ML20099K932

Amend 30 to License DPR-22,changing Tech Specs to Incorporate Limiting Conditions of Operation & Surveillance Requirements for l0-l0 Setpoint Logic Mod

ML20099K932
Person / Time
Site:	Monticello
Issue date:	11/16/1984
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Northern States Power Co
Shared Package
ML20099K933	List: ML20099K932 ML20099K950
References
DPR-22-A-030 NUDOCS 8411290628
Download: ML20099K932 (21)

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION L

.E WASHINGTON, D. C. 20555 j

NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 30 License No. DPR-22 1.

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

A.

The application for amendment by Northern States Power Company (the licensee)datedSeptember7,1984,complieswiththestandardsand requirementsoftheAtomicEnergyActof1954,asamended(theAct),

and the Comission's rules and regulations ~ 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 Comission;

~

C.

There is reasonable assurance (i) 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 conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, i

and paragraph 2.C.2 of Facility Operating License No. DPR-22 is hereby amended to read as follows:

2 Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 30, are hereby incorporated in the license.

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

8411290628 841116 PDR ADOCK 05000263 P

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This license amendment is effective.as of the date of its issuance.

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FOR THE NUC EAR REGULATORY COMISSION l

Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical 4

Specifications Date of Issuance:

November 1.6,,1984 i

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ATTACHMENT TO LICENSE AMENDMENT NO. 30 FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert 4

j vi vi 23 23 24 24 I

48 48 j

60b 60c 63 63-63a 69 69 69a i

70 70 I

71 71 71a 127 127 150 150

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151 151 157 157 4

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o TAI *I CF CCC Fage 1.0 Du.u.. 0N5 1

e 2.0 SAII:T R. Ts AND 'm= SAICT STS?i Sr.... 03 6

2.1 and 2.3. Yuel Cladding !=:eg:1:7 6

2.1 3ases 10 2.3 3ases 14 2.2 and 2.4 Raacur Coolant Sys =m 21 2.2 3asas 22 2.4 3asas 24 3.0 m CONDCONS ?DE OPIRATION A2:D 4.0 sugyc:,:,A::c gggg"W 26 i

3.1 and 4.1 Reacur ?:stec. ion Systsa 26 1

,4.1 3asas 3.1 3ases 1

35 41 1.2 and 4.2 ?:stec.1ve Instrumentation 45 t,

A.

Primary Containment Isolatina 7anc 1ons 45 ~

3.

Imergency Core Cooling Subsystems Actuation 44 C.

Control lod 31ock Actuation 46 D.- Deleted E.

Reac:or 3uilding 7ent11ation Isolation and 47 Standby Gas Treatment System Initiation i

7.

Racir:ulation Pu=p Trip Initta:1on 48 G.

Safeguards 3us 7al: age ? stection 43 l

H.

Ins::umentacion for S/RV Low-Low Sec Logic 48 l

3.2 3ases 4.3 3ases 64 72 3.3 and 4.3 Con:rol Rod Systen i

76 A.

Reactivity L1=1tations 76 3.

Control Rod *Jt:hd:sval 77 l

C.

Scram Insertion Times 31 D.

Con::al Rod Ac=-Scors 32 E.

Raactivity Anomalies 33

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Sc 2m 11senarge 7alume laa C.

Recuired Action 3;;

4 3.3 and 4.3 3ases 34 T

Anendment !!o. 30

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?!ST OF TAM.IS Table "o.

Psee 3.1.1 Rese:or Protection System (Scram) Ins::umen-Requiremen:s 23 4.1.1 Scram Instrument Func:ional Tests - Minimum Functional 32 Test Frequencies for Safety Instrue.entation and Control Circuits 4.1.2' Scram Instrument Calibration - M*"4-an Calibra:1on 34 Frequencies for Reac:or Protection Instrument Channels

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3.2.1 Instruments:1on that Initiates Pri=an Cont.ainmen:

49 Isolation Functions 3.2.I Instru=entation that Initiates Emergency Core Cooling Systems 52 l

3.2.3 Instrumenta:1on that Ini:1stes Rod Block 57 3.2.1 Instrumentation that Initiates Reactor 3uilding Ventilation 59

{-

Isolation and Standby Cas Treatment System Initiation 3.2.3 Instrumentation that'Initia'tes a Recirculation Pump Trip 60 l

