Text

Proposed Tech Specs,Removing Main Steam Line High Radiation Monitor Reactor Scram & Group Isolation Functions to Improve Availability of Main Condenser for Removal of Decay Heat & to Aid in Minimizing Inadvertent Reactor Scrams
	ML20115G679
Person / Time
Site:	Hatch
Issue date:	10/19/1992
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20115G669	List: HL-2006, Application for Amends to Licenses DPR-57 & NPF-5,revising TS to Remove Main Steam Line Radiation Monitor Reactor Scram & Group Isolation Functions,Per BWROG Topical Rept NEDO-31400, SE for Eliminating BWR MSIV Closure..; ML20115G679;
References
	NUDOCS 9210260297
	Download: ML20115G679 (63)
	v • d • e

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ENCLOSURE 3 PLAT 41 HATCH - UtilTS 1 AND 2 NRC DOCKETS 50-321 AND 50-366 OPERATING LICENSES DPR-57 AND NPF-5 REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO REMOVE THE MAIN STEAM LINE HIGH RADIATION MONITOR REACJOR $3 RAM ANDl@UP ISOLATION FUNCTIONS REVISION INSERTION INSTRUCTIONS UNIT 1 Eggg Instruction 3.1-5 Repl ace 3.1-5 Replace 3.1-6a Replace 3.1-8 Replace

, 3.1-13 Replace 3.2-3 Replace 3.2-4 Replace 3.2-19 Replace 3.2-51 Replace 3.2-66 Replace 3.7-19 Replace 3.12-4 Replace UNIT 2 2-4 Replace B 2-11 Replace 3/4 3-2 Replace 3/4 3-4 Replace 3/4 3-5 Replace 3/4 3-6 Replace 3/4 3-7 Replace e 3/4 3-8 Replace 3/4 3-11 Replace 3/4 3-15 Replace 3/4 3-15a Replace 3/4 3-16 Replace 3/4 7-8 Replace 3/4 3-58a Replace 3/4 3-58b Replace ;

3/4 3-58c Replace 3/4 3-58d Replace HL-2006 003397 E3-1 9210260297 921019 DR ADCCK 0500 1

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.% Table 3.1-1 (Cont'd) l a- r

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SC'am .

. Operstde i Number Source of Scram Top Signal Chernets Scram Tnp Settmg Source of Scram Signalis j

(a) Required Per Reqummi to be Operside '

Except es indicate f Below

.$- Trip System tb)

[

i L

8 APRM Downscale 2 >3/125 cf fuit scate The APRM downseale tre is actrve ordy udwn the Mode !

Smtch is ^m RUN. The ~

APRM downscCe trip is sut M, bypassed when

{

the IRM mstru .a.wiis -

operable and not tnpped. ;

15% nux 2 < 15/125 wi fui: The APRM 15% Scrarn is auto- I scale Tech Spec meccally bypassed when the '

2.1.A.1.b. Mode Switch b in the RUM [

.w .

- 3 (Deleted) .

E l .

e

.m .

10 Main Steam Une Isolation '4 . $10% vaive ch Automancany bypassed when

Pt- Velve Closure from fuS epen the Mode Switch is not in

" , *, Tech Spec 2.1.A.5. the RLas pestion. The dessgr*

j.f ' pearruts cfosure of any two . i

- unesw,%ut.e w n

irstiated. !

I'

r m 11 Turbine Control Valve 2 tvtNn 30 truk Auta. 24, bypessed when j

" Fest Coeure weends of the start turbine steam flow is beiew ~l y

,3* of control ve8ve - that c.,.+4.v to 30% of ta f ast desure rated therrmi pw as eneesured

.% Tec5 Spec 2.1.A.4 by turtune first stage

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t T Table 3.1-1 (Cont *d)

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Z

.,, Scram .

Operable Number . Source of Scram Trip Signal Channels scram Trip Settmg Source of Scram Signet ie z (a) Recrired Per Requwed to be Cperable [

Trip System Escept as In6ce**d Below *

(bt I

12 Turbine Stop Velve 4 110% verve closure Automaticepy bypassed when l Ctesure from fuS open herbine steem flow is below i Toch Spec 2.1.A.3. that corresconding to 30% of f reted thermet power es I measured by turbme first stege pressure.

i Notes fee Table 3.1-1 t

a. The coturr n entitled " Scram Nurreer" is for w..u...~e so that e on: : : : retetionship een be estabEshed between items ir Table 3.1-1 and items in Table 4.1-1.

f b. There shall be two opereble of tripped trip sys+ ems for each pctential scram signal. If the rusmber of y operat4e channels cannot be met for one of the trip systems, the inoperable channetts) or the assocseted a trip system shall be tnpped.14eweves, one trip signet channel of a tnp system may be inoperable for up to two (2) hours during periods of required surveit!ance testmg w thout tnpping the channet or essociated trip system peovided that the other sw.-.-.vchannefis) r=wmitoring that parameter witNo that trip system [

is (are) operetMe.-

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[. Notes for Tet4e 3.1-1 (cont)

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-l 1 :c for SCRAMS 1 thru 7 ervi 8 APRM 15% Fkrx. 4f the number of coeratde channels is not r,et for both tnp sys rems; 2

Q initiate insertHm of all control rods capable of being moved try centrol rod drive pressure and termiete their ea w witNn four (4) hours.

', For SCRAM 8 (APRM High Tnps, laoperative, and Downsedel, if the nurr" er of operable channels 3 ret met for tm,th tnp systems; initiate insertum of all control rods capable of being rnoved by control rod drivo wessure eM complete thdr insert >on within four hours or reduce power to the IRM range entf go to the

. *J' ART & HOT STAND 8Y posrtron af the Mode Switch within eight hours.

For St iAM 10. if the rW of operatAe channels is not met for both trip sys' ems, re bee tterbire I load * ' el se mair. steem Ene iseletion valves wittun eig!-t hours or initiate insertion of all control a rods capable of being e.msd by control rod drive pressure end complete tfwir imertion withm fove hours.

Fee SCRAMS 11 and 11, if the nurmber of operable channels is not met for both trip systems, redtree reactor power to 25% of rated ttv emal p-wer or less withe.e eight hours.

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, Scrwn instrumant Chm:k Pvtrumant Functional Test Instmmect Cahbesten 'f Number Source of Scram Tnp Signal Group TArwnsm Frequency IWwnum Frequency fArwnum Fretriency l C L (M 'b) k) .

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9 (Deletedt l t

.to Main Steam Line isolation Vais A NA Every 3 months W !

Cosure

-[

11 Turbme Control Vatwe Fest A- fdA Every 3 moinhs W OnceOperstmg -

Closure Cycle N !

12 Turbine Stop Valve Closure A NA Every 3 moitts N RPS Channet Switch A NA Once/Oper.irig Cycle Not A;mEcebie l Turbine First Stage Pressue e A NA Every 3 months Every 6 months Fermissive ;

a. . The column entitled

Scram Number

es for convenience os that a one-to.one re!atronshe can be esteeishad f' between items in Tebb 4.1-1 and items in Table 3.1-1.

b. The definition for ear, . of the four groups is as f- Rowr. l I

t w Group A - On-off sensees that provide a scran.tnp signal. [

- Group B. Analog detaces coupled witis bi-stable t@s that provide a scram tnp :ignet.

Y 03 Group C. ' Dewees which ordy serve e useful funct>on during some restritted mode of operatsors such e* startup er shutdown, or for which the ordy practical test is one that can be

[

perfom _ i < shutdown.

Group D. Analog transmitters and tnp units that provide e scram tnp function. [

w

-/ c. Ftmetional tests are net retc.ared when the systems are net required to be operable or are tnpped 4

e However, if functional tests are trussed, they sheit be performed prior to returrung the systems to en operable states.

e+

o <f. Cahbrations are net required when the systems are not reqthred to be cperable or are tnpped.

n 7 However, if cahbrations e e missed, they shall be performed s,nor to retummg the system to 'I

" an operable status.

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e. This insta ..w.ah. is exempted from the instrumert fu vetionei ;sst denrunort. This 4-I, aa j w functional test wiH consist of inrectmg a simulated e4ectreal signet mto the mer a a -

M channels.

a C

- f. Deleted i H !

U g. The water level ir< the reactor wiB be perturbed eM the cortsJe.c level irwJicatoe changes wdl f be monitored. This perturbation test will be per"ormed every 3 months efter completmo of the t

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functional test program. i f

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h. Physicalinspection and actuetson of these posit >on switches we M cerformed once per overstmg cycle.

j O 4. (Deleted! I j

Measufe 15.39 miefval ffom EHC pressure smtch actuation to RPS relay K14 #mgiretron. l f

[ P.A$ls IOR tlM111NG CONDl110NS FOR OPIRAllON_.

3.).A.8.b. Incoerativy An APRM is inoperab1) if there are less than two LPRM input per icval or there are less than 11 LPRM inputs to the APRM channel,

c. Downscah The APRM downscale is active only when the Mode $ witch is in the RUN position. The APRM downstale trip is autetati: ally bypassed when the r IRM instrumentation is operable and not trMed. Because of the APRM downscale limit of 13/125 of full scale when in the Run Mode and high ';

level limit of $15/125 of full scale when the Start & Mot $tandby '

Hode, the transition between the Start & Hot Standby and Run Modes must be made with the APRM instrumentation indicating between 3/125 and .

15/125 of full scale or a control rod scram will occur. In addition, the IRM system must be indicating below the High High flux setting (120/125 of full- scale) c" a stram will occur when in the Start & Hot Standby Mode. For noemal operating conditions, these limits provide assurance of overlap between the IRM system and APRM system so that there are no .

" gaps' in the power level indications (i.e., the power level is con- i tinuously monitored from beginning of startup to full power and from ful' ;

power to shutdown). When power is being reduced, if a transfer to i the Start & Hot Standby Mode is made and the IRM's have not been fully- t inserted (a malooerational but not impossible condition) a control !

rod block immediately occurs so that reactivity insertion by control rod withdrawal cannot occur. ;

d 15% Flux The bases for the APRM 15% riux Scram Trip Setting is discussed in the bases for Specification 2.1.A.I.b.

9. (Deleted)

10. Main Steam line isolation Valve Clot,qtg .

The bases fc/ the Main 3 team Line isolation Valve Closure Scram Trip Setting is discussed in the ses-for Specification 2.1.A.5.

11. Turbine Control Valve Fast Closure i _The bases for the Turbine Control Valve Fast Closure Scram Trip Settine j is discussed in the bases for specification 2.1,A.4..

f 12. Turbine Stoo Valve Closqte

! The bases for the Turbine Stop Valve Closure scram Trip Setting is dis-j cussed in the bases for Specification 2.1.A,3.

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--O Table 3.21 (Cont)

O Jaquered Z Ref. Trip Operable Action to be tekee if a No. InstrJment Condit,on Chrnnels Trp Settmg msmbe* of channes se (e) Nomenclature perTnp ret met for both tnp Romerite td)

System (b) evstems (c)

E.

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-4 4 L en Steam Une Ngh 2 53 twnee reemel Irvnete closure of reector instietes closure of

" Radiation ' fut power bach- water semple valves. reactor water ser#

ground ** vefwes 2B31.F019 ano 2831-F020.

5 Mwn Steam Une Low 2 at825 peig Irvttete en onderry Iced Irvtietes Group 1 Pressure reduction and close sootet6en. Orey MSNs wetten 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . vequwed on RUN mode, therefore actrweted when Mode Swatch is in j RUN peeroon.

7 6 Mein Steam Une High 2 $138% reted flow Brutiste en orderfy load inrtsstas Group 1 Flow (s115 pud) redaction and ?80ee MSNe isolenon.

j within 8 heure.

!' 7 Man Steam Une Mgh 2 s194*F Ininets en om load Irvtutre Geoup 1 Tunnel Temperature reducten end ebee MSive isoletion.

wittwe 8 houne.

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- 8 Reector Water Mgh 1 20-80 gum lootste resetor we:er (f1 N Oeanup System cleerm.p system O Dfferential Flow 9 Reactor Water Mgh 2 s150*F lootste reactor weter Oeenup Aree eseenup system.

y Temperature

{ 10 Reactor Water Wgh 2 sE7'F leolate reeeter water y C esnup Area cleerme optem.

/ VentAetion

{ Offererstiel r, Temperef=re r

- 11 Condew wscuum low 2 at7* H2.weeuum Irv'iste en orderly load truttete Group 1 3y reduccon end clore Mrwe withm 8 hre.

lootecon

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12 Drywan Radiation Wgh 1 s135 R.HR. Coose tte effected loodates w esoLeo wahree wit',in csa . . -,; purge

.N 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in M end went wahree.

3

?

Shutdown wrtNn the newt 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months eru* its w

. Cold Shutdswn wit.Jn

, g the nelft M hours.

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a.

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

Notes for Table 3.2-1

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1 g- e. The column er:otted *Ref. No.* is ordy for corwermence so that e oteto-one refetnmeh, can be established between Enes in Table 3.2-1 and items in Teble 4.2-1 2:

b. Pnmary containment integrity shell be memtamed at all twnes prior to withdrawing corrtrol rods for the purpose of going critical, when the reactor is criticat, or whod the reactor water t-vmas is above 212'F and fuelis in the reactor vessel except while g dm.....,, low-power phyues tests at etmosphene pressure et power jeweis not to exceed 5 MWt, or perfomung en inserwee vesses hydrostatse or leeltage test.

When pnmery ma < ; ca integrity is required, thees shall be two operable or tnsped tnp systone for each funenon.

W5en performmg irwervice hy.frostacc or leakage tes.mg on the reactor vessel with the reactor cootent temperature above 212'F, reactor vessst water level instrumentation associated with the low low (Level 2) trip requsres two operable or inpped channels. The drywell pressure tre is not required buceuse pnmery containment irdegnty is not reluired.

I c. If the nurr6er of operable chermels connot be met for one of the trip systeme, the 'noperable charmans) or the escocrated trip system shnll be insped. However, one trip signal charmel of a tnp system.ney be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during penods of required survelience testmg wqthout tnppeg the f channel or essociated tnp system, provided that the other timeinmg ebennel(e' moestormg that same to parameter within that tnp system is tere) operable.

