ML19209B763

Proposed Tech Spec Changes to Accommodate Reload 3,Cycle 4 Operation for Unit 1 & to Clarify Ambiguous Specs for Units 1 & 2

ML19209B763
Person / Time
Site:	Browns Ferry
Issue date:	10/04/1979
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML19209B762	List: ML19209B761 ML19209B763 ML19209B765
References
NUDOCS 7910100407
Download: ML19209B763 (55)

Text

.

3 l-ENCLOSURE 1 PROPOSED CHANGES TO BROWNS FERRY UNIT 1 TECHNICAL SPECIFICATIONS 1138 318 Yo7 79101 00 P

I i

9-GUIDE TO PROPOSED CHANGES TO BROWNS FERRY UNIT 1 TECHNICAL SPECIFICATIONS Page vii - Reload Page 219 - Reload Page 9 - Reload Page 220 - Reload Page 10 - Clarification Page 2?1 - Clarification Page 11 - Clarification Page 254 - Clarification Page 16 - Reload Page 255 - Clarification Page 17 - Reload Page 257 - Clarification Page 25 - Reload Page 330 - Reload Page 29 - Reload Page 30 - Reload Page 97 - Clarification Page 111 - Clarification Page 112 - Clarification Page 131 - Reload Page 134 - Reload Page 145-- Clarification Page 147 - Clarification Page 148 - Clarification Page 149 - Clarification Page 150 - Clarification Page 157 - Clarification Page 159 - Reload Page 160 - Reload Page 168 - Reload Page 169 - Reload Page 171 - Clarification Page 172 - Clarification Page 172a - Clarification Page 172b - Reload Page 182 - Clarification 1138 319

i LIST OF TABLES (Cont'd)

Table Title Page No.

4.2.F Hinimum Test and Calibration frequency for Surveillance Instrumentation 105 4.2.G Surveillance Requirements for Control Room Isolation Instrumentation 106 4.2.H Hinimum Test and Calibration Frequency for Flood Protection Instrumentation 107 4.2.J.

Seismic Honitoring Instrument Surveillance 108 3.5.I MAPLHCB vs Average Planar Exposure............... 171,172, 172-a' 172-b 3.6.H Shock Suppressors (Snubbers) 190 4.6.A Reactor Ccolant System Inservice Inspection Schedule 209 250 3.7.A Primary Containment Isolation Valves.

3.7.B 7estable Penetrations with Double 0-Ring Seals......................

256 3.7.C Testable Penetrations with Testabic Cellows....

257 3.7.0 Primary Containment Testable Isolation Valves...

258 3.7.E Suppression Chamber influent Lines Stop-Check Globe Valve Leakage Rates.

263 3.7.F Check Valves on Suppression Chamber influent 263 Lines 3.7.H Testable Electrical Penetrations 26S 4.8.A Radioactive Liquid Waste Sampling and Analysis 287 4.8.B Radioactive Gaseous Waste Sampling and Analysis..

283 3.ll. A,

Fire Protection System Hydraulic Requirements...

324 6.3.A Protection factors for Respirators 343 6.8.A Minimum Shif t Crew Requirements..........

360 vii 1138 120

1 g.n.7 3,3 yy.;.

1.il1TT1tJG r./JFTY SYSTDI SETTIf.C

.)

pir:i.ciff3ttr.INThItiTY P.1 RJF.I. C1.ADDilIC INTECl: TTY In the evar:t of operation with the core maxi', f raction of limiting power dens.tj (CHFLPD) r,rcater than fraction of rated thernal power (FRP) the setting shall be modified as follows:

SS.(0.66W + 54%) FRP CMFLPD For no combination of loop recircu-lation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120%

of rated thermal pover.

(Note: These settings assume operation within the basic thermal hydraulic design criteria. These criteria arc Ll!GP, f 18.5 kw/f t for 7X7 fuel and6 13.4 Lw/ft for 8X8, 8x8R, and P8x8R fuel, MCPR limits of Spec 3.5.k.

If it is determined that either of these design criteria is being violated during operation, actica s,ha'1 bc initiated within 15 ninutas to rc: tore operation within prescriLod lir:it:.

Surveillance requirements for APF/

scram setpoint are given in specification 4.1.B.

P00R ORRp.

2.

APRM--When the reactor c. ode r.vitch is in the STARTUP POSITION, the APRM scrsm shall be set at less than or equal to 15% of rated p er.

3.

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

b.

C;re '.'hemal Powe-Limit (Pearto-Pressure <800 psin)

The APRt; Rod block trip setting r.h213 p.

D er the reactor pressure is less thsn or equal to 800 psia, 9

1138 121

f;AFETY 1,IMIT LIMITIIIG CAFETY CYSTEM CFTTING

.1 FITEL CLADDING INTEGRITl 2.3 FUEL CLADDIrlG INTEGIIITY or core coolant flow is lesc RB< (0.66W + b'2%)

than 10% of rated, the core thermal power shall not ex-where:

ceed 823 MWt (about 25% of U

rated thermal power).

p3 = Rod block setting i: pet re en t, of rated thermal power (3N)3 MWt.)

W

= Loop recirculation flow rate in percent of rated (rated loop reci rculap' ion flow rate equalr.

3h.2 X 10 lb/hr)

In the event of operation with with the core maximum fraction of limitinI'on of ratedpower density (CMFLPD) greater than fract thermal power (FRP) the setting shall be modified as follows:

Hb< { 0. 66W + h2% } FRP U

CMFLPD C.

Whenever the reactor is in C.

Scran and isol uat. ion--l '; d : n.

ab,ve the shutdown condition with reactor low water ve:

u ro.

irrininted fuel in the reac-ter vencel, the wat,er level shall not be lecc than l'(.7 in ribove the tc. of the D.

Scran--turbine stop < 10 pt :. en t.

nornf. active fue. zone.

valve clocure valve eb m

E.

Scram--turbine control valve U>,i t.ri p o r 1

Fast c lonure t o f :m t, a c i.

,. e;;0 i < t

.Iv.

00

] l

,I'l I

2.

I,ons of control > Zo ; 1; l

oil presnure F.

Scram--low con-

> JB i nche:

denser vacuum 111. cacuum G.

Scram--main steten <,10 peri <cu t line isolation val t.

cir-e ll, l'iin st eam isolation g B:",

i r i;-

valve closure--nucle u

, o t e.1 l<,w prcosure 10 1138 522

SAFLW LIMIT LIMITING SAFETY SYSTDi SETTING 1.1 Fuel Cladding Integrity 2.1 Fuel Cladding Integrity I.

Core spray and LPCI 1 378 in.

actuation--reactor above ven,c1 low water level zero J.

HPCI and RCIC 1470 in, actuation--reactor above vessel low water level zero 1

K.

Main steam isola-1470 in.

l}00q tion valve closure-- above vessel j

l re2ctor low water zero level r

r 1138 423 22

.1.1 BAS ES Because the boiling transition correlation is based on a large quantity of full scale data there is a very high confidence that operation of a fuel assembly at the condition of HCPR =1.07, vould not produce boiling tran-sition. Thus, although it is not required to establish the safety limit additional margin exists between the safety limit and the actual occurence of loss of cladding integrity.

However, if boiling transition vere to occur, clad perforation vould not be expected. Cladding temperatures vould increase to approximately 0

1100 F which is below the perforation temperature of the cladding material. This has been verified by tests in the General Electric Test Reactor (CETR) where fuel similar in design to BFNP operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation.

If reactor pressure should ever exceed 1h00 psia during normal power operating (the limit of applicability of the boiling transition corre-lation) it would be assumed that the fuel cladding integrity Safety Limit has been violated.

In addition to the boiling transition limit (MCPR = 1.o6) operation is constrained to a maximum Ll!GR of 18.5 kv/f t for 7x7 fuel and 13.4 kv/f t for n() 8x8 fuels. This limit is reached when the Core Maximum Fraction of Lieiting Power Density equals 1.0 (CMTLPD = 1.0).

For the case where Core Maximun Fraction of Limiting vover Density exceeds the Fraction of Rated Thernal Power, operation is permitted only at less than 100% of rated power and only with reduced APRM scram settings as required by specification 2.1.A.l.

At pressures belov 800 psia, the core elevation pressure drop (0 pover, O flow) is greater than b.56 pai.

At low powers and flows this pressure differentia? is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essen'f ally all elevation head, the core pressure drop at lov powers and flow vill always be greater than 4.56 psi. Analyses show that vith a flow of 28X105 lbs/hr bundle flov, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.

Thus,3the bundle flov vith a k.56 psi driving head vill be greater than 28x10 lbs/hr. Pull scale ATIAS teut data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flov is approximately 3.35 MWt.

With the design peaking factors this corresponds to a core ther=al power of more than 50%. Thus, a core thermal power limit of 25% for reactor pressures belov 800 psia is conservative.

For the fuel in the core during periods when the reactor is shut dovn, con-sideration must aise be given to water level requirements due to the effect of decay heat.

If water level should drop below the top of the fuel during this time, the ability to remove decay heat is reduced. This reduction in cooling. capability could lead to elevated cladding temperatures and clad perforation. As long as the fuel remains covered <;th vnter, sufficient cooling is available to prevent fuel clad perforation.

P00RDi E l ms m

I.I B A,1FJ 1he safet) limit has been established at 17.7 in, above the top of the

. treadiated fuel to provide a point which can be monitored and also pro-vide adequate margin. This point corresponds approxiraately to the top of the actual fuel assemblies and also to the lower reactor low water level trip (378" above vessel sero).

U.WMBCE 1.

General Electric BW Thermal Analysis Basis (CETAB) Dat a, Correlation and Design Application, NEDO 10958 and NEDE 10958.

P00R OR11RR ll38 Da o

17

2.1 AAfff

,1.

J. 6 K.

f,cactor los yater level set point for initiation of LPCI and RC I C, c l o r, a n r. m i t n steam isolation valves, and startitr LPCI and cor eyray pu'opa.

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

The design of these s stems to adequ.st ely perform the intended fu' c-

/

tion is based on the specified low level scram set point and initia-tion set points. Transient analyses reported in Section 14 of the FSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.

L.

Peferences 1.

Linford, R.

B., " Analytical Methods of Plant Transient Evaloations for the Cencral Elect ric Boiling Llater Reactor," bTDO-10602. Feb.,1973.

2. Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant Unit 1 Reload 3, NED0-24209, August 1979.

7g 4\\M1 -

1138 326 25

1.2 BASES _

There fore,.~ollowine 4.y trancier.t pressute moni'.or higher in the vessel, d

that is severe enough to cause concern that this safety limit was violate,

a calculation will be perforned using all available in'cr-atier. to de ce-eine if the safety limit was violated.

