shall be locked in the "Refuel" position during core alterations and the refueling interlocks shall be operablc except as specified in 3.10.A. 6 and 3.10.A. 7 beloM.

l.

Prior to any fuel hand-ling Mith the head off the reactor 'vessel, the refueling interlocks shall be functionally teated.

They shall be tcstcd at Meekly inter-vals thereafter until no longer required.

They shall also be tested fol-lovfng any repair, Mork associated Mith the inter-locks.

2.

Pucl shall not bc loaded into thc reactor core unloss all control rods are fully inserted.

No additional surveillance required.

302

L I ~(IT INC CONI)ITIONS FOR OP FRATION SURVEILLANCE RE tjIREHENTS t

o. IO.A Rcfue1 in Interlocks

~

3.

The tuel grapple hoist load suitch shall be set at

< 1,000 lbs.

4.

If the frame-mounted auxi-liary hoist, thc monora'il-mounted auxiliary hoist, or the service

'platform hoist le be used for handllnc fuel Mith thc head off the reactor vessel, the load limit switch on the hoist to be used shall be set at i 400 lbs.

5.

Maintenance may be performed on a single contiol rod or control rod drive without removing the fuel in the con-trol cell if the following conditions are met:

4.10.A Refuclin Interlocks 3.

No additional surveillance required.

4, No additional surveillance required.,

5.

Prior to performing control rod or control rod drive maint.enance on a control cell without,".removing fuel assemblies the surveillance requirements of specification 4.10.A.l shall be performed and all rods face adjacent o'r diagonally adjacent to to the maintenance rod shall be electrically disarmed per specification 3.10.A.5.b.

a.

The requirements of specification 3.10.A.1 arei~et, and b.

All control rods diagonally or face ad)'acent to the maintenance rod are fully inserted and. have their directional control.valves hlectrically disarmed.

303

G ON I 0

P 3.10.A.6 A maximum of two non-adjacent

~

control rods may simultaneously be withdrawn from the core for the purpose of performing control rod and/or control rod drive maintenance without removing the fuel from the cells pro-vided the following conditions are satisfied:

4.10.A.6 Prior to performing control rod'r control rod drive maintenance on two control cells simultaneously without removing the fuel from the

cells, two SRO's shall verify that the requirements of specification 3.10.A.6 are satisfied.

The reactor mode switch shall be locked in the "refuel" position.

The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed for one of the control rods on which maintenance is being performed.

All other refueling inter-locks shall be operable.

303A

~ 'llllrl>>s co>>nlTlo>>:;

rr>>t orvtATlo>>

3. lO,A Ncfuclin~lnlor lncko

,",URVKILLLNCY. NYI IIIIIIUIFNYS 4.10.A,Rc(uel in inter locko 6.

(Continued) b.

All directional control valves for remaining control rods shall be disarmed electrically except as specified in 3.10.AN7 and sufficient margin to criticality shall be demonstrated.

c.

The two maintenance cells must be separated by more than two control cells in any direction.

d.

An appropriate number of SRM's are available as defined in specification 3N10.3.

Any n<<mhrr n( control rode YYYay hr uithdrnun or removed froYo the reactor coro pro-vidinlt the fo1 loving condi-tiono arc aatioficd:

a.

Tha reactor mode ouitch ia lockcYl in thc "ro-fucl<<position.

The rcfuoling interlock uhich prcvcnto morc t)mn ono control rod from

" th th<<~de selec

>en s ltth the refuel or shutdown

mode, ro more than one control 'rod may be withdrawn without first removing fuel from the cell except as specified in 4N10.A.6.

Any number'f rods may be withdrawn once verified by two licensed operators that the fuel has been removed from each cell.

304

) o 10 BASES rnds snd cl<c refueling platform provid~

<'edundanc m>>chod>> of prevent ing inadvertent criticality even af ter procedural violations.

Thc interiocks on hoists provide yet another method of avoiding inadvertent criticality.

Fuel handling is normally conducted Mich the fuel grapp)e hoist.

Thc

'otal load on this hoist Mhen the interlock is required consists of the Weight of thc fuel grapple and the fuel assembly.

This total is approxi-mately 1,500 Ibs, in comparison to the load-trip setting of 1,000 lbs.

provisions have >>lao bccn made co alloM fuel handling Mith either nf chc three auxi)i'<ry hoists snd still maintain thc refueling interlocks.

The 400-lb load-trip setting on the>>c hoi>>ts ia adequate to trip the interlock vhcn nnc of the morc chan 600-lb fuel bundles is being )<andlcd.

During certain periods, it is dcsirablo to perform maintenance on tuo control rods and/or control rod drives at, the sane time without removing fuel from the cells.

The maintenance is performed with the mode.

switch in the "refuel" position to provide the refueling interlocks normally available during refueling operations.

In order to with-draw a second control rod after withdrawal of'he first rod, it is

~ necessary to bypass the refueling interlock on the first*control rod wHich prevents more than one control rod from being withdrawn 'at the same time.

The requirement that an adequate shutdown margin be demon-st'rated and that all remaining control rods have there directional control valves electrically disarmed ensures that inadvertent criticality

~ cannot occur during this maintenance.

The adequacy of the shutdown margin is verified 'by demonstrating that at least 0.38%

<<k shutdown margin is available.

Disarming the directional co'ntrol valvessddes not inhibit control rod scram capability.

Spccif ication 3.10.A.7 sllous unloading of a significant portion of che reactor core

~

This operation ia performed Mich thc mod>> suit>>h in thc "rc(ucl" position to provide thc refueling interlocks normally available during refueling operations.

In order to vichdrau morc chan one control rod, it io n<ccssnry co bypass the refueling interlock on each vitl<draun control rod vl<ich prevents morc than one control rod from being Mithdrsvn at <<c i<<<e.

Thc requirement chsc the fuel assemblies in che cell controlled by tl<c control rod bc removed from thc rcsccor core before chc interlock can be bypassed ensures that. vithdraval of anochcr control rod does noc result in inadvertent criticality.

Each control rod provides primary reactivity concrol for the fuel assemblics in rhc cell associated Mith chac control rod.

Thus, removal of an entire cell (fuel aa>>cmblics plus concrol rod) results in a lover rractivicy potential of thc core.

The rcquircmcncr for SR.'i operability during chase core alterations assure sufficient core monitoring.

3IO LOCAL MONlTORING STATIONS BRQV/MS FERRY NUCLEAR PLANT Fi

. 4.2-1 ATH'"NS U.S.

HWY 72 r5 BOP ALA. HWY 20 Legend G

Air Monitor Air Monitor 8 TLD Station TLD Station h.

Automatic Well Sampler H

Dairy Form DECATUR Scale O

l 2

a 4

6 Miles

UNIT 2 PROPOSED CHANGES

1.0 DEYINITIONS continued 20 Mhan a system, subsystem, train, component or device is determined to be Snoperablo solely because its onsitc povcr a'ourca is inoperable, or solelybecause its offsite povar source is inopcrablcf it msy be conoidared operable for the purpose of satisfying the requirements of its applicable Limiting Condition For Operation, provided:

(1) its corresponding offaite or diesel povar source is operable; and (2) all of its redundant system(o),

subsystem(s),

train(s)

~

cozqmncnt(s) and dance(s) are opcrnblc, or likevisc satisfy these requirements.

". This provision describes chat additional conditionsmust be satisfied to permit operation to continue consistent vith the specifications fox pover sources, vhcn an offsite or onsitc povcr source is not operable.

It specifically prohibits operation vhen one division is inoperable because its offsitc or diesel paver source is inoperable and a,system, subsystem, train, component or device in another division is inoperable for another reason.

This provision permits the requirements associated vith individual systems, subsystems,

trains, components or devices to.be consistent with the requircreats of the associated electrical povcr source.

It allovs operation to be governed by thc time limit of the requirements associated vith the Limiting Condition For Operation for the offsite or diesel povar source, not thc individual rcquircraats for each

system, subsystem, train, component or device that is determined to be inoperable solely because of thc inopcr bility of its offsite or diesel power source.

D.

DELETED

1.0 nFFINITIONS (Cont 'd) 2.

Run Mode -

Tn this mode the reactor system pressure,is at or above 825 psi1l and the reactor protection system is encrpfzcd with APRM protection (cxcludinp th<

15/ h5ph fl<<x trip) and the RAM Interlocks in servic<<.

Shutdown Mode

" Placinp the mode switch to the shutdown position initiates a reactor scram and power to the control rod drives is removed.

After a short time period (about 10 sec),

the scram slpnal is removed allowlnp a scram reset and restorinp the normal valve lineup in the control rod drive hydraulic system: also, the main steam line isolation scram and main condenser low vacuum scram are bypassed if reactor vessel prcssure is below 1055 psip.

0.

Re.'ur.l Mode - With the mode switch in the refuel position interlocks are cstablishc<l so that one control rod on)y mny bc withdrawn when thc Source Range Monitor indicate at least 3 cps no<i ch<

r<<fu<.'1 fn<; cram<:

Js n<<r ovc r ihc r< n<:r<~r; a]< o the main ste<<m ii<1>> fs<>lation scram and

<na5n cond<.'user l<>w vacuum scram arc bypass<<d if thc r<<actor vess<1l press<<re is h< lo<<

1055 psip.

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

Rated power - Rated power refers to 1002 of the nominal value of the full, steady-state licensed power level of 3293 N<t authorized by the operating license.

Small, non-steady state power excursions to above 3293 N<t are permissible for brief periods of time provided

'haY the average power level for an eight hour period does not exceed the full, steady-state licensed power level of 3293 Sent.

Rated steam flow, rated coolant flow, rated neutron flux'nd rated nuclear system prcssure refer(to the values. of these parameters when thc reactor is at rated power.

Design power, thc power to which the safety analysis applies, corresponds to 3,440 Wt.

Primary Cnntainm<.nt Int<<prirv - Primary containment intceritv m<an..

that the drywell and pressur<:

suppression chamb<<r arc intact.

and al 1

of the following conditions arc satisfied:

1.

All non-automatic containment isolation valves on 1ines connected to the, reactor coolant systems or containment which are not required to bc open during accident conditions are closed.

These valves may he opened to perform necessary operational activities.

2.

At least one door in <.ach airlock,is closed and sealed.

