Text

Revised Tech Spec Pages of License DPR-50,App A,Reflecting NRC Review Comments
	ML20033B032
Person / Time
Site:	Crane
Issue date:	11/13/1981
From:	; METROPOLITAN EDISON CO.
To:	;
Shared Package
ML20033B027	List: ML20033B026; ML20033B029; ML20033B032;
References
	NUDOCS 8111300301
	Download: ML20033B032 (39)
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' Reactor' qoolant system pressure i

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r*-%;.k_ _ _.. Duringla sEhp' accident from lobower or a slow rod ' withdrawal f rom

.:.highipowerilthe system high pressure -trip set point.is reached before m sfii8.

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AdcNuclear*ovirpower'tripTset poistE The irip setting" limit for high',

,5 ukN;P.CE.G:.23E..~ireactor coolant;-system pressure has been established to maintain t

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j;s dd=% Gdsystem pressure :belovthe safety limit (2750 psig) for any design tran-sient.

(6)Due to calibration and instrument errors, the safe:y analy-sis assumed a 45 psi pressure error in the high reactor coolant system pressure trip setting.

Tne high pressure trip se:poin was subsequently lowered from 2390 psig to 2300 psig.

Ine lowering of the high pressure : rip se: point and raising of the se: point for the Power Operated Relief Valve (PO1V),

from 2255 psig to 2450 psig, has the effect of reducing the challenge ra:e to the POrlV while maintaining ASE Code Saf e:y Valve capability.

The low pressure (1900 psig) and variable low pressure (11.75 Tyg7 5103) : rip setpoin: were es:ablished to maintain ene DNS ratio grea:er that or equal to 1.3 for : hose design accidents tha: result in a pre s-sure reduction (3,4).

Tne 362 generic ECCS analysis however, assumed a low pressure trip of 1900 psig and is therefore the basis of low pres-sure reactor trip.

Figure 2.3-1 shows the nign pressure, low pressure, and variable low pressure trips.

d.

Coolan: outle temperature

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The hign reactor coolant outle: temperature trip setting limit (619 F) shown in Figure 2.3-1 has been established to prevent excessive core coolant t empera ture in :he opera:ing range.

The calibrated range of the temperature channels of the RID is 320* :e 620*F.

Tne trip se point of the channel is 619'I.

Under the wors:

case environment, power supply perturba: ions, and drif t, the accuracy C

of the trip string is I F.

This accu 3acy was arrived at by summing the wors: case accuracies of each module. < This is a conservative me: hod of error analysis since the nor=al procedure is to use the root mean i

square method.

Ine re f ore, it is assured tna a trip will occur a: a value no higner l

than 620*F even under worst case conditions. The safety analysis used a high temperature trip set point of 620*F.

The calibra:ed range of the channel is that portion of tne span of indication which has been qualified with regard to drif t, linearity, repeatability, etc.

Inis does not imply that the equipment is restricted to operation within the calibrated range.

Additional test-ing has demonstrated that in fact, the temperature channel is fully operational approximately 10% above the calibrated ran;e.

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.Since.it?haslbeen established that the enannel will trip at a value of 32 _ ' _g _. g;EC outlet: temperature no higher than 620*F even in the. worst case, and e._. d.._; _:. _.-.u_JBince _theshannel is fully operational approximately. 10 7. above.the _ :

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ns:-..-e= 'ScalihEatedfrangejand. exhibits no hysteresis or foldover characterls ai.n.

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.n a:c..zar.2 r-u. ratics r.2r : ts moncluded tha t the instrument design is acceptable.

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The high reactor building pressure trip se: ting limit (4 psig) provides positive assurance that a rea::or trip vill occur in the unlikely event of a steam line f ailure in the reactor building or a less-of-coolan:

accident, even in the absence of a low reactor coolan: systet pressure trip.

f.

Shutdown bypass In order to provide for control rod drive tests, cero power pnysics testing, and s:ar:up procedures, :here is provision for bypassing certain segments of :he reac:c: protection system.

Tne reactor protection systec seg=ents which can be bypassed are shown in Table 2.3-1.

Two condi: ions are imposed when :ne bypass is used:

1.

By ad=inistrative control the nuclear overpower trip se: poin: must be reduced :o value $ 5.0 percent of rated power du.ing reae:c shutdown.

2.

A nigh reac:or coolan: systet pressure : rip se: point of 1720 psig is automatically imposed.

The purpose of the 1720 psig high pressure trip se: point is to preven:

normal operation with par: of the reactor pro:ection sys:e= bypassed.

This high pressure trip se: point is lower than :ne nor=al low pressure trip se: point so that the reactor cast oe : ripped before the bypass is initia:ed.

Tne overpower trip sa: point of < 5. 0 perc ent prevents any significan: reactor power fro: being produced unen },erf or=ing the physics tes:s..Sufficien natural circulation (5) would De available to remove 5.0 percent of ra:ed power if none of the reactor coclan pu=ps were operating.

References (1)

FSAR, Section 14.1.2.3 (2) TSAR, Section 14.1.2.2 (3) FSAR, Section 14.1.2.7 (4)

FSAR, Section 14.1.2.8 (5)

FSAR, Section 14.1.2.6 E

--. (6) Technical Specification Cnange Request No. 31, January 16, 1976, and Technical Specification Change Request -No. 84, June 23,1978. _

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.'(7)

"ECCS Analysis of -B&W's 177-FA Lowered Loop NSS," BAW-10lO3, Rev. 2, Babcock and Wilcox, April 1976.

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TAnl.E 2.3-1 1

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REACTOR PROIECTinti SYSIEli TitlP SErlitlC l.IlllTS Four Renctor Coolnitt Tlirce Renct or Coolnnt One Renctor Coolant pumprt Opernt(ng Pumpn Opernt Ing Pump OperniIng in (Itom l un t Opernting

( flom i n a l Opernting Ench I.nop (flominal Shutdown power - 1001)

P,ower - 75%)

operntIng Power - 49%)

Bypann

1. Nuclear power, finx.

105.5 105.5 105.5 5.0(3).*

% of rnted power

2. Nucient pc'scr bnned on 1.0fl tiren flow 1.pl1 timen flow I.0f1 timen flow minus Bypassed flow (2) ni.d imhninnce minun reduction due minun reduction due reduction due to mnx. of rated power to I mbn is.nc e to Imbninnco Imbalance

3. Nuctent power honed f3A 11A 91%

Bypassed (5) on pump monitorn, Max. % of rated power

4. Illgh renc t or coolnnt nyn-2100 2.100 2100 1720(4) tem prenure, pntg max.

S. I,ow renetor coolant nyn-1900 1900 1900 Bypanned tem prennure, pntg min.

6. Varinhle low renctor (11.75 Tout-5101)(1)

(11.15 Tout-5103)(1)

(11.75 Tout-5103)(1)

Bypassed coolant nynt em pren-nure pnig. min.

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7. Renctor coolnnt temp.

619 619 619 619 F., Hnx.

fl. liigh Renctor lhallding 4

4 4

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prennure, pntg, max.

(1) Totit in in degreen Fnhrenheit (F)

(2) Renctor coolnnt nyntem flow, I (3) AdminIntretively controlled reduction net only during reactor nhutdown (4) Automatically net when other negmento of the RFS (an npecified) nre hypnn*cd erlp on:

(n) lonn of two reactor coolnnt pumpn in one renetor coolant (5) The pump monitora nine produce n

loop, nnd (b) lonn of one or Iwo renctor coninnt pumpn during two pump operntIon (6) Trip nettings limitn are netting limit n on the net point nido of t he prot ect ion syntem hintrthic connectors.

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I 3PERATION T = 619 F r_

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1500 540 560 580 600 620 640 Reactor Outlet Te=perature, F

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i1NIMUt! CONDITIONS FOR CRITICALITY r

Applicability

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Applies to reactor-coolant system conditions required. prior to criticality.

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themagn[tudeofanypowerexcursionsresultingfromreactivity a.

To limit insertion due to moderator pressure and moderator te=perature coef fi-cients.

b.

To assure that the reactor coolant sys:em will not go solid in the even:

of a rod withdrawal or startup accident.

Tc assure sufficient pressuri:er heater capacity to main:ain natural c.

circulation conditions during a loss of effsite power.

Specification 3.1.3.1 Ine reactor coolan: te=perature shall oe above 5250F excep: for por: ions of low power physics testing when :he requiements of Specification 3.1.9 shall apply.

3.1.3.2 Reactor coolant tempera:ure shall be abovc DT! -10' F.

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3.1.3.3 L' hen the reactor coolan: temperature is below the minimum te=-

.. _ perature specified in 3.1.3.1 above, excep: for portions of low

,7 power physics testin; when the requiremen:s of Specifica: ion 3.1.9 shall apply, the reactor shall be suberitical by an amount equal to or greater than :he calcula:ed reactivity in-ser: ion due to depressurization.

3.1.3.4 Pressurizer 3.1.3.4.1 The reactor shall be maintained suberitical by a: leas: one percent Ak/k until a steam bubole is forned and an incica:ed water level between 60 and 385 inches is established in :ne pressurizer.

(a) k'ith the pressuri er level outside the required band, be in at least E0! SHUTDOk'N with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3.1.3.4.2 A minimum of 107 kw of pressurizer heatr.rs, from each of two pressurizer heater groups shall be OPEEAELE. Each OPERAELE 107 n ^--

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kw of pressurizer heaters shall be capable of receiving power from a 480 volt ES bus via the established manual transfer seneme.

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(*a) k'ith the pressurizer inoperaole due to one (1) inoperable '

' emergency power supply to the pressurizer heaters either

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r-restore the inoperable emergency power sup' ply within 7 r

O-Jaysier?.be in -at least. HOT STANDBY. within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ~

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(b).Vith -Ibe' pfessurizer inoperable due to two (2) inoperable

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emergency power supplies to the pressurizer heaters either restore the inoperable emergency power supplies within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or oe i.n at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.1.3.5 Safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality with the follow-in; exceptions:

a.

Inoperable rod per 3.5.2.2.

b.

Physic s Testing per 3.1. 9.

Shutdown margin may not be reduced below 1% b.k/k per 3.5.2.1.

c.

d.

Exercising rods per 4.1.2.

Following saf ety rod witndrawal, the regulating rods snall oe

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positioned witnin their position limits as defined by Specifi-L_

_ cation.3.5.2.5 prio-to deboration.

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At the beginning of l.._i.fe._of. t_he initial. fuel cycle,j:the moderator tempera.

ture coeffa.cient'Is expe,c,,ted_-to,be slightly, positive at,toperating tempera a b..-

_- tures vita the operating cozifiguration of-control. + ds.'