REPT.PJNCES__

(EETF. GAR Section IL,'))

1.

Plnnt SsTaty *inslysis

/S::E ~.4011er on1 Pressure Vessel Code Secti:n III 2

LT.3 Pipir.c Code, Occticn 331.1 3

itescto.- *::sel and Appurter.an es !!echenicol AnL_a (F..

? f 3.a p,

1.,

Sa' sec t ica L.2) 5.

Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant Unac 1 Reload 3, NEDO-24209, August 1979.

700'l DIElE 1138

~127 29

4 2.2 BAStu BEACTOB COOU NT SYSTilt INTFERITY To neet the safety design basis, thirteen relief valves have been installed on the unit with a total capacity of 82.6% of nuclear boiler rated stean flew. The antdysis of the vorst overpressure transient,

( 3-second cler,ure of all nain steam line isolation valves ) neg'.ecting the direct scran (vrdve position scran) results in a maximun vesrel pressure of 1272 psic if a neutron flux scram is assumed considering 12 valves operable. This results in an 103 psig nargin to the code allovable overpressure lirait of 1375 psis.

To meet the operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) is presented in Pererence 5 on pace 29.

This analysis shows that 12 of the 13 relief valves limit pressure in the stean line to 1199 psic. This analysinshows that peak system pressure is limited to 1227 psig which is 148 psiC below the allowed vessel over-pressure of /375 psic.

1138 323 30

TA3LE 4.2.B (Cunt.aued)

Instruwne check Calibration Punctional Test Function none once/3 months (1)

Instrument Channel Reactor Low Pressure (PS-68-93 & 94) none once/ operating cycle (4)

Core Spray Auto Sequencing Timers (Normal Power) none once/ operating cycle (4)

Core Spray Auto Sequencing Timers (Diesel Power) none ooce/ operating cycle (4)

LPCI Auto Sequencing Timers (Mo mal Power) none once/ operating cycle d

(4)

LPCI Auto Sequencing Timers (Diesel Power) none once/ operating cycle (4)

RfDtSV A3. Bl. C3, D1 Timer a (Bormal Power) none once/ operating cycle (4)

RImSV A3. Bl. C3, Di Timers (Diesel Power) a CO

.. ~

N cone once/operatins cycle

~O (4)

ADS Timer

3_. 2 B M. t.f.

instrumentation which initiates a reactor prntection inst runentat ion han been previded which in addition tn reactor scram, protective connequences of accidents which are initiates action to mitignte the beyond the operator's ability to control, or terminates operator er-This set of speci-in nerious consequences.

rors before they result limiting conditions of operation for the primsry fications provid7n the initiation of the cere cooling nystecs, con-system isolath n function, The objectives of treatment systems.

trol rod block and standby gas effectiveness of the protec-are (1) to assure the the Specifications bility to tive instrumentation when required by preserving its capasuch nye, tens even during tolerate a ningle tittore of any emponent of periods when porttona of unch syntemn are out of service for maintenance, adequate per-and (11) to prencribe the trip settings requir ed to assure one channel may be made inoperable for brief formance. When necessary, and calibrations.

intervals to conduct required functtonal tests Some of the octtinga on the instrun.*ntation that initiate or contrni core the high and containment cooling have tolerances explicitly stated where both critical and may have a subatantial effect en enfety. The set points of other instrumentation, where only the hip,h or and low values are a direct bearing on safety, are chosen at n low end of the wett ing hsn level away f rom the normal operating range to prevent inadvertent actu.-

i (P*%

tion of the saf ety system involved anl exposure to abnormal attuat onn.

zalves is initiated by protective instru-Actuation of primary containvient mentatien shovu in Table 3.2.A which senses the conditions f or which iso-be available whenever pri-lation is required. Such instrumentation must mary containment integrity is required.

isolation th connected The instrumentatton which initlatee primary system ia i dual boi n r e.ua cment.

The low water level tontrumentation set to trip at 177.7" (51A" abov" ven.cl cero) above the top af tho active fuel closes i sol a t ion valva 9

• n a: d drains and the RHR Sy9 tem, Drvwell uni w ppr nion Chanher exhaust, Reactor k'ater rie mop Linen (Group ? and 1 inolarten valves).

The low reactor water level ins t r utnen t al t on that tu get to trip when reactor witer eensel zero) above tho tnp of the active fuel level is 109.7" '(470" above closen the *tain ';t eam 1. i n e J elstfon Valvea and Main Steam, RCIC, and HPCI Drain valven (Group 1 and 7).

Deta119 of vaive n ouping and icquire:

closing times are given in Specification 3.7.

These trip settings are adequate to prevent core uncovery in the cane of a break in the largest line annuming, the maximum cloning tiee.

The low reactor water 1cvel instronentation that is set to trip when rtactor water level la 109.7" (470" above vessel zero) above the top or the scrive fuel (Table 3.2.8) niso initiate the RCIC and HPCI, 111 I

i i 3s 33o

3.2 B AS E_S_

and tripn the recirculation pumps.

The low reactor water icvel inntrumentation that in net to trip when reactor water Icvc1 is 17.7" (378" above vennel 7.erol above the top of the active fuel (Tabic 3. 2. B) initiate, the LPCI, Core Spray Punpo, contributer, to ADS initiation.i n.1 9terts the diesel generatorn.

Thege trip setting levcis were chosen to to-high enmigh to prevent n;>u riou, act ua t ion bu t low enough to initiate C5CS operation no that post accident cooling can be accomplished and the guidelinen of 10 CFR 100 will not tic violate.l.

For large breaks up to the complete c.ircunferential brcnk of a 26-inch rectreulation line anil wit h the trip settinr, given above, CSC5 initi.ation to initiated in t ime-to meet the above criterta.

The high dryvell prc%ure instrumentation in a diver,e signal to the water level inntrumen ation and in addition to initiating CSCS, it cnunca Inalatton nf Group 4 1 a n.t 8 t e.n l a t t on valve,.

For the breaks dincunned above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus the resultn ;;iv en above are applicable here alno.

Venturis are provided in the nain steam linen ns a menna of meanuring steam flow and alno limiting the losa of mans inventory from tho vennel during a steam line treak accident.

The primary function of the in,tto-mentation is to detect a break in the r-ain ocean line.

For the worst case accident, nain stean line break outside the drywell, a trip art ling of 140% of rated st en n flov in conjunction with the flow l icli t e r n ani main steam 11ae valve closure, limits the naun invc ' t.or y loss such t. Sat remain below 1000*F fuel is not uncovered, fuel cladding temperaturen and releane of radionctivity to the environe is well below 10 CFR 100 guideline 9.

Reference Sectton 14.6.5 FSAR.

i teintore m o l t o r t o,(

initan+cntalton in provided in the main ot em lia pi.erided on t I.

i i n,i i i.

t nioie I e...le i. t lenk. in t h. ic asenn.

Tripn nie loiute of 1,nintinn volvei.

..u-mentatton n n.I when e.. cil e '. ctune a n e t t i n d, nf

.MH' f f a s tle-main alcom Iine tunnel deterror 1y low enou.u (o

.le t ec t lenk o of the ider of li gpm; t i.u n, it in c a ;ia b l e of cov. ring toe entire spectrum of breaks.

For large brenks, the high stesa flow instru-eentation is a backup to the tempernture instrumentation.

liigh radiation mon i t o r n in the taain stean line tunnel have been provioed as in the control rod drop-accident. With to detect r.rons fuel failure the establinhed settini, of 3 t ine9 normal backgrcond, and main nttaa line teolntion valve clonore. ftanten proJuct releaue 19 limited. that

• o 10 CFR 100 guidelincas are not axceeded for this acc ident. Reference a mr.al sot point of 15 x 6

An al m, witt a I p. 2 FS AH.

norndi nf Sect ower nich round,

.s tirnviood a no, Pressure inntrumentation is provided to close the main steam isolu t ion valves in Run Mode when the main steam line pressure drops below tb")

peig.

n P001OPJIML 112 1138 U1

W*3

?M*

he npera:or with a visual indtratioa of neu-doce,. n iid e tren 3 rect.

The conseque,cca o: re.:t t vi t y act id:r.t 3 are functionn of ihr tr.itial nce:r na fln.

'* h e r e qu i r c rr.

.S c o f aasure) that.i y

r. r c r. r. i c n t,

et icolt 3 counts per necono onould it occur, ru < ins at or above :he-In:r?il value of 10

• of rnt ra pw. r i.. d in t'.: ensty,e nf t rani ent s f r o,3

~

cold randitions.

Oaa oyarabir tPIt ch in,e l vo sid he adequite to raunitor the approach (n.c:lt!:ality usin.; hczov,rncous A iir.leev patterna of scattered con rol cod v?:hd:. val.

of two opersble SR.*t's are proviiled as an codeo conuervatixn.

5.

The Rc4 Block Monitor (RBM) is dest;:ned to auto:stically prevent f uel da -age in tne even: of er or.cous rod vi:Sdraval fro 3 locatier.:, of high power den *1ty during high pover Icvel operacion. Two channels are provided, ino one of :hese ciay be byp4ssed frco the console f or cuintenance and/or testin6-Tripping of one of the channels vill block error.cous rod withdra.<al soon ensuah to prevent fuci da. age.

Iha epect-fied restrictions with one channel t.ut of re vice conserva-tiv:ly a3,. ore t h.s t fuel d.vaage vill not occur due to rod withdrawal errors when this cond'.t f on exts:s.

A limiting control rod pattern is a pattern which r e r, u l t 3 in the core being on a thermal hydraulic l i t, i t, (ie,

?!C P R given by Spe c. 3. 6. K or Ll!G R of 18.5 for 7x7 e r 13.4 for 8x8,8x8R,6 P8x8R). During use of cuch pe.tterns, it is judged that testing of the RBt! systen prior to with-drawal of such rods to assure its operability will assure that improper withdrawal does not occur.

It is nornally the responsibility of the flu c l e a r Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns.

Other personnel qualified to per-form these functions may be designated by the plant superintendent to perform these functions.

Scram Insertion Times The control rod system is designated to bring the reactor suberitical at the rate fast enough to prevent fuel dacage; ic, to prevent the !!CPR from becoming less than 1.07.

The limiting power transient is given in Reference 1.

Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification provide the required protection, and ?!CPR remains greater than 1. 0 7.

On an early BUR, sone degradation of control rod scram petfornance occured during plant startup and was determined to be canc.J by 131 lI38 332

3.3/4.4 BASFS:

D.

