Text

Proposed Tech Specs Clarifying Terms Sys & Subsystems as Terms Apply to Core & Containment Cooling Sys,Addl Safety Related Plant Capabilities & Reactor Water Level Trip Settings
	ML20079J215
Person / Time
Site:	Cooper
Issue date:	09/30/1991
From:	; NEBRASKA PUBLIC POWER DISTRICT
To:	;
Shared Package
ML20079J213	List: ML20079J215; NLS9100058, Application for Amend to License DPR-46,consisting of Proposed Change 93 to Tech Specs Clarifying Terms Sys & Subsystems;
References
	NUDOCS 9110150346
	Download: ML20079J215 (67)
	v • d • e

_ _ _ . ___

f' APPENDIX A 91101EIO346 910930 hDR ADOCK 05000298 PDR

_ _ _ _ _ _ _ - _ = _ _ _ _ _ _ _ _ ._ __ . _ _ _ _ _. _ _ _ _ _ _ _ _ _

9 TABLE OF CONTENTS (cont'd)

Page No.

SURVEILLANCE LIMITISG CONDITIONS FOR OPERATION REOUIREMENTS 3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 114 - 131 A. Core Spray and LPCI Subsystems Sptem A 114

3. Con in nt Occlia;; SubIys::s (RER Service Waterb 3 116

^ 117 C. HPC1 M y stem C D. RCIC Mystem D 113 E. Automatic Depressurization System E 119

7. Minimum Lov Pressure Cooling System Diesel Generator Availability F 11G G. Maintenance of Tilled Discharge Pipe G 122 H. Engineered Safeguards Compartments Cooling .I 123 3.6 ?R2d.ARY SYSTEM 3OUNDARY 4.6 in - 15 3 A. Thermal and Pressuritation Limitations A 132

3. Coolant Chemistry 3 13 3a L Coolant Leakage C 135 D. Safety and Relief Valies D 13e E. Jet Pumps E !37 F. Rect:culation Fump Flow Mi th F 137 i G. Inservice Inspect'.on G 137 H. Shock Suppressors (Snubbers) H 1372 3.7 CCNTAINMENT SYSTIMS 4.7 159 - 19 A. ? imary Containment A 159

3. Standby Gas Treatmen: System 3 165 C. Secondary Containment C 165a D. ?:1 mary Containment Isolation "alves D 166 3.3 MISCELLANEOUS RADICACT!VI MA O;AL SOURCES 4.3 135 - 136 3.9 ACI!L;ARY ILEC"".CCAI. SYSTIMS 4.9 193 - 20; A. Auxiliarr Electrical Iqu.ement A 193

3. Operation with Inoperable Esaipmen: 3 195

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A. Refueling In:arlocks A '.C 3

3. 3 205

... Cora spen Monitor

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.. .4ew . e, _me .4 e4... . ens WW 8 E. 3 pen

Iuni CJ:k 3*ndling E )C6 3nse *... ... "

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.... 2.L. .W wa A. Aver:ge 213na Linea 3ea: $d".e!!:isn 13:2 A2E3C2' A $..

3. Linear Haa: kneration la:s ' ICT 3 Ill

. Mini um 2:1-1:a1 ? war la:i: "C2?D : 212

-ti- ,:... .. , u-.

- ~ ~ - _ , _ _ , - - _ _ _ _ _ _ _ - - _ _ , _ _ _ _ _ _ _ _ , _ , _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

~. - - -

. l 1

- - -- TABLE OF CONTENTS (cont ' d) " -

,. _Page No.

( SliRVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.12 ADDITIONAL SAFETY RELATED PLANT CAPA3ILITIES 4.12 215 - 215f A. Main Control Room Ventilation A 215 B. Reactor 11 ding Cl;;;d Cooling 3 Weee System B 215b C. Service Water System 4;gg C 215c D. Battery Room Vent D 215c 3.13 RIVER LEVEL 4.13 216 3,14 FIRE DETECTION SYSTEM 4.14 216b 3.15 FIRE SUPPRESSION WATER SYSTEM 4.15 .

216b 3.16 SPRAY AND/OR SPRINT.LER SYSTEM (FIRE PROTECTION) 4.16 216e 3.17 CARBON DIOXIDE AND RALON SYSTEMS 4.17 216f 3.18 FIRE HOSE STATIONS 4.13 216g 3.19 FIRF. BARRIER PENETRATION FIRE SEALS 4.19 216h 3.20 DELE *ED 2161 l

3.21 ENVIRCW. ENTAL / RADIOLOGICAL EFFLUENTS 4.21 216n A. Instru entation 216n B. Liquid Effluents 216x C. Gaseous Effluents 216aa D. Effluent Dose Liquid /Gesecus 216all E .- Solid Radioactive Vaste 216a12 F. Monitoring Progra 216a13 G. Interlaboratory Co:parisen ?r:: as 216a20 3.22 SPECIAL TESTS /EXCE?!!ONS 4.22 216bl A. Shutdown Margin De=onstration 216bi

-3. Trairing Scartup 216b2 C. Physics Tests 216b3-D, Startup Tast Progra: 216b3 5.0 MA.iOR DE2!d5 FEA~URES 3.1 Site Features 2 '. 7 5.2 React:r 217 3.3 Reac:ar vessel 217 5.1 Cca:ai =ent 217 5.3 Fuel Sc: rage 213 3.6 3eismic Oesign 213 3.7 3arge Traffi: 213 6.0 AOMINISTRAT!*lE CONTRCLS 3.1 0:;ani:ation 21?

6.1.1 lesponsibili:7 2 '. ?-

6 . '. . : Jffsi a 2'?.

5...] 7 '.a n : Staff - 3h.f: c::'_a en: 2.-

4 . '. . . ?;an: 3:aff - Cua; :::2:. :na 2.31

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. 0

, 2.1 Bases: (Cont'd)

5. Main Steam Line Isolation Valve Closure on Lov Pressure The low pressure isolation of the main eteam liner (Specifi-cation 2.1.A.6) was prcvided to protect against rapid reactor depressuri:ation.

3. Reactor Vater Level Trip Settings Which Initiate Core Standby Cooling Systems (CSCS)

A The core standby cooling +wksystems are designed to provide suf-ficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit fuel clad temperature. to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1*. To accomplish their intended function, the capacity of each Core Standby Cooling System component was established baoed on the reactor lov vater level scram set point. To lover the set point of the lov vater level scrom would increase :he :apacity requirement for each of the CSCS components.

Thus, the reactor vessel lov vater level scram was set lov enough to permit margin Ior operation, yet will not be set lover because of CSCS capacity requirements.

The design for the CSCS components to meet the above guidelines was dependent upon three previously set parameters: The maximum break si:e lov vater level scran set point and,the CSCS int:iation set point. To icwor the set po:nt for initiation of the CSCS may lead to a decraase in ef f ec:1ve core cooling. To raise the CSCS initia-tion set point vould be in a saf e direction, but it vould reduco the margin established to prevent ac:uation of the CSCS during normal opera: ion or during normally e:tpected transients.

Transient and acciden: analyses repor:ed in See: ion 14 of the Safety l Analyses Repor: demonstra:e tha: these conditions resul; in adequate safety =argins for the fuel.

C. Refersness for 2.! 3ases j i

1. " Generic Reload Fuel Application." NEDE-240ll-?, (most current approved submi::al).

2. "Caoper Nuclear Station Single-Loop dperacian," NTOC-24258, May L980.

3. "Supplemen:al leicad

icensin; Submittal f ar Cooper Nuclaar Station Uni: '. ,"

(applicable reload document).

1 Safa:" Analysis Report (Section !!?).

-11~ o ;, 4

l

'OTES FOR TABLE 3.2.A ,

1. Whenever Primary Containment integrity is required there shall be evo operable or tripped trip systems for each function.

2. If the minimum number of operable instrument chant.els per trip system requirement cannot be met by a trip system, that trip system shall be tripped. If the requirements cannot be met by both trip systems, the appropriate action listed below shall be taken.

A. Initiate an orderly shutdown and have the eactor in a cold shutdown condition in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3. Initiate an orderly load reduction and have the Main Steam Isolation Valves shut within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

C. Isolate the Reactor Water Cleanup System.

D. Isolate the chutdown Cooling Oys:c , modt, o4 the. MCl Sy ste.m.

3. Two required for each steam Jine.

4 These signals also start the Standby Gas Treatment System and initiate Secondary Containment isolation.

3. Not required in the refuel, shutdown, and startup/ hot standby modes (interlocked with the mode switch).

6. Requires one channel from each physical locacian for each crfp system.

7. Lov vacuum isolation is bypassed when the turbina stop is act full open, manual bypass switches are in bypass and sede switch is not in R L'N .

3. The instruments on this table produce primary containment and system isolations. The following 11ating groups the system signals and the system isolated.

Groun 1

!aolation Signals:

1. Raactor Lov Low Lov Water Level (1-143.3 in.)

2. Main Steam Line High Radiation (3 times full power background)

3. Main Stea= Line Low ?ressura (1823 psig in the RCN mode)

e. Main Iteam Line Leak Detection (<200*?}

5. Condenser Lov 7acuum (17" Hg vac5um)

Main Stass Line H1;h Flow (<i30% of raced flow) I

5. ~ L

'.solatiens:

1. RS!7's

2. '. lain S team Line Drains

,... ~

NOTES FOR TABLE 3.2. A (cont'd. )

Croup 2 Isolation Signals:

1. Reactor Lov Water Level (24.5 inches)

2. liigh Dry Well Pressure ($ 2 pstr,)

Isolations:

1. RHR Shutdown Cooling '7- - mode, of whc. 9.MS- Syst em .

2. Dryvell floor and equipment drain sump discharge Lines.

3. TIP ball valves

4. Group 6 isolation relays Group 3 Isolation Slgnals:

1. Reactor Low Unter I.evel (24.5 inches)

2. Reactor Vater Cleanup System High Flow (5200% of system flow)

3. Reactor Vater Cleanup System High Area Temperature ($ 200* F) u.'

Isolations:

1. Reactor Vater Cleanup System Group :.

Isolation Signals:

Trwided by instruments on Table 3.2.3 (HPCI)

Isolations:

Isolates the HPC: steam line Group 5

!.iolation Signali:

ais ' c j :.3 7.C n;

'rovided bv instrumenta on Isolations:

1.10 l a t # 3 iho TCIC St?am . i n e ..

c e nnp ;

, * > t -T f_ i7 n i ! g".a . 6

'* ,a ;

3r St*D 2

30'1'On ~ .

taict,c ,'o He % o,n- '

" , 6 , a. ,

5

3.2 BASES in addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and gtandby was greatment gystem2T' The objectives of the specifications are (1) to assure tYe effectiveness of the protective instrumentation when required even during periods when portions of such systems are out of service for maintenance, and (2) to prescribe the trip settings required to assure adequate performance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situat ions.

A. Primarv Co- - ;n t Isolation Functions Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2. A which senses the conditions for which isolation is required.

Such instrumentation must be available wneneser primary containment integrity is required.

The instrumentation which initiates primary system isolation is ennnected in a dual bus arrangement.

The low water level instrumentation, set to trip at 168.5 inches (54.5 inches) above :he top of the active fuel, closes all isolation valves except those in Groups 1, 4, 5. and 7. Decalls of valve grouping and required closing timas are given in Specification 3.7. For valves which isolata at thin level this trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second valve closing time. Required closing times are less than this.

The low low low reactor water level instrumentation is set to trip when the water level is 19 inches (-145.3 inches) above the top of the active fuel. This trip closes Groups 1 2nd 7 Isolation Valves fReference 1), activates the remainder of the C5C3 subsystems, and starts the emer;ency diesel generators, These trip level set:ings were chosen to be high enough to prevant spurious actuation but low enough :o initiata CSC3 operation and primary system isolation so that post accident cooling can be accomplished.

~ '-

.c.

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

_,_ 3.4 STAyDBY LIQUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM

\

Applicability: Applicability:

' Applies to the operating status of Applies to the surveillance r* quire-the Scandby Liquid Control System. ments of'the Standby Liquid Control System.

Objective: Objective:

To assure the availability of a sys. To verify the operability of the tem with the capability to shutdown Standby Liquid Control System, the reactor and maintain the shutdown condition without the use of control rods.

Specification: Specification:

A. Normal System Availability A. Normal System Availability During periods when fuel.is in the The operability of the Standby Liquid reactor and prior to startup from a Control System shall be shown by the Cold Condition, the Standby Liquid performance of the following tests:

Control System shall be operable, su bsyste rn except as- specified in 3.4.3 below. 1. At least once per month eachgewep>-

-This system need not be operable ;y?cep shall be tested for operability when the reactor is in the Cold by recirculating deminerali:ed water to the test tink.

x Condition and all control rods are fully iaserted and Specification 3.3.A is met. 2. At least once during each operating evele:

s

$cc

a. Check that tha 3ettings of the systam relief valves are 1450 1300 psig.

5. Manually initiaca che aystem, except explosive valves, and pump boron solution from the Standby Liquid Control 3torage Tank througa the l

recirculation path. Minimum pump fles rata af 33.2 gpm againat a Jy3 tem head af L3CC psig shall be l Verittec. Af ter pumping boron aclution the systam vill be

1ushec vit:: 24m:nera11:ea vater.

. .:anuC _- in : tat, m. af :he Standb.

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.ou.: n=rs a t m w cs 2ne l

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l LIMITING CONDITIONS FOR OPEPATION SUPNEILIANCE REOUIREMENTS .

-3.4 4.4.A.2.c (Cont'd.)

pump demineralited water into the reactor vessel from the test tank.

These tests check the actuation of the explosive charge of the tested loop, proper operation of the valves, and pump operability. The replacement charges to be installed will be selected from the same manufactured batch as a previously tested charge.

d. Both

'h systems, including both explosdre valves, shall be tested in the course of two operating cycles.

B. Qp, era don with Inocerablg B. S.grveillange with Inonernble comoonents: --- Comoonents: we.operwas

{% wt,$pP -

subsy 3 tt:,rn sub+r tern -

1. From and after the date tha t k- 1. When a err; ncar is found to be r Fnd: :: x +-an; is made or found inope rable ,t i t; = ' e ::rper .c OPNO4 to be inoperable, Specification shall be verified to be o p e.rabl e

- / "

ce, rna s % 3.4.A.1 shall be considered immediately and dally thereafter g.,g fulfilled and continued operation tatil the i p c r ^ '. c:rp r- i:

era mc permitted provided that the q-i :' meperable. wbyvern '3 .

moperade e.. .m. 41s returned to an operable re%rud to a.s cpe.robt enatu s.

M 54 *Ae"' condition within seven days.

C. Sodium Pentaborate Solution C. Sodium Pentaborate Solution Ac all times when the Standby Liquid The following tests shall be Control System is required to be performed to verify the availability operable the following conditions of the Liquid Control Solutton:

shall be met:

1. The net volume versus concentration L. Volume: Check and record at least of the Liquid Control Solution in once per day.

the liquid control tank shall be maintained as required in Figure 3.4.1.

2. The temperature of the liquid 2. Tamocrature- Check a nti recard at control solution shall be maintained leas: once per day.

above the curve shown in Figure 3.4.2. 2. Cancentration: Check and record at least once per month Also sheck conc $ntration anv :ime ateror %r,n Is

'. M - ,

2

3.6 DM ES.

f STANDBY LIOUID CONTROL SYSTE'i A. 4 The conditions under which the Standhv Liquid Control System must provide gk shutdown enpability areIf identified via the Plant Nuclear Safory Operational j no more than one operable control rod is withdrawn, i

-- , Analysis (Appendix G),

h NO)' the 'hasic shutdown reactivity requirement for the core is satisfied and the [

WM u CocM Standby -

Liquid Control system is not required. Thus, the haste reactivity N we -

hvemconws, requirement for the ccreais the prirnary determinant of when theg quid gentrol ' y6 i

e he, dec . system is required. -

1, l Sub f Wm5 3, ;;s 7

l N tonhoa frhe purpose of$ thgl iquid (ontenl, gystem is to provide the capability oi jCM PCV'W" # bringing the reac' tor Trom full power to a cold, sennn fren shutdownennditionl i N" assumine that none of the withd.rawn control rods can he inserted. To me e t this 4T,aIo Pu ord a ob}ective, the Hquid c nm-E system is designed to inject a quanttty of boron g todepeden.w -

that produces a concentration of 600 ppm of boron in the reactor core tu less Sumn ecom m gg enan I?S minutes. The 600 ppm conecntration in the reactor core is required legt % to bring the reactor considering t he hot ftorom fullreact cold power todif tvity a 3.0 ferrpercent nre. xenon Ak subc ritical condition.

poinnuinr,. crc the f Q3%gd yago time requirement for inserting the boron solution was selected to override the!

rate of reactiVlty inSerrion caused by Cooldown of the reactor iollowing thell y

- xenon poison peak. ^

i n x)ct s ce m

'f"W f.wb the mintmmn limitation on the relie f valve setting is intended to prevent the

'd recycling of liquid control solution via the lif ting of a relie f valve at too low a pressure. The upper limit on the tell f valve netting $ provides vratem protaction from overprassure. 6 Imeccewic w y e l b%WWl 3,josf'y e ms

3. Only nue oc the t' o grandby p uid ontral -- g

& is neaded for operating the system. 'One inoperab1( -

it does not itame d i a t e l v threaran shutdewn capabilitv, and reactor operat!nn -

. can

. weenntinue while the

- ,. La being rapaired. Assurance that the rematntag gystem aill per f orm i ts intended f unction and that *.he long term averaz,e avallantlity of the vestam is not reduced is obtained for a one out of two system by an allowabla equipment out o t' se rvice time of one rhird )f tne normal survet.lan:o frequene: This metnod derarmines an equipment out of ser' ice time of ran d'ys. Additional conservatism t.; tntrnduced by nducine, the 11lowan! aut af service rime to seven davs.

3 I0 Vel indL0ation and alarm indicat0 whether *he solutton Vo[ume la s > hanimi .

wnich might indicato a posathie aalu:Lon conentration chane. The t%-

inte rVal has oeen eS tablished I,n confideration o f these factor 3. Temperatura Mm! quid l e v s.1 alarms '9r ~he i t .n ira ants ne ' a r ed in the 'not rn ! wm De Co$ution i3 he?t it [e33: Il'" II,nVD h e Sar' ration *' N rr O e r a a r a :a ;a *i S i.e 1'ain 10 inc Ndd u igliaSt bor7n pr9ciDLtatiln. 9 piytra a 2

[ LIMITING CONDITIONS FOR OPERATION ! SURVEILLANCE REQUIREMENT J

3.5 CORE AND CONTAIWENT COOLING SYSTEMS 4.5 CORE AND CCNTAINMENT COOL!"O SYSTD!S Aeolicability: Aeolicabilitv:

Applies to the operational status of Applies to the Surveillance Requiremen:s the core and containment cooling of the core and containment cooling ystems, ytebsystems which are required when the corresponding Limitiog Condi:Lon for Operatien is in effect.

Ob i e ctive: Obiective:

To assure :he coerability of,the core To verify the operabil'.ty of, .ae core and containment cooling sehsys- and containment cooling e-d sys:ams tems under all conditions for which under all conditions for shi:h this this cooling capabill:7 is an essential cooling caeability is ar essential response to sta:ica abnormali:1es. response :o station abnornalities.

Sneetfication: S oe ci fic a t i.pn.:

W Q A. Core Sorav and LPC! 4ebsystems A. Core fasi- and L?CI-4essvstems

(

l. Both gore spray subsystems shall be 1. Ce.e Spray SvSrfstem resting.

operacle: --

Itau Frecuency (1) prior :o reactor star:up from a Cold Shu:down, or

~

a. S'1 1>ced. Once/ Opera:ing Automatic Cycle (2) when there is irradiated fuel 1: Actuation Tes:.

he vessel and when :he reac:or vessel pressure is grea:er than b. Pump Operability Once/ non:h atsospheric pressure, encept as specified in 3.3.A.2 and 3.5.7.3 c. Motor Opera:ed Once/=en:n below. Valve Operability

d. Puso flew ra:2. Once/0 men:5s 3cch loops shall deliver at least 1700 gpm agains: a

ys:em head c:rres-pending := a differ-i en:isi pressura of l > 113 psi berveen

! :ne reae:or vessel I and :he primary i

con:ainmen:.

