ML19257D639
| ML19257D639 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/04/1980 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19257D640 | List: |
| References | |
| NUDOCS 8002050225 | |
| Download: ML19257D639 (20) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION yg s,
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JERSEY CENTRAL P0k'ER & LIGHT COMPANY DOCKET NO.53-219 OYSTER CREEK N'JCLEAR GENERATIhG STATI0tl, UtlIT NO.1
' AMENDMENT TO PR0"ISIONAL CPERATING LICENSE klendment No. M License No. DPR-16 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Jersey Central Power & Light Company (the licensee) dated November 16, 1979, com;1ies with the standards and recuirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confor--ity with the applicatinn, the provisions of the Act, and the rales and regulations of the Cocmssion; C.
There is reasonable assurance (i', trat the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) -hat such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amend ent will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amend:elt is in accordance with 10 CFR Part 51 of the Conraission's rege'atior.s and all applicable requirenents have been satisfied.
1871 307
%$ N 8002050 2_.2._5
. 2.
Accordingly, the license is a. ended b. c anges to the Technical Specifications as indicated in the at act.ent to this license amendment, and paragraph 3.B. of Provisi:qal Operating License No. DPR-16 is hereby a. ended to read as follows:
(B) Technical Specifications The Technical Specifications con aired in Appendices A and B, as revised throuch Amend ent No. M, are hereby incorporated in the license.
The l'censee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effectives as f the date of its issuance.
FC. T'-E ti'JCLEAR REGULATORY COMMISSION S0 ph Q L,%
m Dennis L. Ziemanh Chief 0 -ara-ing Reactors Branch #2 Di vis'on of Operating Reactors Attach. ment:
Changes to the Technical Specifications Date of Issuance: January 4, icio 1Bi1 308 PJOR ORigut
ATTACHMENT TO LICENSE AME"DMENT NO. "
PROVISIONAL OPERATING LICENSE N0. DPR-16 DOCKET NO. 50-219 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
PAGES 1.0-1 1.0-2 2.1-3 3.1-6*
3.1-6a 3.1-7 3.1-7a 3.1-8 3.1-9 3.1-10 3.1-11 3.1 12 3.1-12a 3.4-lb 3.5-2 37-18i1 309 3.7-2
- There are no chan]es to the provisions on this page.
It has been retyped for administrative purposes only.
1.0-1 SECTION I DEFINITIONS The folleving frequently used ter=s are defined to aid in the uniform inter-pretatica of :he specifications.
1.1 OPERA 3LE A sys:cm or compenent shall be censidered operable when it is capable of performing its required func:1cn in its required manner.
1.2. C?ERATING Cpccating means that a system or cenponent is performing its required function.
1.3 PC'.eER OPE nT!ON Pcuer cperatica is any cperatien when the reac:or is in the startup mode or' rur. cede ax:ept wnen pri.1:/ centain: cat inte;rity is not recuird.
S.a.,-..,
. m.s D. e 1.,
.a The reactor is in the startup mode when the reactor :cde switch is in the startup mode position.
In this mode, the reactor protection sys:em scram trips initiated by condenser low vacuum and main steam line isola-tion valve closure are bypassed when reac:c pressure is less than c00 psig; the icw pressure cain steanline isolation valve closure is by-passed; the IRM trips for red block an'd scram are operable; and :he SRM trips for red block are cperab12.
1.5 RUM MODE The reactor is in the run ecde when the reac er =ede switch is in the run =cdc pesi:1cn.
In this =cde, the reac Or protection sys:cm is energi:cd w1:h APRM protec:ica and the control rod withdrawal intericcks are in acevice.
1.6 SHUT CNN CCNDITICN The r acc:cr is in a shutdcun condition when the resetor node svi:ch is
'in the ahutdcun ode positica and there is fuel in the reactor vessel.
In this condition, the reactor is suberitical, a control rod block is initiated.
all cperable centrol rods are fully inserted, and specification 3.2-A is met.
1871 310 1.7 COLD SHUTOCNN The reactor is at cold shutdown when the mode switch is in the shut-I dcun ecdc posi:icn. there is fuel in the reactor vessel, all operable centrol rods are fully inserted, and the rea:: r ecolan: system maintained at less than 212*F and ventea.