6 m

d. The valves associated with each Group isolation are om in Table 3.71.

e. Prior to the hydrogen injection system startup e af with resetor power greater then 20% reted power, the normal fus power radiation tripialann metpomes twey be changed based on calcuteted w expected radietion levels during hydrogen injection system operation. Associated trip /elerm i

, setpoints may be adjusted during injection based on either calcu8ecorw er mew-.a. of 5~ actual radiation levels resultmg from hydrogen inrectron. Followmg a reactor startsp e

)- background radiation level udl te determined end the sesocseted trio /aleem setpomts adjusted g within e 72-hour period. The radiation level shell be detemwned and associated trip / alarm !

o setpoints shall be set within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of re+stablishmg noemel radiscon levels after a j

[

"O reductson in, or a completion et, hydrogen insection end pnor to estalAshmg reacter power leveis below 20% ef reted power.

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'I f. The high differential flow signal to the RWCU isolation velves may be bypassad for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

] during periods of system restoestion, mainteneree, or testmg.

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[' Table 3.2-8 (contJ {

t !

e n .. Trip Settmg Action to be taken if Romerke

t Ref. Instrument Trip Required Condition Were oro not two ope @ t s No. Operable tal Nomencle- Channels or tnpped t<p systems !

C per Trip i

z ture Z _

System (bl _

~

5. Mern Steam Line Hi 2 s3 times isolate the me-:horuces One tnp per tnp Radiation Morutor normal full vacuum purvo end the logic system w,N j pome background gisnd seat condere isolete the (e) exhauster enecherucel vocoum purup and the g8end 3 seatcondenser exhauster.

e. The crAumn entitled "Ref. No.* is ordy for conve.dence so that e one-tm relationship can be estabbshed ,

between items in Table 3.2-8 end items in Table .+.2-8. j

b. Whenever the systems are regiired to be operable, there shall be two opeeable or tnoped tne systerne. i q

ff tNs cannot be met, the indicated action shall be taken.

i .t.as c. In the event that both off-gas post treatment trufastson mor9ers become inopers.ble, the rescAor shell be 5

7 placed in the Cold Stastdown witNn 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months urdess one monitor is sooner made ope.able, or adequate attemetrve morutoring faciEties are evadable a

d. From and after the dets that one of the two off$es poet treatment radia* ion morutors is made or found to be inoperable,64 ,+d reactor power operation is pemussible during trie rient fourteen days (the ellowable reprir time), provided that the iruverable morator is tnpped. l 7C

,, e. Prior to the hydrogen irgecnon system startup and with reactor power greater then 20% rated power, the normal fus power rad.eoon tnp/alerm serpomes eney be changed ;

E' bened on calculated expected endistien levels durmg hydrogen ingection system operstson. Assocrited trip /aismo setpoets may be edpasted durmg ingoetsen based on eether "p calculations or n eesurements of actual radiation levels resutting from hydrogen irgection. Fonowmg a reactor startup, a tv-Aground radia%eilevel wit be detemened and g r%e associated triplafarm setpomes adjusted witNn a 72+our period. The vediation level shall be detemuned and associated trip /aform setpoints she!I be set witNn 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months n of twabEsNng normal radiation levels after a reductson in, er a carryletion of, hydrogen iniectrori and prior to estathshmg reactor power leve6s beforv 20% of rated powee. k

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_ BA'!S f 0R LjRiflHG CONDlil0NS f 0R OPLRA110N ,

3.2.A.P. F/t.ittE.yessel Steam Dome Pressure (Shutdown Coolino Mde)..LoyLEermissive 4 ,

i This setpoint is chosen to preserve the pressure integrity of the RHR

system under conditions of increasing reactor pressure (startup). The j RHR suction valves from the reactor (shutdown cooling mode) would be

closed when the 145 psig sc! point is reached. This function protects

against RHR system pipe breaks during the shutdown cooling mode of op-

j eration. Additionally, at reactor pressures below this setpoint the primary containment isolation signals are permitted to close the in-

board motor operated injection valve (LPCI mode).

3. Drywell Pressure Hioh

}:

i The Bases for Drywell Pressure High are discussed in the Bases for Specif-i ication 3.1.A.S. Pressure above the trip setting starts the SGTS and in-itiates primary and secondary containment isolation.

4. Main Steam Line Radiation Hioh i

i Radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident. This in-strumentation causes isolation of the reactor water sample valves. With the l e*tablished setting of approximately three times normal full power i b.ckground, fission product release is limited so that 10 CFR 100 guidelines j are not exceeded for this accident. l f 5. Main Steam Line Pressure low

] The Bases for Hain Steam Line Pressure Low are discussed in the Bases for 1 Specification 2.1.A.6.

6. Main Steam Line Flow Hioh Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a j steam line break accident. In addition to monitoring steam flow, instru-

mentation is provided which initiates Group 1 isolation. The primary func-4 tion of the ins'.rumentation is to detect a break in th6 main steam line.

For the worst case accident, a main steam line break outside the drywell, 4

the trip setting of 115 psid, corresponding to 138% of rated steam flow,

, in conjunction with the flow limiters and main st n m isolation valve clo-l sure, limits the mass inventory loss such that fuel is not uncovered. fuel temperatures remain approximately 1000 4 and release of radioactivity to

the er/ irons is weli below 10 CFR 100 guidelines. Ref. Section 14.6.5 of i the FSAR.

, 7. Main Steam ,Line Tunnel Temoerature Hioh 1

l Temperature monitoring instrumentation is provided in the main stea'n line tunnel to detect leaks in this area. Trips are pro

. tation and when exceeded cause a Group 1 isolation,'/ided its setting on this is low instrumen-enough to detect leaks of the order of five to 10 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, it is a back-

. up to high steam finw instrumentation discussed above, nd for small breaks.

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BASES FOR LlHITING CONDll10NS f0R OPfRA110N I occurs with each monitor indicating Hi H! H!, one monitor HI H! Hl.and the '

I other downscale. or with both monitors downstale. The Hi HI H! setpoint. cor.

responds to the instantaneous release limit. 3 l

1

2. Refuelina floor Exhtust Vent Radiation Monitors I- four radiation monitors are provided which initiate isoletion of the second- i

ary containment and operation of the standby gas treatment system. The in- ;

I strument channels monitor the radiation from the refueling area ventilation j exhaust ducts. 8 l Two instrument channels with two radiat'.cn monitors in each channel are ar.

I ranged in a two upscale (either channel) trip logic. Trip settings for the t monitors in the refueling floor exhaust ventilation ducts are based upon ini-

tiating normal ventilation isolation and standby gas treatment system or ra- }

tion so that none of the activity released during the refueling accident leaves ;

the reactor building via the e rmal ventilation path but rather all the ac- t tivity is processed by the standby gas treatment system.

i

! 3. Reactor Buildina Exhaus't Vent Radiation Monitors 1

! Four radiation monitors are provided which initiate secondary containment. iso-- -

i lation, primary containment purge and vent valves isolation and standby gas '

I treatment system actuation. The instrument channels monitor the radiation j from the reactor building lower level ventilation exhaust duct.

i i, Two instrument channels with two radiation detectors in each channel are ar-ranged in a two upscale (either channel) trip logic. The trip settings are -

{ based on limiting the release of radioactivity via the normal ventilation i path and rerouting this activity to be processed through the standby gas t treatment system.

j l 4. Control Room Intate Radiation Monitors

! Two radiation monitors are provided to initiate pressur13ation of the main I control room and recirculation of control room air through filters. - The i instrument (nannels monitor radiation.from the control room ventilation-l intake duct.

Two instrument channels are arranged in one upscale, two downscale trip log-

[

j ic. The trip settings are based on limiting the radioactivity from entering the control room from outside.

i . - .

5. Main Steam Line Radiation Moniton

[

l The four Main Steam 1.ine radiation monitors initiate isolation of the l

! mechanical vacuum pump and the gland seal exhauster condenser. The ,

instrument channels monitor the radiation in the main steam line tunnel, The purpose of automatically isolating the mechanical vacuum pump line ,

, is to provide timely protection against the release of radioactive 4

materials from the main condenser. Upon receipt'of main steam line-d high radiation signals, the primary containment and reactor vessel isolation control system initiates closure of the whanical vacuum pump line valve. This isolation precludes or limits the release of fission

. product radioactivity which; upon fuel failure would be transported from 1'

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0 Table 3.7-1 Primary Containment isolation Valves Which j

Receive a Primry Containment isolatio'. Yalve Signal l These riotes refer to the lower case letters in parentheses on the previous page.

f40TES:

4. Key: 0 Open SC = Stays closed C - Closed GC . Goes closed

b. Isolation Groupings are as follows:

GROUP 1: The valves in Group 1 are actuated by any Rag of the following conditions:

1. Reactor vessel water level Low Low Low (Level 1) i 2. Main steam line radiation high* l

3. Main steam line flow high j 4. Main steam line tunnel temperature high

! 5. Main steam line pressure low

6. Condenser vacuum low

7. Turbine building temperature at the steam lines high-CROUP 2: The valves in group 2 are actuated by any gtg of the following 1 conditions:

4 l 1. Reactor vessel water level low (Level 3)

2. Drywell pressure high t CROUP 3: Isolation valves in the high pressure coolant injection (HPCI)

system are actuated by any one of the following conditions

?

1. HPCI steam line flow high i 2. High temperature in the vicinity of the HPCI steam line j 3. HPCI steam supply pressure low

4. HPCI turbine exhaust diaphragm pressure

5. Torus room differential temperature high f GROUP 4: Primary Containment Isolation valves in the reactor core isolation cooling (RCIC) system are actuated by any one of the following conditions:

l. RCIC steam line flow high

2. High temperature in the vicinity of the RCIC steam line

3. RCIC steam line pressure low j 4. RCIC turbine exhaust diaphragm pressure high

5. Torus room differential temperature high i

)

Initiates closure of B31-F019 and B31-F020 only. l HATCH - UNIT 1 3.7-19 L:\wp\techsp\h\3_7-19UI. Pro.\l29..

' ~

3.12. MAIN CONTPOL ROOM [NVIPONMENTAL SYSTEM The control room air treatment system is designed to filter the control room atmosphere for intake air and/or for recirculation during #

pressurization conditions.

A. Ventilation System operability Reagirements The control room air treatment system operates on emergency power and is designed to filter the control room atmosphere for intake air and or recirculation air during control room pressurization conditions.

1he control room air treatment system is designed to automatically start upon receipt of an initiation signal and to align the system dampers to provide for pressurization of the control room.

Pressurization will be initiated upon receipt of any one of the follow.

ing signals: High radiation at control room intake, LOCA signal from Unit 1 or 2, main steam line hig5 flow from Unit 1 or 2, or l refueling floor high ratitation from Unit 1 or 2. In this mode the normal control room ehust fan is stopped and outside air is taken in through one of the charcoal filters to pressurize the control room with respect to the surrounding turbine building. High efficiency particulate air (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal-adsorbers are installed to reduce the potential intake of radiciodine to the control room. Bypass leakage for the charcoal adsorbers and partir. late removal efficiency for HEFA filters are determined by halogenated hydrocarbon and DOP, resp utively. The; laboratory carbon sample test results indicate a r4dioactive mothyl iodida removal efficiency for expected accident conditions. Operation of the fans significantly different from the design flow will chan efficiency of the HEPA filters and charcoal adsorbers ge the if the removal performances are as specified, the. calculated doses would be less than the allowable levels statsd in Critorion 19 of the General Design Criteria for Nuclear Power plants, Appendix A to 10 CFR Part 50 HATCH - UNIT 1 .3.12-4 .techsp\h\90-01ul\l56

m -

A Table 3.1-1 (Con t'd ) .

I M

O Sc ram Number Source or Scram Trip Signal Operable Channels Sene Trip setting Source of" Scram Signal is .

I Required Per Requi red tc= be Operable (a) Trip System Except as Indicated Below (b) 2

The APRM downscale trip is c-o 8 APRM Downscale 2 23/125 or rull scale H setive only when thre Mode w Switch Is In RUN. the APRM downscale trip is automatically bypassed when the IRM instrumentation is operable and not tripped.

157. Flux 2 515/125 or rull The APRM 15% Scram es auto-scale Tech Spec maticallf bypassed when the i 2.1.A.1.b. Mode Switch is in the RUN position.

C L i L ')

9 44aln3 team _LinAladla t lon 2 <1 times norest Net required-Jr a11_ steam-backg round _aL rs tad. 4+ net-ars isolated.

therast-powsrJ'8 l 10 Main Steam Line Isolation 4 580% valve closure Automatically by,,assed when Vaive Closure reom rulI open the Mode Switch Is not 8n P

g Tech Spec 2.1.A.5. the Run position. The design permits closure er any two 8 lines-without a scram being

'

Initiated.

11 Turbine Control Valve 2 Within 30 milli- Automatically bypassed when Fast Closure seconds or ti- sta rt turvine steam flow is below or control valve that corresponding to 30% of*

. rast closure ra ted the rma l power as measured Tech Spec 2.1.A.4. by turbine first stage p re s su re.

F cc ~

3 O-B tv 3

0 .

Z L5 e

W N

UT

~

%( _, .

~

j

~

iable 1.1-1 (Cont'd) ac , ..

Ej Se sem Opereble c

r3 K.mber Soorte of Stram Trip Signe! Channels Strs= Trip Setting Snurs e of Steam Signal is UC Required Per Required to be Operable (a) a Irip System [acept as Indita'ed Below c - . _ _ _ ___. ibi ___

[] I/ forhine Stop Valve 4 110% valve closure Aut pe.a t i c al l y bypas sed when CInsure from full apen turbine steam flow is below

~' tect. Spec 2.1.A.t. that corresponding to 30% of rated thermal power as i measured by turbine first stage pressure.