(1.). Calculations. - -

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show that above 5250F.'the positive moderator coefficient is acceptable.

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Since the moderator temperature coefficient at lower temperatures will be less negative or more positive than at operating temperature, (2) startup and operation of the reactor when reactor coolant temperature is less than 525 F is prohibited except where necessary for low power physics tests.

Tne potential reactivity insertion due to the moderator pressure coefficient (2) that could result f ro= depressuricing the coolant f ro= 2100 psia to sa:uration pressure of 900 psia is approximately 0.1 percen: Ak/k.

During physics :es:, special opera:ing precautions will be taken.

In addi-tion, the s:rong nyative Doppler coefficien: (1) and :he small integrated A k/k would limit tne magni:ude of a power excursion resulting fro: a recuc-tion of modera:or densi:y.

The requirement tha: the reactor is not to be made cri:ical below DT!- 10 F provide: increased assurances tna: :ne proper relationship between primary coolan: pressure and temperatures will be main:ained relative to the 2TI of the primary coolan: system.

Hea:up to this te=perature will be acco:clished by operating the reac:or coolan: pu=ps.

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If the shu:down margin required by Specification 3.5.2 is maintainec, enere C =_

is no possibility of an accidental criticality as a result of a decrease of coolant pressu:4.

The availability of at least 107 kw in pressuricer heater capability is sufficient Oc maintain primary systet pressure assuming normal syste heat losses.

Emergen:y power to heater groups 5 or 9, supplied via a manual transfer scheme, -assures redundant capability upon loss of off site power.

Tne requiremen:s that the saf e:y rod groups be fully withdrawn before criti-cality ensures shutdown capability during startup.

This does not prohioit rod latch confirmation, i. e., withdrawal oy group to a maxi =u: of 3 inches withdrawn of all seven groups prio to safety rod withdrawal.

Tne requirements for regula:in; rods being within their rod position limits c

ensures that the shutdown margin and ejected rod criteria at hot zere power are not violated.

c REFERENCES L -~

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FSAR, Section 3.

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FSArt,, Sec tion 3.2.2.1.

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3.1.12 Pras:urie:r Power Oparctcd R21iof Valva (PORV) cnd Bleck Valva "j:: -

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ldepressurir.ingtheReacterCoolant To prevent the possibility of inadvertently overpressurizing or Syste=.

Soecification 3.1.12.1 The PORV shall not be taken out of service, nor shall it be isolated fror. the system (excep: tha: :he PORV say be iso-lated to li=i: leakage to within the licits of specification 3.1.6) unless one of :ne folloeing is in effect:

Hign Pressure Injection Pucp breakers are racked out a.

or KJ-V16A/3/C/D and KJ-V217 are closed, b.

Head of :he Reactor Vessel is removed.

c.

T is above 320*F.

aye 3.1.12.2 The PORV se::ings shall be as follows, wi:hin :he tolerances of _ 25 psi and - 12*F:

Above 275'T - 2450 psig --

Below 275'T - 4E5 psig 3.1. 12.3 If the reactor vessel head is installed and T is $275'F, aye High Pressure Injection Pump breakers shall ne: ce racked in unless:

EJ-V16 A/3/C/D and MU-V117 are closed, and a.

b.

Pressurizer level is < 220 inches.

3.1.12.4 PORV and 3 Lock Valve The PORV and the associated block valve shall be OPERAELE during HOT STidO3Y, START UP, AND P0k?.R OPERATION:

Witn the PORV inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore a.

the PORV to OPERA 3LE status or close tne associa:ed olock valve and remove power from the block vsive; otherwist:,

be in at least HOT STANDBY witnin the nex: 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the PORV block valve inoperable, within I hour either restore :ne PORV block valve to OPERA 3LE status or close :ne PORY (verify closed) and remove power fro = the PORV; otherwise, be in at least HOT STANDBY within the

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next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 y~

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c.* With either the PORV or Block Valve inoperable, restore the operable valve to operable status prior to;startup _

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.from th.e next cold shutdown unless the cold shutdown..n_.;;,._

2

~ :5 Z; occur,sXvithin.90 Effective ~ Full Poweir Days T(EFPDF of ud-2"E 9 m :..

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WW.theYenE'o02the*fdeFeycleTIf a sold ~ish~ut'dsvn occurs"'~~~i-' " '~~'~

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within!this90,dayperiod,restoretheinoperablevalve 5."-

--to-operable-status @rior-to the startup' for the-next - --

fuel _ cycle.

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If the PORV is removed from service, sufficien: measures are' incorporated to i

prevent overpressurization by either eliminating :ne high pressure sources or flowpa:ns or assuring that the RCS is open to atmospnere.

In order :o preven: exceeding leakage rates specified in T.S. 3.1.6. the PORV may be iscla:ed.

Tne PORV se:poin:s are specified with tolerances assumed in :he bases for Technical Specification 3.1.2.

k'i:n RCS te=peratures less than 275'T and the makeup pumps running, the high pressure injec: ion valves are closed and pressuricer level is maintained less than 220 inches to preven overpressurization in the event of any single failure.

Both the PORV and the PORV block valve should be cperable during tne HOT STAND 3Y, STARTUP, and POWER OPERATION.

If eitner the FORV or the PORV block valve are inoperable the PORV discharge line should oe isola:ed to prevent po:ential uncontrolled RCS depressurization.

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3.4 DECAT ESAT REMOVAL - TURBIriE CYCLE

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E Applicability "tX;4_p c:

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r cr::t Applies: to the operatangytatus of ' equipment that functions.to remove decay.

_.c-D bear',"utilising ttij7econdary -sidiof thiStiaiiIGeneraters7 ' '

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Ob iec tive To define the conditions necessary to assure immediate availability of the Emergency Feedwater (EFW) Syste= and Main Stea= Safety Valves.

Soecification 3 4.1 With the Acactor Coolant Syste= te=perature greater than 250*F, three independen: EFV pu=ps and associated flow paths

snail be OPERASLI with:

Two EFW pu=ps, each capable of being powered froc, an OPERAELI a.

emergency bus, and one EFW pu=p capable of being powered fro = an OPERABLE stea= supply syste=.

te.

With one pu=p or flow path

inoperable, restore the inoperable pu=p er flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in COLD SHL*TDORi within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With = ore than one EFW pump cr flow path

inoperable, restore the inoperable pu=ps or flow paths *

o operable status or be suberitical within I hour, in at least HOT SHUTD0k"; within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOR; within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

Four of six turbine bypass valves are OPERABLE.

d.

The condensate storage tanks (CST) shall be OPERABLE with a =ini=ur of 150,000 gallons of condensate available in each CST. With a CST inoperable, restore the CST to operability within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOR; within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and COLD SHUTDOWN l

vithin the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . With = ore than one CST inoperable, l

restore the inoperable CST to operable status or be subcritical

(

vithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , in at least HOT SHUTDOR; within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , l

and in COLD SHUTDOR; within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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For'the purpose of this require =ent, an OPERABLE flow path shall =can

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un unobstructed path frc= the water source to the pu=p and fro = the pu=p

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to a stea= generator W.% w ;; ~-

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3.4.3 With tbt reactor coolant sys:ec temperature greater than 250*F, all f

eighteen (18) main; steam safety _ valves shall be operable or,- if any.

are not operable,;the maximum overpower trip serpoint' (see. Table C.. _. m J 2.3-1). shal Mbe] reset)s; f oliows i s.;G...

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GenerateF '

(T of* Rated Power) 1 92.4 2

79.4 3

66.3 With more than 3 main s: cam safety valves inoperable, restore at leas: fif teen (15) main steam safety salves to operable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months er be in at leas: HO~ STANDSY within the nex: 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SCDOW within the f ollowing 30 heurs.

Eases A reactor shutdown following power operation requires removal of core decay nest.

Nor=al decay hea: removal is by the 'stea= genera: ors wi:h the steam du=p :: the condenser when RCS te=perature is above 250*F and by :he decay heat re= oval sys:e= below 250*F.

Core dacay heat can be continuously dissipa:ed up to 15 percent of full power via d e s:eae bypass :o :he con-denser as fewedva:er in the s:cac genera:or is coverted c s:ca= by heat absorp: ion.

Normally, the capabili:y to return feedwater flow :: :ne stea:

genera: ors is provided by the main feedwater systec.

7.,::

Tne main steam safety valves will be able to relieve to atmosphere the :otal stea= flow if necessary.

If Main Steam Safety Valves are inoperable, the

~

povar level mus: be reduced, as stated in Technical Specification' 3.4.3, such that :he remaining saf ety valves can accomodate the decay heat.

In :ne unlikely event of complete loss of ef f-site electrical power to the station, decay heat removal is by either the steac--driven emergency feed-wa:er pu=p, or two half-sized motor-driven pu=ps.

S team cischarge is to :he atmosphere via the main stea= safe:y valves and con:rciled atmospneric relief valves, and in :he case of the :urbine driven pu=p, froc the :urbine exhaus:. (1)

Both motor-driven pumps are required initially :e remove decay neat with one eventually sufficing.

Tne minimu= amount of water in the condensate s:orage tanks, contained in Technical Specifica:ic,n 3.4.2, will allow cooldown to l

-~

250*F wi:h stean being discharged to the atmosphere.

After cooling to g.- _

250*F, the decay heat removal system is used to achieve f urther cooling.

~

~~

An 'un-limited emergency feedwater supply is available from the river via j

either of the two motor-driven reactor building emergency cooling water f.

pumps for an indefinite period to time.

Q_... _ _ _ _

.... - ~. -

w = w :.=-us w w : :.- e x::-c.=: m.-=- *

. w. ' W.-

2--

,e 6

e T.1 ~*. ~

-.~

, -d;, ' 'J L1 *.' Z.0"."'"10' -- T'."'JC.G.C':'LC C211 L;. ' ~

l f

... ~

...u.-enwe.-va

.'sM.-

2 - " "

ms. e... +.%.. : v.~.4.

6

=... =.

~.. -

^^ *

,~

W f.

.5P4W4..-

" +-

O*-*

I q

. l 4-e

_..,y...

s'. 7 s

a4,4

.b 3-1 h* * *- '

-.. d;* % =

-Ih s.&. p.++ v a%A.G N)$

e%' - *

&^r.-*-*=

O e.w

~~1.E*'~=

The rcquirements of Tcchnical Sp2cification 3.4.1 nscure th t b2for2 th2 reactor is hested to above 250 0 T, adequate anviliary feedvater enp city

--L is available.