Amen ment No. )6,. 44
1.0-2 1.8 PLACE IN SHCOCk'N CCNDITION Proceed with and maintain an uninterrupted normal plant shutdown opera-tion entil the shutdown condition is =et.
1.9 PLACE IN CO:.D SHCD0k'N CONDITION Proceed with and =aintain an uninterrupted normal plant shutdown opera-tion un til the cold shutdewn condition is =et.
1.10 PLACE IN ISOLATED CONDITION Proceed with and =aintain an uninterrupted normal isolation of the reactor fro = the turbine condenser syste= including closure of the
=ain s tea = isolation valves.
1.11 REFL'EL M.0DE The reactor is in :he refuel = ode when the reactor mode switch is in the refuel = ode position.and there is fuel in :ne reactor.yessel.
In this node the refueling platfor= interlochs are in operation.
l.12 REFUELING OLTAGE For the purpose of designating frequency of testing and surveillance, a refueling outage shall =ean a regularly scheduled refueling ou: age; however, where such cutages occur within 8 =enths of the end of the previous re fueling outage, the tes t or surveillance need not be per-formed until the next regularly scheduled outage.
Following the first refueling outage,the ti=e between successive tests or surveillance shall not exceed 20 ocn:hs.
1.13 PRIM >RY CONTAINMEr INTEGRIN Primary containment integrity =eans that the drywell and adsorption cha=ber are closed and all of the following conditions are satisfied:
A.
All non-aute=atic pri=ary contain=ent isolaticn valves which are not required to be open for plant operation are closed.
3.
At least one door in the airlock is closed and sealed.
C.
All au:o=atic containment isolation valves are operable or are secured in the closed position.
D.
All blind flanges and =anways are closed.
}h[l
}jj 1.14 SECCNDARY COWAINET IEECRI'"?
Secondary containment integrity =esns tha: the reactor building _s closed and the following condi:ior.s are =et:
AmendmentNo./,44
2.1-3 The design basis critical heat flux correlation is based on an interrelationship of reac:or coolan: flew and s:ea: cuality.
Seca: quality is determined by reac er pcuer, pressure, and coolant inlet erthalpy which in : urn is a f untrien of feedvater terpe rature and water level.
This correlation is based cocn experi ental da:a tcken ever the entire p c:sure range of interest in a IVR, cnc :ne correlating line was deter =ined by the statistical =ean of the experimental da:a.
Curves are presented for e o different pressures in Figure 2.1.1.
The upper curve is based en ncminal cperating pressure of 1035 psia. The lever curve is based en a pressure c: 1250 psia.
In is reac:or pressure ever c:: rec:ad to exceed 1250 psia no case because of protecticn sys:e= se::incs well belev :his value and, therefore, :he curves vill cover all cperating conci:icns wit.m interpolation.
For pressures be:veen 600 psia (the lower end of the critical hes: flux correla::en da:a) and 1035 psia, :he upper curve is applicable vi:h increased =argin.
The pever shape used in the calculation of Figure 2.1.1 is given in Table 3.2 of Reizrence 10 for a peak to average pcVer l
of 1.5 vi:h a peak loca:ica a: the core nidplane (X/L = 0.5).
Table 3.2 further shovs an axial pever shape vi:h en exial peak of the snna ragni:ude bu: vi:n a peak loc::icn abcve :ha core midplane (::/L = 0.65).
These power shapes result in total peaking fac:crs for each fuel type as shcwn in Specifi-cation 2.1.A.l.
The to:al peaking fac:ce for each fuel type is to be less than that specified in Cccalce 2.1.A.1 at ra:ed power. When ccara:inz balev rated scver with hicher peakins f actors as durir; ccatrol rod manipula:ica er ne:: end cf cere life, applicabili:y of the safety _ini: is assurec by applying the reduc: ion fac crs specified in 2.1.A.2.
The feedvater temperature assu=ed was the maxi =u: desi:n te::er-ature output of the feecva:er heaters at the given pressures and firvs (e.g., 334'? at 1035 psis and 100% flev;.
cr anv icver feedvater :enperature, subcooling is increased and :ha curves provide increasad cargin.