Notes for table 3.1-1

4. Ibe tolemme entitleil "Stram Number" is f or conveniente so that a one-to-one erlat ionsteip e en iw established between i* ems in Table 3.!-I and items ie table 4.1-l.

h. Therr shall be two operable or t ripped trip systems f or each potential scram signal. If the number of operable chan;.els cannot be met for one of the trip syste=s. the inoperable thaneelts) nr th+ assotsated l trip system shell be tripped. However.one trip signal channel o* 3 trip system may Se innperable f or up to two (2) hours during periods of required surveillante testing without tripping the a hann+ or assoc.ieted l trip system, pruvided that the other remaining thannel(s) monitoring that parameier within that trip system is (arel operable, e inr to t he planned start of the hydrogen injection test with the reactor power awute.-tIa,1

-w /' 207. sated power. the i ower radiation background level and associated r'p-setpottiny I.e thanged

$" based on a calculated value of the r Uta, v 1 espected durin a-4ert7' ebackgroundradiationleveland.l ch associated trip selpoints may be adjusted durin a n either <altw1ations or measurements of actual radiatinn levels result'ng from f.

( con. The backgrown r vel shall be determined and associated tr'

~

sia I be set within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of re-establishing norma vels after ydrogen injection and prior to establishing reactor power levels below 201 rated powe. . _

bY

'e

-3 C1.

El o 8 23 2"

O M

N

- V>

,----a---a,_,--

_ 1

T r

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mh b et -

t e . r se yt tv e s ot ei nl r -

~

se mre iou c pp l

dh brtr o. a t t e un t im od r ro noo tor tc rt cu e hd s

il o

e clf o cu tn og ul d oa sr dan e b l td e i r e enn rrh rr noa ot ,

ou nc ;i s fs s

a hyg e snw mo me te cbn ef n t er mp a tto s edr sri y l e yet s te bvM ssr ov aoR ne ip ni rmI pI s

) r e i n r t sd pge rei t

- n i tenh tt

- o d it ar h

- c so fe hii t

( l r e

obo tte o 1 nl rr t oih b

- bnt

- no eore i r 1 ar b mew r e o ort f 3 _ht cn o

ul o nbp.

' Je le ec a s te s b

l by eper erm a s c hacu muo T r ab tcuo oc t .

r ed pe rsedh th o d nr os o ov idrt ntn u o h f o ,rrg sgha soih s rm o

) oi ii e

os e er el st e g l s u lh t

o rn ei arrn l ns eg ctui eis ni N be snoh nhe ne .

mb t

noht wc ntrai ap hn i

nr o rw hwc e ci o Dl u h e l oh sv et he dafc li tl n t leei vr bw b arni bld a fa ,oiw aa rs ip hS rvd es

,c

>: ens voie pnr e e pe x ol us il wd oo ld tt o il rr Fo a.nM rtooar roo oo r esie l t rr

% pnth ron ee 5l oirt 1 o. n e t Pso ic bw mo Mtr rs iesf a up Rnu ,tno'nnbuey aI n l Poo Ach si n i ea p t r o. held iii e tr hm 8l) rnet tmv e l st Tihi ao fh da( ts fem it n h; o it arr gsep ,s g ,d ou imt n 2e 7 nf o HeeY 0ni 1t tlB 1i e a uo M s pD ab dr rin RymN dm n hti PsoA n r ar trh et A (p cT S eeo s 1 o

1si nw .d nT yoe 1%

SI M n M *uaO $cb S25 H M a M ll Aeo Ahe Rti Rtrr Adp Ao a Rna Rt Cst Cou Cac C Stter JbsT S Sr ris sf; ds e onn oorT rr eA rad oo rw FII F f pS ono Fir Fp hg

e:2 } w r 7mD ,

. M=tf . v

, I ,

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

,. Table 4.1-1 (Cont.)

O

Scram s ns t reement Check instrument functional Test inst rument Ca l ibra t ion ~

Humber Source of Scram Trip Signal Group Minimum frequency Minimum f requency Minimum frequency e (a) .

s _j bl,, (c)

C z 9

( L-

_ " ,

WWh) m uo disi, ied4etian n 9 rvorv t -nnehe r: r _ ,y 3 m g ,q:;

w 10 Maan Steam Line isolation Valve A NA Every 3 months (h) e-* Closure i

11 Turbine Cont rol Va lve f a s t A NA Every 3 months (J) Once/ Opera ting Closu re . Cyc l e ( k )

12 Turbine Stop Valve Closure A NA Evury 3 months (h) l RPS Channel Switch A NA Once/ Ope ra t i ng Cyc le Not Applicable Turbine first Stage Pn 1re A NA fvery 3 months Every 6 months Permissive

a. T he < a l umn en t i t l ed "Sc ra m Numbe r" is for convenience so that. a one-to-onc relationship can be established i

betseen items in Table 4.1-1 and items in Table 3.1-1.

b. 'ha definition for each of the four groups is as follows:

'j' Group A. Dr.-of f sensors that provide. a scram trip signal, m Group D. Analog devices coupled with bi-stable trips that provido a scram trip signal.

Croup C. Devices which only serve a useful function during some restricted mode of operation, such as startup'or shutdown, or for which the only practical test is one that can be performed at shutdown.

Croup D. Analog transmitters and trip units that provide a scram trip function.

c. functional tests are 03t required when the syste:as are not required to be operable or are tripped.

lioweve r, if functionst tests a re missed, they shall be performed prior to returning the systems to an operable status.

k

. ra s

d. Calibrations are not required when the systems are not required to be operable or are tripped.

However, if calibestions are missed, they shall be performed prior to returning the system to Q an operable status, to a e, this instrumentation is exempted from the instrument functiona! test definition. This instrument functional test will consist of injecting a simulated electrical signal into the measurement 2 channeis.

O

' ~

f. Deleted w

E

g. T he ve'.e r l eve l in the reactor will be perturbed and the corresponding level indicator changes will be monitored. This perturbation test will be performed every 3 months of ter completion of the

% functional test program.

m 7 h. Phys; cal inspection and actuation of these position swit.ches will be performed orce per operating cycle.

e-* -

1. Standa5Uc //hnt source used-which .providevan4nstrument .hannet-allgrr -libration-using a

$ Tedlet4en-source-shs4! be ==de ence per-cperat4ag-cyc4%

J. Measure time interval f rom EHC pressure swi tch actuation to RPS re'ay . Kill de-energization, i

t i

p ,

BASES FOR LIMITING CONDITIONS FOR OPERATION v

3.1.A.B.b. Inooerative An APRM is inoperable if there are less than two LPRM inputs per level or there are less than 11 LPRM inputs to the APRM channel.-

c. Downscale f

u/ The APRM downscale is active only when the Mode Switch is in the RUN position. The APRM downscale trip is automatically bypassed when the d

IRM instrumentation is operable and not tripped. Because of the APRM downscale limit of 13/125 of full scale hen in the Run Mode and igh level limit of $15/125 of full scale when the Start & Hot tandt...

Mode, the transition between the Start & Hot Standby and Run Modes must be made with the APRM instruentat'on indicating between 3/125 and 15/125 of full scale or a control rod scram will occer. In addition, the IRM system must be indicating below the High High Flux setting (120/125 of full scale) or a scram will occur when in the Start & Hot Standby Mode. For normal operating conditions, these limits provide assurance of overlap between the IRM system and APRM system so that there are no

" gaps' in the power level indications (i.e., the power level is evn-tinuously monitored f rom beginning of startup to full power and f rom f ull 4

power to shutdown). When power is being reduced, if a transfer to the Start & Hot Standby Mode is made and the IRM's have not been fully inserted (a maloperational but not imponible condition) a control rod block immediately occurs so that reactivity insertion by control rod withdrawal cannot occur.

4

d. 15% Flux .

The bases for the APRM 15% Flux Scram Trip Setting is discussed in the bases for Specification 2.1.A.1.b.

9. +4ain-Steam Line Mich Radiation (. At A d HighTradiation~1eiels in tha main steam ^line~ tun:Kabove that~due~to the#

I normal' nitrogen and oxygen radioactivity is an indic,r. ion of leaking-fsel.

A scram is init' lated-whenever such radiation level exceeds-s'iredetermined l 1evel above normal background, sThe purpose of this scram is to reduce ithe source of such radiation to the extent ~necessary to prevent excessive lturbinecontamination. Discharge of'exce'ssive amounts of radioactivity .

jto thy site environs,is-prevented by the off-gas pc3t' treatment radiation 6

! monitors which-ca n e an isolation of the main condenser off; gas-line provided

~

he;_ limit specified in the Environmental Technical Specifications i[exceede.d.3

10. Main Steam line Isolation valve Closure lhe bases for the Main Steam Line Isolation Valve Closure Scram Trip Setting is discussed in the bases for Specification 2.1.A.S.

11. Tur$ine Contrgl Valve Fast Closure The bases for the Turbine Control Valve Fast Closure Scram Trip Settin, '

is discussed in the bases for Specification 2.1.A.4.

12. Turbine Stoo Valve Closure The bases for the Turbine Stop Valve Closure Scram Trip Setting is dis-cussed in the bases for Specification 2.1. A.3.

HATCH - UNIT 1 3.1-13

Table 3.2-1 (Cont.)

I Required . --

$ lee t . Trip Operable Ac t ic,n t o I.e t ak evi it r7 No.

1 Instrement Condition Channels Trop Setting number of channels is (al Nomenclature per Irip not met for both trip Remarks (di

- _ _ _ . _ Srilem_1hl ___ __srsie*1_ LCL c-

a *.y*. n g *gW' *f ga n h s fe, < lo s .r ** s of res s $e **

x ps .r .ta **

, , 9, , 4, Jgt.papy G e reactre w* L te r J L*rS t /*5* & 6&/xq 4 Main Steam tiene fligh 2 13 times normal

tttat ;: ; d -1, lu.J f ;ii.iv> Group i-Radiation full power hacit- -*4 &t'.!:n ;- .; siow n3fn ^1r bt!;n ground m! W- n L- %

5 Main Steam line low 2 1825 psig Initiate an orderly load. Initiates Group i Pressure reduction and close isolation. NI y MSIVs within ti hours. required in RtM mode. .

therefore activated when Mode Switch is in RUN position,

t. Hein Steam line leigh 2 1138% rated flow initiate an orderly load Initiates Group I flow. (15 psidi reduction and close ttSivs isolation.

within 8 Loors.

7 Main Steam line High 2 1194*f Initiate an arderly load initiates Group I funnel Temperature reduction and close M51Vs isolation, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

w il Reactor Water High 1 20 B0 9pm Isolate reactor water (f}

Cleanup System cleanup system. f 7

w Differential Flow 9 Reactor Water High 2 5150*F lsolate reactor water Cleanup Area cleanup system.

leeperature I 10 Reactor Water liigh 2 167*r Isolate reactor water

$ Cleanup Area cleanup system.

cl. Ventilation

$3 DiIferential

. Temperature r+

z 11 Condenser Vacuum low 2 27" lig. vacuum Initiate an orderly load Initiate Group i O

reduction and close M51Vs isolation c, within 8 hrs.

U 12 Drywell Radiation liigh I il38 R/HR. Close the affected asulates isolation valves within containment pur p 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in Hot .and went valvet.

Shutdown within the cent 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shutdown within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Notes frr Table 3.2-1 .

-4 a. The cieluun entitled "Ref. No." is only for convenience 50 that a one-to-one relationship can be established ~ ~

2 between lines in Table 3.2-1 and items in Table 4.2-1.

b. primary conta*nment integrity shall be maintained at all times prior to e ithdrawing control rods for the Q purpose of going critical when the reactor is critical, or when the reactor water temperature is abose

-i 212*F aM fuel is in the reactor vessel encept while performing low-power physics tests at atmospheric pressure at power levels not to enceco 5 MWt. or performing an inservice vessel hydrostatic or leanage

- test.

When primary containment integrity is required. Chare shall be two operable or tripped trip systems fcr each function.

When performing inservice hydrastatic or leakage testing on the reactor vessel with the reactor coolant temperature above 212*f. reactor vessel water level instrumentation associated with the low low (Level 2) trip requires t e operable or tripped channelt.. The drywell pressure trip is not required because primary containment integrity is not required.

c. If the number of operable chn nels cannot be met for one of the trip systems the inoperable channel (s) or the associated trip system shall be tripped. Howcwer, one trip signal channel of a trip system may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during periods of requirect survelilance test sng without tripping the channel or associated trip system, provided that the other rersin'ng channel (s) monitoring that same parameter within that trip system is (are) operable.

F d. The valves associated with each Group isolation are given in Table 3.7-I.

N b e. Pit hln 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months rier to t'he plann tart of '5e hydr at eater th 0% }a,ted rnwer he n mal eu.!-pow enjection d14 tion bac te TNwith ap6 roupyewel We~h reactor sociate powe[r trip letpoit s may be chan~ ased on a .alculat valueofMhe lation leyl p4pect duriAg Ag 4cp gM W U* g [#q),. pd the test -The background diation level )qd ociated trip pointsmaybedjustedd i p'g p the tep based on eithe calculations or ma rements of ac al radiation ley 41s resulting om gh M O-hyd yrin ingction.Ae background radia)4'on Yawel shal e determiped a ' associated tr' se) points shar( bVset within 24gours,sf re-estal ish g normal radsat n levels aller emplets n g of hydrogen injection and prior tomfablishing re(atter powee levels b low 201 rated p r.

en g f. The,high dif f erential flow signal t's the INCO isolation valves may be bypassed f or up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months y during periods of system restoration, maintenance, or testing.

o z

P

%l W

I 'l s

i W

c

.A h Prior to the hydrogen injection system startup and with reactor power greater

than 20% rated power, the normal full power radiation trip / alarm setpoints may be changed based on calculated expected radiation levels during hydrogen injection system operation. Associated trip / alarm setpoints may be adjusted during injection based on either calculations or measurements of actual radiation levels resulting from hydrogen in#ction. Folicsing a reactor startup, a background radiation level will be determined and the associated trip / alarm setpoints adjusted within a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period. The radiation level shall be determined and associated trip / alarm setpreints shall be set within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of re-establishing normal radiation levels after a reducticn in, or a completion of hydrogen injection and prior to establishing reactor power levels below 20% rated power.

J t

k 4

E BA psME

. .. _ _. . __m _ .

~

p r. ^ m Table 3.2-6 (cont.)

I

~

n Ref.

Mo.