One turbine' driven pu=p full capacity (920 gpm) and the g

rvo half-capacity motor-driven pumps (460 gym,each) are specified. How-T ever, only.one. half-capacity 21notor-driven pump is.necessary_.to_ supply

~

_ c:_ :r

[h.T--95 auxiliary ~feedvateirMcE*Moibe steam generators in the' onsefof r:small # ~ 'l.1 -

~

break loss-of-coolant accident..

=f--

~

=

-dgg=~.i ay.

j

,r gggg (1).

FSAR Section 10.2.1.3.

1

=7 l

l l

i l

l I

1.

NC,-

l{:-

?..

.C ~

h

. ~;

.' j.] --~ ' ~ p*h*:' y f r ] ' ' ' D ~ y { ~h'"'M,yT ' F - '

C _^

~

~;.

_,.~ =::::.- ;- -

-m ::- =-

d f['b.*[b

~ L

~

_., '.. = - _

.~k ' * ' 1.-

.' ~. -;...R f.

Hr.-

.:P. 5

qq:;-

g

- f

,', y';

A

}-2hg 2'

= ( _..

4 +.

-3.=_.'3:;i c,.

-4.%..

--s1.~. - i

.-} w+ gg

., -, %:%.~ 4,

~. a i r.o; F,,w w.

e v

reactor coolant fempera*.ure instrument channels, four reactor coolant flov 74 ;.1 instreent chnnnels,- four reactor coolant pressure instrument channels, four pressure-temperature instrument channels, _ four flur-imbd ance flow instru-f@p=_

5Huc. -. ment r hsmnelss. fourt.powersumberiof pumps instrumenL channels..and four high. _

"'" reactor ~buildidiiressu'rMnstrument' chanHels.MTheTesetibrNrip',*$n~' loss of"~ ~ ~

t pr feedvater may be bypasse kelow; 7% : reactor power. The hypass is automatically W':-

removea when reactor Powec ishaised above.7% 6_ The reactor trip, on-.

turbine trip may be bypassed below 20% reactor power. The safety features actuation system must have two analog channels functioning correctly prior to startup.

The anticipatory reactor trips on loss of feedvater pumps and turbine trip have been added to reduce the nu=ber of challenges to the safety valves and power operated relief valve but have not been credited in the safety analyses.

Operation a: rated power is per=ited as long as :be systerts have at leas:

ne recundancy recuirements of Celu n "I" (Ta:le 3.5-1).

Inis is in agree-vi:h reduncan:y and single f ai.ure criteria cf IEII 27c as ces:ribed in men:

F5AK Se::icn 7.

ou ci Tnere are f our reactor protec:icn channels.

lic-=al trip logic is tv four.

Recuired : rip logic for :ne power range :.ns:rumen:a: ion enannels is v: ou: cf :nree.

Minica: trip logi: on other instrumentatien :nanne'.s is one ou: of :vo.

f our reacter prote:: ion channels were previded with key ope:::ed oypass The allow on 'ine testing er maintenan:e on only one enannel a:

suitenes ::

i=e durin; power operation.

Each :nannel is provic ed alar: and '.ign:s ::

indica:e unen :na: cnannel is by passed. Tnere vill be one rea::c prote:-

tion sys:e= bypass switch key per=i::ed in :ne control room.

Each rea :c pro:ec: ion ct.annel key opera:ed shu:down bypass svi; h is pre-vided with alar: and lignts c indica:e unen :he shutdown bypass switch is oein; used.

power is ner.:. ally supplied :c :ne con:rel Tod drive mechanis=s fro tv0 separate parallel 460 vol: sou rc e s.

Recundant. trip devices are employed in ea:h cf these sources.

If any one of :nese trip cevices i s.: i s in :ne un-tripped state on-line repairs :: :ne f ailed devi:e, when pra::ical, vill be cace, and :ne remaining trip devices vill be es:ed.

Eight nours is a:ple time te tes: the remaining trip dev::es anc in cany cases make on-line repairs.

r. : L z.: Lw;

~

F5AE, Sect. ion 7.1 a-g_......

=~,-.

p.3 y

s n eph-. -

' pe.

q s.

m

=

...L L.

_._w

..:r.,,

m g.., 6 '

2

' g: g. n.;

' w.p;r re g...

7.-

^~

^

ur..

= =:

W 3.m_ s

-: : X:.,.

f

,2-

_z_.,

P.

~u.'

.+ -.

..,.*m....

w.

~ ~

e- ---. w. w v.

- ~_..

-*Ji*.

N 3-28 "N

}

a. 3 s-v.

e s

y.

5

.s J'.

e w v

me.

r.. _,.

$1

'q h", 7 ; e M,i fd.

fr1H f J.$ r ihil

9. ': b j,,'

j@

.A..,.

q,4 TAHl.l; l.5-1

]

Us

,]

4 k, y" '

p,,. U n INSTitlif1EtlTS OPERATING COllDITIONS

{' ",

y![ jiT.

{..'p; j,*,.j'h{[!

..m 8

' ?:d-(A)

(H)

(C)

.i' o

I

@' Uj flinimum Operable Plin imum llegree Operat or Act ion if Conditions U

beHety.y,'

,[ [;l,! Functionni linit Channeln i of itedundanc y of Column A Cannot

! i

.9

.d.'

5L g ]'

l

[

A.'

Reactor.rotection Syntem

,pp il$'.iT 1

0 flaintain hot shutdownt'

'f !N*

y

h.

i

2. '! ' !

!!q :l,..",I'Hanual pushbut ton j-

,,.inM4 +.

.J.

i

'}

lr I,Y i,[2.[Powerrangeinstrument 1(n) 1(a)

Itaint nin hot shtstdoNq'IDf? dp ', '

r t -

4

. S' l,.

KM ". q(lf.! ;i ' fl s,-

['

channel

.1

lo

' tr.

shu[down,.

j(b)

. M. i, ' 8 Nly]

' 3.

Intermediate range 1

0 Plaint.nin hot l

inntrument clinnee l n f

iib: 8) !

nhutdown (c)n U t

- y 4.

Source range inntiument l'

0 tlalut.nin hot i,

f, cluutncin 3

P':

5..

Itenetor coolant temperat us e 2

1 Plaintnin hot nhutdown.'

ia'

. f, '

- ?,

inntrumLnt channeln irF

((I

, yI

~,

6.

Prennure-temperature 2

1 Plaintnin hot shutdown i*

inntrument chnuneln

. L

[,i.

' ((. '

7.

Flux / imbalance / flow 2

1

!!nInt nin hot shutdown, inntrument channeln 8.

Hcact or coolant prennure ii

'1l'l n.

liigh reactor coolant 2

1 Plaintnin hot shutdown' prennure instrument

?;

f;

%- /'

(;'

channels V

s b.

l.ow reactor coolant 2

I flaint ain hot shutdown

,. 7 I

prennure innt rument

,k channeln

\\;

e,t v -

4, k ')I!

i i ;...

i

1 ef j

. p.mtiR'i,

, i T, e W;W,,.h. i,

,.a-Jg!';$'k-@J.Mt.

,n; if '

e e

~,h.

s h, y r

'f jh p

$i) * ':

h'.;W' p b

'klotet M' R N;.df':[J T Altl.E 1. 'i - 1 t'on t i nne.I

',M,W f

3

,g r ar

- s p;,

IflSTittitlEtt rS OpElf A rifH: Coilu l T I OrlS

e

.4

c f (A)

(ll)

(C)

N

. I'. N h.,',1 h]hFunctioHIntUnit

"'*"""I"'f"i'i"

" I " I "" "n llen i ec Operat or Art lon i f Ceni ll t f onn Clinnn e l s of Itcilun.Inney of Column A Cansint lieillet

f

p p

i btl.

. d; h,d. '

t M

s i

A

,' g ;

.j A.

It.e.n_c.t..o..r...l'.r.o t e c t l.ini. S.y n t ein (con ' t. )

j a.

cg1 9..

Power /inunlier. of giumpn 7

1 tin l u t n i n lio t-nluitilotin ! 4 p

g 4 TddV i.

% bliintrument clinnneln

j hJP3 p8M 2 t",I a

~,

Qt,

d di' 2

1 lin ing n li, 3,ot' nlus I h

'l

10. ;'llikli renctor luillillon

. E f yd

},

q

.yll!q,rennure clinnneln Jp. Kh p

I r! 4 jiij uj.

(p*}'

'{ !(of For clintinel t ent Inn. cul lbrat init, or maintenance, t he minlmnm number of opernli.le clinittlej n h b' hj fi gr i

lh

j?

y may be two anil n itenree of s edunelancy of one for a maximum of four hoorn.

(li tJhen 2 of 4 power ranne Innernment channeln are nrenter than 10 percent full power,

.9i i

is hot nhnt down in ont i c pil e cel.

A

,t

.*. t.

Y N}.NI o

, h ;' '

' (c) tJhen I of 2 In t ei rneill nl e rnnne innt riement rhnnueln in greater ilinn 10"IO or 2 of'

-?

nmp rt,

ninitdown f(h h b y 3 power range inntrument channeln are grent er thnn 10 percent full power, lio t.

/ p.g,

'[ ;dy;,0[ ;i

-i

)

not requited.

31-

'1 11.

Other Itenctor Tripn

.t

.,i y,

g&. 'M, V s

J 1.

1.nnn of Fecilwnt er 7(n) 1(a)

!!aintnin lenn than T 7%j intlicated reactor power, h'.];,p:

i I

2.

Tiltlilian Tt Ip 2(b) 1(b) tinintalu lenn llutn. 0%?,inillented reactor power. !

.?

t-tj a (n) Itypann of t he f eedwnt es pump t rip ninnat may be placed in effect when Ind icat ed tenctor powerp, jlh Y' If hib' in lenn than 7%. The bypann will be iemoved wluo rearlor power in rained above 7%

i

'(b) The main turblue I rlp bypasr may lie placed in effect when Indient ed reactor power is leds Lljati. 20%

M The liypann will lie iemove.1 wiien the s ene t.o r powe r in inc renne.1 niinvc 70%.

,m 0

<l m

i b

',i l

ldi

'MP fp f H7 't hpg p 9:p

M,,

a.I s p$

n t,1,11 1

t 7 cl::i..

3. i..

hi-5 Tant,E 3.5-1 Continued y,[.

li INSTRilMENTS OPEltATING CONDITIONS f.

1

,; p d.

$jy;,

gji,,', Funct tonal linit (A)

(R)

(C)

. p, ' i Mp Minimum operable Minimum Degree of OperntorActionifCondItions Q'%

$4 p;.9 j

C.q Engbicered Safet.V Featuren Analon Channeln Redundancy of Column A and B Cannot be,llet(a)

{

I 3jftLi f

~ ':i ^ :.