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

All blin<l fiances an<1 manways arc closed.

P.

Sccondarv Contaknmcnt Tnt<.nritv - S<.con<lary containm<,nt in<<<<1 ity means that the r<<actor huildinp is intact and th<< fo)]'o'winP.

condit ious are m<1t:

BASES:

LIMITING SAFETY SYSTEM SETTINGS RELATED TQ FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed throughout the spectrum of planned operating conditions up to the design thermal power condition of 3440 MWt.

The analyses were based upon plant operation in accordance with the operating map given in Fig'ure 3.7-1 of the FSAR.

Due to the 'tatistical nature of the process points to the heat balance c lculation, a certain number of heat balances will in-dicate a core thermal oower above the full, steady-state licensed power level of 3293 KAt when, in fact, the uni.t is operating at rated power.

This is eqpecially true following perturbation" to steady-state when operating close to rated power.

Smail, non-steady state power excursions to above 3293 iQt are permissible for brief periods of time provided that the average power level for an eight hour period does not exceed the full, steady-state licensed power'level of 3293 1Qt.

In addition, these excursions are further limited to 100 1/2% of rated power for one hour, 101% of rated power for 1/2 hour, or 102% of rated power for 15 minutes.

But, under no condition should power exceed 102% of rated power.

Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes.

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

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

Results obtained from a General Electric boiling water reactor have been compared with predictions made by the model.

The comparisions and results are summarized in References 1, 2, and 3.

The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to be about 25% greater than the nominal maximum value expected to occur during the core lifetime.

The scram worth used has been derated to be equivalent to approximately 80% of the total scram worth of the control rods.

The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay, and slowest insertion rate acceptable by Technical Specifications as further described in reference '4..

The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion.

The rapid insertion of negative reactivity is assured by the time requirements for 5% and 20% insertion.

By the time the rods are 60% inserted, approximately four dollars of negative reactivity has been inserted which strongly turns the transient, and accomplishes the desired effect.

The times for 50% and 90% insertion.are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.

19

2.1 Bases For analyses of the thermal consequences of the transients a

MCPR > limits specified in specification 3.5.K is conservatively assumed to exist prior to initiation of the transients.

This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

In Summary The full, steady-state 3icensed power level is 3293 MWt and is the maximum time-averaged (8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) power level.

D.

Non-steady-state power excurs3ons are further limited to:

a.

< 100 1/2

. of. rate'd power excursions for not more than one hour.

b.

< 101%

of. rated power for not more than 1/2 hour.

c.

102% of rated power for not more than 15 minutes.

19a

2. 1 BASES 2.

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

3.

The abnormal operational transients were analyzed to a: power Level of 3,440 MWt.

4.

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

The bases for individual set points are discussed below:

A.

Neutron Flux Scram 1.

APRM High Flux Scram Trip Setting (Run Mode)

The average power range monltori'ng (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated power (3,293 MWt).

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

During transients, th instantaneous rate of heat transfer from the fuel (reactor e

d thermal power) is less than the instantaneous neutron flux ue to the time constant of the fuel.

Therefore, guring transients induced by disturbances, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses reported in Section 14 of the Final Safety Analysis Report demonstrated that with a 120 percent scram trip 'setting, none of the abnormal operational transients analyzed violates the fuel safety limit and there is a substantial margin from fuel damage.

Therefore, use of a flow-biased scram provides even additional margin.

Figure 2.1.2 shows the flow biased scram as a function of core flow.

An increase in the APRM scram setting would decrease the margin pre-sent before the fuel cladding integrity safety limit is reached.

The APRM scram setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation.

Reducing this operating margin would increase the frequency of spurious

scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses.

Thus, the APRM setting was selected because it provides adequate margin for the fuel cladding integrity safety limit yet allows operatinp margin that reduces the possibility of unnecessary scrams.

II I

20

~

~

h Yinit!up'ABLE 3.2.A (C ntinued)

Fls'.".i0 n In trunent Channel

'Vain Stean Line Tuxael

.High Temperature Trio Level Set in.

<2CO F

Action (1)

Remarks Above.rip s tting initiates Vain Stea1 Line'Isolction 2 (11)

"rs rur cnt Cha".ncl-Reector Meter Clee:.up Sjs ew Floor Drain High Te.=oerature 16o - 18ooF l.

Above trip setting initiates Isolation of Reactor Mater Cleanup Line fro-. Reactor and Reactor Rater Return Line.

Instrument Channel-R a"tor Meter Cleanup SJ ste.

Space High T.-.ccraturc 160 - 180oF 1.

Sa.

e as above Instru~en Channel-Reec cr Bu'ding Venti-lat'on High Radiation-Re"c or Zone Instre.".en Channcl-c or Building V--'-

'e ion High B dieticn Be.'~cirg Zone

< 1CO nr/hr or dovnscalc G

< 100 nr/hr or dovnscele F

I 1 upscale cr 2 doiwsce~e vill a.

Ini iatc SOTS b.

Isol te re"cto.

tone anc.

refule'g floor, c.

Close at."os'!re control syst =.

l.

1 upsc lc or 2 dovnscale vill b.

Isolate refuc'ing floor.

c.

Clcse et"csphe.c cont"ol s('st c.l.

2 (7)(c), Instr ".ent Channel SG.S Flov T. a'n A

Hce ers 2 (7)(o) Instruaen hanne SG;S Flov - Train B

Hee ers 2 (7).E,) Instrunent Channel SG.~ Flov - Tr~in C Hca crs Cherco 1 Heaters<

2000 cfh R.

H.

He.crs< 2000 cf Charcoal Heaters

< 2000 cia R.H. Hca.: s< 20CO Chcrcc i'.eaters<

2000 c.=.

R.H.

Hea cr

< 2COO cfn H end (A or F)

H end (A or l.

F) 2.

l.

F)

H and (A or Bclov 2000 cfn, trip s=-t'ing

?:es c>s vill:u:n on.

2.

Belov 2000 cf...,

rip setting heatCrS v'lh Shut Off Belov 2000 cfn, tr p setting heaters vill tu"n cn.

Belov 2"00 c', trip setting heaters v'll shu'ff:

telos 2000 c~t, rip setting heaters vi'1 rn on.

2.

Belov 2000 c~, tr'p setting hcatcrs vill shut off.

charcoal cherco 1

R. E.

chh <<coe!

R. H

~ ~

Channel shared by RPS and Primary Containment G Reactor Vessel Isolation Control System.

A channel failure may be a

channel failure in each system.

7.

A train is mnsidered a trip syst,em.

e.

Two out of three MTS trains required.

A failure of more than ono will require action A and F.

Ther'e is only ono trip system with auto t ansfe to two power sources'0.

Refer to Table 3.7.A end its notec for a liating of Xaoletldh Valve Groups and the1r initiating signala.

11.

Requires two independent channels from each physical location, there are two locations.

TABLE 3.2.0 (Conttnu<C)

.'linta 'n IIG, Ope rao 1 <

Pe r 5

lj Funct ion lnstrua<nt Channel Reactor lAw Prcssure (YS-68-9 3 6 9t, SM Ii)

Tri Level Setttr 100 paly

+

1S Action Remarks l.

Below trip setting in con)unction with contsfisae<<t isolation signal and both suction valves open will. close RHR (LPCl) adrufssion valves.

Core Spr>v Auto Sequencing 6< t < 8 secs.

Timers

{5).

l.

Pith diesel power 2r One p! r. Rotor Cs LPC1 Auto Scqueaciag Tio<rs (i)

RiRS'M A1, B3, C1, and 0 3 Tiaers 0< t <1 cec.

13

< t <15 sec.

1.

Mith diesel power 2.

One per eator

'4fth diesel pover Ooc per pump Core Spray and LFCl Auto Sequencing Timers (6) 0<

6<

12 E

18 t < 1 sec.

t << 8 aec.

t < 16 sec.

t < Qr sec.

1, With normal power 2.

One per CSS notor 3.

Two per RllR tnotor RRRSM h].,

B 3, C3.,

and 0 3 Tieers 27< t <29 aec.

A 1.

arith normal power Cne per peep

.ABLE 3.2.b (Continued)

Mintauo Vo.

Operable Per

~Tr t

~?

~ (I)

Fun cion Trio Level Sect tn Acc ion Renarka Core Spray Trip Systee?

bus pover c<?ntcor 1.

Monitors availability of pover co logtc systens.

ADS Trip Syscen bus pover nontcor H/A C

1.

'Nonicors availability of pover to logtc systee?s and valves.

HFCI Trip Syste?e bua pouer noni ter C

1.

Monitors availability of pouer to logic systone.

RCIC Trip System bus pover nonttor

?I/A 1.

Monitors evailabiltty oi pover to logic systee?s.

Cll o?

l(2)

Ioscruoent han?ael

> Kiev.

551'ondensate Storage Tank Lou Level (LS-73-55A 6. B)

A-Belou trip sett'ng ?Itll open IP~'uccion valves to che suppression chamber.

1(2)

~

Inst ruoent Channel Suppressioa Cha"-ber High Leve) l

< 7" above instrument zero Above trip setting vill open HPCI auction valves to the suppresston chaaber.

2(2)

Instrunent Channel Reactor High Mater Level il Instrument Channel RCIC Turbine Stean Line High Flov

< 583" 'above vessel zero

< 6SO" H 0 (7) l.

Above trio setting trips R IC tuAtne.

1.

Above trip set ttng iaolates RC'lC syste!

and tripe RCIC turbine.

'L NOTES FOR TABLE 3.2. B 1.

Mhonovor any CSCS S

c n ia required by oection 3.5 to b<<opirablo, yi eel cd.

Ef a thoro shall be cwo oparabla czip oyotemo excopt ao noted.

roquiremonc of the first column io roduccd by one, the indicated action shall bo taken.

Ef tho aomo function is inoperabla in moro chan one trip ayocom or the first column roducad by ~re than one, action 5 shall ba takeo.

hccion:

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

?f the function ia not operable in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, tako action 8 ~

5.

Declare the system or ccmponent inoperable.

C.

Euraediacely take action 3 until power ia vorified on tho trip system.

2 ~

D.

Ho action required, indicators aro considered redundant.

En only one trip ayscea.

3.

Hot considered in a trip ayatom.

Requires one channel iree each physical location {there ara 4 loca-tiona) in tho steam line space.