I y

e. Core 3cray Header

_? Ins trumen:2: :n Chack Once/dav l

2a;;Sra:2 :nca : :n: ;

MM

4 0 I

LIttITING CONDITIONS FOR OPEFATION SURVEILLANCE REOUIREMENTS

~

3.5.A (cont'd.) 4.5.A (cont' d. )

i

2. From and after the date that one of 2. When it is determined that one p, ore the gore gpray subsystems is made or jpray subsystem is inoperable. the found to be inoperable for any operablo gore p ray subsystem and reason, continued reactor operation shall be verified l 1s permissible during the succeeding ke-ene to be LPCI subsyste(imrnediately.The operable seven days provided that during such operable gore Lpray subsystem shall

(

seven days all active components 4ot be verified to be operable daily !

I the W cg;re 3 spray subsystem and thereafter.

N"'*N, sa ac tive componeItts tof -e m 4 LPCI I subsystenP and the diesel generatorsb""I are operAle.

3. LPCI subsystem rastinr, shall he as

3. Both LPCI subsystems shall be follows:

operable:

Itam frequency (1) prior to reactor startup from

a. ,.lmulated a Cnce/ Operating a Cold Condition, except as Automatic Actuat. ton Cycle specified in 3.22.B.1, or .

.est (2) when there is trradiated fuel b. Pump Operability once/ month in the /essel and when the reactor vessel pressure is c. :totor Operated once/ month greater than atmospheric 'laive Ope rab tlity pressure, except as specified in 3.3.A.4 and 3.5.A.5 below. d. Pump Flow Rate Onen/3 mnnths

, During s in e,l e pump LPCI, each PHR pump snail deliver at least ~700 GPM but no more than 3600 GPM agni.ns- a sys tem head etiivalent to a react'r "assel praesure of 2n n.:id ann-a fir'?W [ [ p r c '; S u r $ VLTh Narer L O ve '.

An a e -es s ro a cr. p cec enej below rhe jet ;my . At : u- som o^ "^e Arvwe4 acd

era hedea C'"diL!""N- '"d. N I I l"* "ha ' ' hC cwd n o a i.'.cs cru . E year, j at ioast 13,000 ; t'M . % m r !

n. Racirtulation cump -tischarr,a vai ms s ha t.1 he testad each rafaelinn OUCO33 LO veC[fy b 1 [ '. Open 'O .

o ' a *'* ' '- Closed iN d 2 6 sCconds, c pen a u ~_, -+-

i '.;h e n it is det e minnd that onc , . '

S. hn anti '.'"Or 'he daC3 ha t Otw s !

the T!R i /C ' y apq t : in ,' e i T

he RiiR '?C;: Oumps is made .. ,

l at a time when it is remti.rr .> w

ound ta he inocerable for am i j ,ne rah i e . the r?mainiut u mi ,7 reason continue : reactor mera tiu -.:nonnnents f o C ,he L?c' sito nvi- r m t" 4

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l '- no r:n i S 3 (h le '

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0 n 1 '. iuring j ~

oc"na.ti u air:- invs prnvi ded - - ! j r a .,n tn n r. . p r *f r m u s f ".:W N' f ine" d n n+- o ' : mue o ' i

< t u t- l u n 'i r v da. ! u i *":. + S G" & aining ac-i"a a cenn n "' r f a r .~ j ! and -

one > ' N no u "g],}

, . ,. sunerscam r and al. ac t ra l i cr.c-*,fra mnn .t3 e w i no t. ,r> m- !

o ** v' temS '

~.One" t*r" , , . hi m A :. c a m 3 ,

1 *' nL'====" - ! [ 3 3 *

,E"- FN - m ., m . . _ _ .

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21MIT!NC CONDITIONS FOR OPERATION SURV G Ll E E REOUIREMENTS me a pa cm 4.5.A. (Con t' d. )

""**Y*

3. 5. A (Cont' d. )

5. From and after the date that one 5. '.Jhe n i t is determined %thats.be"LPCI f.rCI subsystem is made or found to subsysum is inoperable.#hoth gore be inoperable for any reason, ypray subsys tems anc tam wule&.ne.-

. cnntinued reactor operation is em , ,- < a-'shall he verified l M j; permissible only during the en be operablo imo.ediately and daily 08 #

succeeding 7 days, un'.ess i t, is thereafter.

-f*g

' P" sooner made nperable , provided that 7 davs all acttve during such -

2 n a kr f w % ' O y * "/ #'

u v d m ; components s e both J gore ypray o- pg ,g ,,,,c mg i subsystems,y, , .

.J t. . - 'N ,

,v ma . A a n. 6 m s . , . . .f L ** A h e, n i and MIdlesel generators ;~,N k + < .: b -

9 -yu uw a shall be operable.

6. All recirculation pump discharza 6. All recirculation pump discharge valves shall be operable prior to valves shall be tested for reactor startup (or closed if operability during any period of e l s e wb a ' , in these Reactor enld Wutdown exceeding permitted specifications), an hourn, i f oria rabi lit y testa have not been performed during the pt eci ding 31 day, 7 The reactor shall not he started up with the RHR system supplying cooling to the funt pool.

3- If the requiramants of 1.;.A l.2.3.4.5,6 or 7 cannot be met, wn . f .s orderly shutdown ot the reacter G e e.d e \ Mo~ DMC#\ L 2. - A l shall be initiated and the reactor y~ g ( A 2., e- h-w shall be in the cold shutdown ./

condition vi thin M houri s

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tHMTU1!jfMDITIONS FOR OPEPATIoN et't"n.Tg Mr tr nontnMENTq 3.5.n 'd.)

4.5.S (Cont'd.)

From and af ter the date that any RilR 2. 'Jben it is determined that any Rl!R 2.

Jervice ynter booster pump is made g gervice gator booster purnp is

% ft""'*" or found to be inoperable for any -Mr,at inererable. the r e n:n i n i n r, active M '* C 3 *

t eason, continued reactor operat ion us *Mstponents t of the A **w=i---% - i s..,,en g m . only durinr, the

~b ntse cha li be ve r i f ied l g c g., ,,,,n, in

. y pe t m l.i s ib l e and to be ope t .ih l e imrtedintely e of.g a.a. :neceeding thirtv d4tys. onless such weekly the renf ter.

i, ems er pump is sooner made opertble '"

gg,W,Q ;st ovided that during, such thirty y,,,g, q,u w ,,, g g., ,,, m # , n, -. .* ,, ,

> ,c.mg .innt a il 4 4- r ac t tve cotponentsfU~ , ic e p er a g pum p t s o 4 ,u.f me,, e,ppem,r* ,

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g, % t ' ! - e t c ~ ' in6- ud" " t " 'f e r .

the opccable e -

gall be

+it made or found to be inoperable veriCled to be nperable ima itate ly l l%er

$wupte ' for any reason, continued reactor oraration i s permissible only dur tor, and rinil y therea f ter.

"***',W '

the succeedinrs sevan davs unle<s C'""*" H <uch snh *;v e r eE 4,w.p. i s mor.nc e mnde C u 'L M r v 4.e ' A ' "

' # ' N' operahle. provided that all active 5 v u 616+C ro 't'*

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r i a:: s oc !a r e d di e s e l ge ne r a t orrJaara and operable. -

a. Il the r eipii r e me n t s of 1.5.3.1 1, 5 . 3 . . o .- 1. 5.11. 3 canna r * * * :n e t , in ort =r ! shutdown shall be initiated inr1 the rencror shall be in a cold shutdnwn modition within ' houre m: ' '- : < v e r o n, , i n" ' -e,-mn 1

1.

The !T^! Mvs tem mhall Sa ,perab'.e HrC mvsrem ,r: w, M.a i . ,e i L we ne"c r P ha'r e is irradiated fuel in par.ytme:1 12 fn i ; v. s the r+ ic tar zalsel . reacror pres:ure

.' n ne m

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. S'!"'!EI' ? ANC" o retti n ru rNT J1MITING CODDI?l0NS POD OPE"AT!M 3,5.C 11PCI &*hsys tem (cont' d. ) 4.5.C. HPCI M system (cont'd.)

a. x-

2. From and after the ds *.e that the Ita.g Frecuerg, ilPCI &iahsystem is made or found to bn inop~e rable for any reason, d. Flow Rare at once/3 months continued reactor operation is approximately 1000 permissible only during too psig Steam Prens.

succeeding seven days unless such e bsystem is sooner made operable, e. Flow Rate at once/ operating 150 providing that during such seven apprnximately 150 cycle

$Q,,yy days all active componentse of the psig Steam Press.

- ADS subryrt , the RCIC gystem,-+h*.

both 1.PCIsubsystem$andbothgore gpray The llPCl pur,.p[shall ha demons tra t ed subsystems are operable. to be capable of delivering at least

/e250 y,im for a rystem head

3. 'Jith the surveillance requirements corresponding to a reactor pressure of 4.5.C not performed at the of 1000 to 150 psig.

required intervals due to reactor ahutdown, a reactor startup may bc ? tihen it in At crmined that the liPCI conducted provided the appropriate M ;yntem is itiope rnble, the RC I C , WWe, surveillance is performed within be* L i.PCI runnyst e{. and hnth got l

e. d hours of achieving ISO psig gpcay subt,y n t mo : -hall be vertflad ranctor steam pressure to be operable im% diately. The RClC gystem shall be verified to se 4 If the requirements of 3,5.C.1 operable daily thereaftsr. In cannot be met, an orderly shutdown addition, the ADS - St- loe!:

]

I shall be initiated and the reacter shall be demnnntrated to be orserabia pressure shall be reduced en 113 immedi a h'i v and da il y there ftar p n i v, or less wit.hin 24 hoorn.

D. P a n c '. , r ': , r e I:, '. i t io n 7 4n l i n!i 1 ' ac ,r ora MP ' o n m " r, i3C[471 % svgtam

, a'*.j* g g g a- 4.

1 The ?CIC hgystam 1 hall be operable 1 RC;F M;y s Mm testinr, shal; w whenever there is i.rradiated fuel in perfarmei is follows:

une reactar 'r e s s e l , the reactor p r a :::ure la graatar than 113 p't i c, , '* m Mue y;,-

md 3r:9r M etacar ':tartup f n m a ,

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LIMIT 1NG CONDITIONS FOR OPFRATION SURVFf f 1ANCE REOUTPFMENT 3.5.0 (cont'd.) 4.5.0 (cont'd.)

lti:n Frecuency

b. Pump Operability nee / month ,

c. Mctor Ortrated Once/ month Valve Operability -

d, Flow Rate at once/3 months lggerl approximately 1000 psig Steain Pressure

2. From and af ter the date that the RCICtf is made or found to be e. Flow Rate at Once/ operating i inoperable for any reason, continued approximately 150 cycle reactor power operation is psig Steam Pressure ,

permissible only during the succeedin5 seven days provided that The RCIC pump shall be demonstrated during such seven days the HPC1961s to be capable of delivering at least operable, q m,,, 400 gpm for a system head corresponding to a reactor pressure

3. With the surve illanc.e requirements of 1000 to 150 psf 5 of 4.5.D not performed at the t required intervals due to reactor 2. When it is determined that the P.C C shutdown, a reactor startup may be +++hystem is inoperable. the HPCISA conducted provided the appropriate shall be verified to be operable survelliance is performed within immediately.

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of achieving 150 psig M* '" i reactor steam pressure.

4 If the requirements of 3.5.0 1 & 2 '

cannot be met, an orderly shutdown shall be initiated and the reactor prorsure shall be reduced to 113 psig or less within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

E. Sutnma tic '>cro s surit a tinn '>s t,m E. Automnet: Depp uurt ation Svn r ODS } L12.L.

1. The Automatic Deprossurization 1. During each operating cycle the Wvstem shall be operable whenever following tests shall be performou there is irradiated fuel in tre on the ADS:

reactor vessel, and the reactor pressure is greater than 113 psig A a lmula t vl autmnatic an tan c ion *. a = -

and prior to a startup from a Cold shall he performed prior to nart.m Condition, except as soecified in after each refueling outage.

3.5.E.2 and 3.5 E.3 below.

i

~-.' 3: , 7

l 1

LIMITING CONDITIONS FOR OPEPATfnN SURVEft.1ANCE REOUTREMENTS

~

3.5.E (cont'd) 4.5.E (cont'd)

2. From and af ter the date that one 2. When it is determined that (na valve I- valve in the g,utomatic of the ADS is inoperable, thw ADS depressurization -subyyste's is made :ub:y:::= actuation logic for the or found to be inoperable for any other ADS valves shall be l reason, continued reactor operation demonstrated to be operable is permissible only during the immediately. In addition, the HPCI I succeeding seven days unless such wbgystem shall be verified to b9 valve is sooner made operable, operable immediately, provided that during such seven days theHPCIegystemisoperable.

3. With the surveillance requirements of 4.6.D.5 not performed at the required intervals due to reactor shutdown. a reactor startup may be conducted provided the appropriate surveillance is performed within 12 '

hours of achieving 113 psig reactor steam pressura.

4 If the requirements of 3.5.E.1 or

3. 5.E.2 cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to at les st 113 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

T. Minimum Low Pressure Cooline and Diesel Generator Availabiliev F. Mbimum Low Pressore Coolinc and

1. During any period when one diesel Diesel Generator AvnJ, lability generator is inoperable, continued 1. When it is determined that one diesel generator is innperable. 4 esad ig' reactor operation is permissible only during the succeeding seven im r e- - - M t"y mA J' ~" ? *

days unless such diesel generator is ---*ntecelin~ r % rt m tshall P/D sooner made operable, provided that be verified to immediately and daily thereafter, be operable l

$", e7'" 41 c' -the ice pr cur- ;er: :-4

,d In addition, the operable diesel w p-c% nt !nrent ::alin; curry:::::

ef- ' the*[F8 ting diesel generator sadi- generator shall be demonstrated to he' operable ar.d the requirements of be operable immediately and every 3.9.A.1 are met. If th.'s three days rhereafter.

requirement cannot be mot, the requirements of 3.5.F.2 shall be < r<. W C. .L Cese. See:iy , o rd i ,%

2. During any period when both diesel M i~/wCt. Wc? " s e V F t M generators are inoperable. continued S hef'Co*d e w t. o p< cMW reactor- operation is peemissible die. w a m ece.c only during the succeeding 24 nours unless one diosel generator i s _

sooner made operable, provided tha:. i i f ., c w e wj a, .mg o: 1 11 ^e i c_ -recrur- ~'

1 l 0 0 4 D M " *

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"tmu r niim _ m ~ m fara '

6perable and the reactor power le"el oc% 2 4;. S e d e W a*:r is reduced to 25' of estad power an i u 3 f e..r,. m

ne requirements of 3.9. A. '. are mo t .

I If this requirement cannot be met I j y .+ 3 ,. wm.,

. <-,4

.u

-+ an orderly shutdown shall Se i

initiatad and ':he reactor piv.ed ;,

l :he old shutdnwn conditi6n vi"i- l

[ *i hours.

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SURVEILLANCE REOUIREMENTS ,

LIMITIfC CONDITIONS FOR e_

OPERATICN -

4.5.F (cont'd.)

3.5.F(cont'd.){[. p . i t.o)

3. Any combina; ion of inoperable compo-nents in the -ear: 2nd -::" M i"- " t- { L ec,1, c H V. i>e rv e W c o < f ,

-ee44*g-+yse+m6v shall no t detest the j g g g, , p.g g capability of the remaining operable components to fulfill the cooling functions.

4. 'Aen irradiated fuel is in the reactor vessel and the reactor is in the Cold ~

Shutdown Condition both ** ora spray cow b t X. A. +#o sy 6*c m5, and wbsys t ems ,9 en.e-* e4, .ms-4enwena+as- N S

W ' Lf W C' D

lin.; eubty :4ma may be inoperable, both provided no work is being done whi:h Sua'"/***

S has the potantial for draining the reacter vessel. Refueling require-ments are as specified in Specift-cation 3.10.F.

5. Wi:h irradiated fuel in the reactor vessel, one control rsd drive housing may be open while the suppression chamber is c:mpletely drained provided that:

a. The reactor vessel head is recoved.

b. The spent fuel pool ga:es are open and the fuel pool water level is maintained at a level ,> 33 feet.

c. The condensata transfer system is operable and 2 minimum of 230,000 gallons of va:er is in :he :enden-sate storage :ank.

d. The au:o:utic mode of the dr.-sell sump pump is disabled.

e. No maintainance ta being c:ndue:edi vnich vill prevent filling the suppressian chamber to a level above :ne ; ora jpray and L?C:

suctions.

f. With :ne excep: en of :he supprss--

non enamber water supply. -s ' aoth I 5 we l port gpray ,3yst4:s and % ?C. .

tu.sys t e : ar a 2peracle.

3 The con:rol rod is vit d:rar. ::

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENT Y - . . . . . . . ,

3.5 F (cont'd) ,' 4.5.T (cont'd)

h. A special flange. capable of sealing a leaking control rod housing, is available for immediato use.

1. The control rod housing is covered with the special flange following the removal of the control rod drive.

J. No work is being performed in the vessel while the housing is open. ,

C. Maintenance of Filled Discharge Pipe G. Maintenance of Filled Discharto Pipe m.

- henever,

gore Jpray subsystems, LPCI The following surveillance requirements subsystems. HPCIb or RCIC4are required shall be adhered to to assure that the 4 Y ,4, ,

to be operable, the discharge piping discharge piping of the gore gpray from the pump disenarge of these sys- subsystems. LPCI subsystems, dPC!nand tems to the last block valve shall RCIC 4 are filled: 6vsme0Uf be filled.

1. Whenever eh:-C::: Sprayr-LPC'

o C.oit. G e m **-AC*C--+yeseme- 4e4 a ma de ope rable , the Subs ydt em , LP C.T. discharge piping snail be vented from 6 o ve z +t m ,

eit ,the high point of the svstem and gp e,, g em,ec water flow observed initially and on a

onthly basis.

t cs c g y ,.g rr . 3

2. The pressure switches which monitor l

the L?C:, sore jpray, HPC and RCIC ,re"*

lines to ensure they are full shall be functionally tested and calibrated every three months.

l i

l l

%,19,36

3.5 BASES _

A. Core $orav and LPCI Subsystems This specification assures that adequate emergency cooling capability is available whenever irradiated fuel is in the reactor vessel. I O l The limiting conditions of operation in Specifications 3.5.A.1 thraugh 3.5.A.f4 specify the combinations of operable subsystems to assure the availability of l the minimum required cooling systems. During reactor shutdown when the l 1 residual heat removal system is realigned from LPCI to the shutdown cooling mode, the LPCL, Sy4r.edconsidered operable.

M.wo f, c'.D The gore pray gyste6*!s designed to provide emergency cooling to the core ,

by spraying in the event of a loss-of-coolant accident. This system >

functions in coe.binat ton with the LPCI pvstem to prevent excessive fuel clad temperature.

-The LPC! sOgystem is designed to provide emergency cooling to the core by flooding in the event of a lors-of-coolant accident. This system functions in combination with the ; ore spray gystem to prevent egeessive fuel clad temperature. The LPCI W and the tore quate cooling for break areas of approximately{0.2 square fee]t up to andpray striisy including the double-ended recirculation line break '.ithout assistance f rom ,

the high pressure ecergency core cooling subsystens.

The' allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in reference (1). Using the results developed in chts reference, ,

the repair period is found to be41 /2 the test interval. This assumes that the Oga -g agioa:e -%a e- l (1) Jacobs, I.M.. " Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards", General Electric Co. A.?.E.'). , April.

1969 (AFED 5736).

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I 3.5.A AM.I.li (cont' d. )

core spray subsystems and LPCI 3 constitute a 1 out of / system: however, the combined j

effect of the two systems to limit excessive clad temperatures must also be considered. The rest interval specified in Specification 4.5 is 1 month, chmttm Should one gore ypray subsystem become inoperable, the remaining ; ore ;praysand the 1.PCipyswa are available should the need for core cooling arise. To assure that the

d **

remaining gore 4 pray and LPCI subsystems are available, they are verified to be l ,

operable immediately. l Should the loss of one LPCI purnp occur, a nearly full complement of core oe*4-4 enc 44nment cooling e utpment is available. Three LPCI pumps in conjunction with the

,o re pray subsyste will pe rform the core cooling function. Because of the '

avail viich t will be verified l ility of thea majority to be operable, thirty dayofrepair the core period coolin6 equipment, If is justified. 74'he LPCI Jubsystem is not available, at least 1 LPCI pump must be available to fulfill the containment coolin5 function. The 7 day repair period is set on this basis, w %.v.+JeerQme.+

3. ' " ;ir ~ Oc!!n J ub " n. M 2u2. % vu.c,. W el $ p %t e l T' - -t! , : -ling-. A yrte !c '"$ ? con !:t ef c -re leep aoh "1.sh Ipm

'LPC*) pnp: : rim; :n: :H: -f ~h: """ '-t erget--end t u e " " : * :t e r--

? ::ter Purp: : r! g the othe: 21 e j-*dte upon the use of one RER Servics Water /ooster/ ump and one RHR heatThe design of the exchanger removal af ter a design basis accident. Thus, there are ample spares for margin above design conditions. Loss of margin should be avoided and the equipment maintained in a state of operation. So a 30 day out of service time is chosen f or this equipment.