3.2.6 Instrumenta.1on for Safeguards 3ns Degraded Voltage and 60a Loss of vol: age ? otection l

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3.2.7 Instrumentation for Safety / Relief Valve Low-Low Set Logic 60b 4.2.1 P4"d - Tes: and Calibration Frequency for Core Cooling, Rod

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31ock and Isolation Instru=entation 61 3.5.1 Safen Related Snubbers 13 1

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3.7.1 Pri=ary Centain=en: Isolation 172 3.3.1 Radioactive Liquid Effluent Monitoring Instruments:1on 1S91 3.3.2 Radioactive Caseous Iffluen: Monitoring Ins::uments: ton 198k 4.3.1 Radioac:1ve Liquid Effluent Monitoring Instrumentation 198n Surveillance Reonire=,ents 4.3.2 Radioactive Caseous Iffluen: Monitoring Instru=en:stion 19En

'i Surveillance Requirements 1.3.3 Radiose:ive Licuid Uaste Sa:nling and Analysis ?regram 1997 4.3.1 Radioactive Caseous Uaste Sa=pling and Analysis Program 193s 3.11.1 M2::'-- - Average Plans: Linear Res: Genera:1on late 314 Ys. I.n oSu:S 3.13.I

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Amendment flo. 30

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lhTE3 8E"l!t inued:

2.2 The normal operating pressure of the reactor coolant system is approxjmately 10!0 psig. The t urbine t rip with f allyre of the bypass system represent.s the most severe primary system i

pressure increase resulting from an abnormal operational transient. The safety / relief valves (S/RV's) are sized assumint; no direct scram during PlSIV closure. The only scram

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se;sumed is from an indirect means (high flux). The analysis assumes that only seven of the vinht S/gy's are operable and that they open at 1% over their setpoint with a 0.4

.. cond dalay. Iteactor pressure remains below the 1375 paix AS$1E Code limit for the 1

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2 2.4 1hc net t ings on the react or high pressure scres, teactor coolant systese safety / relief valves, turbine control valve feet closure scram, asul turbine stop valve closure scrs= have been establielied i

in n inure never scathing the reactor conlant system pressure safety ll:ntt no well as assuring the synt ra pressure does not encecil the range of the fuel clailding Integr ity saf ety limilt. The APRh flus scram and the turbine bypese systeen also provide protection for these safety limite.

cutenn In :id.ll t i on to prevent ing power operation above 1975 pelg, slee preesiste scra'ai hacks up tlee APRH llum scram for steam escut i on line teolation type tranalente.

Ihr scactor coolant system natety/ relief valves mesure that the reactor coolant system pressure j

enlety limit le.never rearheit.

In compilence witle Section lit of time ASHE Boller and Pressurei

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Vennel code, 1965 edi t ion, the safety / relief valves must be set to open at a pressure no higher than.

l lil5 g*e rcent of Jealgn pressure, a nd t hey snue t Ilmit time reactor. pressure to no smore tisen 110 percent nl design pressure. The safety / relief valves are elsed scenrding to the Code for a condition of i

mly rionure while operating at 1670 HWt, followed by no HSIV cicoute scram but scram from en 4

in.Il s ert. (high f lus) means.

Witle' the safety / relief valves set as specified leerein, the maulmum ven:ial pressure remains below the 1375 psig ASHF. Code limit.

O'lly seven of the ca ht valves are assumed to be operable in thle analysis and the valves are amenamed to open at II c

l al.nve their setpoint with a 0.4 second delay. '

1 11.c npe rat or wil l set the reactor coolant high pressure scram trip setting at 1975 pelg or lower.

Ib.weve r, the actual netpoint can be as nucle se 10 pet above the 1075 pela Indicated set poliit due to

' the deviatione discussed in alie basis of Specificet ton 2.3 on Page 18.

In a li ke mie nne r, the opesator vill act the react or coulent system safety / relief valve initiation trip setting at 1808 p*ilg or Inwer.

Ilowever, the actual ee't point can be as much as ll.1 pel above the Il08 pelg i

In.llrated set potest elue to the devlatione discussed in the heale of Specification 2.3 on Page IR.