Reactivity Anomalies During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned.

The ria g n i t u d e of this excess reactivity may be inferred from the eritIcal rod configuratfon.

As fucI burnup pro-gresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod nattern at selected base states to the predicted rod inventory at that state.

Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity.

Furthermore, using power operating base conditions permits frequent reactivity comparisons.

Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% ok Deviations in core reactivfty greater than 17.d k a r e not expected and require thorough evaluation.

One percent reactivity into the core would not lead to transients exceeding design ccnditions of the reactor system.

References

1. Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant Unit 1 Reload 3, NEDO-24209, August 1979.

d 134 1138

33

LTtGTUtr, c0*1DIT!C:13 P34 OPP ATIC'i StavEILL/JirE PFO'JIEq{.

3,5.R Reeldual Heat Removal Svatem 4.5.B Residual Hrnt Recoval Syst-m (RHRS1 (LPCI and Cont ain:.ent (RHRS) (LPCI and Contain=ent Coolln;d Cooling) 1.

The RilRS shall be operable:

1.

a.

Simulated Once/

Automatic O p e r r. t irig (1) prior to a reacter Actuation Cyc1:a startup from a Cold Test Condition; or (2) when there is irra-b.

Pt=p Op e ra-Cace/

disted fuel in the bility sonth reactor vessel and when the reactor vessel pres-c.

Motor Opera-Oac:/

sure is greater than ted valve conth operability atmonpheric, extept en specified in specifica-tions 3.5.B.2, through d.

Fue? 71ov Rate ence/3

acnthn 3.5.B.7 and 3.9.B.3.

8' e.

Test Check Valve Once/

2.

With f.h e reactor vemsel pres-0;cratir sure team than 105 pair.. the C yc le etHRS may be renoved from ser-vice (except that two RHR pumps-c on t s t r..,e n t coolin;; mode and Erh IKI ; r.:

hn _

s-t-agaociated heat exchsngers r.uot 3,000 g n 2,.gi-,.

i,

,.i., 2

, $ s' hg remain operabic) for a period syste prm _ra

< ; m, not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while

".M I pu.ps in the : _... 17 7 g g,'

delive: 15,CCO 3g yair..;t.-

being drained of suppression indicate 1 r/r*c-r'. : n r e-char.ber quality water and 200 psig.

filled with primary coolant quality water previded that 2.

An air t at a the d p ell c.2 during cooldo n two locps vith torus heade s and no::len chall one pump per loop or one loop vith be conh eted once/5 yerce.

i.

two pumps, and associated diesel water tcat ray be perfern2d o.

generators, in the core spray systec the torus header in lieu of t'

are operable.

air test.

3.

If one RllR pump (LPCI ciode) 3.

When it is determined that oni-Rll R la innperable, the reactor pump (LPCI mode) is i n ope rab l e at a may remain in operation for a b

re llin b m ai M.

days the remaining RilR punns (L"Cl nade) period not to exceed e provided the ressining R)iR and active compenents in both wre s pumpa (LPCI mode) and both P

( I' accesa patha of the RHRS t

and the dies 1 nenerators (LPCI mode) and the CSS and shall be demonstrated to be opera-the diesel generators r e:sa in ble immediately and daily

operable, thereafter.

r ni,i

?0u.,ll Dghist a

1138 334

'll T l yr. U M D I T I nti'. FoH OW.tulloH 1.U R V E ! LL A? H. l. H t@. I,R L*t t S t. :.

J. 's. B Wealdual llent k r mo v e '6

%vste, 4.$.B Residssi liest R emva l Sys t em

_RHP.3) (LPCI and Containment (R K3 5_) (LPCI and Containment

(

Cooling)

Cooling) 4 If any 2 RSR penps (LPCI mode),

4.

No additional surveillance become inoperable, the reactor l

required.

shall be placed in the cold shutdown condition within 24 hoars.

5.

If one R H '4 pux? (con:atn-

5. When it in dsterniaed that one ncnt cooltnr. n>de) or ao-RER puup (containment co o l in,z nociated heat eschanger is node) or t.ssociat1d heat inc pe rable, the reactor exchanger is inoperable at n

[*h usy remain in operation f or ti=e when operability la re-a period not t> exceed 30 quired, the ramat ring MIR days prov(ded the reeiininA pumps (containment coolini. node).

PhR puaps (containeent the aasociated heat exchana.eri cooling code) and asso-and diesel genetstors. and all clated heat exchangers snd active componenta in the cccesa diesel g?neritnes and all paths of the MlAS (contain,-at access paths of the RHRS cooling rode) shall ha de mn-(containment cooling node) otrated to be operable in:wd t a t el-/

are operable.

an! waeily therea f ter until the in;perable RHR pump (contaitcent coaiina mode) and anacciatad heat exchanAer la r.! turned to nonaal service.

6.

If two RHR pumps (containment

6. 'a~h e n it is deternined chat t '.'o cooling mode) or associatad RH4 puaps (containment caoling hest exchangers are inopars-node) or nasociated hear ecchangers bic. the reactor may remain are inoperable at a time when in operstton f. r a pe riod coersaility is required, sS-not to exceed 7 days pro-rc=a taing R'.i2 pumpa (containmene vided the retaining EHR punps c oo lin g Leo d e), tae associated (containment caoling unde) he.t exchsngers. and diesa'.

and associated heat e.xchangers gir.erators, and all active cen-and all access paths of the ponents in the accras patha of RHR5 (containment cooling code) ths RHRS (conta in=ent cooling k4hhY i

7Q 7zD r

UU

)s

O SURVEILLANCE REQ'JIREMENTS

.L_!

TING CONDITIONS FOR OPERATION 3.5.5 Residual Heat Removal System 4.5.5 Residual Heat Removal Svitem (RHRS) (LPCI and Containment (RKRS) (LFCI snd Containment Cooling)

Cooling) mode) shall be demonstrated are operable.

to be operable immediataly and daily thereafter until at least three RHR pumps (containment cooling mode) and associated heat exchangers are returned to normal service.

7.

If two accean paths of the 7.

When it is determined that one RHRS (containment cooling or este acetse paths of the node) for each phase of the RHRS (containment coolinz ecode) mode (drywell sprays, sup-are inoperable when accc,9 in pression chanber spraya, required, all active cor.ponents and suppression pool cooling) in the access paths of the RHRS are not operable, the unit (containment cooling meda) shall may remain in operation for a bo demonstrated to be operable period no' to exceed 7 days immediately and all active con-provided at least one path ponents in the cccess paths

/*\\

or each phase of the mode which are not backed by c cec nd remains operable, operable accese path for tha sane phase of the mode (iryvell sprays, supprescion chambr. nprays and suppresnion pool coolinz) thall be de.monotrated to be cpera-ble daily thereafter until the second path is returned te ncc-mal service.

8.

If specifications 3,5.3.1 8.

No additienal surveillance through 3.5.0.7 are not met, required.

en orderly shutdown shall be initiated and the reactor shall be shutdevn and placed in the cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

9.

When the risceor vessel pres-9.

When the reactor veso?1 prescur:

sure la atmospheric and irra-is atmospheric, the ?.Si pu::a diated fuel is in the reactor and valves that are recuirsd te vessel at least one RHR Icop bo operabla shall ba demonstrated with two punpa or two loops to be operable monthly.

with one pump per loop shall be operable. The pumpa' asso-ciated diesel generators =ust

("N also be operable.

10.

No additional surveillance 10.

If the conditions of specifica-

required, tion 3.5.A.3 are met, LPCI and containment cooling are not 148 required.

]

t1MITING CONDITIONS FOR uPERATpM4 SURVEILI R C PECUIrTMyg.

3.5.3 Retsidual Heat Removal Systen 4.5.B P.a.idual. ant P.

,i (RHRS) (LPCI and Containment (R}iL) (LPCI and Qt.t s ii.. rnt Cooling)

Cooling)

11. When there 1 :. Irradiated fue.

11. The RHR inm;r no,

Qo in the reactoi and the reactor cent units whici.

,op 1<

vessel pressure is greater than cross-conn-ct in t. :it atmospheric, 2 IGR pu:aps and shall be dcnon.Inat.'

associated heat exchangers and operable nooth15 wtma a,

valves on an adjacent unit must cross-connut.npub1 Lily be operable and capable of is required.

supplying cross-connect capabil-ity except as specified in speci-fication 3.5.P..lZ below.

(Note: Because cross-connect capability is not a chort term requirement, a component is not cor.s idered inoperabic if cross-connect capability can be re-stored to service within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.)

12, 4

D If one RHR pum,' or associ-

12. When it is de t e r:tr.ed ated heat exchanger located that one Rh? pump or on the unit cross-connection associated

.ea t exc h ar.ge r in the adjacent unit is in-located on the unit cross-operable for any reason (in-connection ir. the adja-ciuding valve inoperability, cent unit la ino;orable pipe break, etc.), the at a t ime when ope ribil-reactor may remain in opera-ity is required, t h:

tion for a period not to remaining rig oump and exceed 30 days provided the associated he : excnanger remaining PJ1A punp and on the unit cross-con-associated diesel generator nection and the asavci-are operable.

ated diesel generator shall be demonstr.ted :n be operable icned tately and every 15 davs th"rentrer uatil the Laoper Ale po r aad associated heat ex-c 1 anger are returned to PD

,,3,,

4 l

P nor=al service.

E 3 LL 1

i D

l 1138 537 149

O LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 5 13.

If RHR cross-connection flow or 13.

No additional surveillance heat removal capability is lost.

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

14. All recirculation pump discharge valves shall be tested for operability during any period of

14. All recirculation pump reactor cold shutdown discharge valves shall exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if f

be operable prior to reactor startup (or operability tests have closed if permitted not been performed elsewhere in these during the preceding speci fications).

31 days.

O

}l38 l}]

150

SURVEII.1MCE %QUIRDiENTS LIMITith CONDITIONS FOR OPERATION 3.5.F Reactor Core Isolation cooling 4.5.F Reactor Core Isolation Cooling 2.

When it is determined that the 2.

If the itCICS is inoperable.

RCICS is inoperabic, the HPCIS the reactor may remain in shall be de:nonstrated to be operation for a period not operable ic:::iediately.

to exceed 7 days if the ILPCIS is operabic during such time.

3.

If specifications 3.5.F.1 or 3.5.r.2 are not met, an orderly shutdown shall be initiated anti the reactor shall be depressurizerd to less than 122 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C.

Automatic Depressurization

.C.

Aut oma t ic_Depressurizc t ion Svstem (AUS)

System (ADS) 1.