If one loop is out of service; reactor operation is permissibic for seven days.

With components or subsystems out of service, overall core and containment cooling reliability is maintained by verifying the operability of the remaining cooling equipment. For routine out of serdce periods caused by preventive maintenance.

etc.. the operability of other systems and components will be verified as j;iven in the Technical Specifications . Ilowever. if a f ailure. design deficiency etc.. caused the out of service period, then a demonse rration of ope rabi.i ty mav oe neetted to assure that a similar proble n does not exist on the re ma t n i n r, componears, _ "n r e: ample, if an out of-service period were caused by fa!!ure of a pump to deliver rated capacity the other pumps of this type might be subjected to a capacity es t ,

The pump capacity sst is a comparison of measured pump performance parameters 1

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3.5.3 BASES (cont'd.) l co shop performance tests. Tests during normal operation vill be performed by measuring the flow and/or the pump discharge pressure. These parameters and its power requirement will be used to establish flow at that pressure.

i C. RPCI 19<t4 cas The limiting conditions for operating the HPCI System are derived from the Station Nuclear Safety Operational Analysis (Appendix G) and a detailed func-tional analysis of the HPCI System (Section VI.).

SpMx The HPCIT is provided to assure that the reac:or core is sdequately cooled to limit fuel clad temperature in the event of a small break in the nuclest system and loss-of-coolant which does not result in rapid depressuri:ation of the

, reactor vessel. The HPCI5rpermits the reactor to be shut down while main-

'TM#$I~~ tsining suf ficient reactor vessel water level inventory until the vessel is ,

depressurized. The HPC;5tcontinues to operate until reactor vessel pressure is below the pressure at which LPCI operation or Core Spray System operation main: sins core cooling.

The capacity of the system is selected to provide this required core cooling.

The HPC pump is designed :o pu=p 4:50 gym at reactor pressures be:vvec 1100 and 150 psig. Two sources of water are available. Initially, desinertlined water fr m the emergency condensate storage tank is used instead of injecting vater from the suppression pool into the reactor.

When the EPCI System begins operation, the reactor depressurizes more rapidly than vould occur if HPCI was not initiated due :o the condenaation of steam by the cold fluid pumped into the reactor vessel by the SPC! Systam. As the reac-

or vessel pressure continues :o decrease, tne RPCI dlow momentarily reaches equilibrium vi:h :he flov :hrough the break. Continued depressuri: anion causes

he break flov :o dectesse belov the HPC: flow and the liquid inventory begins

o rise. This :yee of response is typical of the small breaks. The este never uncovers and is continuously cooled throughout :he transient so that no core damage of any kind oc:urs for breaks : hat lie vi:hin the capacity rante of :he HPC;q Syster.

The analysis in the TSAR. Appendin G. shows that the ADS provides a single failure oroof ca:5 for denrassuri:ation for postulated transients and accidents.

l6 ex,y H The 'tC;Caserves as an al:ernate to :ne R@cniv f or decav heat removal when feed vacar is los:. Considering the 3PCitand the ADS plu's)RCICf'is'#Eedundant paths. ref erence (1) e: hods vould give an astimated allowable repair time of 13 days based on :he :ne conth :ss:ing frequency. Mcvever,smaximus,23'gvable rapair :ime of 7 days is selsc:ed f or conservatism. The RPCI and ICIC'as dell as all other Core Standby Cooling 3ystams :ust be operable when star:ing up from a ;old Condition. :: is reali:ed that :heHPC:$?I#50t designed .1 operate until reac:or pressurs anceeds if 0 psig and ta aut:ma:ically isolated before l

the emactor pressure decreases belev 100 psig. :: is the intent of :nis spect-1 l

a * $

I. . . . _ _ . _, . , . , - . . , . , . _ _.

. - -_. .. . - - - - - - - . - - - -.. - _~_-_ - - - _-

L5,C 3ASES (cont'd.)

i fication tg assure that when the reactor is being started up from a Cold Condition, the HPCthf("n~ot known to be inoperable.

A D. [tCIC System S

The RCIC;eteviis designed to provide makeup to the nuclear system as part of the planned operation for periods when the main condensey is unavailable. The nuclear safety analysis, FSAR Appendix G, shows thaQCIC[r3vides watert to cool the fuel when feed l water is lost. In all other postulated accidents and transiente, the ADS provides >

redundancyfortheHPCff a' sed on this and judgements on the reliability of the llPCI Imtnediate verifications of jystemgn allowable re[ pair g]iggf 7 days is specified l

HPCI78perabt11tyduringRCICoutage is considered adequate based nn judgament and practicality, E. Automatie Deoressurizajien System / ADS)

The limiting conditions for operating the ADS are derived from the Station Nuclear [

Operations 1 Analysis (Appendix G) and a detailed functional analysis of the ADS (Section ~11.).

i Tn, e specification ensures the operability of the ADS under all conditions for which the automatic or manual depressuritation of the nuclear system is an essential response to station abnormalities.

The nuclear system pressure relief system provides automatic nuclear - system depressurization for small breaks in the nuclear system so that the '=; ne -

, rint !nject' (1.PCI( and 4+ gora gpray ma*e,ystemn can oparate to protect the fuel barrier.

Because the Automatic Depressuritation System does not provide makeup to 'the reactor ,

primary vessel, no credit is taken for the steam cooling of the core caused by the ;

system actuation te provide further conservatism to the CSCS, Performance analysis of the Automatic Depressuri:ation System i considered oq1v with respect to in depressurizing effect in conjunction with%sCI or Core SpraffI$ere are sir, valves provided and each has a capacity of 300,000 lb/hr at a set pressure of 1080 psig.

The s '.lowable out of service time for one ADS valve is determined as sevan davs ~

because of the redundancy and because thellPCdILY~Iarified to be ocarable during l this period, Therefore, redundant protection for the core with a small breah in the nuclear system is still available.

D e ADS test circut': parmits continued aurveillanco on the operacia rn'ie f valves u assure that they will be available if required.

7. :11ntmm ' ow Pras mre r'oolly ,nd 7 t a ni 'bne r m e willabili-"

The pur ose of Specification ? is to. assure that adequate core no t ing +1ui p-i JJ_}' 4

e 3.$ BASES (cont'd) ment is available at all times. It is during refueling outages that major maintenance is performed and during such time that all icv pressure core cooling systems may be out of service. Specification 3.5.F.4 provides that sheuld this occur, no work vill be performed on the primary system which could lead to draining the vessel. This work would include work on certain control rod drive components and recirculation system. Thus, the specifica-tion precludes the events which could require core cooling. Specification 3.5.F.5 recognizes that, concurrent with control rod drive maintenance during the refueling outage, it may be necessary to drain the suppression chamber for maintenance or for the inspection required by Specification 4.7.A.2.h. In this case, if excassive control rod housing leakage occurred, three levels of protection against loss of core cooling would exist. First, a special flange would be used to stop the leak. Second, sufficient inven-cory of water is maintained to provide, under wordt case leak conditions, approximately 60 ainutes of core cooling while attempts to secure the leak are made. This inventory includes water in the reactor well, spent fuel pool, and condensate storage tank. If a leak should occur, manually operated valves in the condensate transfer system can be opened to supply either the 4cre gpray gystem or the spent fuel pool. Third, sufficient inventory or water is maintained to permit the water which has drained f rom the vessel to fill the torus to a level above the ore gpray and LPCI suction strainers. These systems could then recycle the water to the vessel. Since the system cannot be pressurized during refueling, the potential need for core floodigg,only exists and the specified combination of the gore gpray ortheLPCI,systeqfeanprovidethis. This specification also provides for the highly unlikely case that both diesel generators are found to be inoper-able. The reduction of-rated power to 25* v11,1 provide a very stable operating condition. The allowable repair time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months vill provide an opportunity to repair the diesel and thereby prevent the necessity of taking the plant down through the less stable shutdown condition. If the necessary repairs cannot be made in the allowed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the plant vill be shutdown in an orderly fashion. This will be accomplished while :he evo of f-site sourceJ of power required by Spect! cation 3.9. A.1 are available.

G. Maintenance of Filled Discharte Pion aptems If :he discharge piping of the jora 4 pray, L?C; ewksy*++m. u?CI, and RC;C,are not filled, a water hammar can develop in this pipinC vhen the pump and/or pumps are started. If a water ha==er vers :o occur at :he time at which the system were required, the system would Jtill perf orm its design functions.

However, to minimi:a damage to the discharge piping and to ensure added mar-gin in the operation of these syscams, this Tachnical Specti; cation requires the discharge lines :o be fil'ad. whenever :ne sys:am is :n 2n aperaals conci-

1on.

3. Engineered Safeguards Campartmen:a Cooling The unit :colar :n each pump ::spartment is :apaole of providing : equata ven-

11ation f1:V 2nd :coling. Ingineering analysas indica:s : hat :he :ampe:2:urs rise in aafeguards compar: men:3 vichcut adequata venti ation flov or 'coling ta such :nat ::n:inued operati:n :f :he Ja!*;uarda acuipment or asacciatad aun:11ary equipmen: tannot be assured.

- C 3- :s,09=13

4.5 DASES I'

gore and Containment'Cooline Svslems Surveillanea Freauencies s or c.

The testing interval gfor the cora and containment cooling systems +r based on industry practice, quantitative reliability analysis, judgement and practicailty.

The core. cooling systems have not been designed to be fully testable during operation. For example, in the case of the 11PCI/*aYe*o'matic initiation during power operation would result in pumping cold water into the reactor vessel, which is not desirable. Complete ADS testing during power operation causes an undesirable loss of coolant inventory. To increase the availability of the core and containment cool!ng systems. - the components which make up the system; i.e., instrumentation, pumps, valves, etc. , are tested frequently. The pumps and motor operated injection vaiva are also tested each month to sssure their operabillty. A simulated automatic -

ac tua .. '.on test once each cycle combined with f requent tests . of the pumps and injection valves is deemed to be adequate testing of these systems.

When components and subsystems are out of service, overall core and containment cooling reliability is maintained by verifying the operability of the remaining equipment. yor routine out of service periods caused by preventative maintenance, etc. ,' the operability of other systems and components will be verified as given in the Technical Specifications. !!owever if a failure or design deficiency caused by

-outa6e, then a demonstration of operability may be needed to assure that a generic problem does not exist. For example, if an out of service period were caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of :his type might be subjected to a flow rate test.

+

j.

l l-l

-i l

[

" I 4.,.

L11RTING CONDITION FOR OPERATION Elf,lLMNCi? PJ.%!11tL'iWT l 3.7.A (cont'd,) fe.7:A (cout'd.)

6. Low;.Lpy,Jg. Lie ngf Function 6. ktF. kO3._2sLEtLief function

n. The low.lew set function of the a. Tbc love low set safety / relief valves safety rettef valves shall be wha!1 he casted and e.alibrated as operable when there 11 trrad(ated speelfled in Table 4.2.n.

fuel In the rent t.or vessel and the reactor coolant temperature it f

/l?'F, except as specified in 3./.A.fi.a.1 and 2 belcw.

1. With the law. low function of one salety/ relief valve (S/RV) lunperabic, restnre the insperable Lt.S S/RV to OPEAABLE within 14 days or be in the llCT STANDBY mode within .

the next 12 hones and in COLD SittlTDOWN wichin the following 24 hones. ,

f

?. Wis h the Inw inw set function of both S/RVs inoperanie, he in at I n n g t . Il0T STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months !

nod in Col.D SHt1TDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. The prassura switches which contral the low low set safety /ralief valves aball have the following setting 4, NRT.PS 11A Open Low % lva 10l3 ! 20 psig (Increasing; ,

NDI-PS 51B Clona Low Valve 4 75 t 20 psig (Decreasing)

MP, f PS. W Open 111 gh Valve 1075 t 20 psig finereasing)

!Pf-PS-31D Close High Valve B, ndbv Gas Tre m ent G vs T3 875 t 20 pgig (Decreasing). 1. At toast ance pe r ope ra t i ng, " v e. l e the following ennditions shall be

,. ;;typelhv das Traatment 9vstam ,1e rno n s t rei t e d .

1. Encept _s a spec t fled in 3.i 3.1 -

a. Prassure drop acrun the comninad Selow. horb 4tandby gst greatment' E M d M nystams shalt be our raul. at att . , h4 M W mg.% ",. ,

'j i me s when secondary containment , g g( g i.ntegricv is rarptired.

2.a Tha ranults of the in place cold DOP 5. Inlar heater inone L1 " a pn h t 'a of ,

!cak ruts in the lle'.PA fil ters shall rn<b m ine, 1.11. from '90 -. o O t ' H=

shnw -190% 00P removal . The resul ;a o f " h e- $nt,nanatad hyrir'-arbon la2k ,

2,3. Tho sta and ,amnie ,nnimi, e t~r, on *hn char mH adsorharn !

occ nc.u ton 4 .- 3.4 y,, t 3,,

ha la r,ena t =,1 narf3rmed u bu- 3nce go.r"

s ha t 1 show So% l

ridrocar5nn ramovnt.: 'he 70P tnd { ;1 months for a ndhv ;cruica ar i

,g r., .q t L 1 : m i' ri n n ( < h yrir me n ; S,_ .) +qt3 3hg(( ,,, .,

em m *o it 1 ganen, ,; u ; ,,g y t on ,ny ;g ,, y,, g,g , , y n c m ,,

y % ,7nr .;.w

. ,m 3 q :ma p 3n,;

. . , i.,,, . , , ,

'" o ni a ; -; s.c n it , in e '

' l a, -

cnim : n ! : .I ' <t2 J'

,)) *

}, 'l

l

+ i 1.1MITtNG COND1' LION FOR OPERATIOp. SURVEIU AMCE REOUIREMENT 3.7.5 (cont'd) 4.7.8 (cont'd)

' b. The results of laboratory carbon b. Cold DOP testing shall be performed sample analysis shall show i99% after each complete nr partial radioactive methyl iodide removal teplacement of the 11 EPA filter bank with inlet conditions of: velocity or af ter any structural nutotenance 227 FPM. 21.75 mg/m 3 inlet mothyl on the system housing.

iodide concentration, 270% R.ll, and l CO'C. c. Halogenated hydrocarbon testin6 ,

shall be performcet after each l

c. Each -fan shall be shown to provhte complete or partial replacement of 1790 CMT t10%. the charcoal adsorber bank or af:er any structural maint enanen nn the

, From and af ter the ws dare that one system boutinr,.

gtandby qns greatment., system is *ndo g nr found to be inoperable for any d. F.ach system shall be nperated with reason. reactor operation is the beaters on at least 10 honr:

permissible only during the overy month, succeeding seven days unless such 6b sys tem is sooner made operable. e. Tant saalitig of gaskets for boo *ine, provided that during sneh seven dav, doors downst. ream of t ha ItEPA f I f f ar a W an active componentM -f ^ C> and charcent adsorbers nhall he o Uct.1 gtandby ps rectmentNa fystem, and perlormed at, and Ln ennformanca

,p with, aach test perrnrmed for gg/ its associated diesel generannr.

shall be operable. compliance with spectricarfnn o p ,co ge, and Specific,cion 4.7.3.?.a Fuel handling r eeluireme n t s ara 3.7.3.2.a.

,pecified in Specification 3.10 F..

Svtitem drains where prenant ni..it ~ 'm

i. If these condi tions cannot be trie r inspected quartarly for adeqw =

procndures shall be initiatad water level in loop seals immediately O establiqh- raneror

. onditions for whtch the 4t..andhv ys

.. a . Ar least once per aporaiin.r. 4

e traatment tvat am is not ,pt i rmt vtrnmatic intrIntinn ni +,cb' J

~ '

tandhV j fl 9 yrffAt$f*14- '

cha i l h{o demons? vat 7d, ~

- s a w onco yr ,pe m ron -.

manoni operabilit' of *" w h~u a

n 1 v r* [nr filinr .noling nita l . %a demonstrated.

i

'..'h o n ono N-tandbv &m E r'.a" m .

Gwc 1 ? ** C e m 1*? c ott1m 9 1nOU*ar1hI q ' . ! n'c...

, pet.* O.n.e?'*"

r ? a r *

n h,I - 'm l LatdhV {11 .

.n* rr;;ieu 'm .e ap~ .

[

! . tmmd i a t e ly and *nily w re n # '

y\ dPUlonS i .' S C [ u n 1[ i [ A T /' ( l'

fli' l

'i r P r J D [ [ "'? !. S *10 C ' A (!' l j I h1

spor w e ,n N

I ,_

k _N

[ 'I - ('i l,F .if ^%Il M IIl } f $l' n() hlh4% il f si ii . o v ..v .

hal' ho n.i arainmi -nir n- un e

,i i an we ,o -ni a-n s (4A I** t t ..s 7 qute;t qt, ra 0 44 4

, . _ _ , _ _ . , _ - , . _ . _ - _.-..e-__m,. ._,m._,-,. . _ . , . _.. , _ _ , - . . _ . .._ , _

LIMITING C0NDITIONS F0R OPERATION SURVEIll.ANCE t!OUIREMENTS 3.7.C (cont'd.) 4. 7.C (cont ' d. )

l

a. The reactor is suberitical and Speci- a. A preoperational secondary containment fication 3.3.A is met, capability test shall be conducted after isolating the reactor building and

b. The reactor water temperature is below placing eithtr gtandby gas (,reatment 212'F and the reactor coolant system sub system filter train in operation. Such is vented. tests shall demonstrate the capability to maintain 1/4 inch of water vacuum No activity is being performed which under calm vind (2<5<5 mph) conditions l c.

can reduce the shutdown margin belov vith a filter train flow rate of not that specified in Specification 3.3.A. more than 100% of building volume per day. (va vind speed)

d. Irradiated fuel is not being handled in the secondary containment, b. Additional t zes shall be performed during the first operating cycle under

e. If secondary containment integrity an adequate. number of different envir-cannot be maintained, restore oneental viad conditions to enable secondary containment integrity valid extrapelation of the test results.

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or;

c. Secondary containment capability to

a. 3e in at lesst Hot Shutdown maintain 1/4 inch of water vacuum within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and under calm vind (2<3<5 mph) conditions in cold shutdovn within the with a filter train flow rate of not following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , more than 100% of building volume per day, shall be demonstrated at

5. Suspend irradiated fuel handling each refueling catage prior to operations in the secondary con- refueling.

tainment and all core alterations After a secondary containment vio'a-and activities which could reduce d.

the shutdevn margin. The pro- tion is determined, the gtandby 4as visions of Specification 1.0.J sreatment system vill be operateu are not applicable. [h=ediately af ter the af f ected :enes are isolate 1 from the remainder of the secondary containment to confirm its ability to maintain the remainder of the secondary . containment at 1/4 inch of vacer negative pressure under calm vind conditions.

Primarn Containment Isolation 7alves D. prima rv Cont sinrent !aolation Valves D.

1. During reactar pcVer operating tendi- 1. The primary containment isolation

1:ns. al' isolation valves listad in valves surveillancs shall be perf:rmed Table 3.7.1 and all ins;rument line as follows:

flow check valves shall be operable

a. At least once por operating :ycle tre except as spectfied in 3.7.3.2.

opersole isolation vsives that ars power operat2d anc aut:mati: ally initiated shall te resesd for atmu*sesd aut:mati: in tiation 2nd tiosurs t_nes.

- d o-19':1,14

- - - - - - - - _ _ - _ - _ - - _ _ _ _ _ ^ ^ - ^ ~ - - ~ - __

1.7.A f, 4.7.A MSE c on t ' d)

The primary enntainment is normally s1 LRhtly pressuritad ditrinr, parinda of ranctor npe t at ism. Nittor,en used for inerting could leak out of the enotainment but air enoid not leak in tm incr*asa oxygen concent ration. Once the cont a licent is fillad w i t h n i t i nyp n to the rctiu t red enneentra t:lon, no moni t or i to, of oxytan c oncent r a t Inn in necennarv. Iloweve r , a l. least twice a week the nogen rmu ent e nt inn will !c determined as added assurance.