Instrument Trip Condition Required Ope rable Trip Setting Action to be taken i f" there are not two operable Remarks 1

(a) flomencia- Channels or trippe:S trip systems a ture pe r . ri p g System (b) z w

-A w

5. Itain Steam Line lli 2 53 times Isointe the mechanical one trip per trip Radiation Monitor normal full vectson pump and the logic system will power background gland sea t condenser isolate t%

- - exhauster mechanicab vacuum l pump and the gland sea t condenser exhauster.

a. The coe uen entitled "Ref. No." !s only for convenience so that a one-to-one relationship can be establir.hed between items in Table 3.2-8 and iteris in Table 4.2-8.

b. Whenever the systees are requirad to be operable, there shall be two operable or tripped trip systems.

If this cannot be met, the indicated action shall be taken.

F c. In the event that both off-gas post treatment radiation monitoes become inoperable, the reactor shall be placed in the Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless one monitor is sooner made operable, or adequate alternative ru 8 monitoring racilities are available.

d. From and arter the date that one of the two off-gas post treatment radiation monitors is made or round to be inoperable, continued reactor power operation is permissible during the next fourteen days (the a l lowab is repa i r t ime ), provided that the inoperable monitor is tripped. . . . ~ . . ~ . , , _ _ . . _ . .

e. ^Williin'2tI ~ hours priorTo~ the pjenned start ofi the hydrogen injection , test with thefreactor powerfi.

et' greater thart,20% reted pow 6r,\the norua P rs1I power / rad (ation becfground ieve V and. a ssoc i a tec '

trip'setpoir ts'may be changed based on a-talculsted value of t*tefadiation level expected dut'ing the test. sThe background'radittion Igvel and assop(ated tri% Setpoints may be adjusted'out'ing the test >Qased on eltatf calcultticns'or measurements of actifs1 radiation Jevels resultireys from

radiation levet shall he' determined and Associated, trip \

hydrogen Injection. .Thes. background setpoints shall be set within 24 pours (or re-establishing normat, radiat)6n levels af ter completion;/

of, hydrogen injection and prior,to establishjhg reactor power levels below 20% rated' power.

/

/ % / ' N / ]

/ .

n 5al* ~

, p9l & -

g y , .g m

3 e) 2 O

M M

N t.n

1.

4

.. E Prior to the hydrogen injection system startup and with reactor power gesater than 20% rated power, the normal full power radiation trip / alarm setpoints may be changed based on calculated expected radiation levels during hydrogen injection system operation, Associated trip / alarm setpoints may be adjusted h during trJection based on either calculations or measurements of actual 4

radiation levels resulting from hydrogen _ injection. Following a reactor startup, a background radiation level will be determined and the associated trip / alarm setpoints adjusted within a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period The radiation level shall be determined and associated trip / alarm setpoints shall be set within 24

' hours of re-establishing normal radiation levels after a reduction in, or a completion of nydrogen injection and prior to establishing reactor power levels below 20% rated power.

4 9

't J

b E E W

4 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2.A.2. Reactor Vessel Steam Dome Pressure iShutdown Coolino Mode) Low Permissive l This setpoint is chosen to preserve the pressure integrity cf the RHR system under conditions of increasing reactor pressure (startup). The RHR suction valves from the reactor (shutdown cooling mode) would be

('

closed when the 145 psig setpoint is reached. This function protects l 2

against RHR system pipe breaks during the shutdown cooling mode of op-e ra tion. Additionally, at reactor pressures below this setpoint the primary containment isolation signals are per1hitted to close the in-board motor operated injection valve (LPCI mode).

3. Drywell Pressure Hich The Bases for Drywell Pressure High are discussed in the Bases for Specif-ication 3.1.A.S. Pressure above the trip setting starts the SGTS ar.d in-itiates primary and secondary containment isolation.

4. Main Steam Line Radiation Hig h< < ep.. J YA

/a L o% e- *

f'"

Radiation monitors infthe main steam line tunnel have been provided to

, detect gross fuel fai1ure 4

as in the control rod drop accident. This in-strumentation causes , Crea Msolatiofr; With the established setting of approximately three times normal full power background, fission product re-lease is limited so that 10 CFp 100 guidelines are not exceeded for this

, accident. -Ref. ,ectier 14.4.4-f3AR-

5. S in Steam Line Presgure low The Bases for Main Steam Line Pressure Low are discussed in the Bases for Specification 2.1.A.6.

6. Main Steam Line Flow Hiah Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. In addition to monitoring st3am flow, instru-l mentation is provided which initiates Group 1 isolation. The primary func-tio's of the instrumentation is to detect a break in the main steam line.

For the worst case accident, a main steam line break outside the'drywell, the trip setting of 115 psid, corresponding to 138% of rateJ steam flow, l in conjunction with the flow limiters and main steam isolation valve clo-sure, limits the mass inventory loss such that fuel is not uncovered. Fuel

' ~

k temperatures remain approximately 1000'F and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Ref. Section 14.6.5 of the FSAR.

7. Main Steam Line Tunnel Temperature Hich

( Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in this area. Trips are provided on this instruman-tation and when exceeded cause a Group 1 isolation. Its setting is low enough to detect leaks of the order of five to 10 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, it is a back-up to high steam flow instrumentation discussed above, and for small breaks.

HATCH.- UNIT 1 3.2-51 Amendment No. 121

Aa_a -E 4eJ - - -

. ;1

i i

~

, BA$C$ FOR LIMlilNG CONDITIONS FOR OPERATION occurs with each monitor indicating Hi Hi Hi, one monitor Hi HI H1 and the __

' other downscale, or with both monitors downscale. The HI Hi H1 setpoint Cor ..-j responds to the instantaneous rclease . limit.

2. Re fuelina Floor Exhaust Vent Radiation Monitors 2

Fourradiationmonitusareprovidedwhichinitiatclisolationofthesecond i

ary ccotainment and operation of the standby gas treatment system. The in- -

strument channels monitor the radiation from the refueling area ventilation j~ exhaust ducts.

Two instrument channels with two radiation monitors in each channel are ar-

, , ranged in a two upscale (either channel) trip logic. Trip settings for the

monitors in the refueling floor exhaust ventilation ducts are based upon ini-tlating normal ventilation isolation and standby gas _ treatment system opera l

tion so that none of the activity released during the refueling accident let ves 4

the reactor building via the normal ventilation path but rather all the ac-j tivity _is processed by the standby gas treatment system.

3. Reactor Buildina Exhaust Vent Radiation Monitor 1 Four radiation mor.itors are provided which initiate secondary containment i!)-

lation, primary containment purge and vent valves isolation and standby gas 4

treatment system actuation. The instrument channels monitor the radiation

trom the reactor building lower level ventilation-exhaust duct.-

Two instrument channels with two radiation' detectors in each channel are ar-ranged in a two upscale (either channel) trip logic. . The trip settings are based on limiting the release of radioactivity via the nonnal v'entilation poth and rerouting this activity to be processed through the standby gas ].

j treatment system.

4. Control Room Intake Radiation Monitors l

Two radiation monitors are provided to. initiate pressurization of the main l f control room and recirculation of control room air through filters. The t instrument channels monitor radiation from the control room ventilation i l

intake' duct. !

! Two instrument channels are arranged in one upscale, two downscale trip log i

ic. The trip settings are based on Ihdting the radioactivity from enterin the control room from outside.

5. Main Steam Line Radiation Monitors l

-Althottskthek-pr4 mary function-is-to clost-the-MSIVs, ihe four Main Steam (

Line radiation monitors -ehe. initiate' isolation of the mechanical vacuum puop and the gland seal exhauster condenser. ' The instrument channels monitor thi '

radiation in ths main steam line tunnel. The purpece of automati; ally iso-lating the mechanical vacuum pump line is to provida timely protection.agairst [

the release of radioactive materials from the main condenser. Upon receipt m F of main steam line high radiation signals, the primary containment and;reac-tor vessel isolation control system initiates closure of the mechanical vac 'J:*

o uun pump-line valve. This isolation ~ precludes or-limits the release of fis-l sion product radioactivity whichi upon fuel _ failure would be transported fr m

? .

l . ;

j HATCH - UNIT 1 3.2-66 Amendment No. 156- ] l 1

l l l

l l I :

L

Table 3.7-1

(

Primary Containment Isolation Valves Which Receive a Primary Containment Isolation Valve Signal These notes refer to the lower case letters in parentheses on the previous page.

(' +

@TES:

a. Key: 0 = Open SC = Stays closcd C = Closed GC = Goes closed

( b. Isolation Groupings are as follows:

GROUP 1: The valves in Group 1 are act rated by any e of the following conditions:

1. Reactor vessel water level Low Low Low (Level 1)

2. Main steam line radiation high*

3. Main steam line flow high

4. Main steam line tunnel temperature high S. Main steam line pressure low

6. Condenser vacuum low

7. Turbine building temperature at the steam lines high l GROUP 2: The valves in group 2 are actuated by any one of that following

( conditiors:

1. Reactor vessel water level low (Level 3)

2. D.'ywell pressure high GROUP 3: Isolation valves in the high pressure coolant injection (HPCI) system are actuated by any one of the following conditions:

1. HPCI steam line flow high

2. High temperature in the vicinity of-the HPCI steam line

3. HPCI steam supply pressure low

4. HPCI turbine exhaust diaphragm pressure C. Torus room differential temperature high l

( GROUP 4: Primary Containment Isolation valves in the reactor core isolation cooling (RCIC) system are actuated by any one of the following conditions:

1. RCIC steam line flow high

2. High temperature in the vicinity of the RCIC steam line

( -

?

4.

RCIC steam line pressure' low RCIC turbine exhaust diaphragm pressure high

5. Torus room dif ferential temperature high

& ) ,m e ., < w . n_ . M F ui at WCL 6- !

k HATCH - UNIT 1 3.7-19 Amendment No. 793, 127, 129 l

i

! ,3.12.

< MAIN CONIROL ROOM ENVIRONMENTAL SYSTE!!

J The centrol room air treatment system is dest ned to filter the control -4 room atmosphere pressurization for intake cir and/or for rec rculation during conditions.

A. 'lentilation System Operability Reouirements

[

- The control room air treatment sys'.em operates on emergency power a.$d 2- is designed to filter the control room atmosphere for intake air and 1

or recirculation air during control room pressurization conditions. -l The control room air treatment system is designed to automatically start upon receipt of an initiation signal and to align the system l dampers to provide for pressurization of the control room.

l

-) y[

. p i

Pressurization will be initiated upon receipt of any one of the-follow-ing signals: High radiation at control room intake,- LOCA signal f rom Unit 1 or 2, ein str Mee-higtrradiaticr. frvo. Un it ' er 2, main steam line high flow from Unit 1 or 2, or refueling floor high -

' radiation f rom Unit 1 or 2. In this mode the normal control room i exhaust fan is stopped and outside air is taken in through one of the '

charcoal filters to pressurize the control room with respect to the

.' surrounding turbine building.

I :

2 High efficiency particulate air (HEPA) filters are installed before !!

l the charcoal adsoripers to prevent clogging of the iodine adsorbers.

The charcoal adsorbers are installed to reduce the potential intme of I,:

radioiodine to the control room. Bypass leakag, for the charcoal- '

adsorbers and-particulate removal efficiency for HEPA filters are <

determined by halogenatcd hydrocarbon and 00P, respectively. The

'.aboratory carbon sample test results indicate a radioactive methyl L iodide reinoval efficiency for expected accident conditions. Opera- !:

j

ion of the fans significantly different from the design flow will i cnange the removal ef ficiency of the HEPA filters and charcoal adsor-i bers. If the performances are as specified, the calculated doses 3} >)

, would be less than the allowable levels stated in Criterion 19 of the j

! General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR Part 50. e

I a

I

)

+

l i

l HATCH - UNIT 1 3.12-4 Amendment No. 22, 51, 156

T

. D TABLE 2.2.1-1 H

n # ..

I REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPo!NTS e

c Z FUNCTIONAL UNIT Tw SETP3 TNT ALLOWABLE VALUES

--4 co 1. intermediate Range Monitor, Neutron Flux-High s 120/125 divisions s 120/125 divisions (2C5 i-K601 A,0.C.D.E.F G.5 0 of fuh scale of full scale

2. Average Power Range Monitor:

(2CE i-K605 A.B.C.D E.F)

a. Neutron Flux-Upscale,15% s 15s125 divisions s 20/125 divisions of fut! scale of ferti scale

b. Flow hier :ced Simulated Thermal s (0.58 W + 59% - 0.58aW)" s (0.58 W + G2% - 0.58AWi" powei-Lyscale - with a maximum with a mexirmwn s 113.5% of RATED 5115.5% of RATED THERMAL POWER THERMAL POWER

c. Fixed Neutron Flux-Upscale,118% s 118% of RATED s 120% of RATED THERMAL POWER THERMAL POWER

3. Reactor Vessel Steam Dome Pressure - High s 1054 psig 51054 psig (2021-N678 A,B.C.D) 7 A

4. Reactor Vessel Water Level- Low (Level 3) 2 O inches above instrument zero*

2 O inches above (2821-N680 A,8.C.D) instrument zero*

5. Main Steam Line isolation Valve - Closure s 10% closed s 10% closed (NA)

6. (Deleted) l x 7. Drywett Pressure - High s 1.92 psig s 1.92 psig 3:C (2C71-N650A,0,C,D) v

/

r+

to g *See Bases Figure B 3/4 3-1.

m V *

W = Totalloop recirculation flow rate in percent of rated. Rated loop recirculation flow is equat to 34.2 MLB/hr. 1

r N AW = Maximum measured difference between two-loop and singe-loop drive flow for the same core flow in percent 1

3 of rated recirculation flow for single-loop operation. The value is zero for two-loop operation.

C to H

w W

7 O

s

>=*

C=d W

U- __ ..w - -. - - . _ _ . -. ,

i' 2.2 LIMITING SAFETY SYSTEM' SETTINGS BASES (Continued) i i REACTOR PROTECTION SYSTEM INSTRUMENTATION-SETPOINTS (Continued 1

~

3. Reactor Vessel Steam Dome Pressure-Hiah High pressure in the nuclear system could cause a rupture to the nuclear system process barrier reskiting in the release of fission

products. A pressure increase while operating will a?no tend to in-

crease the power of the. reactor by compressing voids thus adding i reactivity. The trip will quickly reduce the neutron flux,' counter-

[ acting the pressure increase by decreasing heat generation. The-trip setting is slightly higher than the operating pressure to permit norma'i operation without spurious trips. The setting provides for a wide margin

to the maximum allowable design pressure and takes into account the

. location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions-when the turbine stop valve closure trip is bypassed. For a turbine trip under these conditions, the transient analysis indicated a considerable margin to the thermal hydraulic limit.