A vlutation of flits specification is assumest to occier only Wleen a device le kunwingly set, outelde of the limit ing trip setting, or when a sief ficient sensber of devices have been af fected by any secans c,

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1. 0 f.Illllll1G COHulT10 tis FOR Ol'ERATION 4.0 SURVEILLANCE REQUIREMENTS

r. I:ccirrulation Pinup Trip and Alternate I:o.1 In]ert ion Initiation

1. -Illienever tlic reactor is in tlie RUN mode, tin-I.imit ing Con.Il t ions for Operat ion for tii-Justrumentation IIsted in Table 1.2.5 sliall lee swt.

. Safeguas.Is Itns Voltage Protection I. IJlien.ver Llie safeguards auxiliary electrical power system is required to be operable by Sperifscatton 3.<3 tiie I.imiting Conditions tier espeiation for tin-Inutrumentation listed in 1..I.l e 1.2.6 shall lie met.

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1:..suli t ioins for Operul lion for tlie inst o nei.cul a t lun I 1st ed. in Tab le

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Total Ho. of Instru-Illu. No. of Oper-Trip of Operable ment cleannels Per able or Ope ating ite gui rc<l huict ton Setting or Operating Trip System Instrussent Clianne tt. Conditions 4 Trip Systeses Per irlp Systen

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105013/97013 psig ior Opening /

1040t3/960t3 psig.

t:I..s i ng (1)

Illsclearge Pipe 5011 paid (3) 2 (2) 2 Ps essis t e Intelle t t 2

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TABI.E 4.2.1 - Continued flinimum Test and Calibration Frequency for Core Cooling, Rod Block and Iaoistion Instrumentation Inntrument cleanne t Test (3)

Calibration (3)

Sensor Check (3) sAIIs;IIAltlis BIIS Vol.TAGE 1

ii. grn.lc.I VoI tage Hate I quarterly Not applicable

.I oscetson J.

I.ons of voltage tiote 1 Once/ Operating Cycle Not applicable l's o r cc t ton

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t i i i i /4:i't.1IV VAlyE IgtJ-pit! SIT, Ip;IC

i. ::.... i..r sc rani s. nsing once/ Shutdown (81 i:.....-...- 1 %.s s u s. - - oisening once/3 months once/ Operating Cycle Once/ day I

1.

I:.. cs.ir l'i cusure - Closing once/3 months Once/ Operating Cycle Once/ day

t.. iii :w l...i ge l'i ie l'eessure once/3 months once/ Operating Cycle t

r 5.

I nle i te il Timer once/3 months Once/ Operating Cycle R

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2 mo s

5 4

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63 x

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TABl.E 4.2.1 - Continued litnimum Tent and Calibratton Frequency for Core Cooling, Rod Block anel Isolation Instrumentation t h i l l 'i:

(1) folefally noce per month isntil exposure hours (if an defined on Figenres 4.1.1) is 2.0 x 10 thereafter accuriling to Figure 4.1.1 with an interval not greater than three anonths.

(1) thillbrate prior to, normal nhutdown and start-sep niid thereafter check once per ahlft and test once per urek until no longe r respit red.

Calibration of this instrisment prior to normal shutdown means a.l j n s t m.fu t of channel trips so that they correspond, within acceptable range and accuracy, to a n l.nn la t e.1 nl gua l injected into the inntrument (not prlinary sensor).

In addition IRil gain adjisstment ulil be perin a meil, as necenna ry, In the april /IRif overlap region.

( t) inne lon.il tvits, calthrations and nennor checks are not required when tlie systems are not required to I.e opesable or nre t r i piied.

If tests are missed, they shall be performed prior :to retistning the '

riyntemn to asi operable nta tten.

('s) uhenever fuel handitng la in process, a sennor check shall he performed once per shif t.

( 'i ) A functlunal test of this instrument incans the injection of a simentated signal into the Instrument (not pa lmary sensor) to verify the proper instrinnent channel response alarm and/or initiating action.

( f. ) llilk in st rismont will be calibrated every tietee months by means of a built in current source, and each iefneling not. ige with a known rndinactive nonrce.

(1) ::ni vo i l lance a l an to be performed on contninment isolation function of this instrumentation at the upectfled intervnis.

(3) Qnce/ shutdown if not tested during preytous 3 nontli perind.

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incrennen core voiding, a negntive reactivity feedback. liigh preneure sennora initiate the pump trip in the event of an inointion transient.