Th, 500 gn11on conservative. Ilmit on the nit rnp.an S rnrae* rank a= =o rc = t ha t ade<waThe re time is svallable to get the tank refilled ananming normal ;) l a n t nperation.

ac r imared mnv.tmoen makeup rate la MOO SCFD which would re<vitre ahnur 160 r,ntions for a 10 dav makeup rerluirement. The normal leak rat e rhould he ahniu /no W o.

1 /.A 6 & S./.A.6 Mt M QWL R LtLg g tj g fJ M The inw-inw set ralle! logic is an automatic sn f a t.y ra t in[ valvo ( ':RV ) concent memicm designed to in t ti gs t e the postulated thrust inad ennearn of sobanquent ariunrinns n f SRV':: during certain transienra (such an i n.ulve r t a n t MSiV "lnaura) .

and mall and intermediate break loan nf coolant veldent (l #A) evaru s , The }

j s a t.pn i n t s used in Sec tion 3. 7. A,6,b are bnsed upnn a minimum h i nw inun rance en l

prov'de ide<1na t e tima herween valve actuarinns to a l l ow the SRV d i wha r r.n line hi rh wat er leg i n lonr , coupled wi th cnnside rat lnn n[ ins'..n e nt s nar u .ie

unt:hn -nain l

< team tsolation valve isolation setpotnt. l The u fnund actpnint for %I PS 51 A, the prenture w i t c h e no r r n l l i n e, +hr nponing o f RV !! D. inns t be 5 1040 psig. The as - Enund c ln* I n /, se r pn i nt fn- Nit t - Ps < l ti nu e t be at l e a n t. "O psig less than SIA, and munt ha ? 350 psi.. The n foo id f wienint mort he for NB1 PS 31C. pres ure switch control l i ne, the opening of R"-llF C 1050 psle,. The as inund clostnr, setootnt for kflt - rs . T I D mon t ha at least on' peig

.uul must he a 350 pqtg. Th{s en9uras " hat the 1na l W ical opp.r n(h below slo <he m'ne for the npm ing ,*rpoint (1050 psig), the analytical lower 1imi- no p-i-or u t poi nt ( W p.lg) and the analytical limit on the hinwdown ranrf a in tha 1.nw Low let Relia f Function are not exceaded. Al thour,h the vari f f. d (no r i manc

,e golut Enlarance is t 20 psig, an instrutnent drift of + 0 ps i g wat m. e d in me

,nalcnis to encore adequata ma qin in de'armining the

alto opcoing ,nd i i nt ' ' it.3 se t eninu . The opening setpoint is set such that, i f Sot h : he inwent set non-:,sa 3, RV and the highest set of the two LLS S/RVs drif- ' S p a t .; in the mu t ilm r -e di rac r inw . 'be LLS ',,975 Jill stili conrrni nuhsequent ( t"/ a r

na t i n t o, et to ensure the U.3 3,RV :Ln. i nc wi poi ut r.%a i e i n .r the - insinn w pnint i3 in;. n The 30 pm i g hi nwdown p rov i du avic p oit u mv. s the '19 f V inw pronsura erip in cle ir 141 a+o tili e p en W f tun the w ni in one:nre : i me for . he wat ei l ev a

- ] (' I I f Il l k nl18%

. i b ). 8I )h bb b N ho hj OJ '( f*

,o - o,a , i.n i , si - - ,<

The -ernna ,r- .nne,inmen (n ae igned m minimi.~.

maiorials which ujaht res il t mn .i et ut v' 'h- 'u- i- it c ru i j o n, . v..

,p qjg . ,p. ,

qf -g gyi r g( ggg ,g9py, j ) I' f) I h f #{ 3 P c n f)d;") f '[ c f)Q f j k h@p '$ f , ht l g' k [}

}

A p t-h!I i ! h 8 ) )'

The raactor holl ding prnut un pr* mar- n o r : i i .. .c s- .i n i- ca!ad atui tn snrvic?. a.i, '. , c, o, ;

nac*nr in thut down and *he d r n., o i ! in noen. ar g e n s ie i ; o r.

he - - - - g w e nnM r v murlinment is an i nt a g t.11 par ~ of the c w p l e t ,* nm ainnin r w u>

'me n t w nnd , r ' nainment is raquired at all -i,nes

hat o r i wa r inno m.o o, u ac!; e at o i n c. ra fue l i ng. S e c nnibi r- annrainment na . Se armon .i in ' ' 1 t'* 1 n90' on 'otil il i ?te ' emf * * 'H** 6, n ,

s !' ime :S 1I l itw i t- rs t ft A *' O '#e

'*

Y * - : h ) de # er I O

14* j %8 h 4 4 t i tSHl 'l)) ; f j ( i' fid [lf ' f1

' e e nu e M r. m e s' a= w . s'.

3 + - -

srbin.*n's~

a

~

ir mi .

m , s

%e. ,-1E'annho E' u w"5 marment 'Cvstwv) s ie s ; e,neo

o. tr oli -7

,'His i nq limnstnera ;n *he iticg du r '. n q iM mndar' '7n r i yen t me,ma- - ,,

11

'n o , -

3 irment ~c s m Dn; ,r= tes;raq 5 4 rannhv *!$ .n T,. A :r' "' '

mi1 a nne n r , ;ai1;.*n itH1 113!'"1;" " " *

1r*'t i' '"4 "C'

a i i ;) -

',' '41y1

as 60i6 io D. -e 't y a i,ei r * .b

p $t .a e a=i? A t

> on, ,, W'* - .m ,. u m, a

,e w g; '*g 9g > " *

= - - -- - - - - - .-, ___. _ , __

3.7.5 & 3.7.C BASES (cont'd) 4 High effielency particulate absolute (HEPA) filters are installed before and I after tne charcoal adsorbers minimize potential release of particulates to i the environment and to prevent clogging of the iodine adsorbers. The charcoal !

adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and HEPA filters. The laboratory carbon sample test results should indicate a radio- ;

active methyl iodide removal efficiency of at least 99 percent fot expected '

accident conditions. If the performance of the HEPA filters and charcoal {

adsorbers are as specified. the resulting doses vill be less than the 10 CFR 100 guidelines for the accidents analyzed. {

Only one of the two standby gas greatmeni~Iystems is needed to cleanup the reactor building atmosphere upon containment isolation.

found to be inoperable, there is no immediate threat tothecon[tainmentIf one s} s system performance and reactor operation or refueling operacian may continue while repairs are being made. If wit.hes-sp t & !? T - M +,4the plant it brought to a condition where the standby gas gaatment jystem ia not requircu.

l. 7.3 i 4.7.C BASES %th tub *y :A.% cu t, mormro9 C. ,

Standby Cas Treatment System and Secondarv Containment Initiating reactor building isolation and operation of the 4candby pas j;reatmer t gystem to maintain at least a 1/4 inch of Vater vacuum within the seconfary containment provides an adequate test of the operation of the reactor building isolation valves. leak tightness of the reactor building and perfor nance of the Functionally casting the initic. ting sensors and gtandbygasgreatmentgvstem.acsociated tr.; channe.s demonstrates the capability for automatic Per3rming these tests prior to refueling vill demonstrats secondary contain'.ent capability prior to the time the primary containment is opened for refueling.

Periodic testing gives suf ficient confidence of reactor building integrity cad Atandby yas g oatment s ystem performance capability.

Pressure r. rep across the combined HE?A filters and char:oal adsorbers of less than 6 inches of water at the system design flow race vill indicate that the filters and adsorbers are not clogged by excessive amounts of forsign metter.

A 7.d kV heater is capable of maintaining relative humidity belou 70';. Nester capacity and pressure drop should be determined at least once per operating cycle to show s:*atem perf ormance capcbility.

The frequency of tests and sample analysis are necessary to show that the NE?A filters and char: sal adscreers can perf orm as evaluated. Testa af :he cnarcoal adsorbers with halogenated hydrocaroon refrigerant iha n be perfaced in sc:or-dance with A"n N510-1930. The ::st :snnisters that sre insta11ec vit5 theIach adsorber trays should be used for the charcoal adsorber ef ficiency use.

sample saould be at '.e a s t two inches in 11ameter and a length squal to the thicAness of :he bed. !! test resulta ari unacceptaole , all adsorbent in ene afstas snail be replaced m g 3,

. 4 4.7.15 & 4.7.C BASES !

with an adsorbent qualified according to. Table 5.1 of ANSI N$09 1930. The l replacement tray for the adsorber tray removed for the test should meet the same

- adsorbene quality. Tests of the llF.PA filters with DOP acrosol shall be performed in 1

accordance to ANSI N510 1980. Any filters found defective shall be replaced with filters qualified pursuant to Regulatory Position C.3.d. of Rer,uintory Guide 1.52, Revision 2, March, 1979.

All elements of the acater should be demonstrated to be functional and operable during the test of heater capacity. Operation of the heatern will prevent moisture buildup in the filters and adsorber system.

With doors closed and fat. In operation, 00P aarnmot shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the f an exhaust shall be considered on unocceptable test result and the Saskets repaired and test repeated.

If system drains are present in the filter /adsorber banks, loop seals most he _used l with adequate water level to prevent by. pass leakage from the banks. l If significant painting, fire or chemical release occurs such that the !!F.PA filter ,

or charcoal adsorber could become. contaminated from the fumee. chemicals nr foretr,n material, the anme tests and mample analvuls shall be porrnrined an re<pi t red One '

operational ute. The determination nf signi Clennce shall be made hv t he operator int

-duty at the time of the incident. Knowledgeable staf f inembers should be consulted prior to making this determination.

Demonstration of the automatic initiation capability and nperability of filter '

cooling is necessary to assure system performance ccpability. If one _standhv ns

reatment#}stemis-inoperable.

s the ar.bm:p m' opertaility isverifieddatYy. !

fhis substantiates the availability of the operable $ stem and thus reactor operation or refueling operation can continue for a limited period of timo. ,

3. L D & h .1. 0 BASYS orcrods Wosn*.edb l Frimarv Contalemen,e isolation Valvas ,

Douole isolation valves ara provided on I!nes pencerating the primary containment and j npen to the f ree space of the containment. Closure of one of the valven in each !!nn ;

oouid Se suf fie leat to maintain the integrity of the prassure suppression r.vst.ct

Automatic initiation is required t2 mininl:o the potential leak.v.e pr tiu, from tho cmotainment in the event of a lonn of conlant accident, The maximum closure times for the automatic isolation- valves of thn primary containment and raactor vessel isolation control system have been scl acted in !

consideration of the design tntent in pre"ont core uncovering fnllowing pipe braus nutside the primary containment -and the need to contain relanned finnion produc w fnilowing pipe breaks inside the primar e nnta i tunen t .

These valves are highly re!!able, hm a a w se rv ice requirament. and wm ca normally. closed. The in i t ia ti n r, ,c o m e , wd a r.::ne ta r s d r:p channe!r ,rc ii m t checked to domonstrate the capaht ;t: tm ,ut mnatic isolation, The test inro -al 4 once par opetating '.ycle for autamarle ini stion s

'~

n4 ~ r 8 s

4.-

-9y -

w-9 9.%e-e.--,,.mi---9 -%,-9.--9w---- y,g-. vp-+y-p7-.,4-w. 9--gig, e m, g.w op,(- pm4.w w9 t._ w-9 m-g Ap,,p.c ,, paw g-m4w., - g.e4----e y m+ mm wome % + +-sp,p - P-' -. -+

St'RVE ILIMClLRE0tflR D1FNTS

_ LIMITING CONDITIONS FOR Of.EP.ATICf1 3,10.B (Cont'd) 4.10 (Cont'd)

[ 4. . Durin6 spiral reload, SRM

\

operability will be verified by using a portable externsi source every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months unti1 the requlted amount of fuel is londed to maintain 3 eps. As an alternative to the above, two fuel assemblies vill be loaded in dif ferent calls containing enntrol blades around each SRM to obtain the required 3 eps. Until these two assemblies have been leaded, the 3 cps requirement is not neeeasary. C. 21rmlJ ,el Poni Vn(fr_f.cyrd C. Seent Fuel Pani Vater Level Vhene.ver irrt.diated fuel is stored When irradiated fuel in stnred in in the spent fuel pool, the pool the spent fuel pool, the water level shall be recorded daily, water level shall be maintained at or above Sti ' above the top of the (nel.

D. I.'a Listit a t lon Irradiated fuel shall not be handled in or above the reactor prior to Fi hnors after reactor shutdown. Standbv Cm T r c,t.1,ma n t System F. .

C. Standbv Cas Treat rent Svstem From and - af ter the date that one

stem la made When one et.tandb a y an mrentmere

- tandby *r E , b sys tem bec oine u ; nope ra ilo , th<Dc.c..c or founu Ms *reatment[ble 1 %

to un inopera for any ^

reason, fuel handling is permissible ;tandby gan arsa tmenty/s tem sha l' be vertflod to ho nperabl* l only during the succeeding seven days unless such7ystem is sooner immediacciv and dativ thereaftar.

A deinons tra ti on o f d iosc l r,ana raca r m made operable, pinvided that during npe r-hi l i ty i n not verlutred hv thit Y

asre<A.

toch saven davs epenmd'l a active specification.

g cornpnnenrnyf the e gtandbv ps ireatmendgystem, and its assoc!ned diesal generator. shall be operable .

\t least one diesel generator shall Se operable during fael handling operations This one diesel shall be espable of . supplying power to an anerable Standbv Gas Treatment

- 5,,. %, e e .

ing J y<lby.EnnD % ramg

% r l iu*, a c e !' io ! L n e, ou t a ge , re fuel f tut : ;% ;g <4,- g 3,., e j g ,

ope r a t ton f inav contint e 1f gg g ,, g. g

.- ~ . . ~%

DMA k re, h F>y b43V8N N m .

j '

, _* l' - -

QQq

'-- ?.c ine : tnc, ta permt-t M mcm e w i t.h "he e.unpres sion eb imbe r 4tra i nen provided .m o pe r.m i e goro gi .

qr 1 ~

% '.pCl g - j t.up >9 A. tr j

'i t i i r.nad n se a mc nn in be 'm u lo n 4 d r a tt9fMq' ^ 11 y,

'on"1(Ming 1F [ e l ei f , ~ d . O f 71 l' nt i ' i nt:12 3 s*ad .$ " a r i

I" )4 *3 t

-- A

o 3.10 3ASES (Cont'd)

D. Time Limitation

(

The radiological consequences of a fuel handling accident are based upon the accident occurring at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor chutdown.

E. Standby Cas Treatment System u

Only one of the two 4tandby 4,as greatment systems is needed to clean ,up thereactorbuildingatmosphereuponcontaInmentisolation. If one , system ,

is found to be inoperable, there is no immediate threat to the containment ,

system performance and refueling operation may continue while repairs are l being mado. If neitk er ayeres i' sper ble,Athe plat.t is brought to a con-dicion where the grandby g,as greatment gysteilis not required.

F. gre Standbv Cooling Systems During refueling the system cannot be pressurized, so only the potential need for core floodintexists ar.d the specified combination of the p re jpray. or LPCI med%%!XC$r8 vide this. A more detailed discussion is con-tained in the bases for 3.$.F.

G. Control Room Air Treatment If the system is found to be inoperable, there is no immediate threat to the control room and refueling operation may continue for a lini:ed part d of time while repairs are betog made. If the sys:em cannot be repaired within seven days, refueling operations vill be tarminated.

4 H. Seent Fuel Cask Randling l The operation of the redundant crane in the Restrie:ed Mode during fuel cask handling operations assures tha: the cask remains within :he controlled area once it has been removed from 1:s transport vehicle (i.e., once 1: is above :he 931' elevation). Handling of the cask on the Refueling Floor in :he Unrestricted Mode ir allowed only in the case of equipment f ailures or emergency conditions when :hc cask is already suspended. The Unres::icted Mode of operation is allowed only to the extent necessary to get :he cask :o a suitable stationary position su :he required repairs can be made. Operation with a failed controlled area microswi::h vill be allowed for a SS-hour period providing an Operator is on the floor in addition to the crane operator :o assure

hat the cask handling is limited to the :entrolled area as marked on the floor. This vill allev adequa:e :ime to make repairs but still vill not rest:10: cask handling operations unduly.

4.10 J3jjl A. Refueling In:erlocks Complete functional testin6 of all esfueling intar*ocks befois any sfueling outage vill provide posi:1va indica:Lon :ha: :he in:arlocks L operata in the sitaations for whi:n : hey vere designed. 3y loading each hois: vi:h a veight equal :o the fusi assemoly post:i:ning :he refueling platf orm and vitacraving :en: 31 ::ds. the interloct,s can de subjec:ad :s valid operational :ss:3. Whera redundancy ia providad in

he logi: :ir:ui:ry. :ests :an be performec :: Isaurs :na: sach :scundan:

i logi: 41ement an independen:1y perf::= 1:s func:1:ns.

-IO9a-3' 3E 3 5

4.10 B AS ES_ (Tout'd)

3. Core Monitg Qng Requiring ths SRM's to be functionally tested prior to any core alteration

~ the SRM's vill be operable at the start of that alteration.

assures that The daily response check (or 12-hour eneck for spirai reload) of the I

SRM's ensures their contit.ued operability.

E. St.tndby Cas Treatment System s.h Only one of the two g,tendby gas ;?eatment;systees isIf needed one to clean up the reactor building atnesphere upon containment isolation. to the there is no immediate threat 532 system is found to be inoperable,containstme If neith:: cy ::: 1:

system performanc while repairs are being made.to a condition where the Leandby gas greatment is brought '

J required. ,botk wbyeJ'ms ore, acproWd, l H. Seent yuel Cask Handling the redundant crane The Surveillance Requirements specified assttre that d d is adequately inspected in accordance vi:h the artepted ANS1 Stan ar (3.30.0.0) and manufacturer's reecamendations to determine that theThe equipment is in satisfactory condition.the crans operation will be limited to area limit svitches assures that The test of the designated area in the Restricted Mode of operation.

the "t.o-block" limit svitch aseures the power to the hoistingcan motor occur.

vill be interrupted bef ore 4n actual "two-blocking" incidentthis mode of load contro The test of the inching hoist assures that is available when required.

Requiring the lifting and holding of the cask for 5 minutes during the of each series of cask handitng operations puts a load braking initial lift test on the entire crane liftent mechanism as well as the sysien.

Perf srming this cost when the cask 1J being lif ted initially cask car assures that to an excessive height.

i

], }J ^

wb"

. -- . - . _ . - - - - _ _ . - _ - . - _ ~ . -. . - _ . - . - - - _ _ _ _ _ - -- - .-

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 9

)

3.12 (cont' d) 4.12 (cont'd)

E t.at'Nat I- mbag (2.6C) 6m.n (,i'.s Q 6y we e ei e . t. t. .x. S. [teactor4 b'> t'"".14" P.ut 1 -mT Cdbjt ~ me- ~ ~ .e u. m

n. Beacteri

---.rp.

E !!." . m ~ m_- ,,_

~~- . - - -

shetm 1.

b e Bothg T r' . %q, c i c e e tr g ,peceet.(col y ce^11ag 1. REC System Test (ng item Frequenev "er leepr - and Lheir associated purnps shall be operable whenever irradiated fuel is in the vessel or s. Pump Operability once/ Month the spent fuel pool, except as b. Motor operated Onen/ Month specified in 3.12. 5. 2 and 3.12.B. 3 Valve Operability below, c. Pump flow rate Once/3 months i Each poinp r. hall and af ter pump nc i tve r l l M gpin ma l ut enanen at 6'; psid.

d. Systern head tar.k Daily level shall be l ht aMD cfemD'W'/ l ]G.G( 3doptem monftored.

i a (C. I- i thnr v

2. Prnm and ,a f ter the date that any 2 ~4 hen an.tt,, e r. 3o SEsv %.in.a coiuponent % in one W,T becomes activo componcos at.d.e.t.h.m..(n an ill ck

" y' inoperable continued reactor inoperable, a llh%?ni>one n% M.t r tsf e rtne A shali be verified

" operation is permissible during the

' '""' ' N succeeding thirt" days provided that operable immediately { and weekly

dur ng such thirty days all 4++- thereafter. I M 48(bnentsc f " - '

' 7 " the ope rWe. 7 E C.