5 ~

6.

7.

8.

Mith diesel

power, each lU!RS pump ia ochedulcd to start immediately and each CSS pump io sequenced to start about 7 soc later.

Mith normal power, ona CSS and onn RHRS pump ia achodulod to start instantaneously, one CSS and one RHRS pump is sequenced to start ofcar about 7 aec vith similar pumps starting after about 14 aoc and 21 ooc, ac which ciao tho full complement of CSS and RHRB pumps would be oporatint.

The RCIC and HPCL steam line high flow trip level settings are

'ivan in terms of differential pressure.

The RCZCS sect<<g of 450" of water corresponds to at least 150% above maximum steady state steam flow co assure that spurious isolation does not occur while ensuring the initiation of isolation following a postulated steam line break.

Similarly, the HPCES setting of 90 psi corresponds to at least l50% above maximum steady state flow while also ensuring the initiation of isolation following a postulated brcak.

Note 1 does not apply to this item.

9.

The head tank is designed to assure that the discharge piping from the CS and HHR pumps are full.

The pressure shall be maintained at or above the values listed in 3.5.H, which ensures water in the discharge piping and up to the head tank.

71

'I

~ ~

TABLt a 2

A SURVEILLAHCE REQOIRENEHTS POR PRIHART COY1'AIHHEHT AHQ REACTOR BOILDIHG IOOIATIOH IHSTRQKEllTATIOH tunction Group 6 Logic Group 8 (Initzating) Logic Reactor Building Isolation (refueling floor) Logic Reactor Buiiding Isolation (reactor xone)

Log),c SOTS Train h Logic sGTs Train B

Logic'OTS Train C Logic Static Pressure Control cs (refueling floor) Logic co Static Pressure Control (reactor zone)

Logic Functional Test one e/0 pe rating cycle (1 d)

Checked during channel funct,iona 1 test.

Ho further test reguired.

once/6 months (1d) once/6 nonths (18) once/4 eonths (19) once/6 nonths (19) once/6 nonths

~ (19) 0nc e/ope r ating cycle (18) once/oper at ing cycle (18)

Calibration Fr uenc H/h (6l (6)

(6)

(6)

Instrument Chec'k H/A H/A H/A H/A Inatrunent Channel Reactor Cieanup Systos rloor Orain High Tenperature Instruaent Channel Reactor Cleanup Syeten Space 8igh Temperature once/qperating cycle j'/A once/operating cycle'/A

TABLE 4.2.5 (Cont~sued)

Function Inst~at Channel meeter Lao Pressure (FS-48-95 a S4) functional Test Cclibratfoa once/5 aLonths Core Spray Auto Sequenciag Tiaaers (Zoril tearer) once/operating cj'cia Core.Spray Auto Sequoia@,

Tfaaars g&ce2 Pcaar) once/ops:rat Lag cycle LOCI Auto Sequencing Ttacra

~ (!++~a P~r)

IPCI Mta emquanctag Tkmtt'a Qmael Parer)

Al, 5'Q Cl, 53 Tiara VReml P~r) ca"ale~roti.~g cyclic Scca/opsrntiQg 4~14 cRca/crgktrating cycle Al, R.3 Cl, 93 Tisaara Ctlceel Favor) ee-a/operating cycla TIA'.3g ooccjaTf<aratil5Lg cfcIAI

kOTES PO'R TASEES 4 ~ 2.A THROUGH 4.2.8 Cont fnucd 14.

Upscale trip 1 s functionally tasted during funct fonal test t fo ~

aa required by sectLon 4.7.b.l.a and 4.7.C.l.c.

1S.

Tha flou bias cocaparator vill bs tasted by put tfng one flou uof t fn "Taac" (producing 1/2 sara<<<)

and ad)use fng th<< ta ~ t input to sbta fn co<sparator rod block.

Tha flov bf ~ ~ upscal ~ vfll bs variffad by observing a local upscale trip light during oparation aad var: ffsd thac Lt vfif produc ~ a.rod block during tha oparatfng cycl ~.

ld, Perfor<acd during operacing cycle.

Portions of the logic Ls chcckad sero frequently during fun'ctfonal taste of tha functions that produce a rod block.

17.

Thfa calibratfon conaf ~ zs of raa<ov<ng cho functLon froa

~ arvics and perfoaaLng an slactronic calibration of cha channel.

18, Punctional ta ~ t is if<sited to the condf,tion vhere

~ scondary contafmant integrity fs not required as specLf lcd in sections 2 ~ 7.C.2 and ),7.C,3.

]9.

Punctfonaf test fs lfs<fted to the tfea uhara the SCTS is rsquirad to s<aet tha requires<ants of section 4,7.C.l.c.

20.

Calfbra(fon o( the co<sparator raqufra ~ tha fnputi fro<s both rscfrcufitfon loops to be fntarruptad, thereby rcs<ovfng the flou bfa ~

~ Lgnal to ths APRH and RL't a >d sera<a<sfng thc rsactnr.

This calfbratfon can only bs pa r formed during an ou cage.

21 ~

logic ta ~ t 1 ~

1%<sf tad to the ti<sa vhaz'e actual opera tLon of the cqufpuant Ls pars<fssfbf

~

22.

Ona channel ot afthar the reactor zona or refueling zona Reactor bufldfng Ventilation Radfation Honitoring Systc<s a<ay bc adnfnistratfvcly bypassed for a period not to exceed.'ours for functional tastfng and <<calibration.

23.

Deleted 24 ~

This fnstru<sent check consists of conparfng the theraocouplc readings for all valves for consistence and for nocLLnal expected values (not required during refueling outages) 2g.

During each refueling outage, all acoustic ~onitorfng channels shall be calibrated.

This calibration L.ncludes verification of acccleronetcr response due to zsochanfcal cxcirarion in the vicinity of the sensor.

26.

This inctrunent check consists of co<sparing the background signal lave's fnr all valves for con'sistency and for nominal expected values (not required dut'ing refueling outages),

110

~ ~

(13!un CnnlltTtOH!n Vdlh !H'K!(hTlnH SVRVEtLLAAC8 HE(>>! XH!H',.'(TS J.l.a naald al Ilaal II a'~ie'.e.

~RIIBS)

(LPCI and Canaalaaaana Cooling) 4, If'ny 0 RHR ~cps (r~ ~e) become inopcxaM.e, R(b reactor shall be placed t(n the col(t shut Jcr~n eon(tit tcsn vkthht gati alOL(rS ~

4~lab hesstdu<<t Ha(Rb Re((soy(Rt S

atcdd

~RKlS (LPC( and Canna(annal Cooling) 4.

%o additional surveillance required.

5. ll one AHR poles (con:aln-mcnt cooltng PP (de) o"

~ a-noctnted hant ex:hanxer ta inoperabl.,

the reactor ray rrraotn tn operation (or a per lod no. tu exceed 30 days prnvtdsad thc reuaintng pHR ay>>ops (containrent

'oottng s(odc) and aaao-ctatcd heat exchongera and d!eeet genera:nrs d(nd all accede paths u( ths RNRS (contatnoent cooLing (aode)

<<r>> operable.

a tt (sdo RHR rue(pa (Contatn:sent cooltnx mode) or aedoctatad h<<>>t exch((nger~

are tnopers-blc, the reactor

~ay rcnatn ln operstton f"r

~

ps rtod not to exceed 7 dayS p

o vlded thc redcap! nina RHR punps (contatnPRent coot tn3 cxsdt) s th associated heat exchangers, diesel generators, and all access paths of the P)3pq (containment cooling mode) 5.

Mnen tt ta det mined that one RHR pu.sp

{containment cool tng node) oa r,saoc{ated heat exch<<niter ia inoperable nt o

t toe uhen operability lo'c-

qutred, tlse rescatntng RHR pu~pa

!contain(dent cool tng node).

the associated heat cxc4ng=ra ard die<<el generators, and nll active conponenta

'n the access path ~ of the R)BS (containment co()l tnt code) shall ba s]ennn-atratcd to ba operable tracdtately anl use'cly there<<acct'n.tt t'e inoperable RHR pump (containment cooiint mdc) and a ~ <<octa rd heat exchanger ie returned to noma)

~ erv lc e.

6 ~ W'en t n id de t crastned tlsdl t n a&

RHA puopa (contatncsent:oollnS c(ode) or aaaociat d heat exchanger

~

ar. inoperable at a ttne ashen opersbtltty ts reliutced, (hc rcna in n(3 R th p(a<<pa

( on c J in((scot coolth'aode),

t'>c saaoctsted heat exchange rss dtcscl Saneratorss and a'.1 active cc"--

ponencs in the access path<< of tha RHRS (cental~ant cooling 147

LIIIITI!IGCONDITIOXS IOR OPERATION I

3. S.C (Continued)

I 4 ~. Thraa of aha Dl, D2, Bl, 82 RHRSh'umps assigned to the RHR. heat exchanger supplying, the standby coolant supply connection may be inoperable for a

. period not to exceed 30 days provided the operable pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are operable.

S1JRVEILLANCE RKQUI~TS 4,5. C (Continued).

4.

Mhen it is determfncd th t three of thc RHl~w pumps sup-plying standby coolant are inoperable at a tiuc vhan operability is required, the opcrablc RHRSM pump t

and its as o-ciatcd diesel generator and "

the RHR heat exchanger header and asaociate4 aosan-tial control valves shall bc damonetratcd to ba operable immadiataly and every l5 days thereafter.

5.

The standby coolant supply capability may be inoperable for a period not to exceed ten daysr 6.

Jf specifications 3.5;C,2 through

3. 5. C. 5 are no t me t, an orderly shutdown shall be initiated and the unit placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7.

There shall be at

'least 2

RHRSW pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.

153

3.,

BASES Should the capability for prov44iag-flow through the cross-connect linco be lost, n ten day repair time is allowed before shutdown is required.

This repair time is )ustified based on the vary small probability for aver nccding IIHR pump'nd heat cxchangero to supply an ad)accnt unit.

REF'EH Ei>> CRS>>

~ l.'esidual Heat..Removal System (BFIIP FSAR subsection I>.8)

?.