I.ow level sensors initiate the trip on loss of feed-water (and the resulting flSIV clonure). The recirculation pump trip is.only required at high renctor power levels, where the anfety/ relief valves have insufficient capacity to relieve the steam which continues to he generated af ter reactor isolation in this unlikely postulated event, requirfug the trip to be operable only when in the RUN mode la therefore conservative.

s Safety / relief valvo low-low set logic is provided to prevent any safety / relief valve from 2

opening uhen there is a elevated water leg in the respective discharge line. A high water i

leg is formed immediately following valve, closure due to the vacuum formed when g

steam comlenses in the 1ine.

If the valve reopens before the discharge line vacuum o

breal:ers act to return water level to normal, water clearing thrust loads on the si i sch.1rge line may exceed their design limit. The logic' reduces the opening setpoint

.E and increasen t he blowdown range of t hree non-APRS valves following a scram. A 15-second int ei val hetueen 'unhsequent valve actuations 19 provided assuming one valve fails to

3. n.\\SI::.

69 t

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open and instrumentation drift has caused the nominal 80-psi blowdown range to be reduced to 60 psi.

Maximum water leg clearing time has been calculated to be.less than 6 seconds for the Nonticello design.

Inhibit timers are provided for each valve to prevent the valve f rom being manually opened less thou.10 seconds following valve closure.

Valve opening is sensed by pressure switches in the valve discharge line.

Each valve is provided with two trip, or actuation, systems. Each system ja provided with two channels of instrumentation for each of the above described functions. A two-out-of-two-once logig scheme ensures that no single' failure will defeat the low-low set functinn and no single failure will cause spurious operation of a safety / relief valve.

At lowable deviations are provided for each specified instrument setpoint. Setpoints within the specifled allowable deviations provide assurance that subsequent safety / relief valve actuations are suf ficiently spaced to allw for discharge line water leg clearing.

Although the operator will set the set. points within the trip settings specified in Tables 3.2.1 through 3.2.7, the actual values of the various set points can differ appreciably from the value the, operator is attempting to set.

The ouviations could be cansed by inherent Instrument error, drif t of the set point ect.

Therefore, these deviations have been g

accounted for in the various transient analyses and the actual trip settings may vary by the foilowing amounts:

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69a O

Trip Function Deviation it. ac t.it Ilulleling Vent ilat ion luolation, and Ventilation Plenum

+0.2 mr/hr

ian.lhy i;;is Treatment System initiation Radiation Honitors

giccil f ea t ion 1.2.E.3 and Table 3.2.4 Refueling Floor Radiation Ifonitora

+5 mr/hr Low Reactor Water Level

-6 inches liigh Drywell Pressure

+ 1 pai Fii2Tu y (:nntalument Isolation Functions Low Low Water Level

-3 inches Table 3.2.1 liigh Flow in Plain Stenin Line 42 %

liigh Temp. in Hain Steam

+10*F Line Tunnel 1

Low pressure in Hain Steam

-10 pst Line Iligh Drywall-Pressure

+1 psL Low Reactor Water Level

-6 inches IIPCI Iligh Steam Flow

+7,500 lb/hr 5

5.

IIPCI Steam Line Area Illgli

+2*F S

Temp.

5

' RCIC liigli Steam Flow

+2250 lb/hr v

RCIC Steam Line Area Illg's Temp

+2*F f3-Shu.tdown Cooling Supply Iso

+7 psi w."

l..' li A:st.:3 70 l

Trip Function Deviation Instinmentatton Tliat Initiatcs Emergency 1.ow-1.ow Reactor Wa ter I.evel

-3 Incties Co n e Con I liit; Sys t emo Talil e 3.2.2 Reactor I.ow Pressure (Pump

-10 psi Start) Permissive liigh Drywell Pressure

+1 psi I.uw Reactor Pressure (Valve

-.10 pai Permissive)

IIIIIsrumentation That Initiates IRH Ilownscale

-2/125 of Scalc 14o.1 niock IHil tipscale

+2/125 of Scale Talile 3.2.3 APRH Downscale

-2/125 of Scale APlel Upscale See Basis 2.3, RHH Downscale

-2/125 of Scale Hilli (fpucaIe Same as APlui tipscaie Scram tilsclearge Volume-liigli

+ 1 gallon Levei lustrumentation That Initiaten liigh Reactor Pressure

+ 12 pai 3

itecirculation Pump Trip I.ow Reactor Water Level

-3 Inches S_,

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v

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u i.2 nasi:s 71 O

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Trip Function Deviation es.nentnt.fon for Safety /Rel'lef Valve Reactor Coolant System

+ 20 psig l.o u Set 1.ogic Pressure for Opening / Closing.