4. Reactor Vessel Water Level-tow l The reactor vessel water level trip setpoint was chosen far enough

, below the normal operating level to avoid spurious trips but high enough

. above the fuel to essure that there is adequate protection for the fuel and pressure barriers.

. 5. Main Steam Line Isolation Valve- Closure S

The main steam line isolation valve closure trip w2 provided to limit

the amount of fission product release for-certain postulated events. The MSIVs are closed automatically from measured parameters such as high steam flow, low reactor water level, high steam tunnel temperature and low l steam line pressure. The MSIV' closure scram _ anticipates the pressure and flux transients which could follow MSIV closure, and thereby protects reactor vessel pressure and fuel-thermal / hydraulic Safety Limits.

6. (Deleted) a HATCH - UNIT 2 8 2-11 k:\wp\techsp\h\B2-1102. Pro \l4

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

l

.%-4 TABLE 3.3.1-1

.' m

= REACTOR PROTECTION SYSTEM INSTRUMEMTATION

.t APPLICABLE A N! MUM NUMBER E OPERATIONAL OPERABLE CHANNELS Q FUNCTIONAL (JNIT - CONDITIONS Pg3 TRtP SYSTEM M ACTION N

1. Intermediate Range Monitors:

(2C51-K601, A, B, C, D, E, F, G, H)

a. Neutron Flux - High .2ICI, SM 3 1 3, 4 2 2

b. Inoperative 2,5 M 3 1 3, 4 2 2

'2. Average Power Ranga Monitor:

(2C51-K605 A, B, C, D, E, F)

a. Neutron Flux - Upscale,15% 2, 5 - 2 1

b. Flow Referenced Simulated Thermal Power - Upscale 1 2 3

c. Fi: ed Neutron Flux -

W 1)nscale,118% 1 2 3 4 d. Inoptrative 1,2,5 2 4 w e. Downscale 1 2 3 N

8 f. LPRM 1, 2, 5 ' (d) NA

3. Reactor Vessel Steam Dome Pressure -

' High (2B21-N678 A, B C, D) 1, 2 I*3 2 5 y _ 4. Reactor Vessel Water Level- .;

L6w (Lev 9 3? (2821-N680 A, B, C, D1 1, 2 2' 5- t

E v

/

" 5. Main Stear., une isolation Valve -

$ Closure (NA) 18#

-4 J 7

m V 6. (Deleteo)

, l

r

( 7. Drywell Pressure - High (2C71-N650 A,8. C D) 1, 2 ICI 2 5 l,

c '

N

=-4 w

M 7

O

/- :I c3 CD

~ - ~ . . .

d

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION i ACTION i ACTION 1 - in OPERATIONAL CONDITION 2, be in at least HOT SHUIDOWN within 6-hours. ,

In OPERATIONAL CONDITION 5, suspend all o)erations involving

CORE ALTERATIONS or positive reactivity c1anges and fully insert all insertable control rods within one hour.

- ACTION 2 -

Lock the reacto mode switch in the Shutdown position within one hour.

]

l ACTION 3 - Be in at least STARTUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

I ACTION 4 - In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUTOOWN l within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 5, suspend all operations involving

CORE ALTERATIONS or positive reactivity changes and fully 4

insert all insertable control rods within one hour.

ACTION 5 -

Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l ACTION o -

(Deleted) l ACTION 7 -

Initiate a reduction in THERMAL POWEl ,ithin 15 minutes and be at less than 30% of RATED THERMAL bc.J9 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l ACTION 8 - In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUTDOWN l within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 3 or 4, immediately and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that all control rods are fully .

. inserted.

In OPERATIONAL CONDITION S, suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods within one hour.

1 4

HATCH - UNIT 2 3/4 3-4 k:\wp\techsp\h\3_4U2T14. Pro \8 7 - ,--m -

4 e n er- , , , m-- --y--n-

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

) Igj,U.3.1-1 (Continudi

! pfACTOR PROTECTION SYSTEM INSTRUMENTATION 4 ACTION 9 - In OPERATIONAL CONDITION 1 or 2, be in at'least HOT SHUTDOWN _

within 6 hourt, j In OpCRATIONAL CONDITION 3 or 4, lock the ractor mode switch

it the Shutdown psition within I hour.

! In OPERATIONAL CONDITION 5. suspend-all operattons involving.

CORE AllERATIONS or positive reactivity changes and fully i insert all insertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , TABLE NOTATIONS

a. A channel may be placed in in inoperable status for up'to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillance withoct placing the trip system in the tripped

condition p.ovided at least one OPERABLE channel in tiie same trip system 1s monitoring that parameter.

!. b. ;he "shcrting' links" shall be removed from the RPS circuitry during CORE ALTERATIONS and shutdown margin demonstrations performed in accordance with Specification 3.10.3.

l

. c. The IRM scrams are autom n ically bypassed when the reactor-vessel mode 1 : witch is in the Run position and all APRM channels are OPERABLE and on y scale.

4 d. An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 11 LPRM inputs to an APRM channel.

e. These functions are not required to be OPERABLE when the reactor j pressure vessel head is unbolted or removed.'

f. This function is automatically bypassed when the reactor mode switch is.

in other than the Run position.

l g. This function is not re. quired to be OPERABLE when PRIMARY CONTAINMENT l- INTEGRITY is not required.

h. With any control rod withdrawn. Not applicable to control rcJs removed

~

l

per Specification ?.9.11.1 or 3.9.11.2.

t

i. These functions are bypassed when turbine first stage' pressure is $250*

psig, equivalent to THERMAL POWER less than 30% of RATED THERMAL F0WER.

4 J. (Deleted) t-4 e

i f

Initial setpoint. Final setpoint to be determined 'during startup testing, i;

i i

i HATCH - UNIT 2 3/4 3-5 k:\wo\techsp\h\3_4U2T14 Pro \l00

. . . . -... - .. . .. - ~ . . - - . - . . . . . . - . . - . = - . .. .. - ..- - . .. .

[

--4 TABLE 3.3.1-2 S REACTOR PROTECTION SYSTEM RESPONSE TIA"ES E FUNCTIONAL UNIT RESPONSE TIME y (Seconds)

1. Intermediate Range Meretors:

a. Neutro t Flux - High* NA

b. Inoperative NA

2. Average Power Range Monitor *

s. Neutron Flux - Upscale.15%- NA

b. Flow Referenced Sirnulated Thermal Power - Upseste s 0.09"*

c. Fixed Neutron Flux - Upscale,118% s 0.09

d. Inoperative NA

e. Downscale NA

f. D'3M NA

3. Rerator Vessel Stearn Dorne Presstrre - High f; O.55 to N

A 4. Re stor Vessel Water Lever - Low s 1.05 W

$ 5. Iain Steam Line leoletion Valve - Closure s 0.06

6. (Deleted) l

7. Drywell Pressure - High NA 7:-

/

-* 8. Scram Discharge Volume Water Level- High NA t

" S. Turbine Stop Valve - Closurs s 0.06 f

<o

$ 10. -Turbine Control VWye Fast Closure,

[ T 'i Pressure - Low s 0.08 #

t I

/

11. Reactor Mode Switch in Shutdown Position NA w

I 12. Manual Scram NA a

c to

-4

[

Neutron detectors are exempt from response time testing. Response time shaff be measured from

, detector output or input of Orst electrorso component fra channel.

"O

$ *

Not including simulated thermal power time constan*.

/

Z # Measured from start of turbine control valve beure.

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

..y

[ TABLE 4.3.1-1 O REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEllLANCE REOUTREMENTS X

4 CHANNEL OPERATIONAL 5 CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH O FUNCTION AL UNIT CHECK TEST CA LIBR ATIOf{'** SURVEftLANCE REQUIRED t m

g 1. Inte nwnfiate Range Monitors:

e. Neestron Flux - High D SM*8 R 2

'O W R 3. 4, 5

b. Inuperative NA W NA 2,3,4,5

2. ' Average Power Range Monitor:

a. Neutron Flux - Upscale,15% S S/tfes, wm Sgfw, wm 2 S W W 5

b. Flow Referenced Simulated S SA/", O W'*"", SA 1 Thermal Power - Upecafe

c. Fixed Neutron Flux - Upsc de. S SAf". O W'*'. SA 1 118 %'

d. . Inoperative NA O NA 1, 2, 5

e. Downscete NA W NA 1

f. LPRM D NA "'

1. 2. S g 3. Reactor Vessel Steam Dome S Q R 1, 2

% Pressure - High i 4

g 4. Reactor Vessel Water Level- S Q R 1, 2 ,

a low (Level 3) i-N

' 5. Main Steam Line isotation Valve -

Closure NA Q R 1

6. (Deleted) .

X !

7. Drywelf Pressure - High - S O R 1, 2

E u 8. Scram Discharge Volume Water NA O RN 1, 2. S

' Level - High r+

rD n

7 m.

T

. /

7

/

W i

a C

ro H

w A-'

7 2

O

/

w O

_..,m ,

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

[

--4 TABLE 4.3.1-1 (Continued) g REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE HEQUiREMENTS CHANNEL OPERATIONAL C CHANNEL FUNCTIONAL CHANNEL CONDITIOtfS IN WHICH b

-4 FUNCTIONAL UNIT CHECK _ TEST Calf 8 RATION SURVEtLLANCE REQUIRED g 9. Turbine Stop Valve - Closure NA Q RN - 1

10. Turbine Control Valve Fast Closure Trip Oil Piessure -

Low NA Q R- 1

11. Reactor Mode Switch in Shutdown NA R NA- 1,2,3,4,S i Position i 12. Manual Scram NA W NA 1,2,3,4,5

a. Neutron detectors may be excluded from CHANNEL CALIBRATION.

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the orevious 7 days.

c. The APRM, IRM and SRM channels shall be compared for overlap during each startup, if not performed within the previous 7 days.

W d. When changing from CONDITION 1 to CONDITION 2, perform the required surveit:ance within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

) after entering CONDITION 2. !

W e. This calibration shall consist of the adjustment of the APRM chee.nel to conform to the power values E calculated by a heat balance during CONDITION 1 when THERMAL POWER 125% of RATED THERMAL POWER.

Adjust the APRM channel if the absolute difference 12%.

f. This calibration shall consist of the adjustment of the APRM flow referenced sirrnslated thermat power channel to conform to a calibrated flow signal.

7t-

g. The LPRM's shall be calibrated at least once per 1000 effactive full power hours (EFPH) using the .i TIP system. '

E" V h. Physical inspection and actuation of switches for instruments 2C11-NO13A. B, C. D.

/

r+

m I n

7 sn l V

i #

7 s

J w

l

.p.

a~

C ro

-4 w

A 1

O

/

w O

O 4

1

_ . .. . _ . . .....e ,_ . . - - - . . . . .-~ m. 4 ... . . . ~ _ ~___.m... . . .._ -. _ . . .m.. . . = . . .

--4 TABLE 3.3.2-1 m ISOLATION ACTUATION IN"TRUMENTATION g

I VALVE GROUPS MIN! MUM NOV*tER tiPPLICABLE 6 OPERATED BY . OPERABLE CF 4NELS OPERATIONAL O TRIP FUNCTION SIGNAlfal PER TR?P SYi EM(o)(c) CONDfTION ACTION

-4

" 1. PR! MARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level

1. Low (Level 31 2,6,10, 2 1,2,3 20 (2B21 N680 A, B, C D) 11,12

2. Low-Love (Level 2) 5.

2 1,2,3 20 (2B21 N682 A, B C, DI

3. Low-Low-Low (Level 11 1 2 1,2,3 20 (2B21-N681 A. B. C. DI

b. Drywell Pressure - High 2,6,7,10, 2 1, 2. 3 20 (2C71-N650 A, B, C, D) 12 *

c. Main Steam Line 1.- Radiation - H.gh 12, N 2 1,2,3,N 30 l (2011-K603 A, B, C, D)

2. Pressure - Low 1 2 1 22 ' I (2B21-NO15 A, B, C, D)

3. Flow - High 1, 2Mine 1,2,3 21 w (2821-N686 A, B, C, D)

N . (2B21-N687 A, B C, Di

(2B21-N688 A, B, C, DI w (2B21 N689 A, B, C, DI I

~ d. Main Steam Line Tunnel Temperature - High 1 2AineM 1,2,3 21 (2B21-N623 A, B, C, DI (2B21-N624 A, B, C, D)

(2821 N625 A, B, C, Di (2B21-N626 A, B, C, D)

,- e. Condenser Vacu em Low 1 2 1, 2,IU,3'O ' 23 E - (2B21-N056 A, B C D)

V

/

r f. Turbine wilding Area r0 Temperature - H~gh 1 2 I*I 1,2,3 21 Q. (2U61-P001, 2U61-r002, 2061-POO3,

~

v, - 2U6bPOO4)

V

'y-

- g.. Drywell Radiation - High (j) 1 1,2,3 29

/ (2D11-K621 A, B) w I

e C

N

= w

.M 9

O

./,

s.-

O to 4

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION ACTION 20 - Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 21 - Be in at least STARTUP with the main steam line isolation _ valves closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 22 - Be in at least STARTUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l ACTION 23 -

Be in at least STARTUP with the Group 1 isolation valves closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , i ACTION 24 - Establish SECONDARY CONTAINMENT INTEGRIlf with the standby gas treatment system operating within one hour.

ACTION 25 - Isolate the reactor water cleanup system.

/.CTION 26 -

Close the affected system isolation valves and declare the affected system inoperable.

ACTION 27 -

Verify power availability to the bus at'least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or close the affected system isolation valves and declare the affected system inoperable.