! !1i'; ;kakeup an'd Pufification iif 1 cd ; {j 'h 1a f,hllhaSystem(highpressure 5 ' ';'s M,N

.T inlection mode)

@ h'

/

ty.

Reactor'Coolnnt Prennure 2

1 Ilot Shutdown E. ',;g ;;'

' I:

6

'i '

3' O

,H Instrument Channels J

f..

f ;. ti.

. \\.n 6

[

t ;i

t

"(/ { u]

i,g,;,.

is P;,

j;;

. h, React 6r Building 4 pn.lg 2

1 Ilot Shutdown pi

,' A >

i.

h.f Frigh}@k,h:

!M 4b' Instrument Chnuneln h,' 4 hl (4' h'ij'Nf if ' ' l I 9 (r

' [1 c.

Manual Pashbutttui (b) 2 1

Ilot. Shutdown 3V sp u

s's 2..

Decay IIcat System (low ic 9(447 pressiere injection modg)

,. Pl o g

3 r d pp H q l.

a

. w.iM.

ja.

Itenctor Coolant Prennure 2

1 Hot Shutdown p' ' p.{!(

c

+ yG Instrument Channeln t) q j!.s

.[

h.

Henctor Btillding 4 psig 2

1 110t Sluit down I

d $ ~.. ~ '

Instrument Channeln

$ F<b. ' ' I '

i !

T.

f tg t

Open circielt breaker; ', ' [

hi' i

c.

Henctor Coolant Prensure 1

0 3g HCC for DII-V1 or -

L-ir nt D.ll. Valve Interlock Dil-V2withtheaffected.;Ij,1 1

Hintable vnive in the cloned j

positton or maintain

' II 9 y;'j,;,..

.t.

R.C. prennure less s

+

,p I

p than 350 psig.

L gif ilj i O

+.

s%u t

'ii

!:.h ir

' O.

ani r,.,<

p

[*{

g.; '. ' g ;-

p i.g,;f '

1 WI

+

TAlli.E 1.bl Cont inneil t

p

'e j

', 4..;'y,

,p! }

IflSl'ltittil'.fil'S Ol'Elt AI Iflu Corliil I Iotan

i, (m

' [i Feinct innal fini t (A)

(n)

(c),

p I

filn imum Ilpei nlile tiln imum Degr ee of Operntor Artlon if Cnndlklons l

nfColumnAnndIlCnnnn((heHet

$"ll!.ll"" F"'I 8"I _t y,,fp a t ni e n con ' t.

Anning Cliannoin Itcitunelancy f

C.

h ( ' ![f I.

sip

}

j. y:llenct or 11ullding Innint lon

i k.'

' inul Itenctor lint id ing Cool lnh lo;N, n' ir.

6 V

e a.

b'

'., WI Rynt em gt >p

.f!)%{}' yIIJh m-l d

H

r u.

ftenctor Bulliling 4 polh (P,)

2 I

Ilot Sinst down

' f-y:;n.

T M.4 $'

~

innt riiment Chann.1 a o 4

I f;fh 9

l,

n,Ii.

ifanun! Punhhutton 2

I Ilot Simi down

,V Thh t W.

"} : p t;d",

n-

i. i. ;

1*

', : J. f+j,

U 2

1 IInt Shutdown

.c.

ItPS Trip

ttT 37:,

'h M.q,k c

d.

Itenctor Bu t tiling in enin 2

1 Ilot Simt down I

t?.

ItCS Prennure lenn than 1600 pnin 2

I flot Simt slown i.,

ti r.

..lf;; h l e,

i f.

Renctor Butiding l' urge line 1

0 (f) i

]f.' h id' Isointion ( AllV-1 A and AllV-ID) f]!

llinh Radintion

',I 4.

Renctor Building Spray Syst.cm I.

),i%

i

; f. p Renctor 11utiding 30 2(d) 1 Ilot Shutdown 4;'

f

,n)w. /a -

n.

psig inntrument Chatine l y

d4 b.

Spray Pump finnun t Swit ches (c) 2 1

llot Shutdown

}.'

'f[,'}i'i;s h, '

I

O S.

4.1r>KV ES Eun Undervoltage Reinys 7

N df 2

1 (c) f Degraded Grid Voltage Reinyn n.

2 1

(c) iI J;

b.

I.ons of Voltage Reiny u

y l '-

..; y 6.

Emergency Feedwater System

'I 7-1 llot Shutdown f

f.

n.

Loss of bot ti l'ecilwater pumpn e

b.

I,on g of n11 RC Pumpn 2

I Ilot Shut down

-J'

,i dnP II '

>H j ! W. {'

2

.l; p,.

.;y<4thP/

jk'

>2 Tant.i.: 1.s-l coniInned f

g 5.s 111STHiltlF.filS OPI7.RA ritH; CutlDITIOtlS

'. ig :

s.

t.

.h., '

t p

/ "-

'{ jt ',

" ! Fudetf loon t llol'l n

N..h 1.[,..'..

i

,4 s,

f C. - - Engitteel ed Sa t el.1~ FC88t "I"'8 (C""'I * )

1 I,.

4 gh,, '

'y M (n)

If minimum conditI"nn nin not met within 24 honin, the un i t. nhall t hen he ptncce1 in n toid I ninitdown condit. inn.

'{

. ;,; p;,

4 y

97 l[g'J.d,Mkj

,i ! !L Alno initinten 1.nw prenourc In. lect lon.

. h it '.ip M.' h(b)

--$'t,a.

I$

e n

llated in I t m 1 niinve.

I c) Spray valven opett"d liy siuniunt punlibut ton Two nit of tlirce nwit riien in cach net ont ion chaninct opernhtc.

T b nin n I s e ' "I

"h'""

I' r_'!

,. Q:

to 2 hottro f or f unr.t lonal t est ing pornunnt to Tnble 4. bl.

i.

Note:

(a) above does not, apply if Discontintie Reactor nullding piirging and clone AllV-l A and 11).

',,#(f) t c1,'

y' AllV-1 A and AllV-ID are closed.

' y,}'.l..J

~

4hb (g)

For hot fiinctional testing, prior to Cycle 5 criticality the 4 psig signal is not required IN Nuclear Service Closed Cycle Cooling water, Intermediate Cooling and Reactor Coolant Pump Seal Injection (return

{

line only).

'Ivo opernbic channels of a 30 pnig Reactor ni:11 ding inolation signal with a minimum degree j.

of redundancy of I are required if the 4 psig signal is not operable for these lines.

..b]g 2

V p

^. c o.

r ij<

Un

.I

[. hi i

', h.

r,

?

, t.e.

4

.. f,

~

Yi e

t

W.

s.

t.

i

'.f, a

!! ~

b:

si.

k..s I.

1 4

DCIm.:.rD SATIL"DARDS ?ROTECTION STSTDi AC7DATION SITPOINTS 3.5.3 c

Acelicabiliev:'

..a. w -.

E:._ This specification app 3.ies~2o zhe~ engineered safeguards proteccion ses:em

~...

Wiw&W++ A "

.~m -

W--

. I ~ aecuation se:pointsh M E9F % ATf T =

7. u,

. p,:n.p =.:. =n..=_._.. _.

-.w~-.-. -,.

. u. _

~-

- Obiecrive:

-i ~ "m% W- - - -' - " ^. i. _ u. ;- ;;. c _.

Tc rrevide f'er aurcsaric inia [aticln of :he engineered safeguards protection systen ef a breach of Reac:cr Coolant Syste= integrity.

in :he even:

See:ifiezeien-

'"he engineered safeguards protec:icn systen actuation se :cin:s and 3.5.3.1 ee nissible bypasses shall be as fellevs:

Sert in:

Initiatint Sirnal Tune:1en Reac:c luilding Spray

< 3C psig Eigh Reac:er Suilding Pressure (1)

Reactor Building 7 30 PsiS

.sola:1on High-Pressure Injection

< 4 psig M v-?ressure Inie::icn

< 4 psig Star: Rea::c: Suildint Coeling & Reat:er Eu:.1 ding Iselatien

< 4 psi;*

-L 1.ov Kea::cr Coclan:

Eigh Pressure injection

> 1600(2) and Syster Pressure

> 500(3) psig 1.cv Pressure 1:4ection

> 1600(2) and

> 50D(3) psig Reactor Building

> 1600 PsiE (')

Isolation 4.16 kv I.S. Euses

.. c..in e: ve.: age ae.mys Degraded Vc1: age (5):

Sri:ch :: Onsi:c ?over Source and icad shedding 3* 3 vel:s (4) 10 se: (5 )

Degraded g-id timer

~

Sei:ch c Onsite Power 7

k ss of voltage Source and Icad shedding 2400 Telts (6) at 4:

c 1.5 see (7) 5 Loss of voltage timer 4

L (1) lizy be bymasted for rea::or building leak ra:e. test.

n.;

J(2)[Say be bypass'ed belov 1725' psig on decreasing pressure and is

.,a..-

~

~ ~

.. n =. ~

C <f I E N l_

automatically reinstated.before 1800 psig on increasing pressure.

.a.

r*

Sav be bypassed belov S75 psig on decreasing pressure and is

. ':y W.

(3)

~

c50 psig on increasing i

au omatically reinstated befcre exceeding pressure.

i-%-

we~

nty w.\\

3-37 u ^

m

3'

~ ('4) M.inimum allowed 3ertingv3560 v.W== aI16ved cctring is 3650 v.

M*nd-allowed time is 3 sec. un4mus allowed time is 12 sec.

.(5)

-. x 'n,-(6)

F4=4

.. allowed setting,1s,2200 volts, -4=.=

allowed serring is 2860 val =s.

.._s-w.

n.-.

F4"4-7 allowed time $s (1.0) second, ~w'. _...-= allowed time is (2.0) seconds.

~

n

= - _ -.

g~7 ; _3,,

i d...(7)

c..- f z:.

For lio: Tune:1onal TestinE prior to Cycle 3 criticality, the 4 psig Reactor Building isolation signal is not required for Nuclear Service Closed Cycle Coeling water, Intermediate coeling varer and Reactor Coolan: Purp seal injection (return line only). Re=ote Manual and 30 psig Reac:or 5uilding iscla: ion signals are required if the 4 psig signal is no: cperable fc:

these lines.

Eases liirn Fase:cr Suildint F~ e s su re Tne basis for the 30 psig and 4 psig se:pcin: fer the hign pressure signal in adecus:e ::.me :: :ne is to establist. a se:.ing unien would oe rea:ne De far enough abeve spe::run cf break s:.re s and "e:

even: of a LOO;., cever a nor=al operation caxi=u in ernal pressure te prevent spur:.ous 2.nt::a :en.