Opening - Closing Pressure

• > 60 psi Dlacharge Pipe Pressure 10 psid Inhibit Timer Inhibit

-3 sec

+ 10 sec N

m S.

A vluintton of this npecification in assumed to occur only when a device is knowingly set outside of the 3

liniit inn t r ip net t lonn, or, when a nuf ficient number of devices have been af fected by any means nucle that 5

t ii.- not omat le function in inenpable of operating within the allowable deviation while in a reactor mode in utilch t he npecified function must be operabic or when actions specified'are not initiated as specified.

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t 1.0 1.118171183 0011111T10!I9 IUR orERAil0H 4.0 SURVEll. LANCE ret lUIREllEll18

- =. enemme e E.

!Intety/Rettet Velves E.

8stety/Rettet Yalves 1.

Ibir Ing power operating ennditions 1.

a.

A minimum of seven natety/rettet and whenever reactor enninnt pressure valves shall be fiench checked or la gr eater than 110 pnis and reptoced with a, bench checked t eroper atur e is great er then 343"F.

• vetve each refueling nutage.

The nominal self-actuation The safety volve function (self-setpoints are specifled in n.

actuation) of neven safety /

Section 2.4.B.

rettet valves shall be operehte.

b.

At least two of the safety / relief h.

The solennld activated rettet valves shall be dienseembled and function (Automatic Freenure inspected each refueling outage.

Relief) shall he operable se required by spe-Igeration 3 5.R.

c.

The integrity of the safety / relief valve bellove ohntt be continuously

2. 't he I.ow-l.nw Set function for three monitored.

non-Automatic Prensure Itelief volves nhnll he Operabic no npecified in d.

The operability of the bellows Section 3.2.1.

' monitoring systese shall be demon-p strated et leset once every three 4

d months.

n 5

to 2.

Low-Low Set 1.ogic surveillance shall S

be performed in'accordance with Table,4.2.1.

5 t

I w

O 111 t

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nnsen contInned 3.6 and 4.6:

D. Coolant f.eakage l

l The allownble leakage rates of coolant from the reactor coolant system have been based on the predicted l

and experimentally observed behavior of cracks in pipes. The normally expected background leakage due j

t o engulpment design and the detection capability of the instrumpntation for determining leakage was a l so cons tilered. The evidence obtained from experiments suggeste that for leakage sgmewhat greater thnn-thot.specified for unidentified leakage, the probability le small that the leperfection or crack annociated with nucle leakage would grow rapidly. Ilowever,' in all. bases, if the leakage rates exceed' the valuce specified or the leakage la located and known to be Pressure Boundary 1.eakage and they cannot be reduced within the allowed times, the reactor will be aliutdown to allow furtlier investigation, and corrective action.

o Two leakage collection sumpe are provided inalde primary containment. Identified leakage is piped from the recirculatina pump seale, valve stem leak-offs, reactor vessel flange leak-off, bulkhead.

anil hellown draina, and vent cooler draine to the drywell equipment ' drain sump. ; All otlier leakage la collected in the drywell floor drain aump.

Both susps are equipped with level and flow trans-mitters connected to recorders in the control room. An annunctetor and computer alarm are pro-vided in t he control room to alert operators when allowable leak rates are approached. Drywell alrhorne particulate radioactivity is continuously monitored as well as drywell atmospheric tem-perature and pressure. Systema connected to the reactor coolant system boundary are also monitored f or leaknne by the Process Liquid Radiation Honitoring System.

lhe sensitivity of the sump leakage detection systems for detection of. leek rate changes la better than one gpm in a one hour period. Other leakage detection metloods provide warning of abnormal leakage j

an.1 nre not directly calibrated to provide leak rate measurements.

E. Safety / Relief Valves

'lesting of all safety / relief valves each refue' ling outage ensures that any valve deterioration is detected.

A t olerance value of 1% for safety / relief valve setpoints is specified in Section 111 of the ASME Bo'ler 3fj anil l'ressure vessel Code.

Analyses have been performed with all valves assumed set 1% liigh. As discussed i

in the Section 2.2 Bases, the 1375 psig Code limit is not exceeded in any case.

So i

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3. 6 /4. 6 11A S ES 15g DJ ca

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l l-linnec Continued 3.6 an1 is.6:

'the nr.fety/ relief valves have two ninctions; 1.e. power relief cr self-actuated by high pressure.

