ML20056G235

Amends 188 & 127 to Licenses DPR-57 & NPF-5,respectively, Deleting Main Steam Isolation Valve Closure,Reactor Scram & Control Room Pressurization Functions of Main Steam Line Radiation Monitors

ML20056G235
Person / Time
Site:	Hatch
Issue date:	08/17/1993
From:	Matthews D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20056G236	List: ML20056G235 ML20056G237
References
NUDOCS 9309020288
Download: ML20056G235 (34)

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S UNITED STATES

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NUCLEAR REGULATORY COMMISSION gs, j e

wAssinatou, o.c. 2csss-ooes GEORGIA POWER COMPANY OGLETHORpE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA I

CITY OF DALTGN. GEORGIA DOCKET NO. 50-321 EDWIN 1. HATCH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.188 i

License No. DPR-57 i

1.

The Nuclear Regulatory Commission (the Commission) has found that:

~

A.

The oppi cation for amendment to the Edwin I. Hatch Nuclear Plant, 4

~ Unit 1 (tr.e facility) Facility Operating License No. DPR-57 filed by the Georgia N r, Company, acting for itself, Oglethorpe Power Corporation, Mi,nicipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated October 19, 1992, as supplemented May 3 and July 27, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set furth in 10 CFR Chapter I; t

i B.

The facility will operate in conformity with the application, the provisiors of the Act, and the rules and regulations of the Commissit n; t

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be i

conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common i

defense and security or to the health and safety of the public; and j

E.

The issuance of this amendment is in accordance with 10 CFR Part 51-of the Commission's regulations and all applicable requirements have

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been satisfied.

i 9309020288 930817 PDR ADOCK 0300 t4 p

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' i 2.

Accordingly, the license is hereby amended by page changes to the Technical Specificat ions as indicated in the attachment to this license j

amendment, and paragraph 2.C.(2) of Facility Operating License No. DpR-57 i

i is hereby amended to read as follows:

j 2

4 Technical-Specifications The Technical Specifite.;

.s contained in Appendices A and B, as revised through Amendment No. 188, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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3.

This license amendment is effective as of its date of issuance and shall i-be implemented no later than 60 days from the date of issuance.

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FOR THE NUCLEAR REGULATORY COMMISSION l

NO7aA-

/

David B. Matthews, Director l

" Project Directorate II-3 l

Division of Reactor Projects - :/II Office of Nuclear Reactor Regulation l

1

Attachment:

l Technical Specification l

Change; J

Date of Issuance: August 17, 1993 1

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E UNITED STATES isikp!

NUCLEAR REGULATORY COMMISSION

( v4f WASHINGTON. D C. 20555.-0001 l

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION f

MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA i

DOCKET NO. 50-366 i

EDWIN 1. HATCH NUCLEAR PLANT. UNIT 2 l

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.127 License No. NPF-5 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Edwin 1. Hatch Nuclear Plant, I

Unit 2 (the facility) Facility Operating License No. NPF-5 filed by l

the Georgia Power Company, acting for itself, Oglethorpe Power i

Corporation, Municipal Electric Authority of Georgia, and City of

)

Dalton, Georgia (the licensees), dated October 19, 1992, as supplemented May 3 and July 27, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission:

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangert the health and safety of the public, and (ii) that such activit c3 will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

i

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-5 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.127, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

j 3.

This license amendment is effective as of its date of issuance and shall be implemented no later than 60 days from.the date of issuance.

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FOR THE NUCLEAR REGULATORY COMMISSION i

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',,f 7Sv A

ai B. Ma thews, Director l

Project Directorate 11-3 i

Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation j

Attachment:

i Technical Specification l

Changes Date of Issuance: August 17, 1993 k

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ATTACHMENT TO LICENSE AMENDMENT NO. 188 FACILITY OPERATING LICENSE NO. DPR-57 l

DOCKET NO. 50-321 AND j

TO LICENSE AMENDMENT NO.127 i

FACILITY OPERATING LICENSE NO. NPF-5 i

j DOCKET NO. 50-366 l

Replace the following pages of the Appendix "A" Technical Specifications with i

the enclosed pages. The revised pages are identified by Amendment number and j

contain vertical lines indicating the areas of change.

L Remove Pages Insert Pages j

Unit 1 3.1-5 3.1-5 l

3.1-6 3.1-6 3.1-6a 3.1-6a 3.1-8 3.1-8 l

l 3.1-13 3.1-13 i

3.2-3 3.2-3 3.2-4 3.2-4 3.2-19 3.2-19 3.2-51 3.2-51 3.2-66 3.2-66 i

3.7-19 3.7-19 j

3.12-4 3.12-4 j

Unit 2 2-4 2-4 l

B 2-11.

B 2-11 l

3/4 3-2 3/4 3-2 l

3/4 3-4 3/4 3-4 3/4 3-5 3/4 3-5 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 3/4 3-11 3/4 3-11 3/4 3-15 3/4 3-15 3/4 3-15a 3/4 3-15a 3/4 3-16 3/4 3-16 3/4 7-8 3/4 7-8 3/4 3-58a 3/4 3-58a 3/4 3-58b 3/4 3-58b 3/4 3-58c.