C>>3re Stan>I)>y Caolin). System

. (BFIIP FSAR ection 6) 3,'!.C RIIR Re via"- !~Inter 5 "t-.ri nna !ace!rene R <<imimnt C<<a!in: ':.'nt r.':v::tan (RRC'::. )

There are twa ZECM headers (north and south) with four automati.c starting RHRSM pumps on each header.

All components requiring emcrgcncy cooling water are fed from both headers thus assuring continuity of operation if either header is operable.

Each header alone can handle the flows to all components.

Two RHRSW pumps can supply the full flow requirements of all essential EECM loads for any abnormal or postaccident situation.

There are four RHR heat exchanger headers (A, B, C,

& D) with one RHR heat exchanger from each unit on each header.

There are two RHRSW pumps on each header; one normally assigned to each header (A2, B2, C2, or D2) and anc an alternate assignment (Al, Bl, Cl, or Dl).

One RHR heat exchanger header can adequately deliver the flow supplied by both RHRSW pumps to any two of the three RHRSW heat exchangers on the header.

One RHRSM pump can supply the full flow requirement of one RHR heat exchanger.

Two RHR heat exchangers can more than adequately handle the cooling requirements of ane unit in any abnormal or postaccident situation, The RHR Service Water Systems was designed as a shared system for three units.

The specificatio'n, as written, is conservati.ve when consideration is given to particular pumps being out of service and to possible valving arrangements.

Zf unusual operating conditions arise such that more pumps are out of service than allowed by this specification, a special case request m'ay be made to the NRC to allow continued operation if the actual system cooling requirements can be assured.

!e Should three of the four RHRSM pumps normally cr alternately assigned to the RHR

'eat exchanger headers

.." supplying the standby coolant supply connection become inoperable,

'apability for long-term fluid makeup to the unit reocto:

and for cooling of the unit containment remains operable.

Because of the availability of

.,.'akeup and cooling capability which i" demonstrated to bc opernbla,>immediately and with specified subsequent surveillance, n 30-day reps! r period "io Justified.

Unit 2 oiay bc supplied standby:calont fram eit))rr of )'~us pumps--M, 92, Dl, o)

D2, Should the capability to provide standby coolant supply be lost, a 10-day repair time is )ustified based on the low probability for ever needing the standby coolant supply.

164

Oi

BASES opening.

If the check nnd green or check light circuit alone is inoperable f

the valve shall bc considered inoperable for full closure.

If the red and check light circuits arc inoperable the valve shall be considered inopern-blc and open greater chan 3'.

For a light circuit to be considered operable the light must go on and oif in proper sequence during the opening-cloninp cycle. Ifnone of Chc lights change indication during the cycle, the valve shall be considered inoperable and open unless the check light stays on and the red light stays off in which case the valve shall be considered inopera-ble for opening.

Thc twelve drywell vacuum breaker valves which connect the suppression chamber and.drywcll arc sized on the basis of the Bodegn pressure suppres-sion system tests.

Ten operable to open vacuum breaker vnlves (18-inch) selected on Chin test basis and confirmed by the green lights are adequate to limit the pressure differential bet~can the suppression chamber and dry-well during post-accident drywell cooling operaCiono to a value which is within suppression system design values.

The containment design has been examined to determine that a leakag~ equi-valent to one drywell vacuum breaker opened to no more'han a nominal 3

as confirmed by Che rcd light is acceptable.

On Chin basis an indcfinirc allowable repair time for nn inoperable'cd light circuit on any valve or nn inoperable check nnd green or check light circuit alone or a malfunction of thc operator or disc (if nearly closed) on onc valve, or an inoperable green and rcd or green light circuit alone on two valves is )ustified, During each operating cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the dryweL and suppress'ion chamber.

The drywell pressure will be increased by at least 1 psi with respect to the suppression chamber pressure and beld constant; Thc 2 poig set point vill not be exceeded.

The subsequent suppression chamber prcssure transient (if any) will be monitored with a sensitive pres-sure gauge.

If thc drywall prcssure cannot be increased by 1 psi over the suppression chcmber prcssure it would be because a significant leakage path exists; in this event the leakage source vill be identified and eliminated before power operation is resumed.

Mich n differential prcssure

o. greater chan 1 psig, the rate of change of the suppression chamber pressure must not excevd 0.38 inches Of wateV per minute 'ns measured over n

10 minute period, which corresponds to about 0.14 ib/sec of containment nir.

In the event the rata of change exceeds this

'alue Chen the source of leakage will be identified and eliminated before power operation is resumed.

The water in Che suppression chambe'r is used for cooling in Che event of an accident i.e.. it is not used for normal operation; therefore, a daily check of thc temperature nnd volume is adequate to assure that adequate heat removal capability in present.

'72 s

P

I.IHITIHr. CnHUITIOWS FOR OI I'.RATION SURVEILLANCE RE UIREHEHTS

3. 10 CORE ALTERATIOHS 4elO CORE ALTERATIONS Applies to the fuel handling and coro reactivity limitations.

Ao licabilit Applies to the periodic testing of those interlocks and instru-mentation used during refueling and core altcrationa.

~OO eeet e

To ensure that core react1vity is Mithin the capability of thc control rods and to prevent criticality during rcfucling.

~Ott ectf e

To verify the operability of instruraentat ion and interlocks used in refueling and core alterations.

S ccification S ccification A.

Refueling Interlocks h.

Refuel in Interlocks 1.

The reactor cade switch shall be locked in thc "Refuel" position during core alterations and the refueling interlocks shall be opcrablc except as specified in 3.10.Ae 6 and 3.10.A. 7 bcloue 1.

Prior to any fuel hand-ling Mith the head off the reactor 'vessel, the refueling interlocks shall be functionally tcstcd.

They shall be tested at weekly inter-vals thereafter until no longer required.

They shall also be tested fol-loving any repair, work associated Mith thc inter-locks.

2.

Fuel sha ll no t be loaded into the reactor core unlcsa all control rods are fully inserted.

2.

No additional surveillance required.'02

I.!Hl'I'IHC CONI)ITIOHS FOR OPERATION SURVEILLANCE RE VIREHENTS

>.10.*

Rctuclin Interlocka 4.10.A Re fuel in Interlocks 4.

5.

The fuel grapple hoist load suirch shall be set at

< 1,000 lbs.

I I the frame-mounted auxi-1iary hoist, the monorail-mounted auxiliary hoist, or thc service 'platform hoist la be used for handlinc fuel vith the head off the reactor vessel, the load liaLit sMitch on the hoist to be used shall be set at

< 400 lbs.

Haintenance may be performed on a single contxol rod or control xod drive without removing the fuel in the con-trol cell if the following conditions are met:

3.

4.

5.

No additional surveillance required.

No additional surveillance required.

prior to performing control rod or control rod drive maintenance on a control cell without"removing fuel assemblies the surveillance requirements of specification 4.10.A.1 shall be performed and all rods face adjacent o'r diagonally adjacent to to the maintenance rod shall be electrically disarmed per specification 3.10.A.5.b.

a.

The requixements of specification 3.10.A.l

areimet, and b.

All control rods diagonally or face adjacent to the maintenance rod are fully inserted and. have their directional control.valves hlectrically disarmed.

303

INC ON IT ON F

3.10.A.6 A maximum of two non-ad)acent control rods may simultaneously be withdrawn from the core for the purpose of performing control rod and/or control rod drive maintenance without removing the fuel from the cells pro-vided the following conditions are satisfied:

4.10.A.6 Prior to performing control rod or control rod drive maintenance on two control cells simultaneously without remqving the fuel from the

cells, two SRO's shall verify that the requirements of specification 3.10.A.G are satisfied.

a.

The reactor mode switch shall be locked in the "refuel" position.

The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed for one of the control rods on which maintenance is being performed.

All other refueling inter-locks shall be operable.

303A

).10.A Re(ue I fng Infor locke I.Ixlrlxs rnxllITlnxs vnu nr I;RATIOII suRVKILLAHCK RK 0IRI&PHTS

'.10.A -Re(uelfn Interlocks 6.

(Continued) b.

All directional control valves for remaining control rods shall be disarmed electrically except as specified in 3.10.A.7 and sufficient margin to criticality shall be demonstrated.

c.

The two maintenance.cells must be separated by more than two control cells in any direction..

d.

An appropriate number of

SRM's are available as defined in speci.fication 3.10.B.

Any niwhrr n( control rode may br vftlwfrn~i or removed from the rcnctnr coro pro-vfdfnR thr follnvinlI condf-tfono arc eatfef fed:

a.

Tha reactor uode evitch fa locked in the "ro-tuel" poeftfon.

Thc refualins interlock uhfch prevcnta nore than ono control rod from 7

With the nJde selec lcn 5

I ch ln the refuel or shutdo:III rlo 0, no more than one control 'rod may be withdrawn without first re'moving fuel from the cell except as specified in 4.10.A.6.

Any number of rods may be withdrawn once verified by two licensed operators that the fuel has been removed from eacll cell.

304

3. 10 gnh I S tndn and th<< refueling pint form provide ~ redundant m<<chod<<of prevent lng ln<<dvertent'crit icalLty even <<f ter procedural vlolat lons.

Thc interlocks on hoioto provide yct another method of avoiding inodvcrccnc critical icy.

Fuel handling is normally conducted utch che fuel grapplu hoist.

The total lo<<d on thlo hotst vhcn che interlock io required conototo of the vclght of the fuel gtapplc and che fuel aoocmbly.

Thts total te approxi-mately 1,500 lbo, in comparison to thc load-trip oec ting of 1,000 lbs.

Ptovtstons have sloo bccn made to nllou fuel handling uith cLther of the chree auxt1 taty hotsto and still mstntntn thc refueling interlocks.

The "

400-Ib load-trip setting on these hoieco io adequate to trip the interlock vhcn nne of thc morc thon 600-lb fuel bundles ls being handled.

During <<etc<<in'eriodo, it io desitablo to perform maintenance on tvo control rods and/ot coricrol rod drives at the sane time ~ithout removing fuel from the cells.

The maintenance is performed with the mode.

switch in the "refuel" position to provide the refueling interlocks normally available during refueling operations.

In order to with-draw a second control rod after withdrawal of the first rod, it is unnecessary to bypass the refueling interlock on the first control rod wHich prevents more than one control rod from being withdrawn 'at the same time.

The requirement that: an adequate shutdown margin be demon-strated and that all remaining control rods have there directional control valves electrically disarmed ensures that inadvertent criticality

~cannot occur during this maintenance.