Lev Reacter'Coelan: Sy s:er. Pressure basis f er :he le00 and 500 psig low reae:er cocian: pressure se:poin:

Ine f er nigh and les pressure injet:icn ini:ia: ion is :c es:.ablish a value unien is high enough suen that pro:e :icn is provided fer :he en: ire spe::ru: cf

~

creak sizes and is far enough below nernal opera:ing pressure te prevent spurieus ini:ia: ion.

bypass of F21 oclow 1725 psig, and 1.P below S25 psig, coolcown.

preven:s ECCS a::ua: ion during nor=al syste:

4.16 I*.' IS Eus Unde:-reltare Relavs is :: pre:e:: :he Tne basis fer :he degraced grid vel age relay se:pcin:

ef f =- less cf fune:icn in the even:

r. 'ery rela ed electrical equip =en:

~he ti:ier a sustained cegraded vel: age cenditien en the effsi:e power sys:en.

se::ing prevents spuricus transf er :c the ensi:e scuree f er ::ansien:

cendi:icns.

less of offsite power cendi:icn Tne less of vel: age relay and timers de:e::

an' d ' 'a:e transf er te the ensi:e source vi:h :.ini=2.1 time delay.

- * ~

b g.,,.

55~.

f'-,=-_

- _r:

-6.aa

),.

.sh

+....

e z

a T

"+

.1 3 37a,m... _c,

%_,.~

-m

,s....

a -

...ro..

-2

~.

3.5.5 ACCfD5NT ~ MON 1 TORI 80 13 STRUciiNTATION

==

F Applicability

..n.

n

==

=:

=. =..

i

- ~. ~ - -..

...,$Appliesho the operability requirements for the < instrument -identi e---

Nn:5$$Ein:. - ~

.= ---

fied in Table 3.5-2 during START UP or POWER OPERATION.

76 E '-

Ob_iective To assure opera..lity o., key instrumenta: ion usefu, In diagnosing ol.

situations unich could lead to inadequa:e core cooling.

Spe cifica: : c.a.

3.5.5.1 The =ini=u: nu=ber of channels identified for the instruments in Table 3.5-2, shall be OPERAE1Z. With the n>mber of instrumentation nhannels less than the mini =urc required, restore the inoperable channel (s) to OPERAEl.E status within seven (7) days (48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for pressuricer level) or be in at least HOT SHI'TDOWN within the next twelve (12) hours. Prior to start-up following a C01.D SHUTDOWN, the minicus nu=ber of channels shown in Table 3.5-2 shall be OPERAELE.

Eases Tne Satura: ion Margin Moniter provides a cuick and reliable means for da:er=ina: ion of saturation te=pers:ure cargins.

The hand calculation of satura:ict pressure and saturation te=perature

=argins can be easily and quickly performed since it only requires knowledge of RC Syste: loop te=peratures and syste: pressure, and the use of stea= tables; accordingly, hand calculation provides a suitable alternate indication for the Saturation Margin Mo itor.

Disenarge flow f ro tie two (2) p:essuricer code safety valves and

.he PORV is measured oy differen:Ial pressure transmit:ers connect-ed across elbow taps downstrea of each valve.

A delta pressure indication fro: each pres sure trans=itter is available in :ne control roo: te indicate ende safety or relief valve line flow.

An alar is also provided i.n the control roo: :c incicate that ci scha rge fro: a pressuricer' code safety or relief vclve is occuring. In addi-lon, an acous:10 moniter is provided Oc cetec flow in the PORV disena ge line.

An -a larc is provided in the con:rol room for the acousti-monitor.

F?

In the event that a delta pressure monitor or the acoustic monitor w.

becomes inoperable, access to the containment would most 114ely.oe

.g _

required; however,' a~ reactor shutdown :o allow containment g:'ig{ '-

access n-

.or enis repair is nc: justi:laple due :o :ne existence o, a,. er-4-

nate means of detecting and =enitoring code safety or relief valve

'a disenarge flow.

Tne al:erna:e means iaciude discharge line thermo-

$~ ~ '1. _

couples and Reactor. Coelant Drain Tank indications.

.A?.

=... T T.'

~'7L;=& = ~~ W':._ w __-

'=

W'8: = 5 k W I N 8..Y +,f '-

~

l':

-3x 5: y 5

=.4 3 355y..

t -

m,- g _.. 2 =,;

- - - " 3-4 0a -

rp. -

7

tA y--

e, f,,. -

r

~***. fh S $.* e.Y l *-' '-

~

r%*bY' h

.&Wahw a e

=. - * -

"~ Y N W ~

6-d' A---4 g

=.,

~

1, x:

Q:.._~'

g; F:: q.'

ih" ^

Th; Emerg:ncy FC:dw ter Systcs is provid:d with tvu chrnn21s of E=# -;

flow i'strumentation on each of the two discharge lines. Local flow 9'?:~

~

indication is also available for the emergency feedwater syste=.

-- tjE

-.. :.. = 25 - -

,.M S.L+

. z.

k_. y - - --

._uam.. Although ethe-pressurizer has multiple level indications, the sepa- - -

- -. = -

rate indications are selectable via a switch for display on a sin-gle display.

Pressurizer level, however, can also be determined y gf; via the patch panel and the computer log.

J Although the instruments identified in Table 3.5-2 are significant in diagnosing situations which could lead to inadequate core cool-ing, loss of any one of the instru=ents in Table 3.5-2 would not prevent continued, safe, reactor operation provided that the alter-nate indication is operable.

Therefore, operation is justified for up to 7 days since alternate indications are available for, Saturation Margin Monitors using hand calculations, the PORV/

Safety Valve position conitors using discharge line thermoccuple and Reactor Coolant Drain Tank indications, and for EW flow using Stea Generator level and EW pu=p' discharge pressure.

Pressurizer level has two channels, one channel f ree ICS/!C;I (3 D/P instrument strings through a single indicator) and on channel independent of the ICS/!C;I.

Operation with these channels is permitted for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months since alternate indication is available through the plant computer.

-e

?& ~

2' v

3-40b

.m.

.#m-2; e k.

~,.,

r 4

ev

~

A Y, f

u.

o'.v t.J

.

n - e! u v eu s4..'

a s* &*%

t& nek

'=*;%

r"** M A--

0*******

E^ '^ M

5W"" '

A"'*'?*~*

d.

d' N

l.

.hk n '

hI k

Np

- h juNh! I 5]jf' '

I Qi o n,,lgdy 0

i(,iei TAnl.E '). 5-2

[t, t,

i i l,!

ACCIDENT ff0NITORING INSTRUMEllTS 1- - -

. L' i'

T' as t,

t 13; E3

' d ! [' $t tt I N!!!UH C

c-

.y

/

FUNCTION'.

IllSTRUMENTS NUttilER Of CllANib'.i.S tillttilER OF CH AN_NEl.S___

i I

f 1, j Saturation Margin Monitor 2

i I

h.

tI q,"

2. 2

!! : ' MU 2,.m.

Safety Valve Differential 1 per <ltschstge line I per til scha rge ilne j' fib,'.;.

Pressure Monitor nl n)ry *L

, s 7: 4 :.5

'.. l k

8

[,

H [&H '

't : l t

.fu p,.,,ih1l.t l

t, : 't1' 1;

)<,7[hl(j j.p.

.r

.y

,k d,l 3

PORV Position-llor.i tors 2

1*

(bf 3

.-P-l a.3 !

.: :?.,

. e a.t sj y

?

r.

j' P' I{ {.{, j.'"i

! ij e)e;

' f i; yf e

4 Emergency Feetlwater Flow 2 per flow path I per flow path

i

, i.

if 'h-

-y L Ni:9 j

i.,

.r

.., d.s:.

p !J. ' * 'o.,

g)

+

5 Pressurizer I.evel 2

I h

e 1

e

$ill t'.

1 ie i :i ;', i fi Jj r}:..

ih '

jr-t, h

.(!

With the PORV Block Valve closeil in accor< lance with Specif ication 3.1.12.4.a. t he 4

l',

j t!

minimum number of channels is zero.

h.

w)i s i,

a :

p.

..,i '

3 1

2 i,

i'

.).

4,;

I

+*-

g mc.

3.20 Separation' of'TMI-1 and TMI-2 a

N i

Applicability

.c. w..x.- m.a,.

..r -.

.3.

..u%..

o..

=

g p..;t=

.m..=

m 3 g

~ ;._.

w wa=

. m.: :m -...=.x:

r. -

....., -c._-- --. - Applies to interconnectio_ns-betweenTf.1-1 and TMI-2 which'havethe '--t' potential.,or transferring _significant. quanti:les c: contamina:1on.,

between units. %.u.._., E... * -

7. '..r:7pr i-~; --

mi a

,,a-Objective To con:rol the ::ansfer of radica::ivi:y fro: IMI 2 t e TMI-i via s.vs:en inte r:ies.

S De c i,_ : c s ::en 3.20.1

.ne iscia::en cevices :or :ne sys:en :n:erconne:: ens ::s:ec in

.,,, s na,a, renain inpla:e unless vri::en approvai nas seen

.a:Ae.:.:r.

received frc: :ne 1;RC.

If approval for use of inter:ennec: ions is receivec, use sha,- proceed under prees:a:'_ishec procecures.

t 3.20.2 No additional TM -1/TMI-2 in:2rconne :: ens, vi:- the po:enzial of

ansf errin;; significan: cuan:i ies of radioz::ivity, shall ne

{

crea:ed witncu: prior NEC approval.

l t

e 1

3.20.3 Observed defeat of an isola icn devicti, which defeats the separation bemcen Units 1 and 2, vitacu: pric: l'EC appreva; shall be reper:ed l

tc the NRC as a specis1 reper vi hin 30 days if not, correcte within t

two (2) hours.

i

.ases m

i i.

u Inter::nnections exist be:veen ~'MI-1 and DC-2 that have the p :ential fer 6

transferring con arina:ic te ~'2-1 as a resui: cf res:cratic: activities a:

TM1-2.

"'hese interconnections should remain iscla:ed unless approval fer thnir use is received frc: the NRC.

l A

-w.

,.ymw ie.i f.-

e=

-.e

'cer

~

.,~~j",-_.g.

Q;"l, \\ 7 m-..

3-95a

.e.~

-;;,~g. _

. -. ~.

~

ca -

- . ;N,

m'k

-A 44 6

.42<

.. '. +

-~

(

,'b*

4f

.# A T '

it.

4*

M -..,

.p...

t Mb. *. *.

.a

. S-w"b U &

'h-

l

.. ~.,,

'h m.g I

I

"4 P

>3 v.