3/4 3-58c 3/4 3-58d 3/4 3-58d i

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Table 3.1-1 (Cont dl

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Setem Operable g

Number Source of Scram Trip Signal Channele Scram Trip Setting Source of Screm signalis 2

le)

Required Per Required to bs Operable Q

Trip System Except es indicated Delow (b) w 8

APRM Downecele 2

13/125 of full scale The APRM downscela trip is active only when the Mode Switch le in RUN. The APRM downecele trip is automatically bypassed when the IRM instrumentation is operable end not tripped.

15% Flux 2

115/125 of full The APRM 15% Scremis auto-scele Tech Spoo metically bypeesed when the 2.1.A.1.b.

Mode Switch is in the RUN position.

W 9

(Deletodi l

us 10 Mein Steam Une lealetion 4

510% velve closure Automatically bwessed when Velve Closure from full open the Mode Switch is not in g

Tech Spec 2.1.A.5, the RUN position. The design m

3 permits closure of any two a

lines without e scram being irdtieted.

r+

z 11 Turbine control Velve 2

WitNn 30 milli-Automatically bypassed when P

Fest cloeure seconde of the etert turbine steem flow le below of control velve that corresponding to 30% of 00 feet closure rated thermal power se measured Tech Spec 2.1.A.4.

by turbine first stage pressure.

Tebte 3.1-1 (Cont'd)

~4 m

Z Scram Operable Number Source of Scram Trip signal Chennels Screm Trip Setting Source of Scram Signalis 3E (el Required Per Required to be Operable Q

Trip System Except es Indicated Below M

12 Turbine Step Valve 4

510% valve closure AutomaticeNy bypeeeed when Docure from fun open furtWne steem flow le below Tech Spee 2.1.A.3.

that corresponding to 30% of reted thermal power se meesured by turtnne fleet stage preeeure.

Notes for Table 3.1 1 e.

The esiumn oneltled *Serem Number" le for convorwence so that e one-to-one relationship een be estetAshed I

between items in Table 3.1 1 end home in Table 4.1 1.

b.1.

There shen be two operable er tripped trip systeme for each potential serem signal. If the number of y

operabio channele eennet be met for one of the trip eystems, the inoperable channeHol or the eseociated trip -

cn evotem ehen be tripped b.2.

One instrument channel may be enoperehle for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to perform required surveillencee prior to entering other eh settone, peovhfed et least one operstde channel in the some trip system le monitor 6ng that perameter.

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Table 4.1-? (Cont.)

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Scram Instrument Check Instrument Functional Test instrument Ceiibestion Hunter Source of Scram Trip Signal Group unemum Frequency Merwnum Frequency Mromum Frequency

.E.

tal -

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(c)

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(Deletodi l

10 Main Steam Une feeletion Velve A

NA Every 3 monthe th)

Doeure 11 Turbine Control Velve Feet A

NA Every 3 months til Once/ Operating Cloovre Cycle (kl

^

12 Turbine Stop Velve Coeure A

NA Every 3 monthe th)

RPS Channel Switch A

NA Once/ Operating Cycle Not Applicable Turbine First Stege Preeeure A

NA Every 3 monthe Every 6 monthe Permissive e.

The column entitled "Screm Norr6er" is for converdence so that e one-to-one relationship con be established between items in Table 4.1 1 and home in Table 3.1 1.

b.

The defiedtion for each of the four groupe le se follows:

w Group A.

Orvoff eenoore that provide e serem trip eignal.

Group 8.

Analog devicoe coupled with bi-etable trips thet provide e scrom trip signol.

Y Group C.

DeWee which urwy serve e useful funct6on during some restricted mode of operation, ce such es stortup or shutdown, or for wNch the ordy proctical test le one that con be performed et shutdown.

Group D.

Analog trenemittore end trip unite that provide e scram trip function.

e.

Functional teste ere not required when the eyeteme are not required to be operable or are tripped.

s However,if functional toets are redeoed, they shall be performed prior to retuming the systeme to g

on operable status.

(D D

d.

Calibeetione are not required wtwn the systems are not required to be operable or ere tripped.

However, if calibretione are miesed, they shell be performed prior to returedng the eyetem to 2'

en operable status.

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e.

TNs instrumentation lo exempted from the instrument functional teet definition. TNe inetrument functional test will consist of inketing e simuleted electrical signalinto the meesurement co chonnels f.

Deleted g.

The water levelin the reector will be perturbed end the corresponding levelinscetor changes will be monitored. This porturbetion test will be performed every 3 enonthe efter completion of the functional test program, h.

Physicalinopoetion and octuation of these position awitches will be performert oncJ per operating cycle.

4.

(Deleted) l l.

Mesoure time interval from EHC pressure awitch actuation to RPS reley K14 de-energiretion.

1 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.1.A.B.b.

Increrative l

An APRM is inoperable if there are less than two LPPM inputs per level or there are less than 11 LPRM inputs to the APRM channel.

c.