The water level assumed in the calculations was ten inches belev the reac:or icv va:er level scra= pcin: (10'-7" above the :cp of the active fuel), which is the loca:ior of :he bottc= of :he stea:
separator skirts. Of course, the reactor could not be operated in this condition.
As len3 as :he va:er level is above this poin:,
the safety lini: curves are applicable.
As icng as the water level is above the hot:c: of the stea= separator skirts, the c:ount of carryunder would not be increased and the core inle: enthalpy vould not be influenced.
The values of the para =erers involved ir. Figure 2.1.1 :an be de: ermined frem informa:1cn available in the centrol r:ce.
Reac:or pressure and fiev are recorded and :he A?RM in-core.uclear instrenentation is calibrated in tern: of percent pcuer.
18n m P00R OPK Amendment No. J3I, #4
3.1-6 High flow in the main steamline is set at 120% of rated flow. At this setting the isolation valves close and in the event of a steam line break limit the loss of inventory so that fuel clad perforation does not occur. The 120%
flow would correspond to the thermal power so this would either indicate a line break or too high a power.
Temperature sensors are provided in the steam line tunnel to p. ovide for closure of the main steamline isolation valves should a break o-leak occur in this area of the plant. The trip is set at 50 F above ambient ter'pera-ture at rated power. This setting will cause isolation to occur for main steamline breaks which result in a flow of a few pounds per minute or greater.
Isolation occurs soon enough to meet the criterion of no clad perforation.
The low-low-low water level trip point is set at 4'8" above the top of the active fuel and will prevent spuricus operation of the automatic relief system. The trin point established will initiate the automatic depressuriza-tion system in time to provide adequate core cooling.
Soecification 3.1.B.1 defines the minimum number of APRM channel inputs re-quired to permit accurate average core power monitoring.
Specifications 3.1.B.2 and 3.i.C.1 further define the distribution of the operable chambers to provide monitoring of local power changes that might be caused by a single rod withdrawal. Any nearby, ooerable LPRM chamber can orovide the required input for average core monitoring. A Travelling Incore Prabe or Probes can be used temporarily to provide APRM input (s) until LPPM replace-61*' of rated power,gce APRM rod block protection is not required below ment is possible.
as discussed in Section 2.3, Limiting Safety System Settinas, operation may continue below 61% as long as Specification 3.1.B.1 187i 313 Amendment No.
9', M
3.1-6a and the requirements of Table 3.1.1 are met.
In order to maintain reliability of core monitoring in that quadrant where an APRM is inoperable, it is permitted to remove the operable APRM from service for calibration and/or est provided that the same core protection is maintained by alternate means.
In the rare event that Travelling In-core Probes (TIPS) are used to meet the requirements 3.1.3 or 3.1.C, the licensee may perform an analysis o f substitute LFRM inputs to the AFRM system using spare (non-AFD1 input) LPRM detectors and change the APRM system as permitted by 10 CFR 30.39.
Under assumed loss-of-coolant accident conditions it is in-advisable to allow the si=ultaneous starting of emergency core cooling and heavy load auxiliary systems in order to minimice the voltage drop across the emergency buses and to protect against a potential diesel generator overload.
The diesel generator load sequence time delay relays provide this protective function and are set accordingly. The repetitive accuracy rating of the timer mechanism as well as parametric analyses to evaluate the maximum acceptable tolerances for the diesel loading sequence timers were considered in the establishment of the appropriate load sequencing.
Manual actuation can be accomplished by the operator and is considered appropriate only when the autcmatic load sequencing has been completed.
This will prevent simultaneous starting of heavy load auxiliary systems and protect against the potential for diesel generator overload.
1871 314 Referencc:
O:
3750-10139 "An Analysis of Functional Ccmmon Mode Failures in CE EWR t rotection and Control Instr-mentation", L. G. Frederick, et.al, July 1970.