The adequacy of the shutdown margin is verified 'by demonstrating that at least 0.38% bk shutdown margin is available.

Disarming the directional control valvessddes not inhibit control rodscram capability.

Spccif icacion 3.10.A.7 sllovs unloading of a stgnif leant portion of the reactor core.

This operation is performed utch ch<<mod<<sittth ln th<<

"refuel" position to provide the refueling interlocks norma)ly available during refueling operations.

In order to vtchdrnv:morc chan one control rod, it Lo necc snnry to bypass the rcfuclLng interlock on cnch vtthdravn control rod vhlch prevents mote t.hon one control rod from being vichdravn at n time.

Thc tequtremcnt chnc thc fuel <<ssemhltes ln che cell controlled by thc control rod be removed from thc re<<ctor coie before the inc<<rlock c<<n be bypassed ensures chat. vlchdrnvnl of anochct control rod does noc result.

Ln inodvcrtcnt criticality.

Each control rod providos primary reactivity control for the fuel aesembltco in thc cell associated uith chat control rod,

Thus, rcmovnl of an entire cell (fuel aoecmblics plus conctol rod) results in a lover teacclvicy potential of thc core

~

The requirements for Sg.'<

opcrabllity during these core alterations assure sufficient core monitoring.

3IO

~,

LOCAL MONiTORiNG STATlONS BROV/NS FERRY NUCLEAR PLANT Fi

. 4.2-l ATH'"HS U.S.

HWY 72 BFNP ALA. HVIY 20 Legend 8

Air Monitor Monitor 8 TLD Station TLD Stotion h.

Automatic Well Sampler H

Dairy Farm DECATUR Scale 0

I 2

3 4

Miles

UNIT 3 PROPOSED CHANGES

1.0 DEFINITIONS cont'd 2.

Mhen a system, subsystem, train, component or device ia,.

determined to be inoperable solely because its onsite power source is inoperable, or solely because its offsite power source is inoperable, it may be considered opercrable for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:

(1) its corresponding offsitc or diesel power source is, operable and (2) all of its redundant system(a),

subsystem(s),

train(s),

component(s) and device(u) are operable, or 1Qcewise satisfy these requirements p

This provision describes what additional conditions must bc satisfied to permit operation to continue consistent with the specifications for power sources, when offsite or oasite power sources are not operable.

It specifically prohibits operation when one division is inoperable because its offsite or diesel power source is inoperablc and a system, subsystemp

train, component or device in another division is inoperable for another reason.

This provision permits the requirements associated with individual systems, subsystems,

trains, components or devices to be consistont with the requirements of the associated electrical power source.

It allows operation to be governed by the time limits of the requirements associated with the Limiting Condition for Operation for the offsite or diesel power source, not the individual requirements for each system, subsystem,

train, component or device that is determined to be inoperable solely because of tha inoperability of its offsite or diesel power sourcda D.

DELETED E.

H.

arable -

erabilit

<<A system, subsystem, train, component or device shall be operable or have operability when it is capable of performing its specified function(s),

Implicit in this definition shell be the assumption that all necessary attendant instrumentationp

controls, normal and emergency electrical power sources, cooling or seal water lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

~erarfn

- Operating mesne that a system or tompoaent is perfotmfng its intended functions in its required manner.

Immediate - Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and and importance of th~ required action.

Reactor power ration - Reactor power operation is any operation with the mode switch in the "Startup" or "Run" position with the reactor critical and above 1l, rated power.

2a

Refuel Node - With the mode switch in the refuel position interlocks are established so that one control rod only may be withdrawn when the Source Range Monitor indicate at least 3 cps and the refueling crane is nrlt over the reactor;. also, the main steam line isola'.ion scram and main condenser low vacuum scram are bypassed if reactor vessel pressure is belo~

1055 psig. If the refueling crane is over the reactor, all rods must be fully inserted and none can be withdrawn.

Rated power - Rated power refers to 100K of the nominal value of thc full, steady-state licensed power level of 3293 Mft authorized by thc operating license.

Small, non-steady state power excursions to above 3293 HMt arc permissible for brief periods of time provided that thc average power level for "an eight hour period does not exceed the full, stcady-state licensed power lcvcl of 3293 Mft. Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system prcssure refer to the values.of thcsc parameters when the reactor is at rated power.

Design power, thc power to which the safety analysis applies, corresponds to 3,440 Ã4t.

Primar Containment Xnte rit - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

I 1.

All non-automatic containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be, open during accident conditions are closed.

These valves may be opened to perform necessary operational activities.

2 0 3 ~

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

All automatic containment isolation valves are operable or deactivated in the isolated po'sition.

4.

All blind flanges, and manways are closed.

Secondar Containment Inte rit - Secondary containment integrity means that the reactor building is intact and the following condi tions are met:

1.

At least one door in each access opening is closed.

2.

The standby gas treatment system is operable.

4

HAS ES:

LIHITI HG SAFETY SYSTEM SETT R F LATED TO FUEL CLADDING IHTEGRITY Iho abnormal on~rational transients applicable to operation of the ttrowns Ferry Nuclear Plant have been ana)yoked throuqhout the

~n~ctrum ot planned operatinq condii:ons up to the design thermal power condition of 3440 Hwt.

The o.l>lyses were based upon plant operation in accordance with ttie operating map given

~ ~ Figure

.7-1 of t:hu I'SAR.

Duu to the statistic'tl n:tture of the process points to tl>e heat balance calculation, a certain number of hedt balances will in-dicate a core thermal power above the full, steady-state licensed power level of 3293 NJt when, in fact, the unit is operatintt ut rated power.

This is especially true foilowinp pert:urbations tn steady-state whett operating close to rnted power.

Small, nor.-steady state power excursions to above 3293 iMt are permissible for brief periods of time provided that the average power level for an eight hour period does not exceed t'ne full steadv-state licensed power level o:

3293 Y<<"t.

'n addition, these excursinns are

. rther limited to 100 1/2% of rated power for one hour, 101% of rated power for 1/2 hour, or 102%

o rated power for 15 minutes.

But, under no cond"'tion should power exceed 102/ of rated power.

Conservatism is incorporated in the transient analyses in estimating the controllinq factor such as void reactivity coef ficient, control rod scram worth, scram delay time, peakinq

factors, and axial power shapes.

These factors are selected conservatively with respect to their effect on the applicalbe transient results as determined by the current analysis model.

This transient

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

Results obtained from a General Flectric boiling ~ater reactor have been compared with predictions made by the model.

The comparisions and results are summarixed in Reference l.

The absolute value of the void reactivity coef ficient used in the analysis is conservatively estimated to ke about 25% greater than the nominal maximum value expected to occur durinq the core lifetime.

The scram ~orth used has been derated to be equivalent to approximately 80% of the total scram worth of the control rods.

The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical specifications.

The effect of scram worth, scram delay time and rod insertion rate, all conservatively

applied, are of greatest siqnificance in the early portion of the negative reactivity insertion.

The rapid insertion of negative reactivity is assured by the time requirements for 5% and 20% insertion.

By the time the rods are 60% inserted, approximately four dollars of negative reactivity has been inserted which strongly turns the transient, and accomplishes the desired effect.

The times for 50% and 90%

insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to "establish the ultimate fully shutdown steady-state condition.

For analyses of the thermal consequences of the transients a

NCPR of is conservatively assumed to exist prior to initiation of the transients.

This chot,ce of usinq conservative values of controllinq parameters and initiati.nq transients at the desiqn power level, produces more pessimistic answers than would result by usinq expected values of control parameters and analyzing at higher power levels.

See Section 3.5. K.

1S

Steady-stat~

oper J or>

~ rntlor> wlthrut for<cd rccictu]atleA >-ill nnt

~c l>cr hatt".d for mnre lhan 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

and tlie start of a recirculation pi>>o from the nlitiiral circulation condition will lint Le peru!i tied ur>less llie tei:;>."r<<ture differ."Ace betiv<<e>>

the loop to be started and the coi e coolant temperature is less than 75oF.

This reduces the Positive ieactivity insertion to an acceptably

)mr value, In summary:

1.A.

The full, steady-state licensed power level is 3293 Ndc and is the maximum time-averaged (8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) power level.

H.

Non-steady-state power excursiono are further limited t.o:

< 300 1/2

% of rated p>>wer excursion for >>ot more t."han one hour.

b.

101% of rated power for not more than 1/2 liour,.

c.

<,1.02% of rated power for not more than 15 niinutes.

2.

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

3.

The abnormal operational transients were analyzed to a power level of 3440 HWt.

4.

The ana~ical procedures now uoed result in a more logical answer than the alternative method of assuming a higher starting power in conjunction with the'xpected values for the parametero.

The bases for individual oet points are discussed below:

A.

Neutron Flux Scram 1.

APRH High Flux Scram Trip Setting (Run Yude)

The average power range monitorinq (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3,293 Wt).

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

During tranoients, the instantaneous rate of heat tranofer from the fuel (reactor thermal power) is leos than the instantaneous neutron flux due to the time constant of the fuel.

Therefore, durinq transients induced by disturbances, the thermal power of the fuel will be leos than that indicated by the neutron flux at the scram setting.

Analyseo reported in Section N14 of the Final Safety Analysis Report demonstrated that w'th a 120 pexcent scram trip set (ng, none of the abnormal operational transients analyzed violate the fuel safety limit and

.there is a substantial margin from fuel damage.

Therefore, use of a flew-biased scram provid s even additional marqin.

Figure'.1.2 shows the flow biased scram as a function of core flcw.

An increase in ttic APRH ocram netting would decrease the margin present before the fuel cladding integrity safety limit io reached.

The hl>RH scram oettinq was dc" errnined by an analyois of margiiio ceqiiired to provide a

reasorinble rariqn for maneuvering durinq c:j~eration.

Reducinq tliio oprratinq'>arqi>>

~ oui'd iiicruase the frequency of or>urious ocramo, which have an adverse ef Eeet ori reactor safety because of tlie reoultinq thermal ot.resoes.

Tliuo, the APRM oetti>>g vao selected

TABLE 3 2 A PRINART CONTAINNENl'ND REACTOR BQILDZNG ISOLATIOH INSTRUNENTATION Niniaua No.