Dewascale The AFRM downstale is active only when the Mode Switch is in the RUN position. The APRM downstale trip is automatically bypassed when the IRM instrumentation is operable and not tripped. Because of the APRM i

downscale limit of 23/125 of full scale when in the Run Mode and high

[

level limit of $15/125 of full scale when the Start & Hot Standby Mode, 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, j

the IRM system must be indicating below the High High Flux setting (120/125 of full scale) or a stram will occur when in the Start & Hot Standby I

Mc.f e. For nomal operating conditions, these limits provide assurance of overlap between the IM system and APRM system so that there are no 4

• gaps" in the power level indications (i.e., the power level is con-i c

11nuously monitored from beginning of startup to full power and from full i

I 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 impossible condition) a control rod block immediately occurs so that reactivity insertion by control rod r

withdrawal cannot occur.

d.

15% Flux The bases for the APRM 15% Flux Scram Trip 5etting is discussed in the bases for Specification 2.1.A.I.b.

9. (Deleted) l I

10.

Main Steam tire Isolation Valve Closure The bases for the Main Steam Line Isolation Valve Closure Scram Trip Setting is discussed in the bases for Specification 2.1.A.S.

II.

Turbine Control Valve Fast Closure The bases for the Turbine Control Valve Fast Closure Scram Trip setting is discussed in the bases for Specification 2.1.A.4.

12.

Turbine Steo Valve Closure The bases for the Turbine Stop Valve Closure Scram Trip 5etting is dis-cussed in the bases for Specification 2.1.A.3.

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HATCH - UNIT 1 3.1-13 Amendment No. 188

+

Table 3.2-1 (Cont.)

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Requtred 2

Ref.

Tnp OpereNo Action to be taken if t

N o.

Instrument Condition Channels Trip Settmg number of channels is (el Nomenclature per Trip not rnet for both trip Remarks (d)

System (b) systems (el N

4 Mein Steem Une High 2

53 times normal Irvtiete closure of reactor initiates closure of Radiation full power back-water semple velves, reactor water semple ground

• volves B31-F019 and 831.F020.

5 Main Steam Une Low 2

2 825 psig trutiste en orderty load initiates Group 1 Presouro reduction erw! close isolation. Only MSIVe witNn 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

required in RUN mode, therefore activated when Mode Switch is in RUN position, 6

Main Steam Une Hegh 2

5139% rated flow levtiete en orderly load Irvtistes Group 1 Flow (s115 paidl reduction and close MSIVs isolation.

witNn 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

7 Main Steem Une High 2

s 194'F Initiate en orderly load trutsetes Group 1 Tunnel Tempareture reduction and close MSIVs isolation.

witNn 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

4.4 8

Reector Water High 1

20-80 gpm isolete reactor water (f)

N Qeenup System cleenup system.

b Differentief Flow 9

Reactor Water Heuh 2

5150'F isolete reactor water Cleanup Area cleanup system.

Temperature g

10 Reactor Weter H*gh 2

sS7'F lsolete reactor water g

Cleanup Area cleeruip system.

p Ventilation C1 Differential

]

Temperature a

c0 11 Condeneer Vacuom Low 2

2 7" H :. vacuum frutiste en orderly load initiate Group 1 t

reduction and close MSIVe isolation go witNn 8 hrs.

12 Orywell Radiation High 1

5138 R/HR.

Close the effected Isoletes cp isolation valves witNn containment purge CD 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in Hot and vent valves.

Shutdown witNn the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown witNn the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

h Notes for Table 3.2-1

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W column entitled "Ref. No.* le ordy for corwenience so that a ov+to-one relationship con be established between lines in Table 3.2-1 end iteme in Toble 4.2-1.

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w b.1.

Primary containment Integrity shall be madntained et all times prior to withdrawing control rods for the d

purpose of going critical, when the reactor le critical. or when the reector water temperature le above 212'F ond fuelle in the roector veseet except while performing low-power physico teste et atmospheric w

preesure et power levele not to suceed 5 MWt. or perforrnng en inoervice veeeel hydrostatic or leakege test.

When primary contairenont integrity le required, there shall be two operable of tripped trip systems for each function.

When performine Inservice hydroetetic or leekoge testing on the reactor vessel with the reector cooient terversture above 212'F, toector vessel weter levelinstrumentation eseociated with the low low (Level 2) trip requires two operable or tripped chennele. The drywell preesure tnp le not required because primary containment integrity le not required.

b.2.

One instrument channet may be inoperable for up to 6 houre to perform required surveellences prior to entering other app 5celde octRme.

c.1.

With the nunter of operstde channele love then required by the 7Aremum Operable Channele per Trip System requirement for one trip system. either w

place the Inoperable chonne tel in the tripped condition

• within 12 houre 1.

r s

Oft 2.

take the action required by Table 3.2-1.

• With a design providin0 only one channel per trip eyetem, en inoperable channel need not be placed in the k

tripped condition where thee would cause the Trip Function to occur. In thee, eeees, the inoperable ehennel she'l be restored to operstdo stetve within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the action required by Table 3.2-1 for that a.

Trip Function ohes be teken.

E3 cva e.2.

One ine~. went channel may be Inoperable for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to perform required survoitances prior to entering ether applicetde actione.

2 P

d.

h voivne===aal-ted with each Group leoletion are given in Table 3.71.

e.