ACTION 28 -

Close the shutdown cooling supply and reactor vessel head spray 1

isolation valves unless reactor steam dome pressure s 145-psig. l ACTION 29 -

Either close the affected isolation valves within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30. hours.

ACTION 30 - Trip and isolate the mechanical vacuum pump and isolate the reactor water sample valves.

NOTES 4

Actuates the standby gas treatment system.

When handling irradiated fuel _in the secondary containment.

When performing inservice hydrostatic or leak testing with the reactor coolant temperature above 212 F.

t. See Specification 3.6.3, Table 3.6.3-1 for valves _in each valve group,

b. A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for l

required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip j system is monitoring that parameter.

, HATCH - UNIT 2 3/4 3-15 k:\wp\techsp\h\3_4U2T14. Pro \l20 l

TABLE 3.3.2-1 (Continued)

c. With a design providing only one channel per trip system, an inoperable

channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken,

d. Trips the mechanical vacuum pumps.

! e. A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE.

f. May be bypassed with all turbine stop valves closed.

g. Closes only RWCU outlet isolation valve 2G31-F004.

h. Alarm only.

1. Adjustable up to 60 minutes, j j. Isolates containment purge and vent valves.

k. Prior to the hydrogen injection system startup and with reactor power greater than 20% rated power, the normal full power radiation trip / alarm setpoints may be changed based on calculated expected radiation levels i during hydrogen injection system operation. Associated trip / alarm

setpoints may be adjusted during injection based on either calculations or measurements of actual radiation levels resulting from hydrogen injection.

Following a startup, a background radiation level will be determined and the arsociated trip / alarm setpoints adjusted within a 72-hour period. The radiation level shall-be determined and associated trip / alarm setpoints shall be set within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of re-establishing normal radiation levels after a reduction in, or a completion of, hydrogen injection and prior to establishing reactor power levels below 20% rated power.

1. The high differential flow isolation signal to the RWCU isolation valves may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during periods of system restorat. ion, maintenance or testing,

m. Isolates reactor water sample valves 2831-F019 and 2831-F020. These are

Group 1 valves.

1 i

HATCH - UNIT 2 3/4 3-15a k:\wp\techsp\h\3_4U2T14. Pro \120

. . . . . . .. . . . _ . .. _. . . . _ m _ _ . . . _. m _ _ . -

-4 TABLE 3.3.2-2 r7 ISOLADON ACTUATtON INSTRUMENTATION SETPOtNTS r

i ALLOWABLE TRIP SETPOINT VALUE

@ TRIP FUNCTION U 1. PRIMAftY CONTAINMENT ISOLATION

" Reactor Vessel Water Level a.

1. Low (Level 31 a O inches

a 0 inches *

2. Low Low (Level 2) 2 -47 inches

2 -47 inches?

3. Low Low Low (Level 1) 2 -113 inches

2 -113 inches *

b. Drywe5 Pressure- High 51.92 psig s 1.92 psig

c. Main Steam Line

1. Radiation - Mgh s 3 x fuupower background" s 3 x funpower background * *

2. Pressure - Low a 825 psig a 825 paig

3. Flow - High s 138% rated flow s 138% rated flow

d. Main Steam Line Tunnel e Temperature - High 5 194*F s 194*F

e. Condenser Vacuum - Low 2 7* Hg vacuum 2 7* Hg vacuum

" s 200*F s 2OO*F

% f. Turbine Building Area Temp.-High

.p.

g g. Drywen Radiation- High 5138 R/hr 5138 R/hr I

w 2. SECONDARY CONTAIP. MENT ISOLATION ci

s. Reactor Suilding Exhaust Rediation - High 5 60 rnt/hr s 60 mr/hr

b. Drywou Pressure- Koh s 1.92 psig s 1.92 peig

c. Reactor Vessel Water r Level- Low Low (Level 2) 2 -47 inches

z -47 inches

r0 O d. Refueling Floor Exhaust V

[ Radiation - High s 20 mr/hr . 5 20 mr/hr

/

3w 'See Bases Figure B 3/4 31.

! " Prior to the hydrogen injection system startup and with reactor power g eater than 20% rated power. the normal fun power rediation tr!p/ alarm

$ setpoints may be changed based on calculated expected radiation levels during hydrogen injection system operation. Associated trip /ala m setpoints may ru be adjusted during injection based on either calculations or measurements of actual radiation levels resulting from hydrogen inje. tion. Following a reeetor

-4 startup, a background radiation level win be determined and the associated trip /aiarm setpoints adjusted within a 72-hour period. The radiation level shau

be determined and associated trip /elarm setpoints shan be set within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of re-establishing normal radiation levels after a reduction in, ce a

. completion of, hydrogen injection and prior to establishing reactor power levels below 20% of rated power.

t 1

O

/

w W

W

. - _ _ . .. . - . m. . . . - ._.. .... _ _ . .. . . _ . . _ . m. _ _ . . _ , . ._m_. . . - ... -

m. ._

l

~

l- ' h. . TABLE 3.3.6.7-1 (SHEET 1 OF 2) i H

b MCRECS ACTUATION INSTitUMENTATION a

5 MINIMUM NUMBER APPLICABLE Q . OPERAdLE CHANNELS OPERATIONAL g TRIP FUNCTION PER TRfP SYSTEM (aHbl CONDITION ACTION

1. Reactor Vessel Water Level - '2' 1, 2, 3 ' 52 i Low Low Low (Level 1) (c) 2821-N691 A,8, C, D

' 2. Drywell Pressure - High (c) 2 1, 2, 3 52 -

.- 2E11-N694 A, B, C, D

3. (Deleted) -

l.

4. Main Steam Une Flow + High (c) 2/ tine - .1, 2, 3 53 2B21-N686 A, B, C, D 2B21-N687 A,8, C, D -

2821-N688 A, B, C D

" 2B21-N689 A, B, C, D

w .5. Refueling Floor Area Radiation - High (c) i 1,2,3,5.* 54 M

m 2D21-KOO2 A. D u .

6. Control Room Air Irdet Radaation - High (c) 1 1,2,3,5,* S4 1Z41-R615 A, B '

+

r0 O

T t)>

.V

/

llT

/

O I

O N

c N

.'C/ .

Ca

4 i

I TABLE 3.3.6.7-1 (SHEET 2 0F 2) i- MCRECS ACTUATION INSTRUMENTATION j ACTION ACTION 52 - Take the ACTION required by Specification 3.3.3.

l ACTION 53 -

Take the ACTION required by Specification 3.3.2.

t ACTION 54 -

.I i a. With one of the required radiation enitors inoperable, restore the i monitor to OPERABLE status within 7 days or, within the-next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , initiate and maintain operation of the MCRECS in the pressurization

. mode of operation.

l b. With no radiation monitors OPERABLE, within I hour initiate and

! maintain operation of the MCRECS in the pressurization mode of operation.

c. The nrovisions of Specification 3.0.4 are not applicable.

NOTES

When handling irradiated fuel in secondary containment.

a. A channel may be placed in an inoperable status for up to 2- hours for required surveillance without placing the trip system in the tripped ccndition, provided at least one other OPERABLE channel in the same trip system is mor,itoring that parameter.

b. With a design providing only one channel per trip system, an ' inoperable channel need not be placed in the tripped condition where this would-cause the-Tc. Function to occur. In these cases the inoperable channel- shall bi restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.6.7-1 for that Trip. Function shall be taken,

c. Actuates the MCRECS in the control room pressurization mode,

d. (Deleted)

e. (Deleted) l 1

a I

. I l

HATCH - UNIT 2 3/4 3-53b techsp\h\90-07AU2\96 a

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

- ..m-.... m . m. ..___m. . . . . . . . . . ., . _ - - .

-4 TABLE 3.3.67/-2

$ MCRECS ACTUATION INSTRUMENTATION SETPOtNTS i c ~ TRIP FUNCTION TRtP Sc7 POINT ALLOWABLE VALUE M 1. Reactor Vesse! Water Level- 2 -113 inches a - 113 inches g Low Low Low (Levet 11

'2. Drywell Pressure - High 51.92 psig s 1.92 psig

~ 3. (Deleted) l 4 Main Steam Line Flow - High s 138% rated flow s 138% ratM flow

5. Refueling Floor Area Radiation - Kgh s 20 mr/ hour 5 20 mr/ hour 6, Control Room Air inlet I s 1 mr/ hour s 1 rnethour .

Radiation - High .

w Ab -[

W 8'

' (.FI CX3 O

4 c+

2 O ,

.7 V"

/

=f"

/ -

g

, o 8

.O N

M

.' c N

/

LO Ch 4

- =. ---f ~

v v _ - - - - - - -

y-4 T ACE 4.3.6.7-1 9 MCRECS ACTb A3 TON INSTRUteENTATION SinVETLLANCE REctrnEVENTS

' CHANNEL OFTRATIONAL CHANNEL rU?tCT!ONAL CHA*sNEL COND4TIONS IN WHICH c C AUEA ATIC** SURVE'LLANCE REQUrPED CHECW TEST 3

-4 TR!p rUNCTIO*J S M R 1, 2. 3 g 1. Reactor Vessel Water tevel -

Lo=, Low Low (Level 1)

Drywell Pressure - Kgh S M M 1. 2. 3 2.

2. (Deleted) l S M R 1. *z. 3

4. Min Steam U-w Thw - Rgh

5. RefueEng Floor Ares Rad >atwin - 0 M'd Q 1. 2. 3. 5

Kgh

6. Control Roorn Air Idet NA M'd R 1. 2. 3. 5.

Rrdatmn - Mgh w

N W

8 tre CD

& 1 l

Mhn handiirq irradiated fael en the secorrjary contsnmertt,

s. Irstrumeest abgmu ss.r g a standard current source.

r+

C O

7 en V

/

7

/

O CD t

CD N

c N

/

tO CD

1 4

PLANT SYSTEMS 1

j SVRVEILLANCE RE00 REMENTS (Continued)

3. Verifying that on each of the below pressurization mode actuation test i gnals, the system automatically switches

!- to the pressurization modo of operation and maintains the j main contr?1 room at a positive pressure of h 0.1-in.

1 W.G. relative to the adjacent turbine building during i system operation at a flow rate s; 400 cfm.

a) Reactor. vessel water level - low low low l b) Drywell pressure - high c) Refueling floor area radiation - high l d) (Deleted) l 1

[

i e) Main steam line flow - high

! f) Control room intake monitors radiation - high

f. After each complete or partial replacement of a !! EPA filter i bank by verifying that the HEPA fi:ter banks-reniove a 99 percent l of the DOP when they are tested in-place in accordance with ANSI N 510- 197'.: while operating the system at a flow rate of 256 cfm i 10 percent.

i

g. After each cornnlete or partial replacement of a charcoal adsorber banP ty verifying that the charcoal adsorbers remove

{ 2 99 percent of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in acccrdance with ANSI N510-1975 l; while operating the system at a ficw rate of 2500 cfm 10 percent.

i i

1 j

i f i 1

i i

HATCH - UNIT 2l 3/4'7-8 techsp\h\90-07u2\96

[

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

u -

T.

3 TABLE A.2.1-1

-4 o

Z REACTO84 PROTECTION SYSTEM tPtSTRf>MENTATTON SETPOINTS e

c

$ FUNCTIONAL 11NfT TR?P SETPotNT ALLOWABLE VALUES

-4 m 1. Intemwdiste Range Monitor. Neutron Flun#gh s 120/125 dveesene s 120/125 dwielore (2C51-K601 A.B.C.D.E.F.G.HI of fur ocale of fur oce8e

2. Average Power Renge A4onitor:

(2C51 K605 A.B.C.D.E.F) l

e. Neutron No-Upecese 15% s 15/125 dn4elone s 20/125 dvis%;

of fun eesie of fus oe se

b. Flow Referewed Simuleted Thermal s 10.58 W + 53% - 0.58AWi" s 90.58 W + 62% - 0.58AW1**

Power 4Jpecele with e enswimurn wtth a membawva j s 113.5% ef RATYD s 115.5% of RATED THERMAL POWER THERMALP@!6TM

e. Fired Neutron Fkse-Upeeste,118% .3118% of RATED s 120% of RATID THERMAL POWER THERMAL PCWER

3. fteactor Veeeel Steam Dome Pressure - High 51054 pois s 1054pe4 (2821-N678 A.B.C.01 7 4. boeter Yeeed Water Level - Low (Level 3) 2 O Inches ebeve P O incfte= above

(2821-N6.80 A,8.C.DI Irm* swnt rero* Instrument rere*

5. P"ein Steem Line ledetion Vefve - Coeure s 10% clesed s 20% cosed (NA) g .. ..._. - ,-. - _ g .e.. . , . - -

o. eenseosA,arcrDI cpu 4 t e . _-J u' M'.-2"*

Er S 7. Drywes Pressure - High s 1.92 peig 51.92 peig

(2C71-N550A.B.C.D) 2 o

b y *See Beees Figure B 3/4 3-1. -

" W = Toter loop recirculation flow rete 6. percent of reted. Reted loop ;&.Je-Jm, flow le egud to 34.2143& .

e AW = Merimum measured difference between tws4eop and G. -@ drive flow for the some core hw in percent of reted recircu8etion hw for .Jngfe4oop operetion. The verse le rero for two4oop operation.

7'*w+=ma24 heure prior to the pierined start of the hydrogen inrection teet with the reectorpearerit; g

f greater Sanb%wer, tte normel fur-power radietion boekground level, and eveociated arty

. eetpointe may be chenged Eaeedau edeutered value of the esteth a tevd emeeted during the t et. {

Ttw background e distion level end esshoc eejpeawrC=y be odrated during the test bened on h : either calculatione or measuremente of er backgreemd redation level e esteen tevela reeutting troen hydrogen snrect!an. The ermined e uf eseociated t41regcante eSeE bs - swa thin 24 w hours of re-est . ' redation leveis efter -,A^a. of hydrogast prior to y attebbehrroJTesetor power levele below 20% rated power.

- M q.,

aL<a _. _ _.-A _._i.A _. ag_4+-m_mmi___.__a_ _4_ - -

__..n._ a ..<_-_u.J _ _.,__.._s._4. e _a,_ .a 4

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) '

a- .