[r active cornponents tof the engineered 9 4.m W:2m ge,n' ! 1a feguards c ornpa r tme n t cooling h a CCec,t system 3, the diesel generator O percM\Q-l assoctared vitb the ooe rable 4*a.

are operrbla. ' */

The allowabia repnir ti me does not appl," when the reactor is !n the shurdown mode and reactor pras.;ure

'is less than 75 psig.

4(. uw.ee <q ' n _a --nir.m 3, nacug . : . , ?

with one pump per W ',o* wee shall be oparanle as stated in 3 12.S.1 and 3.12.3.2 above during ranetor head-off operat!nns empt i r ! n* ' .*m i a - Core spray /mem l m/a l l ah i '. My or ervica ;p

~

'. a r coni!ng anali % ava il an te .

6, If *. h ** !* f'j u i rP men C 1 7 L' 3.12,3 5.

i arnogn LJ 3 2 canrnP ba met M

+-a,' p r ma i' 5e ;hutdown ." .tn O r e

1 t Alifie t* "I n ti in One ' .

8 ) il

lio r = hn o nd i. ; [ o n .r 1 :3jn 14 lim t r ; ;

,c w r ; : ena reoi riq ' ?C: or gre j l '

g. e - v ~ .m vni'ani;. ; .

nnni^ %

1- t

4 SitRVEttlANCE REOUIREMENTS

_ LIMITING CONDITIONS FOR OpEFATION 3.12 (cont' d) 4.12 (cont'd)

( C, Service Vater System C. Istvice Water System

1. Service Water Sys, tem Testing

1. Both gervice gater subsystems with both purnps in each subsystem shall c'unc t inna1 be operable whenever irradiated fuel item is in the vessel or spent fuel pool and prior to reactor startup except a. Pump Operability once/ Month as specified in 3.12.C.2 below.

b. Motor Operated Once/ Month Valve operability

c. Pump discharge Once/3 months

-t N t o CCe<.A.s head tests opernoJ hr oO When it is determined that any

2. From and laf ter the date that anv ar.M 2. required p rvice ps t ern component i i r. th [ p rvice ga t e'r later l inoperable. 4 the subsystem is made or found to be cornponent is inoperable for any reason, continued operable ;prvice g,ter a subsvntem reactor operation is permissible c ornponen ts shall be verified to be l during the succeeding thirty days operable immediately and weekiy provided that durin6 such thirty the re f ter ,

cd o* C-days all active c ornponents A n f the -

grvice pter subsystem and %crt ca iket' li Co '"P*TW -

o gro les h e associated diesel gene'rator are

~

o fd ros h 'y wt W ut ,

O W **' "

operable.

1. If the requirement of 3.12.C.1 and 3.12.C.2 cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D. 1,a r a r r 'tonm ?cnt!.inticu D. En.j;;;e r-r 't nn m V a nj;.t i , i n

1. 3attery roor ventilation shall be 1. The spara battery room ventilatt on operable on a continuous basis f an shall be checked for operabli

hanever specification 3,9.A is once< weak.

required to be satisfied.

2. From and after the dats that either of the two battery room ven fans 13 macie or found to be inoperable for any reason, continued reactor operation is permissible during .5e succeedin6 7 da?"-

3. If :ha raquirements af 3 12,0.i s 2

) :annot be met. an oc<ie r'" shutdnun nf the raactor shall be iniciarH 2nd the reactar sba;; be in N.-

l t :hutdown within :4 hours.

i

- 12' .

l l

w --

3.1 3ASES Main Control Room Ventilation System A.

s The control room ventilation system is designed to filter the control roem atmosphere for intake air and/or for recirculation during eencrol room isolation conditions. The system is designed to automatically start upon control room isolation and to maintain the control room pressure to the design posi:ive pressure so that all leakage should be out leakage.

High efficiency particulate absolute (REPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential intake of radioiodice to the control room. The in-place :est results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and HEPA fil:ers. The laboratory carbon sample :est results should indicate l a radioac:1ve methyl iodide removal ef ficiency of at least 99 percent for expected accident conditions. If the performance of the HEPA filters and charcoal adsorbers are as specified, the resulting doses vill be less than

he allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power Plants. Appendix A to 10 CFR Part 50.

If :he sys:em is found to be inoperable, there is no i= mediate threat to the control room and reactor operation =ay continue for a limited period of tLee while repairs are being made. If the system cannot be repaired vichin seven days, the reactor is shutdown and brought to cold shutdovn within 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months .

G t.') c w ,y s u, a C % o, d , w r,c.*.

3. Reacror %Pmes%

3d idic: C'.eced Ceo%lC Min "ner- Svs t e- m ., g e. m , w s _,3r.m nin.n

he swetoe r : 2%qmentLeotm r Ni f ~; r ' m f y^%' y.,%:.m

u . 9 r mr.4 ._ hut tvo pu=p s and one heat exchanger,in: :ch f to: 12 c r+ . sch 4+e@/*ia'p"able of supplying

he cooling requirements of the essential services following de:ign accident conditions vi:h only one pu=p in either 44+o, subsyute n .

EG4 Theu avs:em has addi:ional m

.f.,lexibili:r provided by :he capabili:y of inter-connoccion of :he :0o,m,++me3 and the backup water supply to the :ritical crobneI loop by :he tarvice gater gvstem. This flexibility and the need for only Cr UD CAI one pu=o in one41 cop :o =ee: he design accident requirements justifies 0* L '*3

he 30 day repair :ime during nor:a! operation and the reduced requirementa during head-of f operations requiring :he availability c:*'I?CI s or 4he ; ore spray sys: ass.

C. Se: rtce Va:ar 3vstam b M *~ ot" wt h m '^ te N h o'!I i ,~ c vertical serrt e water pumps

he {ervi:s a:arL lystem :onsista off fewts located in the in::aa struc:trs, and associa:ad s::siners, ;1 ping, valving and instruments: ion. The pu=ps discharge :o a co= en header from which independen: piping aupplies :vo Se a=1: Class : :coling va:er 1: ops and ens I

turbine build ng loop. Aut::sti: valvin; is provided o shucoff all supp;7 l

s the turbine building loop :n dr:p in !.eader prassurs :hus assuring suppl:r to the 3eissi: Class ! Loops each of vai:h feeds one diesel genera:or. :vo IER

<ervice ~va:ar boostar pu=es, na :ent;oi ::c: basement fan :cil uni: anc one '2~~

l

" ~' 2g l

I

' -::Ic- ;c 3 34

I 3.12 BASES (cont'd) . - --.. \

heat exchanger. Valves are included in the comcon discharge header to

+ permit the Seismic C ass I larvice water Lvstem to be operated as two

' irdependent le W 4 Ti heat exchange *rs are valved such that they can be individually backvashed without interrupting system operation. ,

I During nor:nal operation two or three pumps vill be required. Three pumps are used for a normal shutdown. .

The loss of all a c power will trip all operating Aervice gater pumps.

The automatic emergency diesel generator start system and emergency equipment starting sequence vill then start ene selected gervice gater pump in 30 40 seconds. In the meantime, the drop in gervice ga, ter header pressure vill close the turbine building cooling wear isolation valve guaranteeing supply to the reactor building, the control room basement, and the diesel generators from the one 4ervice gater pump.

Due to the redundance of pumps and the requirement of only one to meet the acciden requirements , the 30 day repair time is justified.

D. Batterv Room 7entila dpm The temperature riss and hydrogen buildup in the battery rooms without adequate ventilat'.on is such that continuous safe operation of equipment in these rooms cannot be assured.

4.12 BASES A. Main Control Room Ventilstion Svstem Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate vill indicate that the filters and adsorbers are not clogged by excessive amounts of foreign ma: er. Pressure drop should be determined at least once per operating eye _e to show system perfocmance capability.

Tests of the charcoal adsorbers with halogenated hydrocarbon rafrigeran:

should be perfornied in accordance with ANSI N510 1980.

The frequency of tes:s and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. The :es:

canistars : hat are installed with the adsorber trays should be used for the char:oal adsorber efficiency test. Each sample should be at leas: cvo inches in dt.:unetar and a leng:n equal :o the thickness o f the bed. If

est results ars unneceptable, all adsorben: in the sys:am shall be replacad with an acsorbene qualified accordin6 to Table 5.1 of ANSI N509-1980. The replacemen: tray for the absorber ::ay removed for the tes:

should mee: :he same adsorben quality. Tests of the HE?A fil:ars vien DOP 2erosol shall be performed in accordance to ANSI N5101230. Any ME?A filters found defec ive shall oe replaced vi:n fil:s:s qua'ified pursuan: .

to Regulatory Posi:Lon .3.d of Replatory hide 1. 30.

Operation of the sys:sm for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months everv month vill demonstra:*

operabili:y of :he fil:a:s and 2dsorbar sys:3m and remove en:2ssiv$

mois:ure 'auil: up m the ads o r:.a r .

t ._1:4- 22 22 .

_ _ ~ .

r

. i s

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS ,

l 3.22 SPECIAL TESTS / EXCEPTIONS (CONT'D) 4.22 SPECIAL TESTS / EXCEPTIONS (CONT'D)

2. Rod Sequence Control System (RSCS) 2. When the constraints imposed on ;

control rod groups by the RSCS The sequence constraints imposed are bypassed. verify: '

on control rod ~ groups by the RSCS may be suspended by means of the 4. That the Rb'M is OPERA 3LE. ;

individual red position bypass l switches er jumpers. provided b. Conformance with this speci-that the rod vorth minimiser is fication and procedures by 4 OPERA 3LE. for this and the second licensed operator or , ,

following special tests. other qualified employee.

a. Control rod scram timing.

b. Control rod friction measure-ments. ;

c. Startup test program with thermal power less than 20:

of rated thermal power.

If the atoye requirement is not satisfied. the RSCS shall be operable.

'3. RHR System ,

The RRR system may be aligned in the shutdown cooling mode with at least one sautdown cooling mode loop CPERA3L2 while performing the Shutdown Margin Demonstration.

4 Contatament 3ysr. ems Primary containment is not re- t quired vn11e performin6 the Shutdown Margin Demonstration vnen rer.ctor water temperature is equal to or less than 212*?.

3. Training Startup

3. Traininz 5tartus

1. L?C -WC ;f ?S? - sygu The tsactor vessel shall be verified to be unpressuri:ed and the thermal 3, ,. 3,,

The -L?C! anube is requi: 2d to be power verifiad ta 5e less than !! ai ,

3 rated :hermai pover at least 2nca aparable with the axreption that the RER system may be aligned in pe r hour during .tra- ning startup s .

the snutdown :ooling mode 72:2 r=

chan 7- *'"! - :; vhile perform-ing training startups at atmos-pneri: pressurs at povar levels lasa ::an ;; si rattd taermal pover.

-:;262- )c<;9 3e- p

~ _ - . . _ - _ _ _ _ _ . _ _ _ . _ . . -. . , _ . . _ _ _ _ _ _ . . . - . . _ . _ , -- ..-_,.,..m.._~...__,-_..__-

.s., __ a.-__ . A A _ A m_. + a_m,a._ _ - _4_ L_.-4s - w_ m_.

I APPENDIX B

- . _ - - ~ ~

~ . ,

I TABLE OF CONTENTS (cont'd) {

Page tio.

ELfRVEILLANCE ,

LIMITING CONDITIONS FOR OPERATION REOUIREMENTS s

3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 114 131 $

A. Core Spray and LPCI Systems A 114 ;

B. RHR Service Water System B 116 !

C. HPC1 System C 117 D. -RCIC System D 118 '

E. Automatic Depressurization System E 119 F. Minimum Low Pressure Cooling System Diesel

Generator Availability F 120 ,

-G. Maintenance of Filled Discharge Pipe G 122 ]

H. Engineered Safegur.rds Compartments Cooling H 123 l

3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 158 -

A. Thermal and Pressurization Li...ations A 132 B .- Coolant Chemistry B 133a C. Coolant Leakage .

C 135 D. . Safety and Relief Valves D 136 E. Jet Pumps .

E 137 F. Recirculation Pump Flow P imatch F 137 ,

G. ' Inservice Inspection -

G 137 H. Shock Suppressors (Snubbers) H 137a i t

3.7 CONTAINMENT SYSTEMS 4.7 159'- 192 ;

r A. Primary Containment A 159 B. Standby Cas Treatment System B 165 C. Secondary Containment 'C 165a D. Primary Containment Isolation Valves D 166 3.3- MISCELLANEOUS-RADIOACTIVE MATERIAL SOURCES 4.8 185 .-186 t 3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 193 - 202~ +

.i A. Auxiliary. Electrical Equipment A 193 B, Operation with Inoperable Equipment B 195 3.10 CORE ALTERATIONS 4.10 203 - 209 ,

A. Refueling Interlocks A 203

3. Core Monitoring B 205 C. Spent Fuel Pool Water Level C ;205 D. Time-Limitation D 206

.E. Spent Fuel-Cask Handling- E 206 3.11 FUEL RODS. 4.11 219 - 214e A. . Average Planar Linear Heat Generation Rate (AFLEGR) A 210 S. Linear Heat Generation Rate (LHGR) 3 210 ;

.C. Minimum Critical Power Ratio (MCPR) C 212

-li-l L

l -^

..__c. _s. . . . . _ _ _ _ , _ . . . . _ , - _ . . _ _ . _ . . - . . , . . . _ . . . , , . _ _ . . _ . . _ . _ . _ . . . , , , , _ _ . . _ , . . _ , . _ . _ _ _ _ . _ . _

4 TABLE OF CONTENTS (cont'd)

Pace No.

SURVE!LLANCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 3.12 ADDITIONAL SAFETY RELATED PLANT CAPABILITIES 4.12 215 215f Main Control Room Ventilation A 215 1 A.

B. Reactor Equipment Cooling System B 215b l C. Service Water System C 215e D. Battery Room Vt.1*. D 215c 3.13 RIVER LEVEL 4.13 216 3.14 FIRE DETECTION SYSTEM 4.14 216b 3.15 F'.st 3UPPRESSION VATER SYSTEM 4.15 216b 3.16 SPRAY AND/OR SPRINY.LER SYSTEM 4.16 216e (FIRE PROTECIION) 3.17 CARBON DIOXIDE AND MALON SYSTEMS 4.17 216f 4.18 216g 3.18 FIDE HOSE STATIONS 3,19 FI1E BARRIER PENETRATION FIRE SEALS 4.19 216h 2161 3.20 DELETED 9.21 216n I 3.21 ENVIRONMENTAL / RADIOLOGICAL EFFLUENTS A, Instrumentation 216n "

va B. Liquid Effluents 216x C. Caseous Effluents 216a4 kk D. Effluent Dose Liquid " ,. nous 216all /

E. Solid Radioactivo Waste 216al2 F. Monitoring Program 216al3 G. Interlaborato*y Comparison Program 216a20 4.22 216bl 3.22 SPECIAL TFSTS/ EXCEPTIONS A. Shutdown Margia Demonstration 215bl _

L. =c+ ^ng Startup 216b2 g C. n e Tests 216b3 D. Ster op Test Program 216b3 5.0 MAJOR DESIGN FEATURES -

Site Features 217 5.1 Reactor 217 5.2 5.3 Reactor Vessel 217 -

5.A Containment 217 218 I 5.5 Fuel Storage 5.6 Seismic Desigt. 213 5.7 Barge Traffic 213 6.0 ADMidISTRATIVE CONTROLS 6.1 Organization 219 6.1.1 Responsibility 219 6.1 2 offsite 219 Plant Staff - Shif Com.p leme nt 219 6.1.3 6.1.4 Plant Staff - Qualifica:Lons 219a 2

- ii ' .

a

- - - - - ~ _ - - _ _ _ ____._ _ _

2,1 Bases: (Cont'd)

5. Main Steam Line Isolation Valve Closure on Low Pressure The low pressure italation of the main steam lines (Specification 2.1.A.6) was provided to protect against rapid reactor depressuritation, B. Reactor Water Level Trio Settines Which Initiate Core Standby Cooline Svstems ]

(QSQSi The core standby cooling systems are designed to provide sufficient cooling to I the core to dissipate the energy associated with the loss of-coolant accident and to limit fuel clad temperature, to assute that core geometry remains intact and to limit any clad metal-water reaction to less than 1%. To accomplish their intended fune. tion, the capacity of each Core Standby Cooling System component was established based or the reactor low water level scram set point.

To lower the set point of the low water level scram would increase the capacity requirement for each of the CSCE components. Thus, the reactor vessel low water level scram was set low enough to permit margin for operation, yet will net be set lower because of CSCS capacity requirettents ,

The design for the CSCS components to meet the above guidelines was dependent

gglS..

S upon three previously set parameters: The maximum break site, low water level

'" a s- am set point and the CSCS initiation set point. To lower the set point for Fy inttiation of the CSCS may lead to a decrease in effective core cooling, To "4 ,

raise the CSCS initiation set point would be in a safe direction, but it would reduce the margin established to prevent actuation of the CSCS during normal operation or during normally expected transtects.

Transient and accident analyses reported in Section 14 of the Safet's analyses Report demonstrate that these conditions result in adequate safety margins for the fuel.

C. References for ?.1 Bases

1. "Generie Reload Fuel Application," NEDE-240ll-P, (mosc current approved submittal),

2. " Cooper Nuclear Station Single-Loop Operation," NEDO-24258, May 1980.

3. " Supplemental Reload Licensing Submittal for Cooper Nucl.e ar Station

,}!!

Unit 1," (applicable reload document) .

I 4

- l- m

. _ . _ . _ ._ _. - _ _ _ _ . . . _ . _ _ _ . - _ . - _ .-_ - - . . _ _ ~ _ _ . _ _ _ . . . _

~. .,

NOTES FOR TABLE 3.2.A-

1. Whenever Primary ' Containment integrity is required there shall be two operable or

. tripped trip systems for each function.

2 .- If the minimum number of operable instrument channels per trip system requirement cannot be met - by a trip . system, that trip system shall be tripped. If the -

requirements cannot be met by both trip systems, the appropriate action listed below ,

shall be taken.

A. Initiate an orderly shutdown a id have the reactor in a cold shutdown condition in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. Initiate an orderly load reduction and have the Main Steam Isolation Valves shut within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

C. Isolate the Reactor Water Cleanup System.

O. Isolate the Shutdown Cooling mode of the RHR System. l

3. -Two required for each steam line.

4 These signals also start the Standby Gas Treatment System and initiate Secondary Containment isolation.

5. Not required in the refuel, shutdown, and startup/ hot standby modas (interlocked with the mode switch).

6. Requires one channel from each physical location for each trip system.

7. . Low vacuum isolation is bypassed when the turbine stop is not full open, manual bypass _ switches are in bypass and mode switch is not in RUN.

8. The instruments on this table produce primary containment and system isolations. The following listing groups the system signals and the system isolated.

Groun 1 Isolation Signals:

-1. Reactor Low Low Low Water Level (2-145.5 in.)

2. Main Steam Line High Radiation (3 times full power background)

3. Main' Steam Line Low Pressure (2825 psig-in the RUN mode) 4 Main Steam Line Leak Detection (5200*F)

5. Condenser Low Vacuum (27" Hg vacuum)

6. Main Steam Line High Flow (5150% of rated flow)

Isolations:

1 MSIV's

2. Main Steam Line Drains l

l^

l

'32-

.a NOTES FOR TABLE 3.2. A (cont'd.)

Group 2 Isolation Signals:

1. Reactor Low Water Level (24.5 inches)

2. High Dry Well Pressure ($ 2 psig) .

Isolations:

1. RRR Shutdown Cooling mode of the RHR System. l

-2. Drywell floor and equipment drain sump discharge lines.

3. TIP ball valves

4. Group 6 isolation relays Group _3 Isolation Signals:

1. Reactor Low Water Level (24.5 inches)

2. Reactor Water Cleanup System High Flow (5200% of system flow)

3. Reactor Water Cleanup System High Area Temperature (s 200*F)

-Isolations:

1. Reactor Water Cleanup System

-Group 4 1 solation Si5nals:

Provided by instruments on Table 3.1.B (HPCI)

Isolations:

Isolates the HPCI steam line Group 5 Isolation Signals:

Provided by instruments on Table 3 2.3 (RCIC)

Isolations:

Isolates the RCIC steam line.

Group 6 Isolation Signals:

1. Group 2 Isolation Signal

2. Reactor Building H&V Exhaust Plenum HLgh Radiation ((100 mr/h r)

-52a-

w l

3. 2 MSIS.