Prior to the hydregon in6sction system etertup and with reactor power greater then 20% rated 00 power, the normal fue power radiation tripleierm eetpointe may be changed beoed on calculated esqposeed redselon levato during hydrogen in$oction system operation. Asoocleted trip /elerm segseines eney be es$seted dudng in$setion beood on either calculations or measurements of someal radieden levais resulting from hydrogen iniection. Following a reactor etertup, e leashground redesion level wig be determined and the seeociated trip /elerm setpointe adhseted within a 72-hour poded. The redieden level shall be determined and seeocieved trip /elerm seapointe shed be est within 24 heure of = a J.:., normal radiation levele efter e reduction in, or a n., _ _. of, hydrogen iriection end pnor to estatdiohing teactor power I

levele below 20% of reted power.

f.

The high efferonelet flew signal to the RWCU isoletion volves mey be bypeeeed for up to 2 houre during periode of system rootora6on. maintenance. of testing.

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Table 3.2-8 leont.)

--4 O

Ref.

Instrument Trip Required Trip Setting Action to be taken if Romerk o No.

Condition Operable there ere not two operable tel Nomencle.

Chennele or tripped trip eyoteme

.z-ture per Trip y

Svetem lb) 1

~

5.

Main Steam Une Hi 2

53 times leolete the mechenical One trip per trip Radiation Monitor normel full vocuum pump eruf the logic evotem will power background giend seal condoneer loolete the tel exhauster mechenical vacuum pump and the glend sealcondeneer enheueter, s.

The column entitled *Ref. No.* is ordy for convorwence so that e one-to-one refetionship can be estabhehed between iteme in Teide 3.2-8 and items in Table 4.2-8.

b.1.

Whenever the evetems are required to be operable, there eheel be two operable of tripped trip systems.

If tNo eennet be mot, the Indested oction shall be taken,

.t s b.2.

One instrument channel moy be Inoperable for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to perform required surveillences prior to entering 7.*

other e actions.

e.

e, in the event that both off-ges poet tfeetment redietion moedtore become inoperable, the reactor shell be piemed in the Cold Shutdown within 24 heure ordees one monitor le sooner made operable, or adequete ottemotive menteering feeleties are swelleMe g-From and after the date that one of the two off-gee poet treetment recietion monitore is made or found to d.

be inspereNo, eentinued veester power opereglen le s. ' _ _ ; during the next fourteen deye Ithe

^

eg alloweMe e,-* timel, prowided that the inoperside moedter le tripped.

l 9

Prior to the indregon infection eyetem etertup and with remotor power greeter then 20% reted power, the normal full power radiation trip /elerm setpointe mey be changed e.

6 bened on oe'oubt -d sequested redletion leveio during hydrogen in$sedon eyetem operation. Aseeciated trip /essem setpointo may be adjueted during injection beoed on y

either esteuMNw or measuremente of octual redlesen levole roeuhing from hydrogen injection. Following a reactor stortup, a backgrourut radiation level will be

?

decomdnad 4 a the esseelsted tripMerm seapointe ediusted within a 72-hour period. The redetion level ehesi be detemuned and secociated trip /elerm setpointe eheel be est within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of = - "" 1., normal radiation loweie efter e reduction in, or e cornpletion of, hydrogen injection and prior to establieNng reector power levele c) below 20% of reced power.

cp

BASES FOR LIMITING CONDITIONS FOR OPEPATION 3.2.A.2.

Reacter Vessel Steam Dome Pressure (Shutdown Coelino Mode) tow Permissive This setpoint is chosen to preserve the pressure integrity of 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 against RHR system pipe breaks during the shutdown cooling mode of op-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 Hich The Bases for Drywell Pressure High are discussed in the Bases for Specif-ication 3.1.A.5.

Pressure above the trip setting starts the SGTS and in-itiates primary and secondary containment isolation.

4.

Main Steam Line Radiation Hich 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 established setting of approximately three times normal full power background, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident.

I 5.

Main Steam Line Pressure 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 Hich 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 steam flow. Instro-mentation is provided which initiates Group 1 isolation. The primary func-tion 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 1385 of rated steam flow.

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 temperatures remain approximately 1000cF and release of radioactivity to the environs is well below 10 CF2100 guidelines. Ref. Section 14.6.5 of the FSAR.

7.

Main Steam line Tunnel Temeerature Hioh Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in this area. Trips are provided on this instrumen-tation and when exceeded cause a Group 1 isolation. Its setting is low enough to detect leaks of the order of five to 10 gps; 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.188

EASES C04 LIMITING CONDITIONS FOR OPEDATION occurs with each monitor indicating HI HI HI, one monitor HI HI HI and the I

other downscale, or with both monitors downscale. The HI HI HI setpoint cor-responds to the instantaneous release limit.

2.

Defuelino Fleer Edaust Veat Dadiation w niters e

Four radiation monitors are provided which initiate isolation of the second-ary containment and operation of the standby gas treatment system. The in-strument channels monitor the radiation from the refueling area ventilation i

exhaust ducts.

e i

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-tiating normal ventilation isolation and standby gas treatment system opera-l tion so that none of the activity released during the refueling accident leaves l

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

t 3.

Peacter Buildine Exhaust Vent ondiation Moniters Four radiation monitors are provided which initiate secondary containment iso-I lation, primary containment purge and vent valves isolation and standby gas treatment system actuation. The instrument channels monitor the radiation from the reactor building lower level ventilation exhaust duct.

l

+

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 treatment system.