Change No. J' AmendmentNo.f,1, 44
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TAlli.E 3.1.1 PPOIECfl VE lilSTRllt!Et4 TAT 10fl l< EQill HillEtiTS (ChilT D)
W flin. ti o. of Reactor flodes flin. tio. of Operable in Which Function Operable or Instrument flus t le Operable Operating Channels Per (Tripped) Trip Operable Action Trip Setting Shutdown Refuel Startup Run Systems Taip Systems Itequ i red
- Fim c t i on See note h l
- 2. 1.w -1.w-low 1 4' fs" above X(v)
X(v)
X(v)
X 2
2 Reactor Water top of 1.ev e l active fuel
- 3. AC Voltage ll A X(v)
X 2
2 Prevent auto depressurization on loss of V power. See note i
, _ _ _ Iso?at ion Condenser Isolation Isolate Affected 11.
- 1. liigh Flw Steam
< 20 psig A P X(s )
X(s )
X X
2 2
isolation con-l Line denser, comply
- 2. liigh Flw Con-
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O_ffgas System Isolation Isolate reacto f
I.
1.
liigh Radiation
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X(s)
X X
1 2
or trip the l
in Offgas Line Release limit inoperable in-(e)
(See 3.6-A.1) strument channel E lteactor llullding Isolation and Isolate Reactor St andby Cas Treata ent System Bldg.6 Initiate Standby Cao Treat-l Initiation ment Systen, or
_,100 Hr/llr X(w)
Xfw)
X X
1 1
- 1. liigh Itadiation Manual Surveill-Henctor Building ance for not more Operation Floor 2.
Feacto r tildg.
,17 Hr/llr M w)
X (W)
X X
1 1
than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l
(total for all in.
Vent i lat ion struments under Exhaust psig M u)
X (u)
X X
1(k) 2(k)
J) in any 30-day l
- 3. liigh luvuell
- 4 period.
Pressure 4.
1.uw I ou Reactor 7'2" above X
X X
X 1
2 Water f.evel top of active fuel Change No. /
Amendment flo. 44
I 3.1-11 CD N
TAhl.E 3.1.1 l'l(OTECflVE Iti!;Thill!EllTATiott Rl3}llIlmtil:tlTS (WrifD) w
!!! n.!!o. o f g
a fli n.
Ilo. of Ope rata le Heactor }1oeles Operable or Instrument in Which Function Operating Channels l'er
!!ua t fle Ope rable (Tripped) Trip Operable Function Trip Settlun Shutdown Refuel Startup Run Systems Trip Systems Action Requi rett a Ha.1 Il l o ck llo control rod withdrawals 1.
Sitt tipscale
< 5 x 10 cpa X
X(1) 1 3 (Y) permitted l
5 1
SDri Douuucale 1 100 cpu X
X(1) 1 3 fYI l
ff) 3.
IDil Deunucule 1 5/125 fullacale(g)
X
.X 2
3 4.
APidi Upacule X (s)
X X
2
.3(c) 3.
APF11 Dounacalo 1 2/150 fullucale X
2 3(c) 6.
Iltlf Ilyset te
_; 108/125 fullacale X
X 2
3 ranil. eu.c r Vacuum
-Insert control pi.y..p 1 s o I a t i uil toda 1.
Ilf gh Haillation
< 10 x llormal During St artop and 2
2 fu li.ifn Steam llackground run when vacut.m Tu n u s:1 pump to operating l.
liienal Generatoc Time delay after 109] Sequeace Timera energiz, of relay 1.
Containment Spray 40 occ + 15%
X X
X X
2(m) 1(n)
Conalder conta tiunent l'u a p opray loop inoperable and comply uith Anendment flo. M, 44 Spec. 3.4.C(See flate q)
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3.1-12a TAlli.li 3.1.1 (CON'D.1 i.
The interlock is not required during the start-up test program and demonst rat ion of plant electrical output but shall be provided following these actions.
- j. flo t required below 40% of turbine rated steam flow.
k.
All fonr (4) drywell pressure i ns t rinnen t channels may be made inoperable during the integrated primary containment leakage rate test (See Specification 4.5), provided that primary containment integrity is not respii red and tha,t no work is performed on the reactor or its connected systems which could result in lowering the reactor water level to less than 4'8" above the top of the active fuel.
1.
llypassett in Illli Ranges 8, 9, f. 10.
'there is one time delay relay associated with each of two pumps.
m.