In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and Standby Gas Treatment System. The objectives of the specifications are (1) to assure l the effectiveness of the protective instrumentation when required even during perious when porti.ons of such systems are out of service for maintenance, and (2) to prescribe the trip settings required to assure adequate performance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

A. Primarv Containment Isolation Functions Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required.

Such instrumentation must be available whenever primary containment integrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation, set to trip at 168.S inches (+4. 5 inches) above the top of the active fuel, closes all isolation valves except those in Groups 1, 4, 5, and 7 Details of valve grouping and required closing times are given in Specification 3.7. For valves which isolate at this level this trip setting is adequata to prevent core uncovery in the case of a break in the largest line assuming a 60 seccnd valve closing time. Required closing timec are less than this.

The low low low reactor water level instrumentation is set to trip when the water level is 19 inchec (-145.5 inches) above the top of the active fuel. This trip closes Groups &

and ? Isolation '!alves (Reference l), activates the remainder of the CSCS subsystems. and starts the emergency diesel generatara These trip level settings were chosen to be aigh enough to prevent spurious actuati.on but low enoue,h to initiate CSCS operation and primar: system isolation so that post accident cooling can be accomplished.

. - - -, . - - - ~.. - . . - _ - - -

LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REOUTREMENTS l 3.4 STANDBY LIOUID CONTROL S~YSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM 6Enlicability: Aeolicability:

Applies to the operating status of Applies to the surveillance the Standby Liquid Control System, requirements of the Standby Liquid Control System. ]

Obiective: Obiective:

To assure the availability of a To verify the opersbility of the I system with the capability to Standby Liquid Control System.

shutdown the reactor and maintain the shutdown condition without the use of control rods.

Soecimication: Soecification:

A. Normal System Availability A. Normal System Availability During periods when fuel is in the The operability of the Standby reactor and prior to startup from a Liquid Control System shall be shown Cold Condition, the Standby Liquid by the performance of the following

'- Control System shall be operable. tests:

except as specified in 3.4.B below.

This system need not be operable .l. At least once per month each when the reactor is in the Cold subsystem shall be tested for l Condition and all control rods are operability by recirculating

-fully inserted and Specification demineralized water to the test 3.3.A is met. tank.

2. At least once during each operating cycle:

a. Check that the settings of the subsystem -relief valvos are l 1450 ( P( 1680=psig and the valves will reset at P 2 1300 psig.

b. Manually initiate the system, except explosive valves, and pump boron solution from the Standby Liquid Control Storage Tank through the recirculation path. Minimum pump flow race of 38.2 gpm against a system head of 1300 psig shall be verified. After pumping boron solution the system will be flushed with demineralized water.

c. Manually initiate one of the Standby i Liquid Control System Pumps and I

- 10 7-

LfMITfNG CONDITIONS FOR OPERATTON SURVEILLANCE REOUIREMENTS 3.4 4.4.A,2,c (Cont'd.)

pump demineralized water into the reactor vesael from the test tank.

l These tests check the actuation of j the explosive charge of the tested ;

loop, proper operation of the i valves , - and pump operability. The replacement charges to be installed l will be selected from the same i manufactured batch as a previously ;

tested charge.

d. Both subsystems, including both l explosive valves, shall be tested in the course of two operating cycles.

B, Overation with Inocerable B. Surveillance with Inonerable Comuonents: Components:

1. From and after the date that one

1. When a subsystem is found to be subsystem is made or found to be inoperable, the . operable subsystem inoperable, Specification 3.4.A.1 shall be verified to be operable shall be considered fulfilled and immediately and daily thereafter continued. operation permitted until the inoperable subsystem is provided that the operable subsystem returned to an opetable condition.

remains operabic and the. inoperable subsystem is returned to an operable condition within seven days.

C. Sodium Pentaborate Solution C. Sodium Pentaborate Solution f.t all times when the Standby Liquid The following tests shall -be Control System is required _ to be performed to verify the availability operable the following conditions of the Liquid Control Solution:

shall be met:

1. The net volume versus concentration 1. Volume: Check and record at least of the Liquid Control Solution in once per day, the liquid- control tank shall be maintained as required in Figure 3,4.1,

! 2. -The temperature of the liquid 2. Temperature: Check and record at control solution shall he maintained least once per-day, above the curve shown in Figure 3.4.2. 3. Concentration: Check and record at least once per month. Also check concentration anytime water or boron is 1

.08-

m, ~

- ~ 3' 413ASE11 STANDBY L10UID CONTROL S,ySICM A. The Standby Liquid Control System consists of two, distinct subsystems,1each containing one positive . displacement pump and independent suction from the liquid ' control tank .and ; discharge to a common injection header through baral3c; squibb valves. The purpose of the Standby Liquid Conttol System is I to: provide the capability of bringing the reactor from full power to a' cold,

-xenon-free shutdown condition assuming that none of the withdrawn control rods can.be-inserted. To meet this objective, the system-is designed to inject a l quantity . of boron that produces a concentration of 600 ppm of boron.in the reactor . core in less than 125 minates. The 600 ppm concentration in the' reactor core is required to bring the reactor from full power to. a 3.0 percent u

ak s'beritical condition, considering the hot to cold reactivity difference, xenon poisoning, etc. The cime requirement for-inserting the boron solution was selected to override the rate of reactivity insertion caused by cooldown of the reactor following the xenon poison peak.

The conditions under which the Scandby Liquid Control System must provide l shutdown capability are- identified via- the Plant Nuclear Safety Operational .

Analysis '(Appendix G)- . If no more than one operable control rod is withdrawn, the basic shutdown reactivity requirement for the core is satisfied and1the-

' Standby Liquid' Control System is not required. Thus,- the basic reactivity.

requirement for the core is the primary determinant of when the Standby Liquid-Control System is required, The ' minimum limitation on the relief valve setting is intended to prevent the recycling' of liquid control solution via the lif ting of a relief valve at too low a pressure. The upper limit on the relief valve setting provides system [

protection from overpressure.

B. Only_ one of the two- Standby Liquid Control subsystems is needed for operating the system. One inoperable subsystem does not immediately threaten shutdown-capability, and reactor operation can continue while the inoperable subsystem is.being repaired. Assurance that the remaining subsystem wil1~ perform its-intended function and that the long term average availability-of-the system is not reduced is' obtained for a one out of two system by an allowable equipment out of service time of one third of the normal surveillance frequency'. This method determines an equipment out of service time of ten days. Additional conservatism is introduced by reducing the allowable out of service - time to seven days.

l L C. Level indication and alarm indicate whether the solution volume has changed,

which might ' indicate a possible solution concentration change. The . tes t ,'

interval-hat been established in consideration of these factors. .Temperaturm and liquid-level alarms for the system are annunciated in the control room, y

The solution.is kept at least 10'F above the saturation temperature to gut.rd against boron precipitation. The margin is included in Figure 3.4.2.

l

-110-ll i . ..- -

. . -- . - - - ~ . .- - - .. - - - _ ~

. 1TMITING CONDITIONS-FOR OPERATION SURVEILIANCE REOUIREMENT 3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 CORE AND CONTAINMENT COOLING SYSTEMS Apolicability: Aeolicability:

Applies to the operational status of Applies to the Surveillance the core and containment cooling Requirements of the core and l systems. containment cooling systems ' which l are required when the corresponding Limiting Condition for Operation is in effect.

Obiective: Obiective:

To assure the operability of the To = verify the operability of the l core and containment cooling systems core and containment cooling systems l-under all conditions for which this under all conditions for which this cooling capability is an essential cooling capability is an essential response to station abnormalities. response to station abnormalities.

Specification: Specification:

l A. Core Sorav and LPCI Svstems A. Core Sorav and LPCI Svstems l l . 1. Both Core Spray subsystems shall be 1. Core Spray Subsystem Testing.

operable:

Item Frecuengs (1) _ prior to reactor startup from a Cold Shutdown, or a. Simulated Once/ Operating Automatic Cycle (2) when there is irradiated fuel Actuation Test.

in the vessel and when the reactor vessel pressure is b. Pump Operability once/ month greater than atmospheric pressure, except as specified c. Motor Operated once/ Month in 3.5:.A.2 and 3.5.F.3 below. Valve Operability

d. Pump flow rate. Once/3 months Both loops shall deliver at least 4720 gpm against a system head corresponding to a differential pressure of 2 113 psi between the reactor vessel and the primary containment.

e. Core Spray Header t.? Instrumentation Check once/ day calibrate Once/3 months

-ils.

w

' LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REOUIRFMFNTS , _ _ _

3,5,A (cont'd.) 4.5,A (cont'd.)

2. From and after the date that one of 2. When it is determined that one Core l

the Core Spray subsystems is made or Spray obsystem is inoperable, the found to be inoperable for any operable Core Spray subsystem anr e r ab il i ty l-l 11-