4.

Control Room Intake Radiation Monitors Two radiation monitors are provided to initiate pressurization of the main control room and recirculation of control room air through filters. The instrument channels monitor radiation from the control room ventilation intake duct.

Two,1cstrument channels are arranged in one upscale, two downscale trip log-ic. The trip settings are based on limiting the radioactivity from entering the control room from outside.

1 5.

Main Steam Line Radiation Monitors The four Main Steam Line radiation monitors initiate isolation of the i

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 materials from the main condenser. Upon receipt of main steam line high radiation signals, the primary containment and reactor vessel isolation control system initiates closure of the mechanical vacuum pump line valve. This isolation precludes or limits the release of fission product radioactivity which, upon fuel failure would be transported from l

l HATCH - UNIT I 3.2-66 Amendment No. 188

l I

Tabic U -l 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.

NOTES:

a. Key:

0 = Open SC = Stays closed C - Closed GC = Goes closed

b. Isolation Groupings are as follows:

i I

GROUP 1: The valves in Gesap 1 are actuated by any n of the following I

conditions:

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

2. Main steam line radiation high*

l

3. Main steam line flow high

4. Main steam line tunnel temperature high

5. Main steam line pressure low l

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 a of the following conditions:

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

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

5. Torus room differential temperature high i

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

3. RCIC steam line pressure low

4. RCIC turbine exhaust diaphragm pressure high

5. Torus room differential temperature high Initiates closure of 831-F019 and B31-F020 only.

l l

i KATCH - UNIT 1 3.7-19 Amendment No. 188

)

3.12. MAIN CONTROL ROOM ENVIDONwENTAL SYSTEM The control room air treatment system is designed to filter the control room atmosphere for intake air and/or for recirculation during i

pressurization conditions.

A. Ventilation System Orerability Pecuirements 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.

l The control room air treatment system is designed to automatically I

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-l ing signals: High radiation at control room intake.LOCA signal from Unit I or 2, main steam line high flow from Unit 1 or 2, or l

refueling floor high radiation from Unit I or 2.

In this mode the normal control room exhaust fan is stopped and outside air is taken t

in through one 9f 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 l

to the control room. Bypass leakage for the charcoal adsorbers and particulate removal efficiency for HEPA filters are deterstned by halogenated hydrocarbon and DOP, respectively. The. laboratory carbon 1

sample test results indicate a radioactive methyl iodide removal efficiency for expected accident conditions. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers. If the performances are as specified, the calculated doses would be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR Part 50.

i i

i HATCH - UNIT 1 3.12-4 Amendment No. 188 i

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2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 3.

Reactor Vessel Steam Dome Pressure-Hiah High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to in-crease the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counter-acting the pressure increase by decreasing heat generatior. The trip setting is slightly higher than the operating pressure tr, permit normal operation without spurious trips. The setting provida: for a widt 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, i

the transient analysis indicated a considerable margin to the thermal hydraulic limit.

4.

Reactor Vessel Water Level-Low 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 assure that there is adequate protection for the fuel and 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 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)

HATCH - UNIT 2 B 2-1]

Amendment No. 127

TABLE 3.3.1 1 w

S REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUM NUMBER E

OPERATIONAL OPERABLE CHANNELS l

Q FUNCTIONAL UNIT EglipfT10NS PER TRIP SYSTEM'd ACTION N

1, intermediate Range Monitors:

12C51-M901, A, B, C, D E, F, G, H) i 2 *I, SH 3

s.

Neutron Fluu. High 3

3 3, 4 2

2 M

b.

e;..

2.5 3

1 3, 4 2

2 2.

Average Power Range Moniter:

12C51 K905 A,5, C, D E, F) e.

Neutron Flow - Upoeste,15%

2, 5 2

1 b.

Flow Referoneed Simuleted Theemal rewer Upecele 1

2 3

e.

Fleed Neutron Flux -

g Upocale,118%

1 2

3

.e.

d.

Inoperative 1, 2, 5 2

4 w

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Downecede 1

2 3

4 f.

LPRM 1, 2, 5 (d)

NA 3.

Reector Vessel Steam Dome Preeeure -

8 High (2821-N678 A,8, C D) 1, 2 'I 2

5 4.

Reactor Vessel Water Level -

p Low (Level al 12821-N680 A, 8. C, D1 1, 2 2

5 M

.o 5.

Main Steam Une feeletlen Velve -

Q Cleeure (NA)

IM 4

3 ro a

6.

(Deseted)

I o

7.

Drywell Proeoute - High 1, 2W 2

5 ro (2C71 Ne50 A, B, C, Di N

I

4 i

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION I - In OPERATIONAL CONDITION 2, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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

ACTION 2 - Lock the reactor mode switch in the Shutdown position within one hour.

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

ACTION 4 In OPERATIONAL CONDITION I or 2, be in at least HOT SHUTDOWN i

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

In OPERATIONAL CONDITION 5, suspend all operations involving l

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

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

ACTION 6 (Deleted) l ACTION 7 - Initiate a reduction in THERMAL POWER within 15 minutes and be at less than 30% of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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

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

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

i HATCH - UNIT 2 3/4 3-4 Amendment No.127

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION 9 -

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

In OPERATIONAL CONDITION 3 or 4 lock the reactor mode switch in the Shutdown position within I hour.