! REACTOR PROTECTION SiSTEM INSTRUMENTATION SETPOINTS (Continued)

( 3. Reactor Vessel Steam Dome pressure-High High pressure in the nuclear system could cause a rupture to the J nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to in-crease tha power of the reactor by compressing voids thus adding

(' reactivity. r ne tr1p will quickly reduce the neutron flux, counter-acting the pressure increase by decreasing heat generation. The trip setting is slightly higher than the aperating pressure to permit normal i operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure

that occurs In the system during a transient. This trip setpoint is

effective at low power / flow conditions wher the turbine stop valys closure trip is bypassed. For a turbine trip under these conditions,

the transient analysis indicated a considerable margin to the *: hermal hydrau!ic limit.

4. Reactor Vessel Water Level-Low

>(-

The reactor vessei water level trip setpnint was chosen far enougn below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection fnr the fusi cnd pressure barriers.

5. Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain ;astulated events. The MSIVs are closed automatically from measured parameters such as high steam flow, -M 9 h-um ' " - " -t4enf low reactor water level, high steam tunnel temperature and low steam line pressure. The MSIV closure scram anticipates the pressure and flux transients which could follow

( MSIV closcre, and thereby protects reactor vessel pressure and fuel thermal /hyraulic Safety Limits.

6. Ma M h/ h tae h d44ti p uich.

f-( pross failiTFQfuel cladding.W main 4 trip is initiated toWucQhe continued _fai-lure ~6f fuel cladding.

When steam the high 'ine ra3 ation radjation~is detec detectors are provi At the same time the main steamTne tA lition valves are closed to imit the releue of fission-products, e etting is high enough j

above backgrop-adtnTon levels to prevent spurious - p et low jnough togromptly detect grosQilures in the fuel cladding.

HATCH - UNIT 2 B 2-11 Amendment No.-14

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

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

T T ABL E 3. 3.1-1 l REACTOR PROTECTION SYSTEM INSTRUMENTATION h

-0 APPLICABLE MINIMUM NUMBER O

OPERATIOMAL OPERABLE CHANNELS FUNCTl0NAL UNIT CohDITIONS PER TRI P SYSTEMt al ACTION e

, c- 1. Intermediatu Range Monitors:

z (2C51-K601, A, B, C, D. E, F. C, H)

e. Neutron Flux - High 2 8 *',5'*8 3 1 I N 3, 4 2 2 I

b. Inope ra tive 2, $ '

8- 3 1 3, 4 2 2

2. Average Power Range Monitor:

(2C51-K605 A, 6, C, D, C. F) t

a. Neutron Flux - Upscale, 15% 2, 5 2 1

b. Flow Referenced Simulated ,

The rma l Powe r - Upsca le 1 2 3

c. Fixed Iseutron Flux -

Upscale, 118% 1 2 3 i

d.

nope ra t Ivo 1,2,5 2 4

e. Dwnscale- 1 2 3 N

to f". LPRM 1, 2, 5 (d) MA >

4 i

.t . Reactor Vessel Steam Dome Pressure -

! y High (2821-3678 A, B, C D) 1, 2'** 2 5 ra

^ '

4. Reactor Vessel Water Level -

y Low (Level 3) (2B21-N6SO A, B, C, D) 1, 2 2 5

5. Main Steam Line Isolation Valve -

Closure (MA) .

18'8 4 3

( D.RJ.JlLT e,i

6. 49e4*-Ste r L'nc .td ;;;;n ";;h , 2.i, 1

7 ,

~

g s

j g (M14-4404 ^., S,-- C, 9 )

$ 7. Drywe l l Pressure - High 1, 2 2 5 g (2CTI-N650 A, B, C, D) rs II

, et Z

.O N

M N

.N N

.M ,

O>

l LV

= ?

+ b

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

i I

i TABLE 3.3.1-1 (Continued) 1.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.

j ACTION i

ACTION 1 In OPERATIONAL CONDITION 2, be in at least HOT SHUTDOWN

)l within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 5, suspend all operations involving !

! CORE ALTERAT!ONS or positive reactivity changes and fully j 1

insert all insertable control rods within one hour.

ACTION 2 -

Lock the reactor mode switch in the Shutdown position within I j one hour.

j ACTION 3 -

Be in at least STARTUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . I ACTION 4 -

In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUTOOWN l within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . l '

j In OPERATIONAL CONDITION 5 suspend all operations involving CORE ALTERATIONS or positivr reactivity changes and fully insert all insertable control rods within one hour.

l ACTION 5 -

Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

L o. td ..O 4

1 ACTION 6 - ;

I h i- STARTUP-with-the-main-steam 44ne-4so44thon--v44ves-cA4 sed with4r 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s-4n-in-at-4 east-HOT-440TDOWM-w(-th4*-4-heum I i ACTION 7 -

Initiate a reduction in THERMAL POWER within 15 minutes and i

be at less than 30% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

i i ACTION 8 -

In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUT 00WN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . l

(

In OPERATIONAL CONDITION 3 or 4, immediately and at least i

once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that all control rods are fully j inserted.

4 l

In OPERATIONAL CONDITION 5, suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully )

i insert all insertable control rods within one hour.

1 i .

J-

)

), ,

HATCH - UNIT 2 3/4 3-4

> Amendment No. 8 1

+

,, -, - , . _ ~ _ _ - - , , . + ,,m.-..y,,- --- 4 ,rw.... , m, -, . v r - - , - - _e,

1 i l TABLE 3.3.1-1 fContinued)

REACTOR PROTECT!0N SYSTEM INSTRLHENTATION ACTION t - 1r OPERATIONAL CONDITION 1 or 2, be in at least HOT $HUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATIONAL CONDITION 3 or 4, lock the reactor mode swit-in the Shutdown position within I hour. l In OPERATIONAL CONDITION 5, suspend all operations involving CORE ALTERAT!ONS or positive reactivity changes and fully 4

j insert all insertable control rods within i hour.

l I

TABLE NOTATIONS

(

' a. A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for 4

required surveillance without placing the trip system in the triprad l I

condition provided at least one OPERABLE channel in the same trip system ,

is monitoring that parneter,

b. The " shorting links" shall be removed from the RPS circuitry duHng CORE j

ALTERATIONS and shutdown margin demonstrations performed in accordance 5

with Specification 3.10.3.

c. The IRH scrams are automatically bypassed when the reactor vessel. mode switch is in the Run position and all /,?RM channels are OPERABLE and on scale,

' d. An APRM channel 1$ inoperable if there are less than ? LPRM inputs per level or less than 11 LPRM irouts to an APRM channel. l

e. These function are not required to be OPERABLE when the reactor pressure vesse. need is unbolted or removed.

f. This function is automatically bypassed when the reactor mode switch is in other than the Run position.

g. This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRlTY is not required.

h. With ary control rod withdrawn. Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.

1. Tnese functions are bypassed when turbine first stage pressure is s250*

psig, equivalent to THERMAL POWER less than 30'. of RATED THERMAL POWER.

j. WC oint 24 t= <h ~

hours prior to thf'pl U sid'sTt W 6flhT hydrogen injection

(, with eactor power at greater than 20% rated power, the no ull-

' power radia b Q ackground level and associated trip se s may be changed based on N Qlated value of the radia evel espect3d during the test. eral and associated trip The bacinpou setpoints may be edjusted d radiation) during he. Jest based on either calculations or messurements of actual radiatiorr-17velh5 ting from hydrogen inje: tion. The backgjr uad iadiation level sha determined and asso:14ted trip Jetpoints shall be set within 24 hou re establishing normtl r pattfn levels after completion of hydrogen inject d prior tJ, wutnishing reactor power levels below 20% rated power.

T tial setpoint. Final setpoint to be determined during startup testing.

HATCH - UNIT 2 3/4 3-5 Amendment No. 8,2),5)l,69,88 100

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=

>- TABLE ts.3.1-1 S REACTOR PROTECTION SYSTEM INSTRtHENTAT!OM SURVElLLANCE REQUIREMEMTS C CHANNEL OPERAT90 MAL

-.4 U CHACMEL CHECM FUNCTIONAL TEST CHANNEL CAllBRATIOM

CONDITIONS IN WHICH SURVEILLANCE REQUIRED f_UNCTIONAL UNIT

" 1 Intermediate Range Monitors:

a. Neutron Flux - High D S/U' R 2 D W R 3, 4, 5

b. Incpe ra tive NA W MA 2,3,4,5

2. Average Power Range Monitor:

s. Neutron Flux - Upscale, 15% S S/U ' ' " * * , W' ' ' ", W'** 2 S V W 5

b. Flow Referenced Simuisted S S/U, Q V , SA 1 TheresI Powe. - UpseeIe

c. Flxed Neutrcn Flux - Upscale, S S/U, Q W, SA 1 118%

d. Inope ra ti ve NA Q NA 1, 2, 5

e. Downscale NA W MA 1

r. LPRM D MA 1, 2, 5 .

Nh

3. Reactor Vessel Steam Dome S Q R 1, 2 Pressure - High u . .

e 4. Reactor vessel Water Level - S Q R 1, 2 tow (Leve1 3) .

5. Main stesa Line Isolstich valve -

2 Closure NA Q R 1 l

6. 9 kS - ' *rd!:0!;n ' ;gn O G 7 *

,2

7. Dryve l i Pressure - High S ta R 1, 2

$ 8. Scram Dische e Volume Water " MA Q R'** 1,2,5 g Level - Hig a

3 m

3 c+

1 Z

.O W

' Oh Jan e M

M e

M D

D.

. _ . . _ . _ . - _- . _ _ _ m ._ . . ____._m._ . . _ . _ - . _ . _ . . . - - - . _ _ . . _ . ._ _ .__m __. ._.

L i TABLE 4.3.1 , (Continued)

[- REACTOR PROTECTION *iYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

-.4 n

CitANNEL. OPERAT IONAL 8 CHANNEL FUNCTeOMAL CHANNEL ColsDITIONS IN WHiCH FUNCTIONAL UNIT- CHECK TEST CALJDRATION SURVEILLANCE REQUIRED c-z R"88 y 9. Turbine Stop Value - Clost:re NA Q 1 l 4

to 10. Turbine control valve e ast

, Closure, Trip Oil Pressure -

Low NA Q R 1

11. ' Reactor Mode Switch in Shutdown MA R NA 1, 2,3,4,5 Position

12. Manual Sc ram NA W MA 1,2,3,4,5 l

a. Neutron detectors may be excluded from CHANNEL CA13BRATION.

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the prevleus 7 days.

c. The APRM, 'IRM and SRM channels shall be compared for ove. lap during each startup, if not performed within the previous 7 days, w

) d. When changing frus CONDITION 1 to CONDITION 2, perform the required surveillance within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months af ter entering CONDITION 2.

w

[p e. This calibration shall consist of the adjustment of' the APPM channel to conf'orm to the power values calculated by a heat balance during CONDITION 1 when THERMAL POWER 2 2% of RATED THERMAL POWER.

Adjust the APRM channel if the absolute difference 2 2%.

f. This calibration shall consist of the adjustment of the APRM flow referenced simulated t'eermal power channel to conform to a calibrated flow signal,

g. The LPRM's shall be calibrated at least once per 1000 effective full pcwer hours (EFPH) using the TIP system.

h. Physica! Inspection and actuation or switches for instruments 2C11-NO13A, B, C, D.

8, 'as m - at e44;-- e !eg r eteM:re-current-+ourco.

m cL J. C*llh!21.lon-using a c44adard,rgd43tlen te' m ,

8 23 et 2

O e

'M 43

+

O +

O w . g-

. _ .. m __ _ . . - . _ . _ __. .m. _ _ _ m - . . ~ .- .__. .-.__._m _ _ _ . . . -.._.m.

j / p '

f,

(

t i

TABLE 3.3.2-1 4

_C L -

3 ISOLATION ACTUAlION INSTRtNENTATION Nx-VALVE GROUPS MINIMUM NUMBER APPLICA3LC OPERATED Bf OPERABLE CHANNELS OPERATIONAL E TRIP FUNCTION SIGNAlfan P'R TRIP SYSTEMfb?fel _ C0f0ITION , ACTION O

-4 1. PRIMARY CONTAIP.NENT ISOLATIQtj N a. Reactor Vessel Water Level

1. tow (Level 3) 2, 6, 10, 2 1, 2, 3 20 +

(2821-N680 A B, C, 0) 11, 12

2. tow-tow (tevel 2) 5.

2 1, 2, 3 20 ,

(2821-N682 A, B C, D)

3. Lew-tow-Low (Level 1) 1 2 1, 2, 3 20 (2B21-N681 A, B, C, D)

? b. Drywell Pressure - High 2, 6, 7, 10, 2 1, 2, 3 20 t (2C71-N650 A, 8. C, c) 12, *

c. Main Steam Line-I. Radiation - High f$ 12, * * 'N 2 1, 2, 3," ' X30 (2Dll-K663 A, 8 C D)

. 2. Pressure - Low 1 2 I 22 w _

(2B21-N015 A, B, C D)

N 3. Flow --High 1, 2/line 1, 2, 3 21

(2821-N686 A, B C - D) 3

! w (2826-No87 A, B, C, D) 1 (2821-N688 A, B C D)

>- (2B21-N689 A B C, D)

d. Main Steam Line Tunnel Temprature - High 1 2/11ne* 1, 2, 3 21 12821-N623 A, B C D)

(2821-N624 A, B, C D):

(2B21-N625 A, B, C, D) 2= (2821-N626 A, B, C D) i g' ,

t*

s e. Condenser vacuum'- tow 1 2 1, 2 ',3' .23 Q (2821-N056 A, 8 C, D'r a

4 55

f. Turbine Building Area Temperature - High I 2* 1, 2, 3 21 o (2U61-P001, 4061-P002, 2U61-P003, 2U61-P004) w "'

w 9 Drywell 8tadiation - High 1 1, 2, 3 29 (2Dll-K621 A B) u Mia w . '

e h

1 f

. ,_~ _ _~ _ _ .

4

. e.

.' Rt 1 9

j, g TABLE 3.3.2-1 (Continued) t 8 i t' ISOLATION ACTUATION INSTRUMENTATION i j' )% ACTION l

I i' ACTION 20 -

I

)4s g

Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ar.d in COLD SHUTDOWa within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

j I ACTION 21 .