During each operating cycle 1.

Four of tim six valves of the following tests shall be the Autonatic Depressuri-perforced on the ADS:

zation System shall be operab]c:

A simulated aute at ic a.

actuation test shall N (1) prior to a startup from a Cold Condition, perforced prior to startup af ter each refuelin;; out-or, age.

Manual surveillance of the relief valves is (2) Wenever there is irra-covered, in 4.6.D.2.

diated fuct in the reactor vessel and the teactor vessel pr.ssure is greater than 105 puig, ercept as spectficd in 3.5.G 2 and 2.

When it is deternined thu rui r 3.5.G.3 below.

than two of the ADS valves ate 2.

If thicc of the six ADS valves incapable of automat ic operation, the IIPCIS shall be demonstra red are known to be incapable of autematic operation, the to b.3 operable immediatcly and daily thereafter as long aa reactor viay remain in opera-Specification 3.5.C.2 applies.

t ion for a period not to cyrced 7 d.r/s, providc<l the 110C1 system is operable.

(Note th4 the pr c: sore rol(ef function of tbene valves to assured bv section 3.6.D of thcae

}

Q9 spectficalions and rhat this sa

,; d v.pecification only alptics l'

AI to the ADS f unct ion.)

If more than thtec of r.ne six ADS vajveu aic kncvu te be incap-able of ::m on n ir opc: a t i on,

an inum itat e ordei ly r.hutdown shall be initi.m d. 'irh the 157 reactnr in a he t <:h"'.dwn con-dition in 6 houvr and in a cold shutdown condition in the following 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

1138 U-

MVF.;L LW3Cr. c.lu t a' SfNTS OPFRATION LIMITINC CONtJITIONS FOR 4.5.H Nsintenuace of Tilled Discharpa Pin 3.5.H Maintenance or Tilled Discharge Piee e auction of the RCIC and HPCI pu=ps m

adsall be a.ligned to the conder. sate of the RXRS (LPCI and Contai ment storage tank, and the pressure suppres-Spray) and core spray systems, the ston chanber head tank shall normally discharge piping of these systens be aligned to serve the discharge pipinF' shall be vented f rom the high point of the RHR and CS pucps. The condensate and water flow determined.

head tank may be used to serve the RHR and CS discharge piping if the PSC head 2.

Following any period where the LPCI tank is unavailable. The pressure or core spray systess have not been indicators on the discharge of the RHR required to te operable, the dis-and CS pumps shall indicate not less charge piping of the ineperabls sys-than listed below.

te:r. shall be vented f rom the high Pl-75-20 L8 psig point prior to the return of the F1-75 L6 L8 psig system to service.

F1-71+- 51 L8 psig Pl-7L-65 be ps18 3.

L'henever the H7Cl or RCIC syste:n is lined up to take suction Itcu the

!. Average Planar Linear Heat Generation condensate storage tank, the dis-charge piping of the HPCI and RCIC Rate During steady state power operation, the shall be vented free the high point l(aximum Average Planar Heat Generation of the systes and water flew observed Rate (MAPLHCR) f or each type of f uel as on a conthly basis.

a function of average planar exposure shall not exceed the limiting value 4

k* hen the PJ'R$ and the CSS ara 'r e-shown in Tables 3. 5.1-2; -2, -33,- % -5, -6,[-Z quired to be operable, the pressure lifatanytimeduringoperationit is indicators which noniter the dis-

.htermined by normal surveillance that charge lines shall be scnitored the limiting value for APLHCR is being daily and the pressure recorded,

exceeded, action shall be initiated with-in 15 minutes to restore operation to

'sithin the prescribed limits.

If the APLRGR is not returned to within the praecribed limits within two (2) hours.

the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Z.

Max 12num Ave r a r e Planar Licear Heat Cenera.

Surveillance and corresponding action tion Rate MAPLHCR) shall continue until reactor operation The MAPLHCR for each type of fuel a; a f u.i c -

is within the prescribed limits.

tion of average planar exposure shall be determined daily during reactor operation

f. Linear Heat Generation Rate (LHCR) at > 25% rated theraal pwer.

During steady state power operation, the linear heat generation rate (LHCR) of J.

1.inear Heat Ceneration Rate (WN -

any rod in any fuel assembly at any axial location shall not e.<ceed the CR as a function of core hei@t n W.

a maximum allowable LHCR as calculated by be checked daily during reactor cpernion at the following equation:

> 25% rated thermal power.

?DDRORjalNAL 159 i l38 40

,fiMIT.hG CONJITIONS FOR OPP.RK!!ON

~

SURVEILLANCE REQUIREMF.NTS LitCR

< LifCR [1 - (AP/P)

(L/LT)]

g LitGR

=DesignLilGR=18TkW/ft for 7x7 fuel

= 13.4 kW/ft for 8x8, 8x8R, and P8x8R fuel (AP/P)""* = lbximum power spiking penalty

= 0.026 for 7x7 fuel

= 0.022 for 8x8,8x8R,and P8x8R fuel LT = Total core length = 12.0 ft for 7x7 fuelanq/ 9xs 12.5 ft for 8x8,8x8R, & P8x8R

=

L = Axial poaltion above bottom of core If 'at any time during operation it is deter-mined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the LHCR is not returned to within the prescribed limits within two (2) hours, the g

reactor shall be brought to the Cold Shutdown i

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

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

K.

Minimum Critical Power Ratio (MCPR)

K*

Minimum Critical Power Ratio

(_MCPR)

The MCPR operating limit for BFNP 1 cycle 4 is 1.23 for 7x7 fuel, 1.24 for 8x8 fuel, MCPR shall be determined daily and 1.25 for 8x8R and P8x8R fuel. These during reacter power operation at limits apply to steady state power operation

> 25% rated thermal power and fol-at rated power and flow.

For core flows lowing any chenbe in power level or

'other than rated the MCPR shall be greater distribution that would cause opera-than the above limits times K K is the value shown in Figure 3.5.2. g.

g tion with a limiting control rod pattern as described in the bases f r Specification 3.3.

If'at any t Lne during operation it is determined by normal surveillance that the limiting value f or MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hcurs.

Surveillance and corresponding action shall continue until reacter operation is within the prescribed limits.

L.

Reporting Requirements If any of the limiting values identified in Specifications 3.5.I, J, or K are exceeded and the specified recedial action is taken, the event shall be logged and reported in a 30-day written report.

1138 441 160

3.S B AS E 5_

).S.H

%intenance_of Filled f)i nchargejife 11 the discharge piping of the core spray. LPCI. HPCIS, and RCICS are not filled, a water hammer can develop in this pipinr. when the pump and/or pumps are started. To minimize damaec to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requirca the discharr.c lines to be filled whenever the system is in an operable condition. If a discharge pipe is not filled, the pumps that supply that line must be assumed to be inoperable for Technical Specification pur-poses.

The core spray and RHR system discharge piping high point vent is visually checked for water flow once a month prior to testing to ensure that the lines are filled. The visual checking will avnid starting the core spray or allR system with a discharge line not (111ed. In addition to the visual observation and to ensure a filled discharge line other than prior to testing, a pressure suppression chember head tank is located approximately 20 feet above the discharge line highpoint to supply naheup vnter for these systems. The condensate head tank located approximately 100 feet above the discharge high point serves as a backup charging 3Ysten when the pressure suppressico chamber head tank is not in service. System discharge pressure indicators are used to determine the water level above the discharge line high point. The indicators villrefleet approximately 30 psig for a water level at the high point and !*5 Psig for a water level in the pressuresuppression chamber head tank and are con-itored daily to ensure that the discharge lines are filled.

When in their normal standby conditinn, the sisction for the flPCI and RCIC

p. amps are 41tene.1 tn the condensate storar.e tank, which is physicilly at a hir.her elevatinn than the @CiS and RCICS pipinn. This agvsres tha t the it PC i and RCIC discharme pipina, renains filled. Further assurance is prnvided by observing water flow from these systems hir.h points monthly.

tisminue /.varage Planar 1.inear Heat, Generation R,atA (MAPLHGR) 3.5.1.

This specification assures that the peak cladding temperature fo11owin5 tho postulated design basis loss-of-coo 1 Ant 8CCident Vill not GXCeed the limit specified in the 10CTR50, Appendix X.

The peak cladding temperature following a postulated loss-of-coolant acci-dont is primaril'y a function of the average hcot generation rate of all the rode of a fuel assembly at any axial location and is only dependent second-erily on the rod to rod power distribution within an assembly. Since ex-pected local variations in power dictribution within a fuel assembly affect the calculated peak clad temperature by less than 1 200F relative to the peak temperature for a typical fuel desiCo. the limit on the averar.e linear heat generation rate is sufficient to assure that calculated temperatures are within the 10CTR$0 Appendix X limit. The limiting value f or MAPMCR is shown in Tables 3.5.I-1, -2,

-3,

-4,

-5,

-6, and

-7.

166 i l38 442

3,$,y, g ar Heat Ceneration Rate (f.1!C R)

This apecification assures that the linear heat generation rate in ar.y rod la less than the design linear hent generation if fuel pellet dengification is postulated. The power spike penalty specified is based on the.snal-ysis prr=ented in Sec tion 3.2.1 of Ref erence 1 as modified in References linently increating variation in axial. gaps be-2 and 3, and a ssuae s a tween core bot tom and trap, and a swre s ul th a M*,.

con f idence, t ha.t no note than one fuel rod cureds the de'.ign linear heat cencration rate due to pover spiking. Tnc LitCH a :, a functlon of core height sh:11 bc checked daily dur-2 25% power to deterninc if fuci burnop, or cun-ing reactor operation at trol rod movrient has caused changes in power distribution. For UICR to be a

a ll:atting value below 25% rated thermal power, the MTPF would have to be than 10 which is precluded by a considerable cargin when employing greater eny perraissibic control rod pattern.,

3.5.K.

Minimus critical Power Ratio (MCTR)

At core ther aal pcver levels less than or equal to 25C, the reactor will be operating at minimus recirculation pump speed and the raaderator void content vi.11 be very small.

For all designated control rod patterns which uay be em-played at this pol.nt, operating planc experience and thermal hydraulic anal-ysis ind ic a t e u t ha t the resulting MCPR value is in e.xce ss of requirecient s by a censiderable r.a sto, *dith this low void content, any inadvertent core flew increase would only place operation in a more conservative code rela-tive to MCPR.