In OPERATIONAL CONDITION 5, suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods within I hour.

TABLE NOTATIONS a.

Deleted.

b.

The " shorting links" shall be removed from the RPS circuitry during CORE ALTERATIONS and shutdown margin demonstrations performed in accordance with Specification 3.10.3.

c.

The IRM scrams are automatically bypassed when the reactor vessel mode switch is in the Run position and all APRM channels are OPERABLE and on scale.

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 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.

g.

This function is not required to be OPERABLE when PRIKARY CONTAINMENT INTEGRITY is not required.

h.

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

r

i. These functions are bypassed when turbine first stage pressure is 1250*

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

j.

(Deleted)

• Initial setpoint. Final setpoint to be determined during startup testing.

i HATCH - UNIT 2 3/4 3-5 Amendment No. 127 I

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n TABLE 4.3.1 1

--e

.g REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEfLLANCE REQUmEMENTS C

CilANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH 3 FUNCTIONAL UNIT CHECK TEST C A LfBR ATION*

SURVEfLLANCE REQUIRED

--4 1, intermodlete Menge Mordtore:

g e.

Neutron Flum High D

S/U **'

R 2

D W

R

3. 4. 5 b.

Inoperative NA W

NA 2,3,4,5

2. Average Power Range Monitor:

e.

Neutron Nu - Upsesfe 15%

S S/U", W" S/U*, WM 2

S W

W 5

b.

Nw Referenced Simulated S

S/U, O W'*"8, S A 1

Thermal Power - Upeesle o.

Flued Neutron Nu - Upocale, S

SAf*', O W"', S A 1

118 %

d, inoperetive NA O

NA 1, 2, 5 e.

Downeesle NA W

NA 1

f.

LPRM D

NA 1, 2, 5

3. Reactor Vessel Steam Dome S

O R

1, 2 w

Pressure a High a

4. Reactor Veeest Water Level.

S Q

R 1, 2 y

Low flevel al N

5. Mein Steam Line Isofstion Velve -

Cleeure NA O

R 1

6.

(Deletodi l

7. Drywell Proeoute - I4h S

Q R

1, 2 8o 8.

Screm Dische'ge Volume Water NA O

R'd 1, 2, 5 Level - H6gh (D

(9 2

P N

N

IABLE 4.3.1-1 fContinuedi 4O REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEfLLANCE REQi)IREMENTS 2;

i CHANNEL OPERATIONAL C

CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHtCH FUNCTIONAL UNIT CHECK TEST CAllBRATION SURVEllLANCE REQUmED

--4 9.

Turbine stop Velve Cloeuro NA Q

RN 1

g

10. Turbine contro! Velve Feet Closure, Trip 011 Pressure -

Low NA O

R 1

11. Reector Mode Switchin Shutdown NA R

NA 1,2.3,4,5 Position

12. Menvoi Sorem NA W

NA 1,2,3.4,5 s.

Neutron detectore may be owesuded from CHANNEL CALIBRATION.

b.

WitNo 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup,if not performed witNn the previoue 7 days.

c.

The APRM,IRM and SRM chennele shen be compared for overtep during each etertup,if not performed witNn the peevioue 7 days.

W d.

When changing from CONDITION 1 to CONDITION 2, perform the rowired surveillence witNn 12 houre

)

after entering CONDfTION 2.

TNe eenbretion ohell conelet of the adjuotrnent of the APRM channel to conform to the power valuee w

e.

E eelculeted by a heet belance during CONDITION 1 when THERMAL POWER 125% of RATED THERMAL POWER.

Adjust the APRM channelif the shooluto difference A 2%.

f.

TNe oeNbretion ohell conelet of the adjustment of the APRM flow refeeenced simulated thermal power thennel to conform to e eslibrated flow signal, p

g.

The LPRM's shall be calibrated at leset once per 1000 effective full power hours (EFPHI using the g

TIP eystem.

SL h.

Phyelesiinspection end actuation of switches for Instrumente 2C11 N013 A, B, C, D.

?

E I

m I

?

4 TABLE 3.3.21

-4g

[ SOLATION ACTUATION INSTRUMENTATION I

VALVE GROUPS MINIMUM NUMBER APPLICABLE OPERATED BY OPERABLE CHANNELS OPERATIONAL Imf' FUNCTION SIGNAL (al PER TRIP SYSTEM (b!tel CONDfTION ACTION

-4 1.