Be in at least STARTUP with the main steam line isolation valves

^

1u ** closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within 6 t hours and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

j ACTION 22 -

Be in at least STARTUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . !

ACTION 23 -

Be in at least STARTUP with the Group 1 isolation valves closed g within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

4a 17 .

3 ACTION 24 .

Establish SECONDARY CONTAINMENT INTEGRITY with the standby j ; Q gas treatment system operating within one hour.

s y ACTION 25 -

Isolate the reactor water cleanup system, i

1M

% ACTION 26 Close the affected system isolation valves and declare the

affected system inoperable.

\

l ACTION 27 -

Verify power availability to the bus at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months h{ 1 D' or close the affected system isolation valves and declare the f t affected system inoperable.

j ACTION 28 -

Close the shutdown cooling supply and reactor vesse isolation valves unless reactor steam dome pressure . ad 45spray psig.

d ACTION 29 -

4 Either close tne affected isolation valves within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or bt<

ih in HOT SMUTDOWN within the next 6 noues and in COLD SHUTDOWN t within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

NOTES Actuates the standby gas treatment system.

When handling irradiated fuel in the secondary containmer.t.

i When performing inservice hydrostatic or leak testing with the reactor coolant temperature above 212* F.

a. See Specification 3.6.3, Table 3.6.3-1 for valves in each valve group.

b. A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip 1

system is monitoring that Darame;er.

4 4ATCH - UNIT 2- 3/4 3-15 Amendment No. 120

i 1* ,

)

i 1 i .

3

c. With a design pioviding only one channel per trip system, an inoperable channel need not be placed in the tripped condition where this would cause i

' the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.2-1 for that Trip function shall be taken.

I d. Trips the s chanical vacuum pumps.

[i a e.

A channel is OPERABLE if 2 of 4 instrument.s in that channel are OPERABLE.

l f. May be bypassed with all turbine stop valves closed.

9 Closes only RNCU outlet isolation valve 2G31-f004.

{ h. Alarm only, i

1. Adjustable up to 60 minutes.

[ 1

j. Isolates containment purge and vent valves.

k. $tthtn 74 ho(TFrpr1Drt07 (NNed start M j;h hydrogen h{ectiontes ) !

! Kththereactorpowerat'g ter than 20% rated ower. t normal ful s j power radiation backgr,ound el and assogJlfted t ip setp ints nyky b i ,, "- .M chanted based'on a4alculated blue of,the radiati n lpfel expect d uring 1 iu [M --@ the test s The batkground radiat(on le9el and assoc ed trip set- ints may l 6* * ,,, e be adjuste fing the test basedsorieither calcula ons or meas e ts of

, actual ra The back und i

radiatig ev on 'evels i shall be resultigf{om dete lned ' nd hydrogen associa ipfe fd tr } tion. setpot ts sha e set wJthin 24 hobrs s of re-p 11shi normal diatto leve}tafter

! c letion of hydr 6 tojecticn and phor tg stablish eactor power els below 20% rate fiower,

\/

l 1. The high '"erential flow isolation signal to the RWCU isolation valves j may be by,e .ed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during periods of. system restoration, j maintenance, or testing. - i i m, \ u, k % tk < <. u % sn % <*~r\' A la r. , l.6P**% '*4-l- 7. 6 314 t o . T* W s m o m ,, - gu , u.

4 i.

i

[

i p

i HATCH -UN!7 2 3/4 3-15a . Amendment No. - 120 l ;

i- 1 j-

i l k *.

4

( . A N '

Prior to the hydrogen injection system startup and with reactor power greater

than 20% rated power, the normal full power radiation trip / alarm setpoints may ba changed based on calculated expected radiation levels during hydrogen injection system operation. Associated trip / alarm setpoints may be adjusted during injection based on either calculations or measurements of actual i radiation levels resulting from hydrogen injection. Following a reactor i

startup, a hackground radiation level will be determined and the associated j

trip / alarm setpoints adjusted within a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period. The radiation level 4

shall be tietermined and associated trip / alarm setpoints shall be set within 24 1

hours of re establishing formal radiation levels after a reduction ir., of a completion of hydrogen injection and prior to establishing reactor power levels below 20% rated power. .

I l

l 4

4 9

T

-,y.. --e.. r- i,_. , ..r .-m..r- --.

i

=

20 TAatE 3 3 2-2

--e n

I (SOL ATION ACTU ATsON If4STRUMENTATtON SETPOO4TS 8 i c

= ALLOWABLE .

Q TRIP Ft>NCT'd TRIP SETPOtNT VALUE !

1. PRIM ARY CONT AINWNT ISOt ATION

e. Reactor Vessed Water Lewes '

1. Low (Level 3) 2 Oinchee* 2 Oinches*

2. Law Low (Lewei 21 l 2 -47 inchce* 2 -47 inchee*

,. Low Law Law (Lowe; 1) 2 -113 inches

2 -113 inches *

b. Drywet Preneuro High s1.90 peg s 1.92 poig

s. Man Ste.ssi Line

1. Reestion - Mgh s 3 x fuSpower background ** s 3 x fuB-power be

ground * *

2. Pr seure - Low 2 825 peng 2 825 ye!g

3. Row-Hgh s 138% rated flow 5138% rated Row k* d. Mein Steam Une Tunnis Temperature - ngh s 194*F s 194*F La

. e. Condenser Vacuum - Low 2 7* Hg wecuurn 2 7* Hg wecuurn on

f. Turbine 8L=i*ng Area Terre.-Moh 5 200*F s 2OO*F

g. Drywes Radiation - Hoh s 138 R/hr s 133 Rhw g

m

2. SECOND ARY CONTAINV NT ISOL ATION

3 e. Reac*;s Buddsng fehaust m

Reestion - H*gh .3 sne/hr s 80 mr/hr l

3

" b. Drywe8 Preme - Mgh 51.32 peig s 1.32 peig 2

f c. Reactor Vessel Water Level- Low Low (Level 23 2 47 inchee* 2 47 inches *

.D S

g d. Refueling Floor Enhaust Redseason bgh

. 5 70 mrAr 5 20 me!ht

'9 O)

See Bones FWo B 3/4 3-1.

~

- 1Gthin 24 hosere pator to the pienned etert of the nydrogen aneoction test with' the'reacter pews 5 ~/* N -

$ !greeser then 20% rated power. rhe normal eust power radiatiosa background level er t meeectated tnp eetpoente ' r

. ,may be changed bened oss e selcadeted venue of the radiation level expected dut"no Se test. The background *'

. s - ? p. ,, ./ , $ . - e-lrmistaan lowet *- 3 eseecasted snp setpdnts sney be edpusted during the test beoed s. eether calculataor:e C

w 4 or armeeurew.e se et actues em -i neweis resada no from hydsogeriiriection. The L Aground red.etion e' f ..-

towei ehes a , stereraned ered eseouster8 tnp eetpo.pte eben be set witMer 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of re-estethehang } '-

normai roo e. e eet- ce of t,, ogen in,ec or. .nd - io estaanarse.tm,ow. ;

. N _

I"*"

s s i

_. s - - - % s

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

0

. A Prior to the hydrogen injection system startup and with reactor p0wer greatei' be changed based on calculated expected radiation levels during injection system operation. Associated trip /tlarm setpoints may be adjusted during injection based on either calculations or measurements of actual radiation levels resulting from hydrogen injection, following a reactor startup, a background radiation level will be determined and the associated trip / alarm setpoints adjusted within a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period. The radiation level shall be determined and assoc.iated trip / alarm setpoints shall be set within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of re-establishing normal radiation levels after a reduction in, or a completion of hydrogen below 20% rated powei. injection and prior to establishing reactor power levels 4

~

s E la

~

n. -

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TAHLT,3 3.6 _ ME(I 1 OF 2)

'y MOREQS_ ACTUATION IMSTRUMENTATt04 I

e MINIMOH MUM 8fR A PPL ICARt.f OPERABLE CHAMNLLS OPER AT ION AL ,

c-. PER TRIP SYSTDa[s)(b) CONDITIOti ACTION l 2 TRIP FUNCTION-

-4

1. Reactor vessel Water tevel - 2 t, 2, 3 52 ]

N Low Low low (Level 11 (c) 2621-Novi A, B, C, D 2.

Dep t i Pressure - High (c) 2 1, 2, 3 52 2011-N69fs A, B, CD urgs ge s -> i g,_4 t*

gg_

3. Main %[al W IUndiataaa 7 an t 1-Jutn 3 a , n r- n 14 . t sin Steam Line riow - High (c) 2/ tine 1, 7, 3 53

?B21-N686 A,'D, C, D 2821-N681 A, B, C, D -

2021-N688 A,.B. C, D 2821-N689 A, B,C,D S. Ref'eseling rioor Area Radiation - High (c) 1 t , 2, 3, 5,

54 2D71-AUO2 A. D

%)

1.2,3,5,* 54 2 6. Cont rot .7nos Air inlet Radiation - High (c) 12!s t-R61'r A, B 1

e, C3 g

B to 3

Q.

l 3

i. (D 3

m O

N e

15 0 l

l

, e i Ot

<n - - -

g ,

.c . . - - . -. - - -

1

] TABLE 3.3.6;J-l(SHEET 2Or2) ,

j i MCR70$ ACTl'ATION INSTRWENTA!!ON

' f 4

l ACTION q

i, lj ACTION $2 -

Take the ACTION required by Specification 3.3.3.

4 ACTION $3 -

Take the ACTION required by Specification 3.3.2. I

l ACTION 54 - ;l

. . a. With one of the requireri radiation monitors inoperable, restore the rnonitor to OPERABLE status within 7 days or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months i

initiate and maintain operation of the MCRECS in the pressurization I i

s mode of operation. -'

b. With no radiation monitars OPERABLE, within I hour initiate and 1 maintain operation of the MCREC$ in the pressurization mode of

] operation. <

c. The provisions of Specification 3.0.4 are not applicable.

NOTES l

n:

When handling irradiated fuel in secondary contairment. ,l'

. a. A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for ) li '

l l

required surveillance without placing the trip system in the tripped >l condition, provided at least one other OPERARLE channel in the same trip f:

system is monitoring that parameter. l j

b. With a design providing only one channel per trip system, an inoperable j, 2

channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable -

channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the

! ACTION required by Table 3.3.6.7-1 for that Trip Function shall be taken. i i

j c. Actuates the MCRECS in the control room pressurization mode. I i

d. (Deleted) {

/- - - - - - - - - - - - - - -

e r V.L.hkML . -- - -prior

- ~\'

ithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the planned stast of the hydrogen injection test ~ ){

with the rc ctor power at greater than 2C prcent rated powerrthe' normal l-full power r~adiation background level and associated trip-setpoints may be ll changed based on a ' calculated value of the radiation ~1evel expected during the test. The background radiation level and' associated trip setpoints may be adjustett during the test tQsed:Ori either calculations or measurements of actual radiation levels'resul. ting from hydrogen ;

injection. The background radiation level shall-bt determined and

~

associateJ trip setpoints shall be set within 24 houfs'ofqe-establishing .

, normal radiation levels af ter completion of hydrogen injectfon-and prior l

[to-establishing react.or power level _s below 20-percent rated power.N w _i HATCH - UNIT 2 3/4 3-58b Amendment No. 71, 28.-96

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.- . _ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ _ ._, _-.. , . ._ -. - . _ _ _ _ . . . . - . . . . . . . ..._..-.m-.mm_.----..._...m__.._.m..-.-

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].

IABLE 4.3.6.7-1 n MCRECS ACTUATION INSTRUMENTATION SU5NE!LLAMCE REQUIRIMt MTS -

C!sANN14. OPfRATiOMAt a .

CalANREL FUNCilONAL CitANNEl CONDliloftS IN WillCH c TRIP FUNCT ION CHECK TEST CALfBRATIOM SURVEILLANCE FLQUIPEC 2

g"* 1 Reactor Vessel Water Level - S M R 1, 2, 3

Low Low Low (Love 4 '

)

fu

j. 2. D ryine l i Pressure - High S M R 1, 7, 3 I L ;: ] Af

3. 44ain-Steam _-Line Radiation-44igla B - W " L-- "-

-1,- 2 J -

is . Main Steam Line flow - High S M R 1, 2, 3

5. Rerueling F1oor Area Radiation - D M'** Q 1, 2, 3, $

l High- l t i

6. Control Room Ai r inlet NA M**8 M 1, 2, 3, 3, e l j Radiation - High .

l. L 1;

a i

% c A [

4 ta I tn r j -- O2 :

a ,

4

when hano e ing a rradiated f uel en the secondary contae m ut.

i.

4

a. Instrumeert alignment using a standard current source.

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I i i PLANT SYSTEMS l l '

r i

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SURVEILLANCE REQUli. INTS (Continued)

l l

3. Verifying that on each of the below pressurization mode actuation test signals, the system automatically switches

)I

~' !

j to the pressurization mode of operation and maintains the i

main control room at a positive pressure of 2 0.1-in. ~!

l

' W.G. relative to the adjacent turbine butiding during system operation at a flow rate s 400 cfm. - I l a) Reactor vessel water level - low low low -

l

q b) Drywell pressure - high !

1 l

! c) Refueling floor area radiation - high c

i. d) Main.nua)

-4teaa_ lina._radiationshigh i

j e) Main sceam line flow - high f) Control room intake monitors radiation - high i i l ';

f. After each complete or partial replacement of a HEPA filter l bank by verifying that the HEPA filter banks remove 2 99 percent of the DOP when they are tested in place in accordance with ANSI l >[!

N510-1975 while operating the system at a flow rate of 2500

! cfm + 10 percent. I h

4 h

r g. After each complete or partial replacement of a charcoal adsor'ar bank by verifying that the charcoal adsorbers remove ti 2 99 percent of a halogenated hydrocarbon refrigerant-test gas I L when they are tested in place in accordance with ANSE N510-1975 H

. while operating the system at a flow rate of 2500 cfm + 10 percent. l

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4 g

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j HATCH - UNIT 2 3/4 7-8 Amr'dment No,./J, 96 l

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ML20115G679

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