-z TABLE 3.20-1

.:::5= _ -

=== _

._TMI-1/TMI-2. Interconnections

.:.:':: :.&q=.:. u.:

.=.=.--.: :-

-~==.2

. L

t&,a,a.,,u 6 ic& r+: e,

.,r, g?+.-if:M,m,. ~= ~.:. T W. h.,2 r

m-a

~ --

w

~ - - -. -. - -...---

.ww g

's. q :

{1) Un:t 2 Reactor coo ant:31eed Holdup Tank - Uni: 1-Reactor Coolant Waste

.4 =-

Evaporator

-~.:.+i *=::...

=a=,--

... - v=== =. := :

~:

~,... =. -

.....~::. :

.. ~ -.---. :

(2)

Unit 1 Miscellaneous Waste Ivaporator - Uni: 2 Ivapora:or Condensate

~

Tes: Tanks

( 3,'-

Uni: 2 Neu::alizer Tanks, Conta=inated Drain "anks, Reae:c: Coolan:

up. nxs, Aux:..,:.ary s.u Acin;;; Su=p.an.ns an:..,iscettaneous ta.

2 22eec nelc.

Was:e Holdup Tanks - Uni: 1 Licuid L'aste Disposal Sys:e=

(C Uni: 1 Evaporator Content ::e - Uni: 2 Evapera:or Concen:ra:e (5;

Uni: 1 Spen: len Exchange Resin - Uni: '. S pent lon Exchange Eesin t,

+

e_

1 C, -

p,c.,. ~

.- =

p.

s.g=gp,.

4g

~. _,,.. _. _

_ -. - -,:- -1.-

c -;

.a..

e 3-95b

~ --

yw Z7,;;;

tY' -

e.*

le',g N

."p 4

$p,I e "

4

.r--- n.v..

s.

. _. ?. w-...

(Cc :inu.i) - &..... _: :,..

m

~... -

-+-

,i

~

A3,e };.A..og

~.

~

.n.

f.:_

.s,.

e : -

.n u.....a..

- x,.

~..-...

m.:..a an..

2.=-

w. :~, c

,L.

aa1.:m.e s.== z. ::: r _ - -

,:w = =.:1 : mn <. w.-

va:: : =. ; z. :..e :.=..

c -.=, =

w.irt t.

? c i.._*.s.~. sah.es s.4 ta J./

' "" tpG. r 4/.aa,s. u., a4._**Dur

.was:e. gas h'e.ldup e-t e

_.w arys:e= cpera::en..- u ~ ;u,-. o e n - w.-

c A

iN pn q..-

- - D per e ~ 1i:7..i. l- -.-. f'UDGrequired when dischrge s ar. e Joositively' cc : rolled. - -

sJ:o:

7 troup the ce;osure e

-VI.7, and IM-AS and 3" -131 are operahlt. t '

~

v 4".

g

^'

.Xi ::in

=-

.v -

AC~~05 23 Vith the' t=s er cf cha::els OvIII - less tha: rectired bv -he Mini== Channels 0?I?E;.I equire=en:, :he es=:en:;s of the task 3,

=rv.de re easec :o de e:vir===c::.:: cp ::

czys prev:cee

:. : a. : g. s e.e i e as e.

.s a.

7..

sa :.es :, n e :t:r. s :: : e :s ce A:

ezs: :ve :.::co.e= e :

anz'7:ed, zed A:

cas: :ve :e:h:ict' '.y c Ali fi ed se=cers of de Ori: s.aff independe::,.y ver.,_,

e re.. case rzte ca. :. a:i::s 22: ver:.fy ne

, :... eu:.

.s::cge vt.ve 2.

'"he Ea age Ori:

s.a '. ; tp-eve each re'. case.

0:.: e-.. se. s us e. c= re etse :: rar:.:a: =ve c::.uen s via mis :.t.:hvzv..

d

---1.-...

.e,,

.1

-....:..a

. v.

.. __ :_..... e..

e..:

a.... c..

m.

L.e.s

.-.m--

.....:.n...,

. s.r..a e..... e z, e, y.<...:s v:.:.

.zy,

_ y:.s.;...... :y

.g.,

s s..

.. -e

.....,v

..e 7

es nr. ate: a:. eas : := e

.>e

- seur s...

....:..e

'.v...e

. -.se

. -oe

.:....z e.s n v...

ess

..z:

a...-y...

Xizi== ".ha::ei s 0?I?4* I : equire=e :, e f fiuen: reieas es via ::.is pa= var =ze :

itue f:: :: :: 25 czys - ce:.c.ee ; a: sz=::es ce

.a'.e : t: lets: esce per ! hee-s and :hese sz=pl es tre z alyze:

r fe-gros s a::ivity s.thi: '. h::: s.

(See also Specifi,:ation 3.5.1

.a..:.e 2.5.,, 4.te a..c )

x A~~.ON 30 Vi:t :he==ber o f a-els ~?I?MLI less that rec.uired by : e v.:. in= ". carne.ls CFI?2 *.I re:Cir e=e : :

.:.q. l -

u.

. a.

Vi

'd-h:urs, take and analyze an '

d :ial gra: sa=:le.

, ; I.

"rab sz=:les sht21 be taken eve v. 72 hou s =:11 d e

=C. a..,.... s Ct.s.

eC

,p e. e w..e.

4 i'

s..

I S/.

b;. If de sa=ple indica:es grea:er han 22 cryge er 2:

Er_

bydrogen, vi:hin cue hour begi: :o add ti:

ges ::

~

y,-U educe the :::ce=:rz:icn te less tha: 2: cryge: c:

F-3: hv.d ogen.

m c.

Prior :: S ar::= vhe: debera.ing fr== a shu:dev: here:

T. 4

..c= ce.s. ra:i== 1:,an::icipati== of going cri ical excen. :

%. =.

.rd a -

f3.._.

... +..,

, p:.-

M.. c.-E. ' dcring' physics tes.dng 1.f the hydroge: c==ce:::a:1==

wp.

. m...,, ;.

.a is greare:.

w.a-3:4 =1: egen sha'1 be added to reduce a

. g_. :y -

.~

.7..

M.

6e hydroge conce:::atic: to 30.

During S ar:te.

.~

.2

-."?

_...m..-

rac sa=o..e s a, se opera:icus as ces cr.,sec amove, a

)

M *, * ^ '

.ahes'and ana}v. Ed eve.v !. hours.-

-eer r

e-a e e...

e-..

h-

.-.W

.**ulk b

e w

- +

.. 3. t.-

4.l. y -,,

,e e

$'ch. *k. Qv0, aiY *:: s.y&Ae.X..-. M e r G.'.

QUG".L

?0*$$0ES.+ ' -d QM% e.* S W~.a W W W. & >

lM.

t

1 a.T.

. _ L. ~

Channel subject only to "d if t" errors induced within the instrumentation itself can tolerate; longer.. intervals between calibrations. Process system.-

=,

instrumentation errors induced by drif t can be expected to remain within

, =

- IEp=_ ~ acceptable molerancsiiFiEr'ecalibrationTis~pe~rformed atithe. intervals of.eacisCE

~

W};&_, b., & refueling period --_ ~ __:m.

x.mm,_ _= s,,.-.e---r-~ws.u.yc n

.c. =: mym'm.?T p. r__.. _----: E.-

-f m.

c m.

5_'*[ 6 ~

z;-.=. :

' %-~..

,=. -..

-- = - - -

7_

~ ' - -

t,ch--J.S

_ Substantial calibratio Fishiftis sithin a ch'annc1 de'ssentially' a channel fail- -. ""

r ure) will be revealed,.durinUoutine checking and testing procedures.

u.

Thus, minimum calibration frequencies set forth are considered acceptable.

Testing On-line testing of reactor protection channelt is required once every four weak s on a ro:ational or perfectly s:aggered basis.

Tne rotation seneme is designed to reduce the probability of an undetected failure existing within the system and t o minimize the likelihood of the same systematic test errc-s being in:rodu;ed into each redundant channel.

Tne rota: ion schedule f or the reactor protection channels is as follows:

Channels A, E, C, &D Sefore S:artup, when shutdown greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Channel A One Week Af ter Startup Channel S Two Weeks Af ter Startup

-f v

ChanneI~ L "hree Week s Af ter Startup Channel D Four Weeks Af te: Startup Tne reactor protection system instrumentation tes cycle is continued with one channel's ins:rumentation tested each week.

Upon detection of a failure that prevents trip action in a channel, the ins:rumentstion associated with the. protection parameter failure will be tbsted in the remaining channels.

If actuation of a safety channel occurs, assurance will e required :1.at actuation was within the limiting safety sys:em se: ting.

The protection channels ccincidence logic and control rod drive trip breakers are : rip tested every four weeks.

Tne trip test checks all logic 3.

combinations ara it to be performed on a rotational basis.

Tne logic and

- breakers o f th,e four protection channels shall be : rip tested prior to

~

startup when the reactor has been shu:down for grea:er than 2l. hours.

Dis-a covery of a f ailure that prevents trip action requires the testing of the 7i instrumentation associated with the protection parameter failure in the F. ;.j' -

remaining channels.. _._.._.

In ra -

+.

. + ^

.r-

.T-.

( _ m..,,

- ~

g_..

I " ' * : ~. ". ' =~

[ll ~ ^

._e I?

IT.' C_,E. 7 -f.a_' 2..d ',I..?

. d

.5 *'*.* *iN.. - -.J"

' ' ~_ ' { *[,~'

j. I

~

.w

!Mh "T. ;....'7'"'

I '" ' ~ N ~ ' '

w:;w

'7.'..".T'....

- - ~rrr n --

4-2 g.

L.

3... -

- ~ ~ -

p.

2; ;'. : _.

L-

f'-

~

T l

& n.s.~.s.,& v..iu o m..

44:'s w w -~+k&s&

=-"m M " =N " -

For purposes of surveillance, reactor trip on loss of feedwater and reactor trip on turbine trip are considered reactor protection syste= cnannels.

e 5

,,;-. a W2 The _ equipment testing;andisystem. sampling frequencies.specified_in Table hdf'-

-l.1-2 and Table-4;1-33re* considered adequate to mIiritain the~eq6ipment'and

~~ ~ "-

~

~~

n_

systems in a safe operational < status..u

...* * * * -s "-

~ m.m qyyy: r,. =.

4==

n -

4

,_,. " 74... - '

REFERENCE

~

(1)

FSAE, Section 7.1.2.3.4

(

e G

9 e

o e

'Y

5.. ^.

I F. '

,a4.u.-4*.