One tinie delay relay per pump must be operable.
n.
u.
'lhene are two time delay relays associated with each of two pum;is.
p.
'lko t ime delay relays per pump must be operable, Hanhal initiation of af fected component can be accomplished af ter the automatic load sequencing is q.
completed.
r.
Time delay starts af ter closing of containment spray pump ci rcui t breake r.
s,
'i h es e funct ions not requi red to be operable with tiie reactor tempe ra t u re less t han 212*F and the vessel head removed or vent ed.
CD N
t.
These funct ions may be inoperable or liypassed when corresponding port ions in the same core spray system logic trniu are Inoperable per Specification 3.4.A.
~
LN u.
't hese funct ions not requi red t o be operable when prima ry cont ainment integrit y is not required to be maintained.
g N
v.
These functions not required to lie operable when the AliS is not requi s ed t o lie ope rab l e,
w.
These funct ions innst be ope rabic only when irradiated fuel is in the fuel pool or reactor vessel and secon lary containment int egri ty is required per specification 3.5.11.
y.
ihe number :f operable channels inay lic reduced t o 2 per Specificalion 3.9-li and I:
Aine ndment M. JS, #
3.4-lb that the pump and any necessary valves can be started or operated from the control room or frem local control stations and the torus is
=cchanically intact.
B.
Automatic Depressuri:stion System 1.
Five electromatic relief valves of the autcmatic depressurizatien systen shall be operable when the reactor water temperature is greater than 21:*F and pressuri:ed above 110 psig, except as specified in,
3.4.B.2.
The automatic pressure relief :unc::.on or these valves (but no: the automatic depressuri:ation function) may be inoperable or bypassed during the system hydrostatic pressure tes: required by ASSE Code Section XI, IS-300 at or near the end of each ten year inspection interval.
1871 323 Ar.endment tio. J2', 2T', 44
3.5-2 4.
, Reactor Buildinc to Suoeres;ien Chamber acuum 3reaker Svste a.
Except as specified in Specifica:icn 3.5.A.4.b below, two reactor building to suppressien chamber vacuu.:
breakers in each line shall be operable at all tires when primary containment in:cgrity is required.
The se: point of the differential pressure ins:ru=entatien which actuates the air-eperated vacuum breakers shall not exceed 0.5 psid.
The vacuum breakers shall =cve frem closed :
fully open when subjected to a force equivalent of ne: greater than 0.5 psid acting en the vacuu breaker disc.
b.
F cm the time that one of the reac cr building to suppressien cha=ber vacuum breakers is made er :aund to be inoperable, the vacuu breaker shall be locked closed and reac:cr cperatica is permissible enly during the succeeding seeen days unless such vacuu: breaker is made operable scener, provided that the precedure does not violate primary ccatainment integrity.
c.
If the limi:s of Specification 3.5.A.3.a are exceeded, reactor shutdown shall be initiated and the reac:cr shall be in a cold shutdown condition within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.
5.
Pressure Sucoressien Cha=ber - Drvvell 's'acuum Breakers a.
'n' hen primary centainment is required, all suppression chamber - drywell vacuum breakers shall be operable except during testing and as stated in Specificatien 3.5.A.4.b and c, below.
Suppressien chamber - drywell vacuum breakers shall be censidered operable if:
(1) The valve is demens: rated to epen frc= closed c fully cpen with the applied force at all valve positions not exceeding that equivalent to 0.5 psi acting on the suppressica chamber face of the valve disk.
(2) The valve disk will clcse by gravity to within not gres:er than 0.10 inch of any point on the seal surface of the disk when released after being opened by remote or manual means.
(3)
The position alarm system will annunciate in the control room if the valve is open more :han 0.10 inch at any point along the seal surface cf the disk.
1871 324 Amendment No. W, E, 4
3.7-1
- 3. 7 AUXI*.! ARY ELEC"FlCAL ?CWER Acc li cab ilit v : Applies to the opera:ing status of the auxiliary electrical pcwer supply.
Cbjective:
To assure the cperability of the auxiliary electrical pcuer supoly.
Scecifiestien:
A.
The reactor shall not be nade critical unless all of the follcwing requirements are satisfied:
1.