PRIMARY CONTAINfMNT ISOLATION g

e. Reector Vessel Weter Level

1. Lew (Level al 2, 6,10, 2

1,2,3 20 (2821 N680 A,8, C, D) 11,12

2. Low-Lew itevel 2) 5*

2 1,2,3 20 (2821 N682 A,8 C. D)

3. Low-Low-Lew flevel 11 1

2 1,2,3 20 (2821-N681 A,8. C, D)

b. Drywe4 Procoute - l6gh 2,6,7,10, 2

1,2,3 20

- (2C71-N660 A,8, C, D) 12, *

o. Mein Steam Une

1. Medletion - High 12, N 2

1, 2, 3,"I 30 l

12D11 K903 A,8, C, D) 2, Pressure. Low 1

2 1

22 (2821 N015 A, B, C, DI

3. Flow - High 1,

2/line 1, 2, 3 21 w

(2821-N686 A,8. C, DI N

(2821-M687 A,8, C, DI (2821-N688 A,8. C, D) w (2821-M689 A,8 C, D) e

d. Main Steam Une Tunnel Termereture. High 1

2Aine '8 1,2,3 21 l

(2821-N623 A,8, C, D)

(2821-N624 A,8. C DI (2821 N625 A,8, C, D)

(2821 N626 A,8 C, D)

e. Condoneer Veeuum - Low 1

2 1, 2,I'8,3"8 23 f

(2821-N056 A,8 C DI S

f. Turbine Bu# ding Aree 9

Temperature High 1

2M 1,2,3 21 (2U61 P001, 2U61-P002, 2U61-P003, e

2U61-P004)

• o

g. Drywell Redletion i H6gh Q1 1

1,2,3 29 (2011 K621 A, Bl tu

~

2.-

-- - --.- -a -

- ---- n--.

a-

---

e--

_-_--,--_-_.__--____---w______

TABLE 3.3.2-1 (Continued)

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

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

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

ACTION 23 - Be in at least STARTUP with the Group 1 isolation valves closed I

within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 24 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within one hour.

ACTION 25 - Isolate the reactor water cleanup system.

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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l

or close the affected system isolation valves and declare the i

affected system inoperable.

1 ACTION 28 - Close the shutdown cooling supply and reactor vessel head spray isolation valves unless reactor steam dome pressure s 145 psig.

]

ACTION 29 - Either close the affected isolation valves within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN i

within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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

liQlEl 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.

a.

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

l b.

Deleted.

HATCH - UNIT 2 3/4 3-15 Amendment No.127

1 TABLE 3.3.2-1 (Continued)

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

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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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.

l g.

Closes only RWCU outlet isolation valve 2G31-F004.

h.

Alarm only.

i. Adjustable up to 60 minutes.

l l

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 l

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 injection.

Following a reactor startup, a background radiation level will be determined and the associated 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-establishing normal radiation levels after a reduction in, or a completion of, hydrogen i

l injection and prior to establishing reactor power levels below 20% sted power.

1.

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

m.

Isolates reactor water sample valves 2B31-F019 and 2B31-F020. These are Group 1 valves.

~

l l

HATCH - UNIT 2 3/4 3-15a Amendment No. 127 1

=

4 TABLE 3.3.2 2

-4Q ISOLAT10N ACTUATION INSTRUMENTATION SETPOINTS i

ALLOWABLE

$ TRr FUNCTION TRF SETPolNT VALUE E

1.

PfuMARY CONTAINMENT ISOLATION N

e.

Rosetor Veeost Weter Lowl 1.

Low (Level al m 0inchee*

a 0inchee' 2.

Low Low (Level 21 m -47 inches' 2 -47 inches' 3.

Lew Lew Lew (Level II m.113 inchee' a.113 inches' b.

Dwywee Preeeure - High 51.92 pelo s 1.82 pois c.

Main Steam Une 1.

Redeshm High s a w fus power boekyound

s 3 x fuspewer back pound'*

2.

Pressure Low a 825 pelg k 525 peig 3.

Flow - High s 133% reted flow s 138% reted flow d.

Main Steam Line Tunnel Terrgwroture - High s 194'F s 194*F e.

Condenser Veeuum - Low a 7* Hg voeuum a 7* Hg voeuum U

f.

Turbine Building Area Temp.-High 5 200*F s 200*F e

w g.

Drywes Redleden High s 13e R/hr s 13e R/hr h

2.

SECOfeARY CONTANENT ISOLATION e.

Roseter Bubding Eshouet Reeselon High 5 90 mr/hr s 90 mr/hr b.

Dryws4 Pressure - High 5 1.92 pelg s 1.92 pelo e.

Remotor Vesoul Water Level-Low Lew (Lm4 21 m -47 inchee' a 47 baches*

g d.

Refusung Floor Eshavet to Radiot'en - High s 20 mr/hr s 20 mrihr ocx 8

M n 314 3-1.

mW

" Prior to the hydrogen ledsetten eyetem etertue end whh reestee ww then 20% reted poww. the nwmst fue poww redetion triples <m z

setpointe be ehenged beood on esseuleted oW tedeWen hydrogen inlootion eyotem operation. Aseeelsted trip /eleem setpointe mey o

bee $seted inleetion bened on sicher ealewiesione or measuremente of eetual redleWen levele coeuhing from hydregon Inisetion. Famowing a reector etermap, e beekground posselon level wm be determined and the esooelsted tetp/sierm setpointe adjusted within a 72-hour period. The redeWon level shed be decomdnad and h tripAsierm seapointe ehes be set witNn 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of to-ootatmahing normal rede6en levele efter e reduction in, or a ay eenytoden of, hydrogen injection and prior to estelmehhg reester power levele below 20% of rated poww.

PLANT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) 3.