4'*

e '

g {' -.=.

emm ew e

" ['

-_,ierm eg p a..,.

,ga w

[.

.?

-g S.

~.

s

- q.: "

-a

.~

-;-Frf u;p:.- ;;.

~._

^

4-2a m.

m. em PJMG. ~. _ __ Z. 6.

lJ. r~_s

^

' ~

'4,,"g

. e ( ', Qt e

,,.,4 g

[

w g

a

.h..,*

b

K.. 5d#

=*g-9

O #=OW,.

'h em f.

'N

1. [h

,.f 44 4

4sp

N#

'I k;Q' G*f

.;k\\g, -

~

p 1 4!

y,4

,j ]iI' u p (h

,' bpk.

ap c

.g q'.,0 Tant.E 4.1-1 (Cont. I nneel) g i'

,1/1

. t3

'g Ni h, / g INSTitt!!!ENTS OPEllATING C' flDITIONS lj,#

O

.7

i., e m

by

' @(; CllANNEl. DESCRIPTION CllECK TEST CAI,Illit ATE REffARKS O

i ip'

6 y p

Pi.,

..f'N;Uh, IReactor Hil!!d'ing Emergency h,

{.

4'19 htCooling and Itiolation j

,M' i M System Chdnnels t

y b h.

Renctor Building S(l) 11(1)

P 11 (1) When C0!!TAINilENT INTEGRITY l's rer bed

,,i ihlQ 4 psIg ChanneIs

' j'.q

)I b

RCS Pressare 1600 psig S(l)

M(1)

NA (l) When RCS prensure>1750 paid k

h l ' b 9};

1 k tiflf,o

>.l l in l

l'

[iO IM 4,

nr 3 g

j: th'ic.

RPS Trip',

S(1)

M(1)

NA (1) When CorlTAINNENT INTEGRITYi rer redi m

,4,',

W"'i]'p.[l{'jp.t>

+t~ ' t,.

I l' C;

WhenCONTAINHENTINTECRITYMLgreeiIN!dn

'j 1

2 6:U

. ;. j f.'G d.

Renctor llullding 00 psig S(1)

H(1).

R (1) 1 h; jgtyM i (1) When CONTAINMENT INTEGRITY :l]o j,reqEred ',

',-]l,N f

~

I is g;the.

Renctor Edilding Purge W(1) 11( 1 )

R ,j /pf* h hk !' I k hf[I V 20 b Reactor Rullding Spray NA Q NA [ j}g'. a n. I' System 1. ogle Channel 2 T , 7 *:, vi ;s 21.;.y eactor nutiding Spray qg y y l jg ; R 'j, '.y a "j; y 31; System Analog Channels i i "l j 1 i t a. Reactor Ilnlidleg NA H R 30 psig Channet,' s; ip l,) D;. 22. Prennurizer Temperature S NA R ,g };j '! r {q IM ]f ' t,[i ? ' Channels

,,j ;jj' 23.

Control Rod Ahnninte Position S(l) N, R (1) Checlk with Relative Position Indit dtor ./' [ .I'j 24. Control Rod Absolute Posillon S(l) NA R (1) Check with Absolute Posttion{Indidator 'j:, i d !y g 4

1, 25.

Core Flooding Tanks 3 ll ' ).t 'l n. Prenure Cimnneln S(l) NA R (1) When Iteac t o r Coolant system pressure l d, :, dj.lf j:, ' h. In greater than 700 psig l lb i li' h. I,evel ChanneIn S(1) NA R j,

n','

t J.] 26. Prensnrlzer I,evel Channeln S NA R g ~ 27. ilakenp Tank 1.evel Channeln D(l) NA R (1) Uhen Hakenp and Puri f icat lon System ) .in in operatinn

5 !N b Efd[$ i [ "') ~ &,kI h p45 q y!' V m,. J ', ~ Tant.E 4.1-1 (Cont i minii) ~;. CilAlltlEL DESCitiPTIO!! CilCCK TEST C A l.I R PATE REllARKS i z,. +,.. 'itc

p h k.39.

h.

q. ? '.!v 4 ,,i ~ Turbine Overnpced Trip 11A R HA U.' ML 1 'Sii 40) ; SoditimLThiodtilidte Tnnk 1.evel 11A h riA R g ',' Wlhidb,Indlentor > M !, !%' l ;. l. n;.li y t-h ' 4'ili Sodlism 1!ydroxide Tank Icvei NA NA R hjf ' i,'l$j}, Indicat6r ... k 3h Diesel Chnerator Piotective NA NA R g U 1, tj;Relnying, jI, i 43.' 4 KV ES Hiin llodervoltne,n Reinyn NA 1t(1) R (1) Relay operatiot will be (Dienel Start) checked by local tent punh-htt t L Onn. ((O o , j d4. Renctor Coolant Prennure S(l) R (1) When reactor crioinnt r.y s t e m In prenntertrc<Inbove 300 e - Dil Vnive Interlock ntstable i. pnir, or Tnves'in grenter ] U,

i.f rj than 200*F.

[ 45 1.onn of Feedunter Renctor Trip S(1) !!(1) R (1) When renetor power execcels 7 %' power, { 46' Ttithine Trip /Renctor Trip S(1) 11( 1 ) R (1) When renctor power exceeds P-20% power .{ 47.n Prennortzer Code Safety Valve and S(l) R (1) When T in greater than nye 525*F ~[ FORV tnilptpc flow monitorn n [h ..I When Taye is gr{ eater than 47.h PORV - Acotant ic/ Flow NA ff(1) R (1) 525*F. ,l - NA ff(l) R (1) Per specification 3.1.12 48. PORV Setpointn Excluding vnivh) operation. i. ,t. 4

A[* } NI If){.s ' ' [i g '.I ar s( W t Ji,1 n,q;. # ,.m u TABLE 4.1-1 (Coo t I numl) ,1 .p, ,b J Q i

<;u e

..; ', o . g,i ',pi.; c p p I' (', CllANNEL 1)ESCRIPTION CllECK T E'iT CAI,I RRATE REffARKS ,.e ? b '

49.,i[Snturation Margin Honitor S(1) tf( t)

R (1) When Tnvc in renter thnn 525"F. !ry I. e o v. 50. ' Emergency Feedwater Flow 11A !!(1) R (1) When nyeis/p,renterthan F 250"F. i

!.inntrumentation s

.. y: t-f 51. Emergency Feedunter Initintion 3 [, ah i g ster thnd 'n Loss of RCP's or Feedunter 11A f)(1) (2) R (1) When 250"F.,ll. Ot} l' (2) Includen,logie' test only T/W - Twice per week it - Ecch Refueling Period. S - Each Shift B/t1 - Every 2 monthn g,g. gg; rp, D - Daily: 4' W - tieekly Q - Qunrterly D/W - Ev d M M n it - !!onthly P - Prior to ench ntnrtup i

g 3

..jh 7 if not donc previoun week 3 t .s

a..,

,t -lmlfi!d 9 'l h* i 0 g i W i l, iip'l D, p' v' 's j - ? , E i V !i. 4! [' 1:. -t v h' [.h! .h l '~ ll, ,[] ! ' n ,.4 -l-j '; ' 'i n }'. i i,s t i. 4

TABLE 4.1-2 x MINIMUM EQUIPMENT IEST FREQUENCY f-l9 3;. _ ~... - 1. Control Rods Rod drop times of all Each refueling shutdown g full length rods 2. Control Rod Movement of each rod Every two weeks, when reac-tor is critical Moveman: 3. Pressurizer Saf ety Setpoint* 30% each refueling period Valves bb' 4 . Main S:eam Safety Setpoint 25L each refueling periof Valves 5. Refueling System Func:ional S:ar: of eacn refueling In:erlocke pe Ioc 6. Main S:eam (See Section 4.E) Isolation Valver 7. Reactor Coolan Evaluate Daily, unen reactor coolant Systo= Leakage sys:e= te=pera:ure is grea:er : nan-525'T I r~ s. Deleted l t . = u_ 9. Spent Fuel Cooling Functional Each refueling period prior

o fuel handling System 10.

Intake Pump House (a) Sil: Accumulation-Each refueling period Floor Visual inspection of (Elevation 262 Ft. Intake Pump House Floor 6 in.) (b) Silt Accumulation Quar:erly Measurement of Pump House 71oor f-11. Pressurizer Block F unc tional** Quarterly Valve (RC-V2) 5':- F== y. =q: bE ~ The set point of the pressuricer code safety valves shall be in 176,- ~ accordance with ASME Boiler and Pressurizer Vessel Code, Section III, EL (( . Article 9, Winter, 1968. c-Function shall be demonstrated by operating the valve through one i-35 co=plete cycle of full travel. T

b. __

The required percentage of Main Steam _ Safety Valves will be tested ~ ' prior to Cycle 5 criticality. __=. - W.- W. ')% J.*/A O- .p

. ' 'hd ' h i{!; ' "I ,b 3 ~~' 'b " p,, TABLE 4.1-2(Conti nucil) f Iton Tent Frequency 12. Isolation devices Vinunt I nn pce ti o n Seminnnually o n IJn i t 1/llni,t 2 Interconnbetions (a) IInit 2 Renctor Onoinnt Biced 11oldop Tank - lini t 1 Itenctor Cool.1nt Wante Evaporator I (b) IJni t 1 Hiscellnncons Wnste Evaporator - lini t 2 Evapor-ntor Condennate Ten t Tnnks (c) lini t 2 Neutralizer Tanks, Con-tainrnent Drnin Tanks, It enc t o r Coo land Illend Ilo lititp Tanks, Auxiliary Building Sump Tankn nnd Hincellancous Wnnte lloidup Tanks - IJnit 1 T.iqtild Unstc Disposal Systnn I$ (d) lini t 1 Evnporntor Concentrate - lini t 2 Evaporator Concentrate (c) lini t 1 Spent Ion Exchange Henin - 11 nit 2 Spent inn Exhnnne Rosin

5 MERGENCY LOADING SEQUENCE AND ?OWER TRANSFER. EMERGENCY CORE ' 1 4.3 COOLING SYSTM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 J.mergency %As Sequence c==. & plicabilitv: .=-.w... ~+- Al n,.a aw_iodici testing requirements for safety actua: ion svs:sms. Applies to per ~ Obire:ive: To varify that the e:nergency loading secuence and automatic power transfer is op2rable. Seccifica:icns:

4. 5.1.1

Secuence and ?over Transfer Tes:

shall be cendue ec := de= ens:ra:e During each' refueling interval, a tes:

he emergency loading secuence and power transf er is eperable.

a. tha: vill be consider 'd sa:isf ac: cry if :he fellering p=ps and f ans Tne tes: b. have been successfully s:ar:ed and the felloving valves have c==rie:ed the emergency power their : rave; on pref erred power and transf erred :: opera:ing lights, and ei:her as evidenced by the cen:r:1 board===penent

he s:stien : =ru:er er pressure /flov indica:icn.