'ihe nolenold netunted funct.lon ( Automatic Pressure' Tlelier) in which external instrumentation si nals of E

coincident high dryvell pressure and lou-low uster level initiate opening of the valves. TI,ile funct.lon is discunned in Specification 3 5.E.

In addition, the valves can be operated manuall,v.

'lhe norety ninct. ion la perfonned by the some safety / relief valve witit self-actuated integral 1cilous nnd pilot valve capeing noin volve operation. Art.icle 9 of the AalE Pressure vessel code Gection III II.ielenr Vesnels recluires that, these bellove be monitored for failure since this voisld defeat the safety fi.nction of the safety /. relief volve.

11. la realized that there la no way to repair or replace the bellows during operation and the plant inunt be shut down to do t.hin.

'lhe thirty-day period to do this allows'the operator flexibility to choose hln time for shutdown; meanwhile, beenuse of the redundancy present in' the dealgn and the continuing siont Loring of the integrity of the other valven, the overpressure pressure protection has not bee'n cierproni ned.

'lhe auto-reller function would not be impaired by a failure of the bellows. Ilowever, the nel f-nctuated overpressure safety Ibnction vould be impaired by such a' failure.

Provinion also has been maile t.o detect failure of the bellows monitoring system. Testing of this nyntem quarterly provides naaurance of bellows integrity.

When the setpoint la being bench checked, it la prudent to dissaaer.,ble one of the safety / relief valves to exnmine for calid buildup, bending of certain actuator members or other algne of possible deterioration.

Low-l. ins Set. I.ogic has been provided on three non-Automatic Pressure itelief System valves.

This logic in discussed in detail in the Sectioni 3.2 Bases. This logic, thrdugh pressure' sensing instrumentatlon, reduces the opening setpoint and increases the blowdown range of the three y

select ed valves folloutng a scram to eliminate the discharge line water leg clearing loads iesultinn isom niultIple valve openisins.

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i ne i.. i ea 8

151 3.6/I. 6 luutt:;

0 4

].0 I.IllITING C0!!DIT10flS FOR OPElt ATION 4.0 SURVE11. LANCE RF.QillREttEfffS d.

During reactor isolation conditions d.

Whenever there in indication of relief the reactor pressure vesset shall be valve operation with a suppression pool l

deprennurized to < 200 psig at normal temperaturell60*P and the primary I

cooldown rates if the suppression coolant system prer.eure > 200 poig; an

. pool temperature exceeds 120*F.

extended visual examination of the suppresalos; chamber shall be conducted e.

The suppression chamber water volume before resuming power operation.

l nhaII be168,000 anili 72,910 cubic 3

feet, e.

The suppreaston chamber unter volume shall be checked once per day, i

f.

Two channels of torus water levet instru-menta tion sha ll be ' operable.

From and f.

ne suppression chamber unter volume a f t e r t he da t e tha t' one channel is made indicators shall'ha calibrated semi-or found to be inoperable for any reason, annually.

reactor operation is permissible only during the onceeeding 30 days unless

.2.

Primdry Containment Integrity nuch channel is sooner made operable.

If both channels are nide or found to be a.

Integrated Primary Containment Leak Test (IPCLT)

I nopera b le for any reason, reactor opera-t ton in permissil.le only during the he containment leakage rates shall be succeeding six hours unless at least demonstrated at the following test schedule one channet is sooner mide operable, and shall be determined in conformance with the criteria specified in Appendix J of 10 2.

Primary Containment Integrity CFR 50 using the methods and provisions of ANSI H45.4-1972:

Primary containment integrity, as defined i

in Sect Ion I, shall be maintained at all t imes when the reactor is critical or when 1.

Three Type A Overall Integrated containment the reactor water temperature is above Leskage Rate testa shall be conducted at 2I2*F anel fuel.is in the reactor vesset 40 + 10 month intervals during shutdown k

except uhtle performing low power physics servlce(41 psig) during each 10-year at P s

tents at atmospheric pressure during or period. The third test of each I

after refueling at power levels not to set shall be conducted during the shut-exceed 5 Hw(t).

down for the 10-year plant inservice inspection.*

I E

• The third test of the first 10-year service g

I period shall be conducted during the 1980 refueling nhutdown. The first test of the

.second 10 year period shall he conducte.1 j

during the 1984 refuelirig shutdown.

157 L il!.. ?

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