The daily require sent for calculating MCPR above 25% rated thermal power is sufficient since power distribution shif ts are very aiov vhen there have oc,e been significant pcver or control rod changes. The requirement for calculating PCPR vhen a II:21 ting control rod pattern is approached ensures that 11CPA vill be kaovo following a change in pcver or power shape (regardless of asa gnitud e) that could place operatten at a thermal limit.

3, $, g,,

Rep o r t in F Requirenents The LCO's associated with monitoring the fuel rod operating conditinns are required to be mat at all t imes,

i.e.,

there is na allowable time in which the plant c an ks evi ng l y e xc ee d t he liniting values f o r MAPGC'R, MCR, cnd st a ted in Specif ic at ions 3.5.I, J, cod MCPR.

It to a requirer.ent, se that if et any t ime durior, s teady st ate pever operat ien, it la determined that the liniting value s for F)?1RCR, LHCR, or MCPR are exceedwi actior, is then initiated to restore operation to within the prescribed limits. This action is initiated as soon as normal surveillance indicates that an cperating lir-it has been reached.

Fach event involving steady state operation beyond specified limit aball be legge.1 and reported quarterly.

It must ce reccanizd

t. t t i.e r e is always an action wh!ch would return any of the psreca te:o (MAPIER, 1BCR, or MCFR) to within prescribed limits, na=ely powt reduction. Under most c ir cu::ta t anc e s, this vill not be the only alternative.

H.

References i

1.

" Fuel Deost!! cation Ff *ects on General Electric Boli.ng War MJctor Puel," Supptecents 6, 7, and 8, bTIP-107 35, August 19 M.

2.

Supplencot 1 to techntesl Report on Den sif icat ion s of Gener t,'.

Electric Reactor Puela, December 14, 1974 (USA Ragulat or y St af f).

3.

Co:::::uoica t ion :

V. A. Moore t o I. S. Hi t c hc 11. "Moc i f ie d CT. too e l f or Pual Densificatico." Docke t 50-321, March 27, 1974, 169 P00R ORBINA ms m

Table 3.5.I-l MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Iype: Initial Core - Type 1 & 3 Average Planar Exposure FUV?LHCR (Mud /t)

(kW/ft) 200 15.0 1,000 15.1 5,000 16.0 10,000 16.3 15,000 16.1 20,000 15.4 25,000 14.2 30,000 13.1 Table 3.5.I-2 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: Initial Core - Type 2 Average Planar Exposure MAPLHGR (Mwd /t)

(kW/ft) 200 15.6 1,000 15.5 5,000 16.2 10,000 16.5 15,000 16.5 20,000 15.8 25,000 14.5 30,000 13.3 1138 344

Table 3.5.I-3 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 8D274L Average Planar Exposure MAPLHGR (Mwd /t)

(kW/ft) 200 11.2 11.3 1,000 11.9 5,000 12.1 10,000 12.2 15,000 12.1 20,000 11.6 25,000 10.9 30,000 Table 3.5.I-4 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 8D274H Average Planar Exposure MAPLEGR (Mwd /t)

(kW/ft) 200 11.1 1,000 11.2 5,000 11.8 10,000 12.1 15,000 12.2 20,000 12.0 25,000 11.5 30,000 10.9

'. ' 2 1138 545

TABLE 3.5.I-5 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 8DR265H Average Planar MAPLHGR Exposure (mwd /t)

(kv/ft))

200 11.5 11.6 1000 5000 11.9 12.1.

10,000 12.1 15,000 20,000 11.9 11.3 25,000 10.7 30,000 TABLE 3.5.I-6 MAPPHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 8DR265L Average Planar MAPLHGR Exposure (mwd /t)

(kw/ft) 11.6 200 11.6 1000 12.1 5000 12.1 10,000 32.1 15,000 20,000 11 9 11.3 25,000 10.7 30,000 l

0

(

172-a 1138 546 t

c TABLE 3.5.I-7 MAPLilGR VERSUS AVERACE PLANAR EXPOSURE Fuci Type: P8DR284L Average Planar Exposure MAPLilGR (Mwd /t)

(kw/ft) 200 11.2 1000 11.3 5000 11.8 10000 12.'0 15000 12.0 20000 11.8 25000 11.2 30000 10.8 1138 347 4

e 172-b t

SU RV EILt.ANc t s t i}UIRLMt N a _

MS FOR OrlpATio?#

8.tMIT,licmts i t t t:

4.6.E Jet Funpa 3,d,3.

• J e t ru,pn

6. The indicated value of core 3,6,7 Jet Purp Flow Missatch flow rate varies f ree the value derived f rom loop flow measurements by core than 101.

The dif f user to lober plenu:n c.

dif f erential pre ssure read-inE on an individual jet pump varies f rom the mean of all jet pu:rp dif f eren-tis 1 pressures by smore than 101.

2.

tihenever there is recirculation flow with the re' actor in the

1. The reactor shall not be Startup or Run Mode and one re--

operated with one recirculation circulati n pu:np 1's operating loop out of service for more arith the equalizer valve closed, than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With the reactor

'the dif f user to lower plen'uri operating, if one recirculation 41fferential pressure shall be loop is out of service, the checked daily and the dif feren-plant shall be placed in a hot tial pressure of an individual shutdown condition within det Puup in a loop shall tiot 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is vary frcm the mean of all jet sooner returned to service, pump <'.if f erential pressures in that loop by more than 10t.

2, Following one pump operation, the discharge valve of the lowg F.

Jet P w o Flow Hiematet speed ptcp may not be opened unless the speed of the faster Recirculation pu:sp epeeds shall 1.

pu=p is less than "K$ of its be checked.and logged at Icast rated speed.

ence per day.

3.

Steady state operation with both recirculation puros out of ser-vice for up to 12 hrs is per-mitted. During such interval restart of the recirculation 2dN"zd lm0,0 h N' l

umps is permitted. provided the k U}lb HML oop discharge temperature is within 750F of the saturation temperature of the reactor vessel water as detemined by do:ne pressure.

C.

Structural Integrity 1.

Table 4.6. A together vith sup-C.

Structural Interrity Piementar7 tates, specifies the The structural integrity of 1.

the primary system shall be 1 82 1138 i48

3.6/h.6 cases detected reasonably in a matter of feu hours utilizing the available leakace detection schemes, and if the cricin ccanot be determined in n reasonably short time the unit should be shut dosn to a* lov further investigation and corrective action.

The total leakege rate consists of all leakage, ident! 'ied and unidenti-fied, which flows to the dryvell flocr drain and eouip1ent drain sutps.

The capacity of the dryuell floer sue:p pucp is 50 spm and the capacity of the dryvell equip,ent surp penp is el.so 50 Ep.. Renovel of 25 gro from either of these sumps can be cc.oaplished with considerable nergin.

REE.RC!CES 1.

Nuclear System LeakaCe Rate Limits (BF;i? TSAR Subsectica L.10) 3.6.D/4.6.D Reller Valves To meet the safety desi n basis, thirteen relief valves have been C

installed on the unit with a total enpacity of 82L6P. of nuclear t oiler rated stean flov, The analysis of the vorst overpressure transient, (3-second clonure of all nain stean line isolation valves) neglecting the direct scram (valve position scran) results in a maxinun vessel pressure of 1272 rsi if a neutron flux nerar is n:suned considering 12 valves opernble. 1 hic results in on 103 psig marsin to the coue allowable overpressure limit of 1375 psig.

To meet the operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) is presented in Reference 5.

This onalysis shows t hat 12 of the 13 relief valves limit pressure in the stean line to 1199 rsic. This analysin-hova that peak system prensure is limited to 1227 rniG vhich is 148 psic below the allowed vessel over pressure of /375 psiG-P00R OR M L

!138 149 219

3.6/4.6 Basts:

P00RORGINA!.

valve operation shows that a testing of Experience in relief to detect failures or 50 percent of the valves per year is adequate valves are benchtested every deteriorations. The relief their set points are within the second operating cycic to ensure that The relief valves are tested in place once per f 1 percent tolerance.

operating cycle to establish that they will open and pass steam.

nuclear system can be The requirements established above apply when theThese requirements are applicable conditions.

pressurized cbove ambient at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed. However, these in terms of pressure, than those starting transients are much less sevcre, The valves need not be functional when the vessel at rated conditions.

head is removed, since the nuclear system cannot be pressurized.

REFERENCES _

Nuclear System Pressure Relief Sys.em (BFNP FSAR Subsection 4.4) 1.

22 in response to AEC Question 4.2 of December 6,1971.

2.

Amendment 3.

" Protection Against Overpressure" (ASHI Boiler and Pressure Vessel Code,Section III, Article 9)

Browns Ferry Nuclear Plant De,ign Deficiency Report--Target Rock 1.

Safety-Relief Valves, transmitted by J. E. G111cland to F. E. Kruesi, August 29, 1973.

Supplemental Reload Licensing Submittal for Browns Ferry 5.

1979.

Nuclear Plant Unit 1 Reload 3, NEDO-24209, August 3.6.E/4.6.E Jet Pumps nozzle assembly Failure of a jet pump nozzle assembly holddown mechanism.

and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break. Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break. Therefore, if a failure occurred, repairs must be made.

The detection technique is as follows. With the two recirculation pumps balanced in speed to within f 5 percent, the ficw rates in boch recircula-If the tion loops will be verified by control room monitoring instruments.

riser and nozzla two flow rate values do not dif fer by more than 10 percent, assembly integrity has been verified.

220 1138 350

3.6/4.6 H AS P.5 :

the core flow rate nessured by the If they do differ by 10 percent or more, jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow If the difference between neasured and derived core flow rate correlation.

(with the de: ved value higher) diffuser measurements is 10 percent or more will be taken to define the locatica within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.

If the potential blowdown flow area ir increased, the system resistance to the recirculation pump is also reduced; hence, the af f ected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a ningle nozzle failure). If tha. two loops are balanced in flow at the name pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow ratea would be indicated by the plant process instrumentation, in addition, the affected jet pu: p would provide a leakar.e path past the core thus reducing the core flow rate.

The reverse flow throur.h the inactive jet pump would still be indicated by a positive differential pressure bat the net effect would be a slight decrease (3 per-cent to 6 percent) in the total core flow measured. This decrease, together wfth the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.

Finally, the affected jet pump diffuser differential p-1sure signal would be reduced because the backflow would be less then the

rmal forward flow, A nnrzle-rtger system failure could also generate the coincident failure of f

a )ct punp diffuger body; however, the converse is not true.

The lack of iny substantial stresn in the jet pump diffuser bo3y makes failure impossible without an initial nozzle-riser systen failure.