Operable Per

~Trf S

• 1 Tri Leve) Setti Action 1

Reaarks 2

Instrument Channel-Reactor Mater Cleanup

'ystea Floor Drain High Tenperature 2 (ll)

Znstruacnt Channel-Reactor Mater Cleanup Systea Space High Tenpera ture Inst unent Channel-Reactor Building Venti-lation Bigh Radiation-Reactor tone Znstrunent Channel-Reactor Building Venti-lation High Radiation-Refueling Zone 160 1BOOF 160 - 1804F 5 100 ar/hr or downscale 100 ar/hr.or downscale

=

F 1.

Above trip setting initiates Isolation of Reactor Mater Cleanup Line froa Reactor an'eactor Mater Return Line.

1.

Sane as above 1

1 up"ca le or 2 dovnscale vill a.

Initiate SGTS b

Isolate reactor zone and refueling floor.

c.

Close ataosphere control systea.

upscale or 2 dovnscale vill a.

Initiate SGTS b.

Isolate refue 5-.g floor-c.

Close ataospn-:re control systea 2 (7) (8)

Instxuaent Channel SGTS Flov - Train A Heaters Charcoal Heaters S 2000 cfa R.B. Heaters 5 2000 cfa B and (A or F) 1.

Belov 2000 cfa, trip setting coal heaters <<ill turn on.

2.

Below 2000 cfa, trip setting heaters vill shut off.

cha r-R H.

2(7) (8) 2(7) (B)

Instruaent Channel SGTS Flov - Train B

Heaters Instrument Channel SGTS Flow - Train C Heaters Charcoal Heaters S 2000 cfa R.H. Heaters 5 2000 cia Charcoal Heaters 52000 cfa R.H. Beaters S 2000 cfa H and (A or Fl H and (A or F) 1.

Belov 2000 ci'a, trip setting coal heaters vill turn on.

2, Below 2000 cfa, trip setting heaters vill shit off.

l.

Belov 2000 cfa, trip setting coal heaters vill turn on.'.

Below 2000 cfa, trip setting heaters <<ill shut off.

char-R. H.

char-R H.

3.

There are four channels pex'team line of which tvo must. bo operable.

4.

Only required in Run Node (interlocked vith Mode-Switch).

5.

Hot required in Run Node (bypassed:iy mode switch).

Channel shared by RPS and Primary Containment G Reactor Vessel Isolation Control System.

A channel failure may be a

charnel failure in each system.

7.

A txain is considered a grip system.

8.

Two out of thxee SGTS,trains required.

A failu o of more than one vill require action A and F.

9.

There is only one trip system with auto transfer to tvo power sources.

0.

Refer to Table 3.7,h and its notes for a listing of Isolation Valve Groups and their initiating signals.

11.

Requires two independent channels from each physical location, there are two locations.

I 63

Table 3.2.8 I)4TR'JHS !TA 10,1 TEXT IHITIATFS uR CON'ROLS Ti 8 COBB A!0 COWTAISHKNT CCOI I !O SYSTKHS Hinimn 1!o.

Operable Per Function Instruaent Channel-Rcactor Lou Prcssure (FG-3-7u A 6 8, SW ti)

(PS-68-95, SW t 2)

(PS-68-96

~

SW t 2)

Instrment Channel-Rcactor Low Pressure (PS-3-70A 6 8, SW t 1)

(PS-68-95

~

SW t 1)

(PS-68-96'W 1 1)

Ins runent Channel-Reactor Low Pressure (PS-68-93 8 94'W t'1)

Tri Lev~l Settino 43'~1g

+

15 230 psig

+

15 l00 psi>>

15 tictlon Rei>>arks 1.

Belou trip set ing ( raissive for opening CsS 8!!d LPCT, adri'. 'sion valves.

~

Recirculation disc)!arge valve actuation.

1 Belch trip setting in con3unction With containnent isolation signal and both suction valves open will close RHR (LPCI) adnission valves.

Core Spray Auto Sequencing Timrs (5)

LPCI Auto Sequercing Timers (5)

RllRSW A3, Bl. C 3 and 0$ Ti>>ers 6<t<8 secs.

O<t<l sec.

13<t.<15 sec.

A 1.

With diesel (u"er 2.

One per motor 1.

With diesel power 2.

One per ~otor 1.

With diesel power 2.

One per puaip

Table 3.2.8 I'HSTRNZNTATIGH TEAT INITIATES OR CORI'ROLS THE CORP ASD CORTAI?INERT COOLItio SYSI'9.S N nisr:n >o.

Operable Per Trip~s~

f?)

Fsnct icn Core Spray and LPCI Auto Sequencing Ti".crs (6)

R.'IRSVP Ag Bg C 3 and D QTi~ers Tri Level Settin OSt$ 1 sec.

6<tS8 sec.

12St<15 sec.

18StS24 sec.

27StS29 sec.

Action Renarks 1.

With normal pmwer 2.

one p.r CSS motor 3.

Two per RHR motor 1.

Pith nor~i ~'er 2.

One per punp 1 (16)

ADS Tiner 120 sec

+ 5 l.

Above trip setting in conjunction with los reactor vater level, high drywall pressu"e and LPCI or CSS pu~ps running initiates ADS.

. Instrument Channel -

100

+

10 psig RHB Di charge Pressure 1.

Delo~ trip 'etting de!ers ADS actuation'

Table 3

2 B

IllSTRIM174TATION THAT IHJTIATKS OR CORK'ROLS TBK CORK ARD CO'.rrhinnKHT COOLING SVSTFAS Ywniaun Ho.

Operablc Per

~Tf S ~l Function Core Spray Trip Systea bus pover aonitor

• DS Trip Systaa bus pover "onitor HPCI Trip Systea bus pover nonitor Tr Leve1 Set t in 8/h

~bet jo Reaarks 1.

Mcnitors availability of pover to logic systeas Hcnitors availability of pover to logic systens and valves.

1.

honitors avai1ability of pover to logic systeaso 1 (2) 2 (2) 2 (2)

RcIc Trip Systen bus pover N/h nonitor Instruaent Channel-2 Kiev. 55'I ~

Couuunsate Storage Tank Lov Levet (LS-73-56A 6 8)

Instruaent Channel S7e above iQSCKUEGQC ZG10 Suppression Chanber High Level lnstruaent Channel-S 583 ~ above vessel zero Reactor High 1later Level

)(oniturs availability of pover to logic systeas.

Helov trip setting vi.ll open HPCI suction valves to the suppression chanber.

above trip setting vill open HPCI suction valves to tbe suppression chaabcr.

1.

Above trip setting trips RCIC turbine.

4 (4)

Inetru-ent channel-RCIC Turbine Steasa Line High Plov Instzuaent Channel-RCIC Steaa Line Space High Teaperature S

450%

H 0 (7)

S200OF 1.

Above trip setting isolates RCIC systea and trips RCIC turbine.

1.

Above trip setting isolates RCIC systea and trips RCIC turbine.

NOTES FOR TABLE 3. 2 B

whenever any CSCS System is required by section 3.5 to be

operable, there shall be two operable trip systems

except, as noted.

Zf a requirement of the first column is reduced by

one, the indicated action shall be taken.

If the same function is inoperable in more than one trip system or the

- first column reduced by more than one, action B shall be taken.

Action:

A.

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

If the function is not operable in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, take action B.

B.

Declare, the system or component inoperable.

C.

Immediately take action B until power is verified on the trip system.

D.

No action required, indicators are considered redundant.

2.

In only one trip system.

3.

Not considered in a trip system.

4.

Requires one channel fxom each physical location (there are 4

locations) in the steam U.ne space.

5.

With diesel

power, each RHRS pump is schedul'ed to start immediately and each CSS pump is sequenced to start about 7

sec later.

6.

with normal power, one CSS and one RHRS pump is scheduled to start instantaneously, one CSS and one RHRS pump is sequenced to start after about 7 sec with similar pumps starting aftex'bout 14 sec and 21 sec, at which time the full complement of CSS and RHRS pumos would be operating.

7.

The RCIC and HPCI steam line high flow trip level settings are given in terms of differential pressure.

The RCICS setting of 450" of water corresponds to at least 150X above maximum steady state steam flow to assure that spurious isolation does not occur while ensuring the initiation of isolation following a postulated steam line break.

Similarly, the HPCIS setting of 90 psi corresponds to at least 150X above maximum steady state flow while also ensuring the initiation of isolation following a postulated break.

8.

Rote l does not apply to this item.

9.

The head tank is designed to assure that the discharge piping from the CS and RHR pumps are full.

The pressure shaU.

be main aired at or above the values listed in 3.5. H, which ensures water 'n the d'scharge piping and up to the head tank.

TABLE a 2~A SURVEILLANCE REQDIRENENTS FOR PRIMARY CONTAINNENF AND REACTOR BUILDING ISOLATIOH IHSTROHEHTATIOH Function Group 6 Logic Group 8 (Initiating) Logic Reactor Building Isolation (refueling floor) Logic Reactor Building Isolation (reactor xone) Logic SGTS Train A Logic SGTS Train B Logic SGTS Train C Logic Static Pressure Control (refueling floor) Logic Static Pressure Control (reactor zone)

Logic Instrument Channel-Reactor Cleanup System Floor Drain High Temperature Instrument Channel-Reactor Cleanup System Space High Temperature nctiona 1 Test ce/operating cle (18) acked during iuannel functional test No further test required.

once/6 months (18) once/6 months (18) once/6 months (19) once/6 months (19) once/6 months (19) once/operating cycle

'(18) once/operating cycle (18)

Cal ibration Fr uenc N/A (6)

(6)

(6)

(6) once/operating cycle once/operating cycle Instrument Check N/A N/A N/A N/A

TABLE 4.2.B SURVEILr

!iCE RE"8 REHENTS FOR INSTROMEHTATION THAT TVITIATE OR CO::THOL THE CSC."