. .e ,

' - LfMITING CONDITIONS FOR OPERATION SURVEILIANGE REOUIREMENT

~

l'.3.5.C HPCI System (cont'd.) 4.5.C. HPCI System .(cont'd,) l Item Frecuency

2. From and after the - date that the-l _HPCI System is- made or found to be -
inoperable for any reason, continued d. Flow Rate at Once/3 months reactor operation is permissible approximately 1000 only during ' the succeeding seven psig Steam Press.
l days unless such system is sooner made operable, providing that during e. Flow Rate at Once/ operating such seven days all active approximistely 150 cycle components that affect cperability psig Steam Press.
of the ADS , the RCIC System, both LPCI subs / stems and both Core Spray The HPCI pump shall be demonstrated l subsyst na are operable. to be capable of delivoring r.t 1 cast 4250 gpm for a system head
3. With the surveillance requirements corresponding to a reactor. pressure of 4.5.C not performed at the of 1000 to 150 psig. .
required intervals due to reactor shutdown, a reactor startup may be 2. When it is determined that the HPCI conducted provided the appropriate System is inoperable, the RCIC surveillance is performed within System, both LPCI subsystems, and
-48 hours of achieving 150 psig both Core Spray subsystems shall be reacto steam pressure. verified to be operable immediately.
The RCIC System shall be verified'to l 4 If the requirements of 3.5.C.1 be operable daily thereaf ter. In cannot be met, an orderly shutdown addition, the ADS logic shall be j shall be initiated and the reactor demonstrated to be operable pressure shall be reduced to 113 immediately and daily thereafter.
psig or less within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
D. Rer.ctor Core Isolation Cooling D. Reactor Core Isolation Coalinc f,RCIC) Svsten (RCIC) Svsten l l
l 1. The RCIC System shall be operable 1. RCIC System testing shall be l whenever there is irradiated fuel in performed as follows:
the reactor vessel,. the reactor pressure is greater than 113 psig, Item Frequenn and prior co-reactor startup from a Cold Condition, except as specified a. Simulated Once/ operating in 3.5.D.2 and 3.5.D.3 below Automatic cycle Actuation Test
-LiS-
. . - . _~ _ . - . - _ -- - .
. LVMITVNC CONDITIONS FOR OPERATION SURVEILIANCE REQUIREMENT 3.5.D (cont'd.) 4;5'.D (cont'd.)
itege Freauency L. Ptunp Operability Once/ month
c. Motor Operated once/ month-Valve Operability
d. Flow Rate at Once/3 months approximately 1090 psig Steam Pressure
2. From and after the date that - the l RCIC System is made or found to be e. Flow Rate at Once/ operating inoperable for any reason, continued approximately 150 cycle reactor power operation is psig Steam Pressure
-permissible only during the succeeding seven days provided that The RCIC pump shall be demonstrated during such seven days the_ HPCI to be_ capable of delivering at least System is operable. 400 gpm for a sys te.n head corresponding to a reactor. pressure
3. With the surveillance requirements of 1000 to 150 psig.
-of .4.5._D not. performed at the required . intervals due to reactor 2. When '.t is determined that the RCIC shutdown, a reactor startup may be System is inoperable, the HPCI conducted provided the appropriate System shall be verified to be '
surveillance is -performed- within operable immediate ly .'
48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of achieving 150 psig-reactor steam pressure.
4. If the requirements of 3.5,D 1 & 2 cannot be met, an orderly shutdown shall be-initiated and the reactor pressure shall be reduced to 113 psig or less within 24 hoars.
- F. , Automatic Deoressurization System E. Automatic Depressuritation Svetem
( ApS T - (ADS)-
1. The- Automatic Depressurization 1. During -each operating cycle the
-l System shall be _ operable whenever following tests shall be performed there - is irradiated fuel in t.ne on the ADS:
reactor vessel .and the reactor pressure _ is greater than 113 psig A simulated automatic actuation test and prior to a startup from a Cold shall be performed prior to startup Condition, . _ except as specified in after each refueling outage.
3.5.E.2 and 3.5.E.3 below.
-119-
l' l
l SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATION 3.5.E (cont' d) 4.5.E (c ont ' d)
2. From and after tue date that one 2. When it is determined that one valve of the ADS is inoperable, the ADS l valve in the Automatic actuation logic for the other ADS Depressurization System is made or found to be inoperable for any valves shall be demonstrated to be reason, continued rea-tor operation operable immediately. In addition, is permissible only during the the HPCI System shall be verified to l succeeding seven days unless such be operable immediately.
valve is sooner made operable, provided that during such seven days l
the HPCI System is operable.
3. With the surveillance requirements of 4.6.D.5 not performed at the required intervals due to reactor shutdown, a reactor startup may be conducted provided the appropriate surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of achieving 113 psig reactor steam pressure.
4. If the requirements of 3. 5. E.1 or 3.5.E.2 cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to at least 113 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
F. Minimum Low Pressure Coeline and F. El.nimum Low Pressure Coolinn and Diesel Generator Availability Diesel Canerator Availabilir-r
1. Nring any period when one diesel 1. When it is determined that one generator is inoperable, continued diesel generator is inoperable, the i reactor operation is permissible LPCI, Core Spray, and RHR Service only during the succeeding seven Water subsystems associated with the days unless such diesel generator is operable diesel generator shall be sooner made operable , provided that verified to be operable immediately and daily thereafter. In addition, the operable diesel generator and its associated LPCI, Core Spray, and the operable diesel generator shall RHR Service Water subsystems are be demonstrated to be operable j
operable and the requirements of immediately and every thtee days are hereafter.
3.9.A.1 met. If thi requirement cannot be met, the requirements of 3.5.F.2 shall be met.
I I
I
-120-
SURVEIL 1ANCE REOUIREMENTS LIMITING CONDITIONS Fff OPERATION 3,5,F (cont'd.) 4.5.F (cont' d. )
2. During any period when both diesel generators are inoperable, continued reactor operation is permissible only during the succeeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless one diesel generator is sooner made operable, provided that both LPCI subsystema, both Core Spray subsys terns , and both RP.R Service Water subsystems are operable and the reactor power level is reduced to 25% of rated power and the requirements of 3.9 A.1 are met.
If this requirement cannot be met, an orderly shutdown shall be l
initiated and the reactor placed in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 3, Any combination of inoperable components in the LPCI, RHR Service Water, and Core Spray systems shall not defeat the capability of the remaining operable components to fulfill the cooling functions.
4. When irradiated fuel is in *he reactor vessel and the reactor is in the Cold Shutdown Condition, both Core Spray subsystems, both LPCI sub sy s tems , and both RHR Service Water subsystems may be inoperable, provided no work is being done which has the potential for draining the reactor vessel. Refueling requirements are as specified in Specification 3,10.F.
5. With irradiated fuel in the reactor vessel, one control rod drive housing may be open while the suppression chamber is completely drained provided that:
a. The reactor vessel head is removed.
b, The spent fuel pool gates are open and the fuel pool water level is maintained at a I
level 2 33 feet.
I
c. The condensate transfer j sys tern is operable and a minimum of 230,000 i;allons of water is in the condensate SCOrage tank
- 1. 21 -
l- .
'E SLIMITING CONDITIONS FOR OPERATION' SURVEILIANGE REOUIREMENT 5
[3. 5. F1(cont' d) 4.5.T (cont'd)
M d. The automatic mode of the f drywell sump pump- is disabled.-
-e. No- maintenance . is being ,
conducted - which will prevent '
filling. the- suppression ,
chambar to a level above the l Core Spray and LPCI= suctions.
l-
f. With the exception of the
-suppression chamber -water ,
supply, _ both ~ Core Spray
subsystems and both LPCI subsystems are operable, t
5 The control rod is withdrawn to the backseat.- i
h. A special flange, capable of- r sealing a' leaking control rod
housing, is available- for immediate use.
i. The . control = rod ' housing is '
covered with the special flange. following the removal-
.cf the control red drive.
J. No work is being performed in the vessel while L the housing.
A s
-i open.
'G. Maintenance of Filled Dischad.ipt. _G. d.;1 1 ntenanc,e of Filled Discharho Pips Whenever the Core Spray subsystems,- The. .following surveillance LPCI - subsyst. ems , HPCI - System, or- requirements shall.be :sdhered to, to 1 RCIC--System are. requireo to be assure that the discharge piping of -
operable, the discharge piping from the Core. Spray subsystems, LPCI the_ pump. discharge of these systems subsystems. HFCI _ System and RCIC to the last block - valve shall be System are filled:
1.ll e d '
l. :Whenever a Core Spray _ subsystem, *
'LPCI subsystem, ' the' HPCI System, or RCIC-System is . made ' operable , the discharge piping shall be vented from the ~ high point of the system and water flow obser 'ed. initially and on a monthly basis.
2. The pressure switches which monitor the LPCI, Core Spray, HPCI'and RCIC System lines to ensure they are full
' shall be functionally tes ted. and calibrated every three months .
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4 -
3' 5- BASES ,
A,- Core Sorav and LPCI Subsystems This specification assures that adequate emergency cooling capability is available whenever irradiated fuel is in the reactor vessel.
The limiting conditions of operation in Specifications 3.5.A.1 through 3.5 A. 8 l specify-- the combinations of operable subsystems to assure the availability of the minimum required ~ cooling' systems. During reactor shutdown when the residual heat removal system is realigned from LPCI to the shutdown cooling mode, the LPCI subsyste:as are considerc 1 operable. .l The. Core Spray System is a low pressure coolant system which is comprised of two,_ i distinct subsystems and designed to provide emergency cooling to the core by spraying in the event of a loss-of coolant accident. This system functions in combination l with the LPCI System to prevent excessive fuel clad temperature.
The LPCI System is an operating mode ~ of- the RHR System and is comprised of two, distinct LPCI subsystems. The LPCI System is designed to provide emergency cooling to the core by flooding in the event of a loss of coolant accident. This-system
' functions in combination with the Core Spray. System to prevent excessive fuel clad temperature. The LPCI. and Core Spray systems provide adequate cooling -for break' areas of approximately C.2 square feet up to and including the double ended recirculation line _ break without assistance from the high pressure emergency core .
cooling subsystems.
The allowable repair times are established so that the average risk rate for repair would be no greater than the basic. risk rate. The method and concept are described in reference (1). Using the results developed in this reference, the repair period
i. is found to be slightly greater than 1/2 the test interval. This assumes that the l (1) Jacobs, I.M. , " Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards", General . Electric Co. A.P.E.D., April, 1969
,(APED-5736).
1 i
I L
l-
. i v. .
.- - . . . - . -.- - - _ . ~~ -
i
'3.5.A BASES (cont'd.)
Core Sptay subsystems and LPCI subsystems constitute 'a 1- out of- 4 system; however. l the combined effect of the two systems to limit excessive clad temperatures must alco -
be considered. The test interval specified in Specification 4.5 is 1 month.
Should one Core Spray subsystem become inoperable, the remaining Core Spray subsystem and the LPCI subsystems are available should the need for core cooling arise. To assure-that the remaining Core Spray and LPCI subsystems are available, they are verified to be operable immediately. ;
Should the loss of one LPCI pump occur, a nearly full complement of core cooling-equipment is available. Three LPCI pumps in conjunction with the Core Spray subsystems will perform the core cooling function. Because of the availability of the majority of the core cooling equipment, which will be. verified to be operable, a thirty day ' repair period is justified. If one LPCI-subsystem is not available, at l 1 east 1 LPCI pump must be available to fulfill the containment cooling function. The 7 day repair. period is set on this basis. ,
B. RHR Service Water Svstem l The RHR Service Water System consists of two, distinct subsystems designed to provide heat removal for -the containment cooling function. Each RHR Service Water subsystem contains two RHR Service Water booster pumps serving one side of one of two RHR Heat Exchangers, while two RHR (LPCI) pumps serve-the other nide. The RHR Service Water System operates in conjunction with the RHR System to provide the containment cooling function.
The design of the RHR Service Water System is pt edicated upon the use of one RHR Service Water booster pump and one RHR heat exchanger for heat removal af ter a design basis accident. Thus, there are ample spares for margin above design conditions. ,
Loss of margin . should be avoided and the equipmentz maintained in a state of operation. So a 30 day out-of service time is chosen for this equipment. If one loop is out-of-service, reactor operation is permissible for seven days . The requirements for availability of the RRR System for support of the containment cooling function are reflected in the associated action statements for the - LPCI-System.
With components or-subsystems out-of service, overall core and containment cooling reliability is maintained by verifying the operability of the remaining cooling equipment. For routine out - o f- s e rvic e periods caused by preventive- maintenance ,
etc., the operability of other systems and components will be verified ~as given in the Technical Specifications . However, if a failure, design deficiency, occ. , caused the out of service period, then a -demonstration . of operability may be needed to assure that a similar problem - does not exist on.tne remaining components. For-example, if an' out-of-service period were caused by failure of a pump to deliver ratec capacity, the other pumps of this type might be subjected to a capacity test.
The pump capacity test is a comparison of measured pump performance parameters
-125-
. . _ , _.- -_ _ - _ __ _ . - . , . __ _ _ ._ __ ___ _ = .
y: -3 s
9
$ 5;B BASES.'(cont'd.)
r to - shop performance tests. Tests during normal operation will be performed by measuring the - flow and/or the pump discharge pressure. These parameters and its power requirement will be used to establish flow at that pressure.
' C. MPCI System The limiting conditions for operating the HPCI System are derived from the Station Nuclear Safety Operational Analysis (Appendix G) and a detailed functional analysis of the HPCI System (Section VI.).
The'HPCI System is provided to assure.that the reactor core _. is adequataly cooled l to limit fuel clad temperature in the event of a small break in the nuclear system and loss-of coolant which does not result in rapid depressuritation of the. reactor vessel. The HPCI System permits the reactor to be shut down while maintaining l sufficient reactor vessel water level inventory until the vessel is depressuri=ed. '
The HPCI System continues to orerate until reactor _ vessel pressure is below the l pressure at which LPCI operation or Core Spray System operation maintains core cooling.
The capacity of the: system is selected to provice this required core cooling. The HPCI - pump is designed to pump 4250 gpm at reactor pressures between 1120 and 150 psig. Two sources of water are available. Initially, demineralized water from-
,the emergency condensate storage- tank is used instead of injecting water from t.ne suppression pool into the reactor.
-When the HPCI System begins operation, the reactor depressurizes more rapidly than would occur if HPCI was not initiated due to the condensation of -steam by the cold fluid _ pumped into the reactor vessel by the HPCI System. As the reactor vessel pressure continues to decrease, the HPCI flow momentarily reaches equilibrium with the flow through_the break. Continued depressurination causes the break flow to decrease below the HPCI flow and the liquid inventory begins to rise, This type of response. is' rypical of the small breaks. The core _never uncovers and is continuously .
ooled throughout the transient so that no core damage of any_ kind occurs for breaks that. lie-vithin the capacity range of the HPCI System. l The analysis in the FSAR, Appendix G, shows that the ADS provides a single failure proof path for cepressurization for postulated transients and accidents, The RCIC System serves as an alternate to the H?CI System only for. decay heat removal when
- feed water is lost, Considering the HPCI System and the ADS plus the RCIC System _as redundant paths, refer 2nce (1) methods would give an estimated allowable repair time of 15 days based or the one month testing frequency. However, a maximum allowable repair time. of 7 days is selected for conservatism. _The HPCI and RCIC~ Systems as [
well. as all other Core Standby Cooling Systems must be operable when starting up from '
l- a Cold Condition. It is realized that the HPCI System -is not designed to operate' I i until reactor pressure exceeds 150 psig and is automatically isolated before the reactor pressut'e decreases below 100 psig. It is .the intent of this specification
(;
-126-
. . - ~ . _ - _ _ . _ -. _ .-
+
4
-3,5.C BASES (cont'd.)'
to assure that when the reactor is being started up from a Cold Condition, the HPCI-System is not-known to be . inoperable. l D. RCIC Svr Di The RCIC System is designed tc' ptovide makeup to the nuclear system as part of the l-planned. operation for periods when the main condenser is unavailable. The nuclear safety analysis, FSAR Appendix G, shows that'the RCIC System provides water to cool 'l the fuel when feed water is lost. .In all other postulated _ accidents and transients, the ADS provides redundancy for the HPCI System. Based on this and judgements on the reliability of the HPCI System, an allowable repair time of 7 days is specified.
Immediate verifications of HPCI System operability- during RCIC System outage is considered adequate based or. judgement and practicality.
E. Automatic Deorassutization Svstem (ADS)
The limiting conditions for operating the ADS are derived from the Station Nuclear Operational Analysis (Appendix G) and a detailed functional . analysis of the ADS (Section VI.).
This specificatio- ensures ti.e operability of the ADS under all conditions for which the automatic or manual deprossurization of the nuclear system is an essential restor.se to station abnormalities.
The. nuclear system. pressure rellef system.-provides automatic nuclear -system depressurization for small breaks in the nuclear system so toat the LPCI and Core Spray _ Systems Man aperate to protect the fuel barrier.
Because che Automatic Depressurization System does not provide makeup to the reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further conservatism to the- CSCS. Performance analysis of the ' Automatic Depressurization - System is considered only with. respect to its depressurizing effect in conjunction with the LPCI or Core Spray Systems. There are l
-six valves provided and each has a capacity of 800,000 lb/hr at a set pressure of
-1080 psig.
Thel allowable out of service time for . one ADS- . valve is determined as seven days-because of the redundancy and because the HPCI System is verified to be operable l-during- this period. Therefore, redundant-protection for the core with a small break in the nuclear system is still available.
The ADS test circuit permits continued surveillance on the operable relief valves to assure that they will be available if required.
~
Minimum Low Pressure Cooline imd Diesel Generator Availabilitv, F.
The: purpose of Specification F is to assure that adequate core cooling equipment
-127-
4 3.5' BASES (cont'd) is available at all times, It is during refueling outages that major maintenance is performed ~ and during such time that all low pressure core coolin8 systems may be out of service. Specification 3.5.F.4 provides that should this occur, no work will be performed on the primary system which could lead to draining the vessel. This work would include work on certain control rod tive components and recirculation system.
Thus,- the specification precludes the- evet.ts which could require core cooling.
Specification 3.3.F.5 recognizes that, concurrent with control rod drive maintenance during the refueling outage, it may be necessary to drain the suppression chan.ber for maintenance or for the inspection required by Specification 4.7. A.2.h. In this case ,
if excessive control rod housing leakage occurred, three levels of protection against loss of core cooling would exist. First, a special flange would be used to stop the leak. Second, sufficient inventory of water is maintained to provide, under worst case . leak ' conditions , approximately 60 minutas of core cooling while attempts to ,
secure the leak..are made. This inventory includes water in the reactor well, spunt-fuel pool, and condensate storage tank. If a leak should occur, manually operated valves in the condensate . transfer system can be opened to supply either the Core 4 Spray System or - the spent fuel pool. Third, s'fficient inventory af water is=
maintained to permit the water which'has drainet . rom the vessel to fill the' torus.
to a level above the Core Spray and LPCI suction strainers. These systems could then i .
recycle the water to the vessel . Since the system cannot be pressuri::ed during refueling, the potential need for core flooding only exists and the specified combination of the Core Spray or the LPCI subsystems can provide this. This {
l. specification also provides for the highly unlikely case that both diesel generators are found to be inoperable. The reduction of rated power to 25% will provide. a very stable operating-condition. The allowable repair time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months will provide an opportunity to repair the diesel and thereby prevent the necessity of taking the plant down through the less stable shutdown condition. It the necessary - repairs I cannot be made in the allowed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the plant will be s'autd wn in an orderly
fashion. This will be accomplished while the two off-site sources of power required by Specification 3.9.A.1 are available.
G. Mainrenance of Filled Discharge Pioe If the discharge piping of the Core Spray, LPCI, HPCI, and RCIC systems are not l' tilled, a water hammer can develop in this piping when the pump-and/or pumps are
- r, tarted. If a water hammer were to occur at the time ati which the system were required, the system would still perform its design functions. However, to minimite damage to the discharge piping and to ensure added margin in the operation of those systems, this Technical Specification requires the discharge lines to be filled whenever the system is in an operable condition.
H. .Encineared Safecuards Comnartments Cooline
' The unit cooler in each pump compartment-is capable of providing adequate ventilation flow and cooling. Engineering analyses indicate that ' the temperature rise in safeguards compartments without adequate ventilation flow or cooling is n.ch that l continued operation of the safeguards equipment or associated auxiliary equipmen; cannot be assured, i-
-123-
,,&. -nmp- -a-xm p- -. ~my r m-o gr-- c-c, p myn --.---g,my--9 y_y y y =-- .,prT*
,. ,. . - . - . - - . - - - . - - - .- - -. - - - - --~ ~ , .-. .
l_f f 4
&4 4
%5 16FJ -,
Core and' Containment Cooling Systems Surveillance Frecuencies 1
. Thel testing intervals for the core and containment cooling systems are; based on l industry practice, quantitative reliability analysis, judgement and practicality.
Tho . core cooling systems have no t _ be en designed to be -fully testable during-operation. For example, in the case of the HPCI System, automatic initiation during l power operation would result in pumping cold water into the reactor vessel, which is not desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant inventog. To increase the availability of the core and containment '
' cooling systems, the components which make up the cystem; i.e., instrumentation, pumps, valves :etc., are tested frequently. The pumps and motor operated injection valves.are also tested'each month to assure their operability. A simulated automatic
actuation test once each cycle combined with frequent tests of the pumps _ and injection valves is deemed to be adequate testing of- these systems, tThen components and subsystems are out- o f- s ervice , overall core and containment cooling reliability is mainte.ined by verifying the operability of the remaining equipment. For routine out-of service periods caused by preventative maintenance, etc., the operability of other systems and' components will be verified as given in the Technical Specifications, However, if a failure or design deficiency caused by
-outage, _then a demonstration of operability may be needed to assure that a generic problem does not exist, For example, if an out-of-service period were caused by -
failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a fkw rate test, d
d l
-131-i
. .. -- .- . - - . -. - . - .- - . . . - = . .. - -. .- .- ~ -
..1
.L )
i l
l LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT ,
1
,3.7.A (cont'd.). 4.7.A (cont'd.)
6. Low-Low Set Relief Function 6. Low-Low Set Relief Function
a. The low-low ' set . function of the a. The low-low set safety / relief valves- .
safety-relief- valves shall be shall be tested and calibrated as !
operable when there is irradiated specified in Table 4.2.B.
fuel in the reactor vessel-and the
-reactor coolant temperature is '
2 212'r, except as specified in 3.7.A.6.a.l'and 2 below.
1. With the low low function of one safety / relief valve (S/RV) inoperable, restore - the inoperable LLS S/RV to OPERABLE-within 14 days or be in the HOT STANDBY mode within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and .in Cold SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
2. With the low-low set function of both S/RVs inoperable, be in ar least HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and !.n COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,
b. The pressure' switches which control the low-low set safety / relief valves shall have the following settings.
NBI-PS-51A Open Low Valve 1015 20 psig (Increasing)
NBI-PS-51B Close Low Valve 875 20 psig (Decreasing)
NBIiPS-51C Open High Valve 1025 .t 20 psig (Increasing)
3. Standhv Gas Treatment Svstem NB1-PS-51D Close High Valve
-875 20 psig (Decreasing) 1. At least once per operating cycle the following conditions shall~be B. Standhv Cas Treatment System demonstrated.
1. Except as specified in -3.7.B.3 a. Pressure drop across the combined-below. both' Standbv Cas Treatment-HEPA .ilters and charcoal adsorber
- subsystems shall be' operable at all banks is less than 6 inches of water times when secondarv c.ontainment at the system design flow rate.
integrity is required'.
2.a. The results of the in-place cold DOP b. Inlet heater input is capable of leak tests on the HEPA filters shall reducing R.H. from 100 to 70% R.H.
.show 299% DOP removal. The results of the- halogenated hydrocarbon leak 2.a. The tests and sample analysis of tests on the charcoal- adsorbers Specification 3.7.B.2 shall be shall show 2997. halogenated performed at least once every hydrocarbon removal. The DOP and 18 months for standby service or i- halogenated hydrocarbon tests shall after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system be performed at a Standby Gas operation and following significant Treatment flourate of $1780 CFM and painting, fire or chemical releace at'a Reactor Suilding pressure of in any ventilation tone 6 . 2 5 Wg . communicating with the system.
-163-
LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT 3.7.3 (cont'd) 4.7.B (cont'd)
b. The results of laboratory carbon b. Cold DOP testing shall be performed sample analysis shall show 199% after each complete or partial radioactive methyl iodide removal replacement of the HEPA filter bank with inlet conditions of: velocity or after any structural maintenance 227 FPM, 21.75 mg/m 3 inlet methyl on the system housing.
iodide concentration, 270% R.H. and
$30'C. c. Halogenated hydrocarbon testing shall be performed after each
c. Each fan shall be shown to provide complete or partial replacement of 1780 CMP 110%. the charcoal adsorber bank or af ter any structural maintenance on the
3. From and after the date that one system housing.
Standby Gas Treatment subsystem is l
made or found to be inoperable for d. Each subsystem shall be sperated l any reason, reactor operation is with the heaters on at least permissible only during the 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month.
succeeding seven days unless such subsystem is sooner made operable, c. Test sealing of gaskets for housing l
provided that during such seven days doors downstream of the HEPA filters all active components that affect and charcoal adsorbers shall be operability of the operable Standby. performed at, and in conformance Gas Treatment subsystea, and its with, each test performed for associated diesel generator, shall compliance with Speelfication be operable. 4.7.B.2.a and Specification 3.7,0.2.a.
Fuel handling requirements are specified in Specification 3.10.E. 3. System drains where present shall be inspected quarterly for adequate 4, If these conditions cannot be met, water level in loop-seals.
procedures shall be initiated immediately to establish reactor 4.a. At laast once per operating cycle conditions for which the Stindby Gas autcratic initiation of each Standby Treatment System is not required. Cas Treatuent subsystem shall be demonstrated,
b. At .s t once per ope rating cycle 1 cperability of the bypass va. /e for filter cooling shall be demonstrated,
c. 'Jhe n onc Standby Cas Treatment subsystem b3cames inope able, the operable Standby Gas Treatment subsystem shall be verified to be
~
operable immediately and daily thereafter. A demonstration of diesel generator operability is not required by this specification.
C. Lecondar" ontainm-n t C. Secondarv Containe nt
1. Secondary containme nt surveillance Secondary containment integrity shall be performed as indicated 1.
shall be maintained during all modes below:
of plant operation except when all of the following conditions are met.
1050-
i LIMITING CONDITIONS FOR OPERATION SURVEILIANCE RENj :REMENTS 3.7.C (cont'd.) 4.7.C (cont'd.)
a; The reactor 'is suberir# . and a. A- preoperational secondary Specification 343.A is'me- containment capability test shall be conducted after isolating the.
-bi The reactor water temperature is reactor building and placing either below 212*F and- the reactor coolant Standby Gas Treatment subsystem ~ l
-system is vented. filter train in operation. Such tests shall demonstrate the