3.6.F/4.6.F Jet Pump Flow Mismatch P00R OPJB!NAL 1138 351 o

221

b NOTES FOR TABLE 3.7.A Key: 0 = Open C = closed SC - Stays Closed GC = Goes Closed Note: Isolation groupings are as follovs:

The valveo in Group 1 are actuated by any one of the following Croup 1:

conditions:

1.

Reactor Vessel Low Water Level 470")

2.

Main Steamline High Radiation 3.

Main Steamline High Flow 4.

Main Steamline Space High Temperature 5.

Main Sr.esuline Low Pressure Group 2: The valves in Group 2 are actuated by any of the following conditions:

(*%

1.

Reactor Vessel Low Water Level (538")

2.

High Drywell Pressure Group 3: The valves in Group 3 are actuated by any of the following conditions:

1.

Reactor Low Water Level (538")

2.

Reactor Water Cleanup System High Temperature 3.

Reactor Water Cleanup System High Drain Temperature Group 4: The valvco in Group 4 are actuated by any of the following conditions:

1.

,HPCl Steamline Space High Temperature 2.

HPCI Steamline High Flow 3.

IIPCl Steamline Low Pressure Group 5: The valves in Group 5 are actuated by any of the following condition:

1.

RCIC Steanline Space High Temperature 2.

RCIC Steamline High Flov 3.

RCIC Steamline Low Pressure Group 6: The valves in Group 6 are actuated by any of the following

['\\

conditions:

1.

Reactor Vessel Low Water Level (538")

2.

liigh Drywell Pressure 3.

Reactor Building Ventilation High Radiation 254 b

h2

O The valves in Group 7 are autonatically actuated by only the Group 7:

following condition:

1.

Reactor vessel lov vater level (470")

Croup 4: The vslves in Group 8 are automatically actuated by only the following condition:

2.

High Dryvell pressure r

ii38

153

,A 255

il AS Ef.

Crour _1 - procesa lines arc isolated by reactor vessel low water level (490") in order to allow for recoval of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cocling systema. The valves in group 1 are also closed when process inotrumentation detects excenoive main steam line flow, high radiation, low prcosure, or main steam space high renperature.

Group _2 - isolation valves are closed by reactor vessel low water level

($3S") or high dryuell pressure. The group 2 iso 1'ation signal also " iso-latco" the reactor building and starts the standby gas treatment system.

It is not desirable to actuate the group 2 iaolation signal by a tran-sient or spurious oignal.

lines are normally in use and it is therefore not der.irable Group _3 - procces to cause spurioun isolation due to high drywell pressure resulting from non-safety reisted causes. To protect the reactor from a possible pipe break in the eyotein, isolation is provided by hif. temperature in the cleanup system area or high flow through the inlet to the cleanup system.

since the vessel could potcatially be drained through the cleanup

Also, system, a low 1cvel isolation is provided.

C r_oug4 and 5 - process lines are designed to re::isin operable and mitignte o

the connequcncen of an accident which resultn in the isolation of other process lines. The cignals which initiate isolation of Group 4 and 5 linen are therefore indicative of a c<.ndition which would render proccos them inoperable.

Croup 6 - lines are connected to the primary containment but not directly to the reactor vessel. These valves are isolated en reactor low water level (538"), high drywell pressure, or reactor b.tilding ventilation high radiation which would indicate a possible accident and necessitate piimary containrent isolation.

Group 7 - process lines are closed only on reactor low water level ( 4 70'$.

These clone on the came signal that initiates HPCIS and TsCICS to ensure that the valves are not open when HPCIS or RCICS action is required.

Cnop 8 - Itne (traveling in-core probe) is isolated on high dryvell pres-sure.

This is to ausure that this line does not provide a leakage path when containment pressure indicates a possible accident condition.

The nuiximun closure time for the au t or.a t ic isolation valves of the prirary containment and reactar vensel 12olation control system have been selected in connider.2tica of the design intent to prevent core uncovering followine, pipe breaks outsice the primary containment and the need to contain released fission products following pipe breaks inside the prieary containment.

In satisfying this design intent an additional margin has been included in specif ying maximum closure tines. This margin per=its '.dentification of b,

degraded valve performance, prior to exceeding the deugn closure tians.

277 P00RBRMAL ma m

5.0 HAJOR DF. SIGN FEATURES 5.1 SiTL FLATilRES Browns Ferry uait 1 is located at Browns Ferry Nuclear Plant site on property owned by the United States and in custody of the TVA.

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

5.2 REACTOR Tb7 reactor core may contain 764 fuel assemblies consisting of 7x7 assemblies having 49 fuel rods cach, 8x8 assemblies having 63 f t.el rods each, and 8x8R (and P8x8R) assemblies having 62 fuel rods each.

B.

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

density.

5.3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2-2 of the FSAR.

Thh applicable design codes shall be as described in Table 4.2-1 of the FSAR.

5.4 CONTAlt(MENT A.

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

The applicable design codes shall be as described in Section 5.2 of the FSAR.

B.

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

C.

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

5.5 FUEL STORA,CE A.

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

gg 33 1138 355

ENCLOSURE 2 PROPOSED Cl!ANGES TO BROWNS FERRY UNIT 2 TECIINICAL SPECIFICATIONS 1138

'156

SAFETY LD(IT LIMITING SAFETY SYSTDi CETTING 1.1 Fuel Cladding Integrity 2.1 Fuel Cladding Intearity 1.

Core spray and LPCI > 378 in.

actuation--reactor above vessel low water level zero J.

HPCI and RCIC 1470 in.

actuation--reactor above vessel low water level zero K.

Main steam isola-1 470lu.

tion valve closure-- above vessel reactor low water coro level D

11 1138 157

TA3LE 4.2.8 (Cunt.aued)

Instrument Check Calibration Functional Test

% -tion none once/3 months (1)

Instrument Channel Reactor Low Pressure (PT-48-93 6 94) none once/ operating cycle (4)

Core Spray Auto Sequencing Timers (Normal Power) none once/ operating cycle (4)

Core Spray Auto Sequencing Timers (Diesel Power) none once/ operating cycle (4)

LPCI Auto Sequencing Timers (Normal Power) none coce/ operating cycle 8

(4)

LPCI Auto Sequencing Timere (Diesel Power) none once/ operating cycle (4)

RERSV A3,

31. C3, D1 Timers (Bormal Power) none once/ operating cycle (4) 121SV A3, Bl. C3, D1 Timers (Diesel Power)

~

'00R BRIGWl.

'~

Cn CX3 none once/ operating cycle (4)

ADS Timer

3.2

BAnyf, instrumentation which initiates a In addition to reactor protectioninstrumentation has been provided which reactor scram, p rot ec t ive consequences of accidents which are the initiates action to mitigate beyond the operator's ability to control, or terminates operator er-This set of speci-in serious consequences.

rors before they result limiting conditions of operation for the primary fications providen the con-initiation of the cere cooling systec9, system isolation function, The objectives of treatment systees.

trol rod block and st.tndby gas the effectiveness of the protec-the Specifications are (1) to assure required by preserving its capability to tive inattumentation when tolerate a ningle tillore of any component of wch ny9 tem 9 even during periods when per t ic.n of such sy,teno are out of service for maintenance, and (ii) to prescribe the trip settings requir ed to assure adequate per-one channel may be made inoperable for brief formance. When necessary, and calibrations.

intervals to conduct required f unct tonal tests Some of the cettinga on the instrunentation that initiate or control core the high have tolerances explicitly stated where and containment cnoting and low values are noth critical and uay have a substantial effect on where only the high or safety. The set points of other instrumentation, bearing on safety, are chosen at a

low end of the sett ing hse a direct to prevent inadvertent actua-level away f rom the normal operating range (f"%

tion of the safety system involved an d exposure to abnor:al oituations.

instru-valves is initinted by protective Actuation of primary containnent conditiens for wh'ich iso-in Table 3.2.A which aen9es thebe available whenever pri-mentation shown lation is required. Such instrumentation must mary containment integrity is required.

isolation is connected The instrumentatian which initiates primary system to i dual hui a r r. uni.erae n t.

The Inw water level inntrumentation set to trip at 171.7" (53N' above vennel zero) above the top of the active fuel closes isolation vaivan in and the RitR Sy st em, Drvwell an4f wpptr<nton ChanSer mhausta and drain 9 Reactor Wnter Cleinup Liacs (Group ? and 1 isclatina valves).

The inv water reactor water level instrumentation that is set to trip when reactcr vessel :ero) above the top of the active fuel level ip 109.7" ( 470" above clonen the ?tain Stran I,i ne Isoistion Valves and Main Steam, RCIC, and itPCl I1 rain Valves (Group i and 7).

De t ai ls of valve groupin;; and require 1 closing times are given in Specification 3.7.

These trip settings are the cane of a breni in the largest adequate tn prevent core uncovery in tice.

line nasumine, the maximum closing The low reactor water level ins t r ument at ion t hat is set to trip when reactor water level la 109.7" 470" above vessel zero) above the top of the ac tive fuel (Table 3.2.8) also initiate the RCIC and HPCI, n

P00RORMAL i i 38 459

3.2 BASES and tripn the recirculation putaps.

The low teactor water levei inntrumentation that la oct to trip when reactor water level to 17.7" (378" above vennel zero) above the top of the active fuel (Table 3.2.8) initiates the LPCI, Core Spray Pumps, contributes to ADS initiation nol 9t art s the die sol generatorn.

The w trip setting level

• vere chosen to be high enouy.h to prevent npuriou, actuation hut low enough to initiate CLC5 operation no that post accident cooling can be accomptinhed and the guidelinen of 10 CFR 100 will not be violated.

For large breaks up to the complete circunferential break of a 28-inch rectreulation line an1 with the trip setting given above, CSCS ini.tiation tu i n i t 14 t eel in time to mort the above criteria.

The high drywell prenure instrumentation in a diverw nignal to toe unter leve1 inntrumentation nnd in addition to initiating CSCS, it causes inoiation of Groupi / and n inointion valves.

Enr the breakn d i n c u n ne.1 above, this inst rument at ion will initiate CSCS operncion at about the e a rne timc as the low water level instrumentation; thus the resultn given above are applicable here alao.

V e n t u r i an are provided in the main steam lines as a r-cans o f mea su ring steam flow and alno limiting the loss of mans inventory from the venoel during a ateam line treak accident.

The primary function of the instru-rnen t a t ion is to detect a bres in the main steam line.

For the wurat case accident, rnain steam line bren out9ide the d r ywe l l, a trip settin6 of 140% of rated 9 team flov

• a ennjunction witn the flow limitern nnl main steam line valve closura, limita the mann inventory lous such that fuel 19 not uncovered, fuel cladding temperatures remain below 1000*F and release of raJioactivity to the environs is well below 10 CFR 100 guidelines.