Verifying that on each of the below pressurization mode actuation test signals, the system automatically switches to the pressurization mode of operation and maintains the main centrol room at a positive pressure of 2: 0.1-in.

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

a) Reactor vessel water level - low low low b) Drywell pressure - high c) Refueling floor area radiation - high d) (Deleted) l e) Hain steam line flow - high f) Control room intake monitors radiation - high f.

After each complete oi partial replacement of a HEPA filter 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 N510-1975 while operating the system at a flow rate of 2500 cfm 10 percent.

g.

After eacH complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove 99 percent of a halogenated hydrocarbon refrigerant test gas 2:

when they are tested in-place in accordance with ANSI H510-1975 while operating the system at a flow rate of 2500 cfm i 10 percent.

HATCH - UNIT 2 3/4 7-8 Amendment No.127

TABLE 3.3.8.71 fSHEET 1 OF 21

--4 Mx:

MCRECS ACTUATION INSTRUfWENTATION I

E MINIMUM NUM8ER APPLICABLE

' Q OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP SYSTEM (eHbl CONDITION ACTION l

1.

Rosetet Veseel Water Level.

2 1,2,3 52 Low Low Low (Level 1) tel i

2821 MG91 A. 8, C, D 2.

Dryweg Preeevre - High fel 2

1, 2, 3 52 2E11-Ne94 A,8, C D 3.

(Deleted) l v

4 Main Steam Une Row High tel 2Aine 1, 2, 3 53 2821-N686 A,8. C. D 2821-Nee 7 A. 8. C, D 2821-Ness A,8. C. D l

2821-N609 A,8. C. D a

w 5.

Refueling Hoor Aree Radiation High (c) 1 1.2,3,5

• 54

.E.

2021-K002 A D 8.

Control Room Alt inlet Radiation - High (c) 1 1, 2, 3, 5.

• 54 1Z41Re15 A,8 i'

E i

aa 8

a i

. 2 O

N N

t i

t m

m.

-...-s.

._...-_.,y

., _,,, _.. +

,mm...-_.__.--_,

,,,ym,,,,,_-,,,.,,..,,..,,,,,,..mu..ca_,,,.~,.mm.,,.

TABLE 3.3.6.7-1 (SHEET 2 0F 2)

MCRECS ACTUATION INSTRUMENTATION ACTION ACTION 52 - Take the ACTION required by Specification 3.3.3.

ACTION 53 - Take the ACTION required by Specification 3.3.2.

ACTION 54 -

With one of the required radiation monitors ir.gerable, restore the a.

monitor to OPERABLE status within 7 days or, wt.hin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintain operation of the MCRECS in the pressurization mode of operation.

j b.

With no radiation monitors OPERABLE, within I hour initiate and maintain operation of the MCRECS in the pressurization mode of operation.

c.

The provisions of Specification 3.0.4 are not applicable.

NOTES i

When handling irradiated fuel in secondary containment.

a.

(Deleted) 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 Trip Function to occur.

In these cases, the inoperable channel shall be restored to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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) a i

HATCH - UNIT 2 3/4 3-58b Amendment No. 127

)

i TABLE 3.3.6.7-2

-4$.

MCRECS ACTUATION INSTRUMENTATION SETPOINTO g

TMIP FUNCTION TRIP SETPose(I ALLOWABLE VALUE M

1.

Reector Veeeel Weter Level.

m -113 inches a 113IEhee Low Low Low (Level 1) m 2.

Drywee Pressure -liigh

,s 1.92 pose s 1.92 peig o

3.

(Deleted) l 4.

Mein Steam Une Flow - High s 138% reted flow s 138% seted flow

- 8.

Refvegn0 Reor Atee Redledon. N s 20 mAcur s 20 methour S.

Control Room Air inlet 5 1 w / hour s 1 mr/ hour Radletion. High

%A 44 i

e

(#1 CD n

3=

ko CL k=

i 2*

O NN r

6 m.

m m- -

-.m

-+w=me..

m---r-

<r-u mm amwm--

+4ee e v e w- *..

4,.

,e.,,we-e-e.sgiw..sw....

s c-e,,i v.,,-.

.....a.,...4--wem.-.,--=

-.-,,=,m--,w,w=

.., -. -. =w

-.--w=

TABLE 4.3.6.71

. g Q.

MCRECS ACTUATION INSTRUMENTATION SURVEfLLANCE REQUIREMENTS 8

CHANNEL OPERATIONAL C

CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRr FUNCTION

. CHECK __

TEST Sakt0 RATION SURVEILLANCE REQUIRED

--4 1.

Meector Veeest Weter Lowel -

S Q

R 1, 2, 3 m

Low Low Low (Levet Il 2.

Drywell Proseure H6gh S

O R

1, 2, 3 3.

(Deleted) l 4.

Main Steam Une Flow. High S

O R

1,2,3 5.

Refue#ne Floor Aree Mediedon.

S QM Q

1,2,3,5

• High 6.

Centrol Room Air 1rdet NA O*

R 1,2,3,5,'

RadleWon H6gh w

N w

Isn om a

i

• When hondung irredleted fuelin the secondary containment.

Sog e inetrument _"..._. using a etenderd current source.

3 Q.

8ac4

<T O,

a NN

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

....-.