Funct.

o>>

Instrument Channel Reactor Low Pressu"e (PS-3-74A 6 &)

(PS-68-95)

(PS-68-96)

Instrument Channel Reactor Low Pres-ure (PS-68-93 6 94)

Functional Tes Calibration once/3 months once/3 months Instrumen Che. k none none Core Spray Auto Sequencing Tim rs (Normal Power)

Core Spray Auto Scgcen ing Timers (Diesel Power)

(4)

(4) once/operating cycle once/operating cycle none none LPCI Auto Sequencing (Normal Pow r)

Q I

LPCI Auto Seq encir.g (Diesel Power)

Timers lmers (4)

(4) once/operating cycle once/operating cycle none none RHRSH A3, BQ C 3 D.$

(Hormal Power)

Timers RHRSW A3, B 3, C. 3 D 1 Timers (Diesel Power)

(4)

(4) once/operating cycle once/operating cycle none none

HOTKS POR Taetgs 4.2.A THROVCII 4,2.H Con<fnuad l4, L5.

Vpscafo trip ts functfonafly tasted durfng functional teat tf<<s as required by section 4.7.B. l.a and 4,7.C.l.c.

Tho flou bfss coaparstor uff1 be reseed by putting one flou ucft fn "T t" (

oducfng I/2 sera<<7 and ad Juscfng the tesC fnpuc to obtain co<<paracor rod block.

The flow bfss upscale vttt bc vari a

y a ifiad b obsecvtng a local upscalo trip lLght during operation and verifLsd Chat Lt vfll produce a rod block during tha operating eyelet L6.

Performed during operating cycle, Portions of the logic Ls checked

<<ore frequently durfng functional teats of the fuaccfons that produce a rod block.

L7.

Thfa calfbratfon conelet'f removing the functfon frou service and pertorsdng an alaccronLc calibration of the channel.

le.

Punctfonal canc fs lf<<iced to the condftton vherc secondary contafnmant tntagrtty ts not rcqutrcd s ~ spccfftad tn sections'.7.C.2 and 5.7.C,5.

19.

Punctfonal eeet Is Lf<<fted to the t fee vhcrs the SCTS ia required to eeet the rrqutrawants of section 4.7.C.l.c.

20.

Caltbratton o( tha conparator requtrea tha inputs fro<< boch rscfrculation loops Co bc incorrupted, thereby rcoevfng Che flov bialy

~ Lgnai Co Cho APIIN and AH a id sera<<<<tng the reactor, This calfbracLon can ocly bs perfor<<ed durtng an outage.

21'ogfc test la ll<<ttcd to the ttae vhare accual. operatfon of the equtp<<ent Ls permissfbLa.

22, 23.

One channel of eI.thar tho reactor conc or refueling tone Reactor Butldfng Yanttlatton Radtatfon Ifonftorfng System

<<ay b>> adufntstratfvcly bypassed for a partod not to cxcoad 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for functional testing and ralLbraCion.

Deleted 24, This instrumcnc check consists of comparing the thermocouple readfngs for all valves for consistence and for nominal expccred values (noc required durLng refueling outagcs).

)

25.

During each refueling outage, all acoustic c:onitoring channels shall be calibrated, This calibr'ation includes verification of accelerometer response duc to mechanical excitation Ln rhc vicinity of the sensor.

26.

This instrument check consists of comparing thc background sign~i levels for all valves for consistency and for nominal expected values (not requf.rcd durfng refueling ourarcs).

ln7 i >>>>>>

4

~

~

e

~

qs ~ P>> w>> >>>e

~ ~ ~ tv s>>

~

~

~

gz

~

LINITINC CONDITIONS FOR OPEPJLTION SUR'JEILI ANCE RFQUIREllENTS 3 ~ 5 COR E AND CONTAIN"4F~'T COOLIth~

S YS".'EHS rr. 5 CORE AND CON'AINHENT COOLING SYSTEMS If one RliR pump (containrnant coolinq l>ode) or associated heat exclranger i" ir:operable, the reactor may remain in operation for a period not'to exceed 30 days provided the remaining RllR pumps (con=a innent cooling mode) and associated heat exchanger"-

and diesel generators and all access path of the RllRS (containment cool'g mode) are operable.

No arlilicional surveillance required.

6.

If two Rl)R pumps (conta inment cooling mode) or associated heat exchangers are inoperable, the reactor may remain in operation for a period not to exceed

'1 days provided the remaining RllR pumps (containment cooling mode) the associa ted heat exchangers, diesel generators, and all access paths of the MRS (con-tainment cooling mode) are operable.

}51 5.

when it is determined that one RllB pump (containment coo'ing mode) or associated heat exchanger is inoperable at a time wlren operability i"

required, the rernaininq HHR pumps (containrnerrt coialing mode),

the associated heat exchanqers diesel gerrerators, arrd all active compon 'nts in tire access patlrs of the RliRS (contairenent cooling mode) shall b>> demonstrated to he oper rble im:nediately and weekly thereafter until the inoperabl.e RllR pump (contain..en cooling Node) and associated heat

The containment, design has been examined to determine that a leakage equivalent to one drywell vacuum breaker opened to no more than a nominal 3o as confirmed by the red liqht is acceptable.

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

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

The drywell, pressure will be increased by at least 1 psi with respect to the suppression chamber pressure and held constant.

The 2 psig set point will not be exceeded.

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

If the drywell pressure cannot be increased by 1

psi over the suppression chamber pressure it would be because a

significant leakage path exists; in this event the leakage source will be identified and eliminated before power operation is resumed.

with a differential pressure of greater than 1 psig, the rate of change of the suppression chamber pressure must not exceed 0.38 inches of water per minute as measured over a 10-minute period, which corresponds to about 0.10 lb/sec of containment air.

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

The water in the suppression chamber is used for cooling in the event of an accident; i.e., it is not used for normal operation; therefore, a daily check of the temperature and volume is adequate to assure that adequate heat, removal capability is present.

The interior of the drywell is painted with an inorganic zinc primer top-coated with an epoxy coating.

This coating provides protection against rusting as well as providing a surface which is decontaminable.

The inspection of the paint during each ma]or refueling outage, approximately once per year, assures the paint is intact.

Experience with this type of paint at fossil fueled generating stations indicates that the inspection interval is adequate.

The interior surfaces of unit 3 suppression chamber is coated with an organic protective coating of the thermosetting resin type.

The inspection of the coating during each refueling outage, approximately once per year, assures the coating is intact.

2&9

LIMITING.CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3

1 0 CORE ALTERATIONS 10 CORE ALTERATIONS A licabilit A licabilit Applies to the fuel handling and core reactivity limitations.

Applies to the periodic testing of those interlocks and instrumentation used during refueling and core alterations.

~ch 'ective To ensure that core reactivity is within the capability of the control rods and to prevent criticality during refueling.

~oh 'ective To verify the operability of instrumentation and interlocks used in refueling and core alterations.

S ecification S ecification A ~

Refuelin Interlocks The reactor mode switch shall be locked in the

<<Refuel<< position during core alterations and the refueling interlocks shall be operable except as specified in 3.'lO.A.6 and 3.10.A. 7 below.

A.

Refuelin Interlocks Prior to any fuel handling >>ith the head off the reactor

vessel, the refueling interlocks shall be functionally tested.

They shall be tested at weekly intervals thereafter unti no longer required.

They shall also be tested following any repair work associated with the interlocks.

331

LIMITING CONDITIONS POR OPERATION SURVEILLANCE REQUIREMENTS

a. 10 CORE ALTERATIONS 4 ~ 1 0 CORE ALTERATIONS 2.

Fuel shall not be loaded into the reactor core unless all control rods are fully inserted.

2.

No additional surveillance required.

3.

No additional surveillance required.

3.

The fuel grapple hoist load switch shall be set at <

1,000 lbs.

332

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

3. 10 CORE AL TIONS 4

10 CORE ALTERATIONS Xf the frame-mounted auxiliary hoist, the monorail-mounted auxiliary hoist, or the service plat. form hoist is 'to be used for handling fuel with the head off the reactor vessel, the load limit switch on the hoist to be used shall be set at ( 400 1 bs.

5-Maintenance may be performed on a single.

control rod or control rod drive without re-moving the fuel in the contxol cell if the following conditions are met:

4 No additional surveillance required.

5.

Prior to performing control rod or control rod drive maintenance on a control cell without removing fuel assemblies the surveillance requirements of specification 4.10.A.1 shall be performed and all rods face adjacent or. diagonally adjacent to the maintenance rod shall be electrically disarmed per specification 3.10.A.5.b.

a.

The requ'irements

'f specification 3.10.A.1 are met, and b.

All control rods diagonally or face adjacent to the maintenance xod are fully inserted and have had their directional control valves electrically disarmed.

333

3.10.A.6 4.10.A A maximum of two nonadjacent control rods may be simultaneously withdrawn from the core for the purpose of performing control rod and/or control rod drive maintenance without removing the fuel from the cells pro-vided the following conditions are satisfied:

6.

Prior to performing control rod or control rod drive maintenance on two control cells simultaneously without, removing the fuel from the

cells, two SRO's shall verify that the requirements of specification 3.10.A.6 are satisfied.

a.

The reactor mode switch shall be locked in the "refuel" position.

The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed for one of the control rods on which maintenance is being performed.

All other refueling interlocks shall be operable.

333A 0

LIHITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 10 CORE ALTERATIONS 4 ~ 10 CORE ALTERATIONS All directional control valves for remaining control rods shall. be disarmed electri-'ally except as specified in 3.10.A.7 and sufficient margin to criticality shall be demon-strated.

c.

The two mainte-nance cells must be separated by more than two control cells in any direction.

d.

An appropriate number of SRM's are available as defined in specification 3.10.B.

334

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUXREHENTS 3 ~ 10 COR E ALTERATIONS 0 ~ 10 CORE ALTERATIONS S

Any number of contxol

~ rods may be withdrawn or removed from the reactor core providinq the following conditions axe satisfied:..

a.

The reactor mode switch is locked in the >>refuel>>

position.

The re fueling interlock which prevents more than one contxol rod fran being withdrawn may be bypassed on a

withdrawn control x od after the fuel assemblies in the cell containing (controlled by) that control rod have been

'emoved from the reactor core.

All other re fuelinq interlocks shall be operable.

Hith the mode selector switch in the refuel or shutdown

mode, no more than one conLrol rod may be withdrawn without first removing fuel. from the cell except as specified in 4.10.A.6.

Any number of rods may be withdrawn once verified by two licensed operators that the fuel has been removed from each cell.

335

v C criticality.

The nuclear characteristics of the core assure that the reactor is subcritical even when the highest worth control rod is fully withdrawn.