Reference Sectinn 14.6.A FSAR.

.. t iatute ainitoring innern -ntatinn in provided to the main at enn lim on th1. i n il r n -

piovide.1 ionnei eo.le t ei t I enk i io t h. ic asenn.

T r i p e.

ns.

meneasjon and when c..

d.-

c n m... l.cin t e of 1 oint ion vulven.

11' n e t t i n ir of Mo* F foi th-main atenn line tunnel det.rror is low e nn n g.,

to detect l ea' n of the rder of li spn. t hu s, it in capable of covrting the e

entire spectrum of breaks.

For lntge ti r c a k s, the hir,h steau flow i n t. t r u-mentation is a b a c 'e u p to the temperature instrumentation, liigh radiation mon i t or s in the main stean line tunnel have been provined to detect P.ross fuel failure a1 in the control rod drop accident.

With the establiahed 9etting of 3 t imes normal background, and ma in c t eam line isolntion valve clonore, f i n.e t o n product relcano i <a limited en that 10 CFR 100 guticlines are not xceeded for this acc ident. Reference n m nal sot point, of 1 5 x witn a s l, r".,3 l

An fu U' 6.2 FSAR.pouer en cW rowd,

Secti norni[n nro riced also.

Pressure instrurnentation is provid d to close the main accan isolar:o valves in Run blode when the main steam line pressure dropo below 88 peig.

P00R ORI K I m8 m 112

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L'idITUTC CO lDITIONG FOR C' FATION Sff".VEILLAN N _"E?II' M 1 _

3.5.R Reeldunt lient Removal Sveten 4.5.B Residual Heat Recoval Syger~

(RHRS) (LPCI and Contain=ent (RRRS) (LPCI and Containment Cooling)

Cooling) 1.

The RilRS shall be operable:

1.

n.

Simulated Once/

Automatic Operating (1) prior to a reactor Actustion Cycle startup from a Cold Test Condition; or (2) when there is irra-b.

Pu.:p Opera-Once/

diated fuel in the bility ronth reactor vessel and when the reactor vessel pres-c.

Motcr Opers-0;icc /

aure to grcoter than ted valve

non th operability atmospheric. except an specified in specifica-tions 3.5.3.2, through d.

Pu=p 71ov Rata Ct.ce/3 caontho 3.5.B.7 and 3.9.B.3.

e.

Test Check Valve once/

2.

With the reactor vegnel pres-sure lean than 105 p a i r., the Operatine C yc le rlHRS may he re oved from ser-vice (except that two RHR pu ps-containment coolin;; rsde and EanLmi 7,p 3g., ; t.,.g a

t r, q i, u,3 aannocia ted heat exchangers ::ust 9,000 g7 y y.n3-renain operable) for a period system pre r:r., y ly; ;3.

Two not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while L?C: pumps 1, t r r. ss' lonP shall being drained of suppression deliar 15,000 gp: yain:t en chanber qeality water and indicated sys r.: precure a:

EO P S i.i.

filled with primar/ coolant quality water provided tha.

2.

An air test on t.ne dryvel2 and during cooldown two locps with ms

<ade s ad nml.:n d_.1 4

one pump per loop or one loop with be cenducted once/5 yr.rt.

I two pumps, and associated diesel water test ray be perferned oa generators, in the core spray syste:

the terus header in lieu of t.c are operable.

air test.

3-If one it!IR pump (LPCI mode) 3.

When it is determined that one IUIR in inoperable. the reactor pump (LPCI mode.) u m nope rab le at a may remain in operation for a time when operability is required, days the rernaining IUIR pumps (LPCI node) period not to exceed /

provided the re.aining UiR anc! actM components in both access pumpa (LPCI mode) and both P"

" E lhC U'

'dCl

""d accesa patha.f the RHRS the CSS and the diesel mnerators (f.PCI mode ) and the CSS and shall be denons t rat ed to be opera-the diesel generators re=ain ble inmediately and daily there.

operable.

after.

O 145 P00R DEEL m9 m

'i t T I'iG CONDITioth FOR fiPEluTION SU RVic t LLANC L lt LQlj K h U < i :.

J. '3. B f esidual llent F.rm vaip p en

4. 5. B R e s i ! gl]i e n t Re-oval Syity (R HP.5 ) (LPCI and Conta1*. ment (R}Lafd iLPCI and Con ta inme n t Cooling)

Cooling) 4 If any 2 PJG pur.:ps (LPCI =cde),

4.

No additional surveillance becorte inoperable, the reactor l

required shall be p bced in the cold g

shutdown condition within 21+ hoars.

5.

If one RHR p urr a (con:ain-

5. vaen it in determined that one acnt coolinr, code) oc ao-RFR pump (containment c o o l in ?,

occiated hunt cy hanxer la node) or t.ssocia ud heat incperable, the reactor exchanzer in inoperable at a h

issy remain in operation for time when operability in re-a period not ta exceed 30 quired, tbc remaining 3HR days provided the reclinina pumps (c on ta inme n t coo li n g

-'od e).

RtiR puq". (cont a incen t the aasociated heat e x cha n a,e r a cooling mode) and asso-cnd diesel generators, and all ciated heat exchangera snd active components in the nccess diesel generators and sll paths of the RitRS (containm-nt access paths of the RHRS cooling nade) shall be de-on-(containment cooling node) strated to be. operable 1-ndiately are op-rable.

and va nly thereafter until the inaperable RiiR pucy (containment cooling mde) and asacciated hest exchan;,er is raturned to no mil service 6.

If two RHR rumps (containment 6.

',th e n it is detemined that two cooling mode) or sesocistad RHA puips (contain=ent molin",

best exchangers are inopars-m3e) or naiociated hn t

?< changers bic, the reactor may re sin are (noperable it a time when in opersttun for a psriod c,-eri3111ty la required, the not to exceed 7 days pro-r e::.s in in g RM p r.p s (c o n t a i nme n t vided the re ra i ning EER pumps coc'.ing tzde), tae sanociated (contain sen t cooling onde) hen exchangers, and diesal and associated heat exchangers ganeritors, and all active com-and all accesa parha of the ponents in the access esthe of RHRS (containment cooling code) ths RHRS (conta Lnsent cociinz 147

SURVEILLANCE REQUIEEMENTS

.L TINC CONDITIONS FOR OPERATION 3.5.B Residual llent Renoval Systen 4.5.B Residual Heat Renovel Svsten (RKRS) (LFCI snd Centainment TKHRS) (LPCI and Containment Cooling)

Cooling) code) ehall be deronstrated are operable.

to be operable immediataly ard daily thereafter until at least three RHR pumps (containment cooling code) and associated heat exchangers are returned to normal service.

7.

If two accean paths of the 7.

When it is determined that one RHRS (containment cooling or more access patha of the mode) for each phase of the RHRS (containment coolin? ende) are inoperable when acce,3 in mode (drywell sprays, eup-pression chanber sprays, required, all active componente and suppression pool cooling) in the acccas paths of the RRRS are not operable, the unit (containnent ecoling ncda) shall may remain in operation for a bo denonstrated to be operab12 period not to exceed 7 days immediately and all active con-provided at least one path ponenta in the eccess path 9

[~5 or each phase of the node which are not backed by c cecond remains operabic.

operable a: cast path for tha sanc phase of the mode (drywell aprays, supprescien chanbe r sprays and ouppression pool cooling) shall be de.onatr'ted to be epera-ble daily theresfter until the second path is returned to ner-nal service.

8.

If specifications 3.5.3.1 8.

No additienal eurveillance throu6h 3.5.B.7 are not net, required.

en orderly shutdevn shall be initiated and the reactor shall be shutdown and placed in the cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

9.

When the reactor vessel pres-9.

When the reactor vessel arcssuro sure is atracopheric and irra-is a nonpheric, the 151 pum7a diated fuel is in the resctor and esivea that are recuir3J. to vessel at least one RHR locp be operable shall bu demonstrated with two punpa or two loepe to be operable monthly.

with one pump per loop chall be operable. The pumps' asso-ciated diesol generators must

("N also be operable.

10.

If the conditions of specifica-10.

No addit ional surveillance required.

tion 3.5.A.5 are met, LPCI and

~-

containment cooling are not 148 required.

PnM 1RGINAL

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIEHF.NT_i.

3.5.B Residual Heat Removal Systen 4.5.B Rr.idual rie n t R nova!

.: tn (RHRS) (LPCI and Containment (RR'IS ) (LPCI and Containrent Cooling)

Cooling)

11. When there is irradiated fuel 11.

The RHR pumps on the adja-in the reactor and the reactor cent units which supply vessel pressure is greater than cross-connect capibility atmospheric, 2 RRR ptraps and l

shall be dccan.itiated te b.-

associated heat exchangers and operable month 1', whea tha valves on an adjacent unit =ust cross-connect e n p:1b i l i t y be operable and capable of is required, supplying cross-connect capabil-ity except as specified in speci-fication 3.5.B.1Z below.

(Note: Because cross-connect capability is not a chort term requirement, a cor.penent is not considered inoperable if cross-connect capability can be re-stored to service within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.)

l O

I i

I l

I t

i i

I I

i I

l

/2, If three R1G pumps or 12.

'#'c " it ! t e ' n

associated heat exchtnp ri that t i r.,- :m < r r

e located on the unit cross-t isoc i.it ed, 'a t

< d t i n r, ' -

I p

connection in the adjacent locitec on rue unit cro l.-

units are inoperable fo r connectton i.. r}.

idir er n -

any reason (including valve unita ara t r.o re r.t r i o

.t' 149 1139 n04

LIMITING CONDITIONS FOR OPERATION SURVElt'.ANCE kEQUIREtT'gh a time when operability inoperability, pipe break, is required, the re-etc), the reactor may remain maining RHR pump and in operation for a period not associated heat exchanner to exceed 30 days providea on the unit cross-connec-the remaining RHR pump and tion and the associated associated diesel generator diesel generator shall be are operable.

demonstrated to be oper-able inmediately and every 15 days thereaf ter until the inoperable pump and associated heat exch&nger 13.

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

service, the unit may remain in operation 1.'. N additional surveillance required.

for a period not to exceed 10 days unless such capability is res tored.

14. All recirculation pump disc',arge valves shall be tes.ed for operability

14. All recirculation pump during any period of reactor cold shutda.vn discharge valves shall p

be operable prior to exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if reactor startup (or operabili ty 'ests have closed if perinitted not been p Nd elsewhere in these during the preu :ing specifications).