c. No activity is being performed which capability to maintain 1/4 inch of =
can reduce the shutdown margin belou water vacuum under calm wind that specified in Specification (2<E<5 mph) conditions with a filter 3.3.A. train flow rate of not more than 100% of building volume per day,
d. Irradiated fuel is not being handled (5- wind speed) in the secondary containment.
b. Additional tests shall be performed
e. If secondary containment integrity during - the first operating cycle cannot be maintained, restore- under an adequate number of secondary- _ containment integrity difforent- environmental wind-within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or; conditions to enable valid extrapolation of the test results,
a. Be in at least Hot Shutdown .
within the: next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and c. Secondary containment capability to in cold shutdown within the maintain 1/4 inch of water vacuum following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , under calm vind (2<E < 5 m p-h )
conditions with a filter s:ain flow
b. Suspend irradiated _ fuel rate of not more than 100% _ of.
handling operations in the building volume per day, saall be secondary containment and all demonstrated at each refueling core alterations and outage prior to refueling, activities which could reduce the shutdown margin. The .d. After a secondary containment provisions of Specification violation is determined, the Standby 1.0.J are not' applicable. Gas Treatment System will be s operated immediately after the affected tones are isolated from the remainder of the .-secondary
- containment to confirm its ability to maintain the - remainder of the secondary containment at 1/4 inch of water negative pressure under calm '
_ wind conditions.
D. -Primarv Conta f nrent Isolation Valves D. Primary Containment Isolat* on Val"as
1. During reactor power operating 1. The primary containment isolation conditions, all isolation valves valves surveillance shall be listed in Table 3.7.1 and all performed as fotlows:
instrument-line flow check valves "' ^E I**SU nce per Sperating cycle shall be operable except as t.he operab.le isolation valves that specified in 3 ~.D.2. are power operated and automnically initiated shall be tested for simulated automatic initiation and
! closure times.
-166-H
~ - - - - - -.. -.- --. . . ~. - - - . - - - - -- -
9 -
3.7.A-& 4.7.A BASES (cont'd)
The primary containment is normally slightly pressurized during periods of reactor
, operation. Nitrogen used for. inertiPS could leak out of the containment but air.
could not leak in to increase oxygen concentration. Once the containment is filled-with nitrogen t_o the required concentration, no monitoring of oxygen concentration is_ . necessary. However, at-least twice a week the oxygen concentration will be determined as added assurance.
The 500 gallon conservative limit on the nitrogen storage tank assures that adequate time ~is available to get the tank refilled assuming normal plant operation. The estimated maximum makeup rate is 1500 SCFD which would require about 160 gallons for a 10 day atakeup requirement. Tne normal 1,ak rate should be about 200 SCFD.
i 3.7.A.6 & 4.7.A.6 LOW.LOV SET RELIEF FUNCTION i The'iow-low set relief logic is an automatic safety relief valve (SRV) control system designed to mitigate the postulated thrust load concern of subsequent actuations of
SRV's during certain transients (such as inadvertant MSIV closure) and small and intermediate break loss-of-coolant accident (LOCA) events. The setpoints used in Section 3.7.A.6.b are based upon a minimum blowdown range to provide adequate time between valve actuations_to allow the SRV discharge line high water leg to clear, coupled with consideration of instrument inaccuracy and the main steam isolation valve isolation setpoint.
The as-found setpoint for NBI-PS 51A, the pressure switch controlling'the opening of RV-71D, must be 5 1040.psig. The as-found closing serpoint for NBI-PS-51B-must be at least 90 psig less than 51A, and must be 2 850 psig.. The as-found setpoint for NBI-PS 51C, pressure switch controlling the openin, of RV-71F must he s 1050 psig.
- The as-found closing setpvint for NBI-PS-51D must be at least 90 psig below 51C and must be- 850 psig, This ensures dat the-analytical upper limit tor the openin5 setpoint (1950 psig), the analytical lower limit on the closing setpoint (850 psig) and the analytical limit on the blowdown range (2 90 psig) for the Low Low Set Relief Function are not exceeded. Although the specified instrument setpoint tolerance is 20 psig, an instrument drift of ' t 25 psig was used in the analysis to ensure adaomte margin in determining the valve opening-and closing setpoints. The opening setpoint-is set such that, if both the lowest set non LLS S/RV and the highest set of the two LLS S/RVs drift 25 psig in the worst case directions, the LLS S/RVs will still_ control subsequent S/RV actuations. Likewise, the closing setpoint is set to ensure thm LLS S/RV closing setpoint remains above the MSIV low pressure trip, The 90 psig blowdown provides adequate energy release f rom the vessel to ensure time for the water -leg to clear between subsequent S/RV actuations.
, 3.7.B & 3.7.C STANDBY CAS TREATMENT SYSTEM AND SECONDARY CONTAINMENT l
The secondary containment is . designed to minimize any ground level release of radioactive c.aterials which might result from a serious accident. The reactor building provides secondary containment during reactor operation when the drywell is sealed and in service. The reactor building provides primary containment when the reactor is r. hut down and the drywell is open. as during refueling. Because the secondary - containment is an integral part of the complete containment system.
secondary containment is required _at all times that primary containment is required as well as during refueling. Secondary containment may be broken for short periods of time to allow access to the reactor building roof to perform necessary inspections and maintenance.
The Standby Gas Treatment System consists of two, distinct subsystems. each l containing one exhaust f an and associated iilter train. wnich is designed to ;ilter I and exhaust the reactor building atmosphere to the stack during secondary containment isolation concitions. Both Standby Cas Treatment system fans are designed to i automatically start upon containment isolation and to maintain the reactor building pressure to the design negative pressure so that all leahage should be in-leakabe Should one subsystem fail to start, the redundant wbsystem is designed tc start i automatically Each of the two fans has 100 percent capacity
-180-
h .,
. i l
(
3.7,B & 3.7,C EbES. (cont'd) :
High afficiency particulate absolute (HEPA) filters are installed before and after the ' chatcoal adsorbers to minimize _-potential release. of particulates to' -the '
. environment and to prevent clogging of the iodine adsorbors, -The charcoal adsorbers L are-installed to reduce the potential release of radioiodine to the environment. The in aplace test results should indicate a r.ystem leak tightness of less than 1 percent
< bypass leakage for the charcoal adsorbers and HEPA filters. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected accident conditions, if the performance of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 Guidelines for the accidents analy::ed.
Only one of the two Standby Gus Treatment subsyste ns is needed to cleanup the reactor building atmosphere upon containment isolation, if one subsystem is found to be inoperable.-there is no immediate threat to the containment system performance and ;
reactor operation or refueling operation may continue while repairs are being made.
If both subsystems are inoperable, the plant is brought to a condition where the Standby Gas Treatment System is not required.
4.75B & 4.7.C BASES Standbv Gas Treatment System and Secondarv Containment Initiating reactor building isolation and operation of the Standby Gas Treatment System to - maintain at least a 1/4 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building-
-isolation valves, leak tightness of the re. actor building and performance of the Standby Gas ' Treatment --System. Functionally testing the initiating sensors and- I ass ociated trip channels demonstrates the capability for automatic actuation..
Performing these tests prior to refueling will demonstrate secondary - containment capability prior to the time the primary containment is opened for refueling, Periodic testing gives sufficient confidence of reactor building integrity and I Standby Gas Treatment System performance capability.
Pressure drop _ across the combined HEPA filters and charcoal adsorbers cf less than
'6 -inches of water at the system design flow rate will indicate that the filters and adsorbers:are not clogged by excessive amounts of foreign matter. A 7.8 kw heater is capable:of maintaining relative humidity below 70%. Heater capacity and pressure drop shoul d be . determined at least onca per perachg cycle to show system performance capability. :
The frequency _of tests and sample analysis . are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with ANSI NS10 1980. The test canisters that are installed with C..e adsorber trays should be used for the charcoal adsorber efficiency test. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If rest
= -results-are unacceptsble,. all adsorbent in the system shall be replaced 182-
l 4.7,8 6 4.7.0 BASES with an adsorbent qualified according to T le 5.1 of ANSI N509 1980. The ;
for the test should meet the same i replacement adsorbent quality. tray for Tests the of adsorber the HEPA tra["ilters recoveu with DOP aerosol shall be performed in !
accordance to ANSI N510 1980. Any filters found defective shall be replaced with i filters qualified pursuant to Regulatory Position C.3 A. of Regulatory Guide 1.52, t i Revision 2, March, 1978. .
All elements of the heater should be demonstrated to be functional and operable during the test of heater capacity. Operation of the heaters will prevent moisture j buildup in the filters and adsorber system. ;
With doors closed and fan in operation DOP aerosol shall be sprayed externally along i the full linear periphery of each respective door to check the gasket seal. Any datcetion of DOP in the tan exhaust shall be considered an unacceptable test result j and the gr.skets repaired and test repested. -
If system drains are prescnt in the filtor/adsorber banks, loop seals must be used with adequate water level to prevent by pass leakage from the banks. ;
If significant paintin5, fire or chemical release occurs such that the HEPA filter !
or charaal adsorber could become contaminated from the futes, chemicals or foreign material, the same tests cnd sample analysis shall be performed as required for ,
operational use. The determination of significance shall be made by the operator on !
duty at the time of;the incident. Kmwledgeable staff members should be consulted ,
prior to making this determination. ;
Demonstration of the automatic initictf on capability and operability of filter i cooling is nicessary to assure system performance capability. If onc Standby Gas Treatment subsystem is inoperable, the operable subsystem's operability is verified
' daily. This substantiates the availability of the operable subsystem and thus reactor operation (/ refueling operation can continue for a limited period of time. ,
3.7.D & 4.7.D fdSES, !
Primary Containment Isolation Valves ,
e Double isolation valves are provided on lines peaetrating the primarv .:ontainment and open to the free space of the containment. Closure of one of the vabes in each line would be sufficient to 'naintain the int.grity of the pressure suppression system, Automatic initiation is recuired to minimite the potential leakage paths from the containment in the event of -a loss of coolant accident.
The maximum closure times for the automatic isolation valves of the primary contcinment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following p u breaks 4
outside the primary containment and the need to contain released fission prc . mt.t .
following pipe breaks inside the primary containment. !
These valves _ are highly reliable, have a low service requirement, and most at i 4 normally closed. The initiatine sensors and associated trip channels are _ als >
checked to demonstrate the capabil,ity for automatic isolation. The test interval ' f once per operating cycle for automatic initiation '
1 Y
-183- ,
_ . - . .. _ _ - _ _ _ _ _ _ _ _ _ . _ -. _ _ . _ . _ _ . - . - ~ _ . _ _ .. - . _ . ._-
._-. . _ _._ _ .._ _ ~ - _ _
s LIMITING CONDITIONS FOR OPERATION OURVEILLANCE REOUIREMENT9
'J.10.B (Cont'd) 4.10 (Cont'd) 4.- During spiral reload, SRM operability will be verified by using a portable external source every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months until the required amount of fuel is loaded to maintain 3 cps. As an alternative to the above. two fuel assemblies will be loaded in different cells containing control blades around each SRM to obtair. the required 3 eps. Until these two assemblies have been loaded, the 3 cps requirement is not necessary,
c. Spent Fuel Pool Water Level _
C. Stien t Fuol Pool Water Level Whenever-irradiated fuel is stored in the spent fuel pool, the pool When irradiated fuel is stored in water level shall be maintained at the : pent fuel pool, the water level or above 8h' above the top of the shall be recorded daily, fuel.
D. time Limitation ,
Irradiated fuel shall not be handled in or above the reactor prior to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor shutdown.
E. Standbv Gas Treatment System E. Standby Gas Treatment System From and after the date that one l Standby Gas Treatment subsystem is When one Standby Gas Treatment made or found to be inoperable for subsystem becomes inoperable, the any reason, fuel handling is operable Standby Gas Treatment permissible only during the subsystem shall be verified to be succeeding seven days unless such operable immediately and daily i subsystem is socner made operable, thereafter. A demonstration of provided that during such seven days diesel generator operability is not all active componene.s that affect required by this specification. l operability of the operable Standby Gas Treatment subsystem, and its ,
associated diesel generator, shall ,
be operable.
1 At least one diesel generator shall be operable during fuel handling ,
operations. This one diesel shall be capable of supplying power to an operable Standby Gas Treatment l subsystem.
F. core Standbv Coolint Svsten During a refueling outage. refueling L operation with-fuel in the reactor vessel may continue with one Core Spray _ and one- LPCI subsystem inoperable, or with both Core Spray subsystems inoperable. Refueling is permitted with the suppression chamber drained provided an operable l Core Spray or LPCI subsystem is ,
aligned to take a suction on the '
condensate storage tank containing at least 150.000 gallons (114 ft.
indicated level).
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3.10 MIES (Cc.nt ' d)
D. Time Limitation The radiological consequences of a fuel handling accident are based upon the accident occurring at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reattor shut.down.
E. Standbv Cas Treatment System-only cae of the two Standby Gas Treatment subsystems is needed to clean up the reactor building atmosphere upon containment isolation. If one subsystem is found to be inoperable, there is no immediate threat to the containment system performance and refueling operation may continue whilo repairs are being made.
If both subsystems are inoperable, the plant is brought to a condition where i the Standby Gas Treatment System is not required. .i F. Core Standby Cooline Systems During refueling, the system cannot be pressurized, so only the potential need '
for core flooding exists and the specified combination of the Core Spray or LPCI subsystems-can provide this. A more detailed discussion is contained in ,
the bases for 3.5.F.
G. Control Room Air Treatment .
If the~ system is found to be inoperable, there is'no immediate threat to the !
control room and refueling operation may ccntinue for a limited period of time while repairs are being made. If the system cannot be repaired within seven days, refueling operations will be terminated.
H. Spent Fuel Cask Handling. , ;
The operation of the redundant crane in'the Restricted Mode during fuel cask ,
handling operations assures that the cask remains within the controlled area i once it has been rerwved from its transport vehicle (i.e. , once it is above the 931' elevation). Handling of the -cask on the Refueling Floor in the ,
Unrestricted Mode isallowed only in the case of equipment. failures or ;
emergency conditions when the cask is already suspended. The Unrestricted Mode
. of- operation is allowed only to tho extent necessary to get the cask to a
suitable stationary position so the required repairs can be made. Operation with a failed controlled area microswitch will be allowed for a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period providing an Operator is on the floor in addition to the-crane operator to assure that the cash handling is limited to the controlled area as marked on >
the floor. This will allow adequate time to make repairs but still will not !
restrict cask handling operations unduly.
4.10 BASET I A. Refueling Interloch ,
Complete functional testing of all refueling interlocks beforo any refueling ;
outage will provide positive indication that' the interlocks operate in the
. situations for which they were designed. By loading each hoist with a weight equal to the fuel assembly positioning tce refueling platform ard withdrawing control rods, the interlocks can be subjected to valid operational tests.
Where redundancy is provided in the logic circuitry, testa can be performed to assure that each redundant logic elettent can independently perform its -
functions.
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4.10 ItASES (Cont'd)
B. Core Monitoring Requirin6 the SRM's to be functionally tested prior to any core alteration assures that the SRM's will be operable at the star of that alteration. The daily response chech (or 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months check for spiral reload) of the SRM's ensures their continued operability.
E. Standbv Cas Treatmtnt System only one of the two Standby cas Treatment subsystems is needed to clean up the reactor building atmosphere upon containment isolation. If one subsystem is l found to be inoperable, there is no immediate threat to the containment system pt.rformance and refueling operations may continue while repairs are being made.
If both subsystems are inoperable, the plant is brought to a condition where the Standby Cas Treatment System is not required.
H. Spent Fuel Cask b ndling The Surveillance Roquirements specified assure that the redundant crane is adequately inspected in accordance with the accepted ANSI Standard (B.30.2.0) and ruanuf ac turer's recommendations to determine that the equipment is in satisfactory condition. The testing of the controlled area limit switches assures that the crane operation will be limited to the designated area in the Restricted Mode of operation. The tast of the "two-block" limit switch assures the power to the hoisting motor will be interrupted before an actual "two. blocking" incident can occur. The test of the inching hoist assures that this mode of load control is available when required.
Requir'ng the lif tina and holding of the casx for 5 minutes during the initial lif t of each series of cask handling operations puts a load test on the entire crane lifting mechanism as well as the braking system.
l'erforming this test when the cask is being lif ted initially from the cask car assures that the system is operable prior to lifting the load to an excessive height.
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ITMITINC CONDITInN9 FOR 01FRATION 9URVEIIlANCE RFOUTREMENT9 +
3.12 (cont'd) 4.12 (cont'd)
B. Ep_ actor Equitment Cooline. (REC 1 B. Reactor Equipment Cooling (REQ 1 System System ,
1, Both Reactor Equipment Cooling 1. REC System Testing I Frecuency subsystems and their associated Itern pumps shall be operable whenever .
irradiated fuel is in the vessel or a. pump Operability once/ Month the spent fuel pool, except as b. Motor operated Once/ Month specified in 3.12.B.2 and 3212.B.3 Valve Operability belov, c. Pump flow rate Once/3 months Each pump shall and after pump deliver 1175 gpm maintenance at 65 psid.
d. System head tank Daily level shall be monitored.
2. From and after the date that any 2. When it is determined that any active component that .affects active component that af fec ts operability of one REC subsystem operability of an REC subsystem is ,
becomes inoperable, continued inoperable, all active components reactor operation is permissible that affect operability of the during the succeeding thirty days operable REC subsystem shall be provided that during such thirty verified operable immediately and days all active components that weekly thereafter. i affect operability of the operable REC subsystem, the active components that affect operability of the engineered safeguards compartment l cooling systems, and the diesel generator associated with the I operable subsystem are operable.
The allowable repair time does not apply when the reactor is in the shutdown mode and reactor pressure is less than 75 psig, l l 3. Both' REC subsyctems , with one pump ,
I per subsystem, shall be operable as stated in 3,12.3.1 and 3.12.B.2 above during reactor head off operations requiring LpCI or Core Spray system availability or Service Water cooling shall be available.
4 If the requirements of 3.12.B,1 ,
through 3.12. B.3 cannot be met , the '
reactor shall be shutdown in an orderly manner and in the Cold Shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or- operations requirint, LpCI or Core Spray system availability shall be halted.
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LIMITINC CONDITIONS FOR OPEPATION SURVFIII).NCE REOUTREMENTS 3.12 (cunt' d) 4.12 (cont'o) t Service Water System C. Service Water System C.
1. Both Service Water subsystems with 1. Service Water System Testing l both pumps in each subsystem shall be operable whenever irradiated fuel Item Eunctional is in the vessel or spent fuel pool and prior to reactor startup except a. pump Operability Once/ Month as specified in 3.12.C.2 below, Motor Operated
b. Once/ Month '
Valve Operability
c. Pump discharge once/3 months head tests Fron. and a iter the Alar , that any 2. When it is determined that any 2.
active con. cone s that affects required Service Water System op9rability of am Service Water component is inoperable, all active si.bsystem is male or found to be components that affect operability of the operable Service Water tuvperable for an; resson, continued subsystem shall be verified to be reactor operaaon is permist.ible immediately and weekly during the succeeding thirty days operable provided that during such thirty thereafter.
days all active components that affect operability of the operable Service Water st.bsy s t em and its associated diesel generator are operable.
3. If the requirement of 3.12.C.1 and 3.12.C.2 cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Batterv Room Ventilation D.
D. Battarv Room Ventilntion The spare battery room ventilation 1.
Battery room ventilation shall be fan shall be checked for operability 1.
operable on a continuous basis once/ week.
whenever specification 3.9.A is required to be satisfied. 1
2. From and after the date that either of the two battery room vont fans is made or found to be inoperable for any reason, continued reactor operation is permissible during the succeeding 7 days.
3. If the requirements of 3.12.D.1 & 2 -
cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
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3,12 BASES A. Main Control Room Ventilation System The control room ventilation system is designed to filter the control room atmosphere for intake air and/or for recirculation during control room isolation conditions.
The system is designed to automatically start upon control room isolation and to maintain the control room pressure to the design positive pressure so that all leakage should be out leakage.
High efficiency particulate absolute (HEPA) filters are installed before the charcoal edsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential intake of radiciodine to the control room. The in-place test results should indicate a system leak tightness ofThe lesslaboratory than 1 percent carbon ,
l bypass leakage for the charcoal adsorbers and HEPA filters.
sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected accident conditions. If the performance of the e
' HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria ,
4 for Nuclear Power Plants, Appendix A to 10 CFR Part 50.
If the system is found to be inoperable, there is no insediate threat to the control room and reactor operation may continue for a limited period of time while repairs '
are being made. If the system cannot be repaired within seven days._the reactor is shutdown and brought to cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
B. Reactor Eaulement Coo 11st (REC) System l The Reactor Equipment Cooling System consists of two, distinct subsystems, each containing two pumps and one heat exchanger. Each subsystem is capable of supplying ,
the cooling requirements of the essential se rvice s following design acciden. l conditions with only one pump in either subsystem.
The REC System has additional flexibility provided by the capability of
-interconnection of the two subsystems and the backup water supply to the critical cooling loop by the Service Water System, This flexibility and the need for only one ,
pump in one critical cooling loop to meet _ the design accident requirements justifies the 30 day repair time during normal operation and the reduced requirements during l
head-off operations requiring the availability of the LPCl or Core Spray systems. l C. jytr/ ice Water Systeg The Service Water System consists of two. distinct subsystems,- each containing two vertical Se rvice Water pumps located in the intake structure, and associated strainers, piping, valving and instrumentation. The pumps discharge to a common )
header from which independent piping supplies two Seismic Class I cooling water . oops and one turbine building loop. Automatic valving is provided to shutof f all supply to the ;urbine building loop on drop in header pressure thus_ assuring supply to the Seismic Class I loops each of which feeds one diesel generator, two RHR Service Water booster pumps, one control room basement fan coil unit and one REC ,
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i 3.12 BASES (cont'd) heat exchanger. Valves are included in the common discharge header to permit the Seismic Class 1 Service Water System to be operated as two independent subsystems. l
- The heat exchangers are valved such that they can be individually backwashed without j interrupting system operation, During normal operation two or three pumps will be required. Three pumps are used l
] j for a normal shutdown. '
Tae loss of all a c power will trip all operating Service Water pumps. The automatic j emergency diesel generator start system and emergency equipment starting sequence
. will then start one- selected Service Water pump in 30 40 seconds. In the meantime, the drop in Service Water header pressure will close the turbine building cooling i
water i olation valve guaranteeing supply to the reactor building, the control room l
basencr., and the diesel generators from the one Service Water pump. r Due to the redundance of pumps and the requirement of only one to "eet the accident '
requirements, the 30 day repair time is justified.
D. Batterv Room Ventilatien t
The temperature rise and hydrogen buildup in the battery rooms without adequate ventilation is such that continuous safe operation of equipment in these rooms cannot t be assured.
4.12 ghgES A. Main-Control Room Ventilation Svstem ,
Pressure drop across the combined HEPA filters and charcoal adsorbers of less than i 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Pressure drop should be determined at least once per operatin5 cycle to show system performance i
capability. .
Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant should be l performed in accordance with ANSI N510 1980.
The frequency of! casts and . sample analysis are necessary to show tha t the HEPA filters and charcoal adsorbers can perform as evaluated. The test canisters that are installed with the adsorber trays should be used for the charcoal adsorber afficiency test. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable, all adsorbent tn the I
system shall.be replaced with an adsorbent qualified according to Table 5.1 of ANSI N509-1980. The replacement tray for the absorber tray removed for the test should meet the same adsorbent quality. Tests of the HEPA filters with DOP aerosol slall be performed in accordance to ANSI N510-1980. Any HEPA filters found defective snall be ' replaced with filters qualified pursaant to - Regulatory Position C.3.1 of Regulatory Guide 1 52.
Operation of the. system for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month will demonstrate operability of the filters and adsorber system and remove excessive moisture built up on the adsorber.
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f SURVEILLANCE REOUIREMENTS LIMITING CONDITIONS FnR OPFRATION 1
3,22 SPECIAL TESTS / EXCEPTIONS (CONT'D) 4.22 SPECIAL TESTS / EXCEPTIONS (CONT'D)
2. k' hen the constraints imposed on
2. Rod Sequence Control System control rod groups by the RSCS are (RSCS) bypassed, verify:
The sequence constraines imposed on control rod groups a. That the R'4M is OPERABLE.
by the RSCS may be suspended bv means of the incividual b. Conformance with this r'od position bypass switches specificati n and procedures or jumpers, 7rovided that the by a second licensed operator red worth minimizer is r other qualified employee.
OPERABLE, for this and the following special tests.
a. Control rod scram timing,
b. Control rod friction measurements.
c. Startup test program with thermal power lass than 20% of rated thermal power.
If the above requireeent is not satisfied, the RSCS shall be operable.
3. RHR System The RHR system may be aligned in the shutdown cooling mode .i with at least one shutdown cooling modo loop OPERABLE while performing the Shutdown Margin Demonstration.
4. Containment Systems Primary containment is not required while perfortc.ing the Shutdown Margin Demonstration when reactor water temperature is equal to or less than 212'F.
B. Traininc Star *up B. Trnining Startun l 1. LPCI System The reactor vessel shall be verified The LPCI System is required to h3 to be unpressurized and the thermal l
operable with the exception that the power verified to be less than 1% of RHR system may be aligned in the rated thermal power at least once j shutdown cooling mode while per hour during training startups.
performing training startupc at atmospheric pressure at power levels less than 1% of rated thermal power.
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ML20079J215

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