The combination of refueling interlocks for control rods and the refueling platform provide redundant methods of preventing inadvertent criticality even after procedural violations.

The interlocks on hoists provide yet another method of avoiding inadvertent criticality.

Fuel handling is normally conducted with the fuel grapple hoist.

The total load on this hoist when the interlock is required consists of the weight of the fuel grapple and the fuel assembly.

This total is approximately 1'00 lbs, in comparison to the load-trip setting of 1,000 lbs.

Provisions have also been made to allow fuel handling with either of the three auxiliary hoists and still maintain the refueling interlocks.

The 000-lb load-trip setting on these hoists is adequate to trip the interlock when one of the more than 600-lb fuel bundles is being handled.

During certain periods, it is desirable to perform maintenance on two control rods and/or control rod drives at the same time without removing fuel from the cells.

The maintenance is performed with the mode switch in the "refuel" position to provide the refueling in'terlocks normally available during refueling operations.

In order to withdraw a second control rod after withdrawal of the first rod, it is necessary to bypass the refueling Interlohk on the first control rod which prevents more than one contxol rod from being with-.

drawn at the same time.

The requirement that an adequate shutdown margin be demonstrated and that all remaining control rods have their directional control valves electrically disaxmed ensures that inadvertent criticality cannot occur during this maintenance.

The adequacy of the shutdown margin is verified by demonstrating that at least 0.38% hk shutdown margin is available.

Disarming the directional control valves does not 'inhibit control rod scram capability.

Specification

3. 10.A.7 allows unloading of a significant portion of the reactor core.

Thi" operation is performed with the mode switch in the "refuel" position to provide the refuelinq interlocks normally available during refuelinq operations.

In order to withdraw more than one control rod, it is necessary to bypass the refueling interlock on each withdrawn control rod which prevents more than one control rod from being withdrawn at a time.

The requirement that the fuel assemblies in the cell controlled by the control rod be.

removed from the reactor core before the interlock can be bypassed ensures that withdrawal of another control rod does not result in advertent criticality.

Each

~ control rod Qi 341 LOCAL MONITORING STATIONS BRQV/NS FERRY NUCLEAR PLANT FX

. 4.2-1 ATH'"HS U.S.

HWY 72 BFNP ALA. HWY 20 Legend G

Air Monitor Monitor 6 TLD Station TLD S totion Automatic Well Sampler H

Dairy Farm DE C ATUR Scale 0

I 2

3 4

5 Miles

ENCLOSURE 2 DESCPRIPTION, JUSTIFICATION, AND SAFETY ANALYSIS BROGANS FERRY NUCLEAR PLANT TVA BFNP TS 173 Units 1

2 and Pa e 2a Descri tion of Chan e

Delete the sentence "Unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least Hot Standby within 6

hours, and in at least 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 />."

Also, delete the sentence, "This is not applicable if the, unit is already in Cold Shutdown or Refueling."

Justification This change allows any particular situation to be judged by its applicable LCO for equipment out of service, or by definition 1.C. 1 if no LCO is specified.

The present definition creates several cases where a

LCO is specified for this situation (T.S. 3.4.D, 3.7.E.4, 3.7.B.3 for example) which conflict with the definition.

If no LCO is specified-, definition 1.C.1 is still applicable, which-contains the requirements for a 6-hour Hot Standby condition, and 30 additional hours to be in cold shutdown.

Safet Anal sis This change clarifies the existing specification and has no effect on the safe operation of the plant.

Units 1 and 2, Pages 4,

19,

19a, 20 Unit Pa es 4

18 19 Descry tion of Chan e

Modifies references to maximum steady state licensed power to indicate that steady state is averaged and to include limitations to non-steady state power levels.

Safet Anal sis This change clarifies the existing specifications and has no effect on the safe operation of the plant.

Units 1 and 2, Pages 56 and 61 Unit Pa es 58 and 6

Descri tion of Chan e

Adds Note 11 to Table 3.2.A.

Justification This note clarifies the minimum number of operable channels per trip system and better reflects the as-installed instrumentation.

Safet Anal sis Adding note 11 eliminates the possibility of technical specification misinterpretation by clarifying the minimum number of operable channels per system.

This change will not effect the operation, safety margins, accident analysis or overall safety of the plant.

Units 1 and 2, Page 64 and 97 Unit Pa es 66 67 and 93 Descri tion of Chan e

Revises the assignment of the RHRSM pump timers.

Justification This reflects the as-built configuration and resolves OPQA audit item No. OPQAA-BF-81TS-02 item A-2.

Safet Anal sis This is an editorial change and has no effect on plant safety.

Units 1 and 2, Page 66 Unit Pa e 69 Descri tion of Chan e

Change setpoint for HPCI suction switchover from seven inches above normal water level to seven inches above instrument zero.

Justification This change is a commitment for corrective action on LERs 259/8185, 260/8165, and 296/8170.

Safet Anal sis This change more clearly establishes the setpoint at its original design level in the torus.

The technical specification is currently referenced to normal water level where normal water level is not defined.

Past technical specification revisions have changed the allowable water levels in the torus several times causing the specification on this setpoint to lose its original meaning.

Since this change establishes the setpoint at its designed level, it has no adverse affect on safety.

Units 1 and 2, Page 71 Unit Pa e 70 Changes incor rect reference from section 3.5.I to 3.5.H.

Safet Anal sis This is an editorial change with no effect on plant safety.

Units 1 and 2, Pages 88 and 110 Unit Pa es 91 and 107 Descr i tion of =Chan es Deletes reference to RMCU space high temperature isolation instrumentation consisting of RTDs in table 4.2.A and deletes note 23.

Justification The referenced RTDs are not part of the primary containment isolation system.

The reference in this table is in error.

Safet Anal sis Since these RTDs are not part of a safety system, their removal from technical specification surveillance requirements does not affect safety.

Units 1 and 2, Page 147 Descri tion of Chan e

Adds "and diesel generators" to the required operable equipment.

I Justification This change makes the limiting condition for operation consistent with the surveillance requirements.

Safet Anal sis This change simply makes the LCO consistent with the surveillance requirements and does not affect operation in any way.

Unit 2 Pa es 153 164 Descri tion The proposed changes will allow operation of unit 2 without performing additional surveillance testing when no more than two standby coolant supply pumps (RHRSW) are out of service.

Of the four available standby coolant pumps on unit 2, only two of these are required for operation without entering a limiting condition for operation.

Consistent with other system surveillance requirements, no additional testing will be required when a full complement of pumps (2) is operable.

Justification In many common situations such as a unit 1 or 3 outage, as many as two standby coolant pumps may not be available to supply unit 2.

The current surveillance requirements for unit 2 result in unnecessary testing of RHRSW pumps, diesel generators, and motor-operated valves in the service water system.

This unnecessary testing increases the probability of system unavailability and diverts operator attention to unnecessary tasks.

.Safet Anal sis The proposed changes will make the unit 2 technical specificatiqns for standby coolant supply consistent with the other units'pecifications, Therefore, any safety implications have been previously considered with regard to having two standby coolant supply pumps available per unit on units 1 and 3.

Since unit 2 coolant supply requirements are similar to units 1 and 3, the same surveillance requirements concerning inoperable standby coolant pumps should apply to all three units.

Units 1 and 2, Page 272 Unit -3 Pa e 289 Descri tion of Chan e

Corrects the referenced rate of change of suppression chamber pressure which corresponds to 0.14 lb./sec. of containment air (current reference is incorrect).

Justification The present rate of 0.25 inches of water per minute corresponds to 0.09 lb./sec.

and not 0.14 lb./sec. of containment air.

Safet Anal sis This change to the technical specifications will not directly affect the safety of the plant because the change is for correction only.

The limit, 0.14 lb./sec. of containment air, will not change, only the reference to inches of water per minute (changing 0.25 to 0.38).

Units 1 and 2, Pages

302, 303,

303a, 304, 310 Unit Pa es 3

1 2

3 3

a 4

35 341 The attached changes affect only the performance of control rod drive maintenance with fuel remaining in the control cell.

The requirement for disarming the directional control valves for all the remaining control rods has been eliminated for single conrol rod maintenance.

Also, the requirement for shutdown margin demonstration with the strongest rod out has been eliminated where related to conrol rod maintenance during refueling and replaced by the requirement for sufficient shutdown margin during the maintenance.

Justification Xt is frequently necessary during refueling outages that control rod drive maintenance be performed while fuel remains in the control cell.

This usually applies to single rod maintenance rather than to the maximum of two rods.

The "strongest rod out" shutdown margin requirement has been eliminated from both the single-and two-rod cases because it is seldom possible to determine the strongest rod in the core without performing possible to determine the strongest rod in the core without performing extensive calculations.

This information is calculated at the beginning of a fuel cycle and may change as the cycle progresses and during core alterations.

The requirement for disarming all remaining directional control valves during single rod maintenance with fuel remaining in the control cell requires a time consuming procedure which results in added personnel exposure and wear on the directional control valve electrical connectors.

Safet Evaluation A.

Removal of "Strongest Rod Out" Shutdown Margin Requirement, During control rod drive maintenance with fuel in the control cell, sufficient requirements exist which prevent additional control rod withdrawal making it unnecessary to account for the effect of an additional (strongest) rod on the reactivity margin.

In the case of maintenance on one control rod only, the one rod permissive interlock is in force to prevent additional rods from being withdrawn.

Also, the ad)acent rods to the affected control cells are electrically disarmed.

In the case of two maintenance

rods, the requirement to disarm all remaining control rods ensures that no further rod withdrawal occurs.

B.

Removal of Requirement for Disarming All Remaining Control Rods for Single Rod Maintenance In order for more than two control rods to be inadvertently withdrawn at the same time during maintenance or one control rod with fuel remaining in the control cell, multiple failures must occur.

The administrative procedures, in addition to the one rod permissive, ensure that no additional rods are withdrawn.

The ad)acent rods are disarmed to further prevent the possibility of adJacent rods being withdrawn.

Units 1

2 and A

endix B Pa e 42 Descri tion of Chan e

Deletes two environmental sampling points for raw milk samples.

Justification The milk sampling locations are being deleted from the surveillance program since milk-producing animals are no longer at these locations.

Safet Anal sis The proposed change is an administrative change with no impact on the safe operation of the plant.

II