The folicwing buses or panels energiced.
a.
4160 vol: buses 1C and 1D in the turbine building stri:chgear roc =.
b.
AoD volt buses lA2, 13 2, LA21, 1321 vital MCC 1A2 and 132 in the reacter building switch gear rece: 1A3 and 133 at the intake structure; 1A21A. 1321A, lA213, and 13213 and isolation valve MCC 1A32 en 23' 6" elevation in the reactor building; 1A24 and 1324 at the stack.
c.
208/120 vol: panels 3, 4, 4A, 43, 4C and VACP-1 in the reactor building switchgear room.
d.
120 volt' protection canel 1 and 2 in :he cable roc =.
e.
125 vol: DC distributien centers A and B, and panel D in the battery roc =; isolation valve motor control center DC-1 en 23'6" eleva:1cn in reactor buildinz and pc el E in the cable room.
f.
24 volt D.C. peser panels A and B in the cable room.
2.
One 230 KV line is fully cperational and switch gear and both startup transf ormers are energiced to carry pcuer to the statien 4160 vol: AC buses and carry pcuer :o or away frc= the plant.
3.
An additional scurce of pcuer consisting of cne cf the followinz is in service connected :c f eed :he appro riate plan: 4160 V bus or buses:
a.
A second 230 K7 line fully coeratienal.
b.
One 34.5 KV line fully opera:icnal.
4.
The station batteries are available for normal service and a battery charger is in ser/ ice f or each bat:ery, except ene battery and associated charger may be removed frem service as recuired fer surveill nce testing as set forth in Specification 4.7.3.
3.
The reac:or shall be placed in :he cold shutdown cendition if the availability of power falls belew tha: required by Specification A above, except that the reae:cr may remain in operation for a period not to exceed 7 days in any 2a day period if a startup transf erner is out of service.
l 1871 325 Amendment t'o, 44
- 3. 7-2 None of the engineered safety feature equipment fed by the remaining transformer may be out of service.
C.
Standby Diesel Generators 1.
The reactor shall not be made critical unless both diesel generators are operable and cacable of f eeding their designated 4160 volt buses.
2.
If one diesel generator becenes incoerable during pcwer operation, repairs shall be ini:iated i==ediately and the other diesel shall be cperated at least one hour every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at greater than 20: ra:ed pcver until repairs are c c=pleted.
The reac:or may remain in cperation for a period not to exceed 7 days in any 30-day period if a diesel generator is cut of service.
During the repair period ncne cf the engineered safe:y features normally fed by the coerational diesel generator may be out of service or the reactor shall be placed in :he cold shut-devn condi:ica.
3.
If both diesel generators beccme inoperable during pcwer cpera:1cn, the reactor shall be placed in the cold shut-dcun condi:1on.
4.
For the diesel generators to be considered operable there shall be a =in1=u= of 14,500 gallens cf diesel fuel in the standby diesel generator fuel tank.
Bases:
The general objective is to assure an adequate supply of pcwer with at least one active and cne standby source of pcwer available for operacion of equip =ent required for a saf e plan: shu:down, to maintain :he plant in a safe shu:dewn condition and :o epera:e the required engineered safety feature equipment folleving an accident.
AC pcVer for shutdcun and operation of engineered safety f eature equipment can be previded by any of four active (two 230 *:7 and two 34.5 KV lines) and either of two standby (two diesel generators) sources of pcwer. Normally all six sources are available.
McVever, to provide for =aintenance and repair of equipment and still have redundancy of peser sources the requirement of one ac:1ve and one standby source of power was established. The plant's =ain generator is not given credit as a source since it is not available during shutdcun. The clan: 125V OC power is nor= ally supplied by two batteries, each with an associated ch arg er.
A :hird charger is available to supply either bat:ery.
These charc.ers are ac:1ve sources and supply the ner=al 125V DC requirener with :he bat:eries as s:a dby sources.
In applying the =inimum requirenent of one active and one standby source of AC pcwer, since bc:h 230 KV lines are en :he sa.-e se: ci
- cwers, either one or both 230 KV lines are considered as a single
- 96
]8,/}
c active scurce.
Amendment No. 4d