31 days.

1139 J05 e

150

Lli41 TIM; CONDITIO% FOR OPERATION SURVEIL.LWCE..XO.UlldltiW rS-..

3.5.F Reactor Core Tsolation Cooline 4.5.F Reactor Core Iso lat imi i: ol ia.a If the.CICS is inoperable, 2.

When ii is deterr it.eJ.aia t Li.-

O 2.

M the reactor may rertain i RCICS in iaoper ao i.e, t h" m :i..

operai. ton for a period,

shall be de:aor.at r. ti:c t o be to exceed 7 days if th~

operable i t.t aed i.. L e l y.

IIPCIS is operabic during such time.

3.

If specifications 3.5.F.1 or 3.5.F.2 are nat met, an orderly shutdown shall be initiated and the reactor shall be depressuri:eed to less than 122 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

G.

Aut oraatic IL:nressitc f.za t ion G.

Autornt i_e_ p.'pf._

a n c. a d o

System (ADS)

" y s t e..i (AD:Q l.

Four of ti.e six valves of 1.

During e.u.h o;mtaitu.

.i the Automat le Derre:.ori-the foitowlan t..I

..h s i zation Systen.; hall be perforted on t he ADE.

operablo:

.i s i:..u t.,

i il i--

(1) prior to a. tat tup ac t i:a t i:>u Lim t from a Cold Cendition, per for r "ri prio) i..

a f ter c.u:b i. t n.

ii,,

or, ar.e. :ta m i1 (2) whencver there is irra-o t b.

ro !, e t

,w

'"t.

a diated ftiel in t'u reactor ve'ssel and s he t eact or vessel pr.- a:o is gre:1te r than 10; m,ig, e,

ept as specitied ia 3.i.r.J and 3.5.G.3 t,elow.

2.

%en.: it vt.i....

i t h u.

t.., of sives i m.i., a. e o tu 2.

If thtee

.t t'n

.n /eX r

are known to b.

.nr.pable of the L Cth u

i-

.o bc

1 cru autem nic optt.

ti reactor c:oy rc: ai, in epera-da i ly LL.i

i. ion for periad ne; to Sp e c i l.' c ; t. c.

,.!i.

ex.:een 7 cay-nov.1 - ! Liie llPCI s):;t"4 is-i.

.t..

(Note that ti.e p i. :

mt..

relief fuectioa o f tite:e valves is as.or._d '

section 3.6.D of

n. e

){,

%(

.i., tS u spectticetiem

,rlies specificarice o...

1.. )

1: ncre t o ti.e W ti.

i than three o f :..e d>

,C.

valves nre 1 no' -* :o be incap-c.ble o f r ut o;.a * : c o a. ca t ion,

an immediate ord.ii-d att h w

[

shall lie init i.it d.

itb tb" 15/

reactor in a hr' ch u. J..,t coi-dition in 6 hour:: and i t. a cold shutdown condinton in the f f )h 1106 follewing 18 heura.

y

• U H V E I L LA'4C P a f QU l P.y f.14T O

i. lg l Nc.

.f para i l l f Pe % FnR O P E FLAT I O'6 i

s. 6. E let ru r 4.6.E Je t P ue.p s, 3 6. r A t Pump F] et-Miccatch b.

The indicated value of ct.re flow rate varies fro., the value derived f rcm loop flow tireturements by core than 101.

c.

The dif f user to Iwer pli cuts differentiel presaurc re.d-ing on an individual j e t pu'p varico frem tha.icate of all jet ptmp dif ferus..

tici presource by core char.

10I.

2.

Whenever there is recirculet;.cn flow with the reactor in the 1,

'Ihe reactor she.H not b.3 Startup or Run Hodr and one ' e-operated with one recirculatica circulation punp is opera;fnt loop out. of service for tsore with the equelizer velve c)cied, the dif fucer to lower plenu-than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With the reactor operat.ity;, if one recirculation differential pt,asure shall i=

checked da!!7 and the diffe..n-loop is out of service, the O

tini pru aure of an individual n

plant shan be placed in a hot shutdown conditior uithin jet pump in e loep shall not

" * "/ from the racan of all jet 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is sooner returned to service, pump cifferential rreseures in that loop by esrc than 104.

2.

Follo'. ring one pump operr. tion, F.

Jet Pump Flow Hiamatch the di; charge valve of.the low speed picp may not be opened 1.

Itecircula' ion pussp speeds sh s11 unicss the speed of the faster be checked and logged at least pun:p is less than 507, of its once per day.

rated speed.

3.

Steady state operation with both recirculation punps out of ser-vice for up to 12 hrs is per-mi t ted. During such interval restart of the recirculation pumps is oermitted, provided the 3fg '

I

[

loop discharge tenperature is

""{'

1 3

i ldg fg within 750F of the saturation temperature of the reactor vessel water as determined by dome pressure. The total elapsed time iii natural circula-tion and one pump oferation must C.

Struccural Interrity be no orcater than 24 nrs.

(;. Structural Int.r,rity 1.

Table 4.6.A together vith sup-O

'17 'l he 6 Es~rTural inter.rity of t he prima ry systeni shall be ple:nen ta r7 n t es, specifies the 182 1139 n07

O 3.6/4.6

!as_rg:

If they do differ by 10 percent or more, the core flow rate measured by the pump diffuser differential pressure system must be checked against the jet rate derived from the measured values of loop flov to core flow core flow correlation. If the difference between measured and derived core flow rate 10 percent or note (with the derived value higher) dif fuser measurernents iswill be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs. If the potential blowdowt,. low area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a ningle nozzle failure). If the two loops are balanced in flow at the name pump speed, the resistance characterist'es cannot have changed. Any mbalance between driv. loop flow rates woe a be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakar,e path past the core thus reducing the core flow rate. The reverse flow throur.h the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 per-cent to 6 percent) in the total core flow measured. This decrease, together with the loop flow increase, would result in a lack of correlation between measured.tnd derived core flow rate.

Fina.ly, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be aess than the normal forward flow, f

A nnr.zle-riner system failure could also generate the coincident failure of a jet pump dirfuger body, however, the converse is not true.

The lack of any substantial stress in the jet pump diffuser body makes f ailure irnpossible without an initial nozzle-riser system failure.

3.6.F/4.6.F Jet l' ump Flow Mismatch 300R ORB E 1139 003 n

221

D NOTES FOR TABLE 3.7.A Key: 0 = Open C = Closed SC = Stay's Closed f"l

}Q h Q h[II ((Ilh tr1L I

GC = Goes Closed UU11 U 3

Note: Isolation groupings are as follove:

Group 1: The valveo in Group are actuated by any one of the following cor.ditions:

1.

Reactor Vessel Lov Water Lovel (470")

2.

Main Steamline High Radiation 3.

Main Steamline High Plev 4.

Main Steamline Space High Temperature 5.

Main Steanline Lcv Pressure Group 2: The valves in Group 2 are actuated by any of the folleving conditions:

(*N 1.

Reactor Vessel Lov Water Level (538")

2.

High Dryvell Pressure Group 3: The valves in Group 3 are actuated by any of the folioving conditions:

1.

Reactor Lov Water Level (538")

2.

Reactor Vater Cleanup System High Temperature 3.

Reactor Water Cleanup System High Drain Temperature Group 4: The valvco in Group 4 are actuated by any of the folloving conditions:

1.

HPCI Steamlino Space Ptgh Temperature 2.

HPCI Steamline High Flov 3.

IlPCI Steamline Lov Pressure Group 5: The valves in Group 5 are actuated by any of tts followin:

condition:

1.

RCIC Stennline Space High Temperature 2.

RCIC Steamline liigh flow 3.

RCIC Steamline Lov Pressure Group 6: The valves in Group 6 are actuated by any of the followin:

O conditions:

1.

Reactor Vessel Low Water Level (538")

2.

liigh Dryvell Pressure 3.

Reactor Building Ventilation High Radiation i

9 007

D Group 7: The valves in Group 7 are autou tically actuated by only the following condition:

1.

Reactor vesse low water level (470")

Group 8: The valves in Group 8 are automatically actuated by only the following condition:

2.

High Drywell pressure bbk ON$fjf)l, r

1139 010 e

255

n P00R DEM DASES Croup 1 - proc:ss lines are isolated by reactor sessel low water level (490") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cocling systemo. The valves in group 1 are also closed when procesa inotrumentation detects excenoive main stese line flow, high radiation, low proonure, or main steam space high tenperature.

Group 2 - isolation valves are closed by reactor vessel low water level (33S") or high dryuell pressure. The group 2 isolation signal also " iso-latco" the reactor building and starts the standby gas treatment systeo.

is not desirable to actuate the group 2 isolation signal by a tran-Itsient or spurious oignal.

is therefore not desirable Groupj - process lines are normally in use and it isolation due to high drys ell prea:mre resulting f rco to cause spurious non-safety related causes. To protect the reactor from a possible pipe break in the eyote:a, isolation is provided by hiSh temperature in the cleanup nystem area or high flow through the inlet to the cleanup system.

since the vessel could potentially be drained through the cleanup

Also, nyntem, a low icvel isolation is provided.

4 and 5 - process lines nre designed to remain operable and mittute Gr_o,uL the conocqucncen of an accident which resultn in the isolation of other procens lines. The olanals which initiate isolation of Croup 4 and 5 proccos linen are therefore indicative of a ecndition which would render them inoperabic.

not directly Croup 6 - lines are connected to the primary c.,ntainment but to the reactor vessel. These valves are isa1.ated on reactor low water level (53S"), high dryvell pressure, or reactor building ventilstion and nececeitate high radiation whicl would indicate a possible accident pt imary containr ent isolation.

Group 7 - process lines are closed only on reactor low water level (470").

These close en the came signal that initiates HPCIS and RCICS to ensure that the valvce are not open when liPCIS or RCICS acticn is required.

Group 8 - line (traveling in-core probe) is isolated on high dryvell pree-G e.

This is to assure that this line does not provide a leeksge path when.conta ment pressure indicates a possible accident condition.

The maximum clonure time for the auto-atic isointion valves of the primary centninment and reactor vennel iaolation control system have tecn selected in consideration of the design latent to prevent core uncovering f o l lew i n.:

pipe breaks outsiac the primary containacnt and the need to contain releaned fisnion products following pJpe becaks inside the prieary containment.

In sa t ie f v in g, this design intent an additional atrgin has been included in specif y ing maximun closure cines. This margin ptruits identification of times.

h, degraded valve performance, prior to exceeding the design closure 277 1139 011

ENCLOSURE 3 1139 n12