-M. U. ? =: -0. E. ?=p and D. E. Injection Valves and D. E. Supply Valves -I. 3. Coeling ?=p -I.. I. Ven:ils:::s -D. E. Closed Cycle Coeling Pc: -N. S. Closed Cycle Cocling ?u=p ( -0. E. River Cocling Pu=: -N. S. P.iver Co: ling Pu=p ( -D. E. and 5. 5. ?unr Area Coeling ?an -Screen Eeuse Area Coeling Tan of 3 7.. E. 30 psig (Initiated in coincidence vith a 2 cc: -Spray Pu=p. Pressure Tes: Signal.) I -Motor Driven Emergency Feedvater Pu=p

he energency diesel, :he diesel 7:*. ice ng successful ::ansf er ::

si=ula:e : rip cf :he genera::: i then c. genera:cr breaker vill be cpened :: reclosed c verify block load ce the reclosure.

Tnis test shall be perfer:ed prier te Cycle 5 criticality.

4.5.1.1 Secuence Tes: shall be condue:ed :c de==n-te exceed 3 =enths, a tes: shall a. A: intervals no: the e=ergency loading secuence is cperable. this ;es: s: rate tha: be perf ormed on either pref erred pcVer or emergency power. vill be considered satisf ae: cry if the pu=ns and f ans listed in Th'e tes: 4.5.1.lb have been successfully s:arred and the valves listed in4.5.1.lb b. have ec=pleted : heir travel as evidenced by the cen:rel board ce=nenen: operating lish:s, and either the s:ction ce=pu:er er pressure /flev indi-cation. '-nams controls the operation Tna E=ergency loading secuence and au:cmatic power transfer \\ 4-30 \\_ n

~ .~- d. Th2 b:ttcry will b2 subjac:Gd to a lecd tis: a: a fr qu2ncy nst to eiceed refueling periods. The battery voltage as a func: ion of time will be monitored to es:ablish tha: the bactery performs N,.wY D l O d_+,2 as expected during this load test. e -.=cm w_wn, ,,w. =, - 3ases_-y . The tests specified -are designed to demons:: ate that one diesel generator vill provide power for operation of saf eguards equipment. They also assure that :he emergency generator control sys:e= and :he con:rol systems for the of a loss of safeguard: equipment will func: ion automa:ically in the even: nor=al a-: station service power or upon :he receipt of an engineered saf e-guards Actuation Signal, The automatie : ripping of manually ::ansferred loads, on an Engineered Safegweds Actuation Signal, protects the diesel generators from a po:ential over-load condition. Tne tes:ing frequency specified is in: ended :o iden:ify and per=i: corree: ion of any =echanical or elec:rical deficiency before i: can resul: in a sys:en fa'ilure. ~he fuel oil supply, star:ing circuits, and con:::ls are continuously =eni:ored and any f aul:s are alarmed and indicated. An abno. al condition is nese sys-te=s would be signaled 4richout having :o place :ne diesel generators on tes:. Tne sur-Precipitous f ailure of the sca: ion ba::ery is ex:remely unlikely. veillance specified is that vnien nas been demons::ated over :he years :o prov'ide an indica: ion of a cell beco=ing unserviceable long before it fails. 3 The PORV has a remo:ely operated block valve :o provide a positive shutof f capabili:y should the relief valve becone inoperable. The elec:rical power .t" f or bo:n ne relief,. valve and :he bicek valves is supplied Iro= an 57 power - = - - - - - source to ensure :he abii:y to seal this possible RCS leakage pa:h.. a minimu= of 107.kw of pressurizer heaters and :neir The recuiremen: :na: associated con:rols be capable of being supplied elec:rical power froc an emergency ous provides assurance :ha: :nese heaters can be energized during F-- a loss of of f size power condition to main:=d-natural cirnulation. REFIRENCE 1 l (1) FSAA, See: ion 8.1 \\ E..O p. l,. .mo l. i o. 4-47

== - e.> y_p 9 W s O M'<

r; e. 15%.= - 4.6.3 Pressurizer Heaters .1 a.. __3

=.- -

y:. gq54.. p ~ r- _. rqa._..Once each refueling pressurizer. heater groups 8. and 9 shall be ' " ~ 7' ~ transfered from the normal power bus to the emergency power bus and energized. Upon completion of this test, the heaters shall g';. _ _^' be returned to their normal power bus. b. That the pressurizer heaters breaker on the emergency bus cannot be closed until the safeguards signal is bypassed and can be closed following bypass. c. That following input of the Engineered Safeguards Signal, it shall be verified that the circuit breakers, supplying power to the manually transferred loads for pressuizer heater Groups 8 and 9, have been tripped. f l 1 4-47a l ~ ..m, .e.m l - _f.--,.. .x. 2m, F. > '" : *e ~

- <a-

's ~ -A c^L " " '- % -.<4'

4.9 EMERGENCY FEEDWATER SYSTEM PERIODIC TESTING Applicability Applies to the periodic testing of the turbine driven and two motor-driven Emergency -feecwater pu=ps, associated actuation signal, and valves. Obj ective To verify tha: the E=ergency Feedwater (EFW) Syste= is capable of perfor=ing its design function. Specification 4.9.1 TEST 4.9.1.1 Whenever the Reactor Coolant Syster tempera:ure is greater than 250 F, the EFW pumps shall be tested in the recirculation = ode in accordance with the requirements and acceptance criteria of ASME Section XI Article IWP-3210. The test frequency shall be a: leas: every 31 days of plant operation a Reactor Coolant Te=perature above 250 F. 4.9.1.2 During testing of the EFW Syste= when the reactor is in STARTUP or POWER OPERATION, if one stea: generator flow path is =ade inoper-able, a dedicated qualified individual who is in co=munication with the control roo= shall be continuously stationed at the EFW local nanual valves (See Table 4.9-1). On instruction fro = the Control Roo: Operator, the individual shall realign the valves fro = the test mode to their operational align =ent. 4.9.1.3 At least once per 31 days each valve listed in Table 4.9-1 shall be verified to be in the status specified in Table 4.9-1, when required to be operable. 4.9.1.4 On a quarterly basis, verify that the manual control (HIC-849/850) valve station functions properly. 4.9.1.5 On a quarterly basis, EFV-30A anc 303 shall be checked for preper operation by cycling'each valve over its full stroke. 4. 9.1. 6 Prior to start-up, following a cold shutdown greater than 30 days, conduct a test to de=onstrate that the =otor driven EFR pu=ps can pu=p water fro = the condensate tanks to the Stea= Generators. l~^

For tha purpose of this require =ent, an OPERABLE flow path shall =can an unobstructed path fro = the water source to the pu=p and f ro= the pu=p to a Stea= Generator.

l 4 fE 4-52 ,.~s;, : ~.a.. n

1t

37. 1

, 4. 9.' 2 ACCEPIANCE CRITERIA zi. ~ _.. - - _A - These tests shall be considered satisf actory if control board indication and M M i...dvisual observation of)the equipmen demonstra:es that all components have . c=-- . _. _ _ ~ properly.....-..-.._ _.. --._. y 7 -. operated. ~ Eaus-.. ~. =.".. =....... .g -E The 31 day testing frequency will be sufficient to verify that the turbine driven and two motor-driven EFW pu=ps are operable and tha: :ne associated valves are in the correct alignment. ASME SEc: ion XI Ar:icle IWP-3210 specifics requiremen:s and acceptance s:andards for the :es:ing cf nuclear safe:y related pumps. Co=pliance with the nor=al acceptance criteria assures tha: the EFW pumps are operating as expected. The test frequency of 31 days (no=inal) has been demonstrated by the B&W Emergency Feedvater Reliabili:y Study to assure an appropria:e level of reliabili:y. If :esting indicates tha: the flow and/or pu=p head for a par:icular pu=o is no: within the nor=al accep:ance standard an evaluation of the pu=p performance shall be co=pleted within 96 hours or the pu=p declared inoperable. For the case of rne IFW Sys:e=, the syste= shall be considered operable if under the _~ case single pu=p f ailure, a =ini=u= of 500 gp: can be delivered vnen wors: stea genera:or pressure is 1050 psig and one stea: genera: ion is isolated. A flow of 500 sp=, a: 1050 psig head, ensures :ha: sufficient flow can be delivered to eitner S:ea Genera:or. The surveillance requirements ensure tha: the overall IFW Syste= func:ional capability'is =ain:ained. .M S 6 .u.. - =- S e m e h 5 F.* e ~ e M d..**,,

[ ' 7.

em -'*.-m.". .~ ~ .~. p. -.. 3._. 1.~ #. C. e S O er 9: = -~ u. = (k $ ' " ' ~

.r---

Y 4-5 2a - - - -~.. ~

2.

.a._ es k NU I O.Aishhe9dg '4 5+ "4 [d - /* I ' ' *M [ N b Nh.m'4Id.:seus.. -* tt 3.5A a.- V T4h 3

s, Table 4.9-1

==:.- Status of IFW Valves r Valve'No.~ = 7.- S t a tus ' - CO-V-10.I_ Open -... _ '. - 1... CO-V-10B Open r EF-V-1A Open f{ ;r n i n. EF-V-13 Open EF-V-2A ~ Open ,p s 4F-V-25 Open - 4 r' MSV-2A Open J a MSV-23 Open / t EF-V Locked Closed, e ',. is

i EF-V5 Loc'ged Closed

, r, .s 2 3'-- EF-V6*. ~ Locked open _. _.... =. - _ .gy_ylogw -_. r L Lock 9 -Open .t z [ 7, i -, y:

J EF-V10B * ^ ~.

Locked Open 'f, - ' ' ~ / ] / EF-V-16A* y 5:._ Locked Open',; rii t Locked Open' EF-V-16B* e / 4 Locked Open f y IF-V-20A ,s EF-V-203 Locked Open, p----.----.- ? y

d.....;._. - r r: -.

IF-V-22_2.12.::. Locked Open t-- i. [ _, {.; _ ~ ~ ~ ~ - =. - -. -. -,,s l @~2~

Mahual valve to-which Specification 4.9.1.2 zpplies

~ s. a. w pa

=a'*

_ I.. y-. t me. 3 O .a d y + ,i h 1 7 .g 4-52b 1 -w~ p l Y 'd' f. t &. &NMly9. 41 St*wa. U.s *1 w"enr 1-c ,}}

ML20033B032

Text

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