ML20028B964

Amends 84 & 77 to Licenses DPR-29 & DPR-30,respectively, Expanding Tech Specs for Scram Discharge Vol to Include Surveillance Requirements for Vent & Drain Valves

ML20028B964
Person / Time
Site:	Quad Cities
Issue date:	12/23/1982
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Commonwealth Edison Co, Iowa Illinois Gas & Electric Company
Shared Package
ML20027A864	List: ML20027A863 ML20028B964 ML20028B965 ML20028B966
References
DPR-29-A-084, DPR-30-A-077 NUDOCS 8301040195
Download: ML20028B964 (14)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION j 4/

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COMMONWEALTH EDISON COWANY AND IOWA ILLINOIS GAS D ELECTRIC COWANY DOCKET NO. 50-254 QUAD CITIES. STATION UNIT NO. 1 AENDMENT TO FACILITY OPERATING LICENSE Amendment No.84 License No. DPR-29 1.

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

A.

The application for amendment by the Comonwealth Edison Company (the licensee) dated October 14,1980, as supplemented October 22, 1981, complies with the standards and requirements of the Atomic Energy Act J

of 1954, as amended (the Act), and the Comission's rules and regu-lations 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 (f) 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 common defense and security or to the health and safety of the public; and l

l 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 Spec-ifications as indicated in the attachment to this ' license amendment and paragraph 3.B of Facility License No. DPR-29 is hereby amended to read as follows:

i B.

Technical Specifications The Technical Specifications contained in Appendices A and l

B, as revised through Amendment No. 84, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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i 8301040195 821223 l PDR ADOCK 05000254 P

PDR

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

FOR THE NUCL REGULATORY COMNISSION Domenic B. Vassallo, Chief l

Operating Reactors Branch #2 l

Division of Licensing

Attachment:

l Changes to the Technical l

Specificctions Date of Issuance: December 23, 1932 l

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ATTACHMENT TO LICENSE AMENDMENT NO. 84 FACILITY OPERATING LICENSE NO. DPR-29 DOCKET NO. 50-254 RehisetheAppendix"A"TechnicalSpecificationsasfollows:

Remove Insert 3.2/4.2-14 3.2/4.2-14 3.2/4.2-16 3.2/4.2-16 3.3/4.3-3 3.3/4.3-3 3.3/4.3-9 3.3/4.3-9 l

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No. I,84 0

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DPR-29 5

In8LE L21 MDuit0M TEST AND CAL'8EAfl0N FRE!15ENCY FOR CORE AND CONTAINMENT C00UNE SYSTEMS INSTR 9MDITATION.

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300 SLOCA1, AND 150LAT10ll5h hutusset tueussut Amamul inseussut samms Tama'D gasmauss28 gassam ses useimmsenes 1.

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Dywes he presswa Q)

Once/3 months Nuo 3.

Reactor hw presswe (D

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Once/3 mor.ths Mene

b. Contamment pressre (1)

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Once/3 montes Nee dischste 5.

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A8IBA flow vanacle u) G)

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358 detector not a startup pantun None

10. 558 downscale (5)0)

(5)Q)

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11. Higle water irvel a scram Chee/3 months Not appacaele Mme decnerte voanne (SW)

12. SW high level trip Refuelir.g Not acglicable Norte '

bypassed cutage imin sessmen musen 1.

Steam tweet he temperatwe Refereg outage Refueling outage None i

2.

Steamine h h Sow u)

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3.

Steantee bw pressre (1)

Once/3 months None 4.

Steamtme he radaten (1) (43 Refusing outage Once/ day 5.

Reactor hw bw water level Q)

Ome/3 mestns Once/ day seit keisnes 1.

Steandine hp new Qice/3 months Once/3 masths Mme 2.

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law reactor presswe Once/3 mor:tas h3 momis None Amendment No.M 04 12n.2-16

l QUAO-CITits DPR-29

3. The controt ted drive housing support synaem shall he in place during reactor

3. The sorrectness of the sentrol rod power operation and when the reactor withdrawal sequence input so the coolant system is pressursard above RWM sompasse shall be venAed afast simospherse peruure with fact in the leading the sequence.

l reactoe vencl. un!cu all conerni rods 88'I'**'

I " 'A are fully inserted and Speenfication hawal wwards emWity, the capaW.'

3 3.A.I as met ity of the rod wonh saisimizer a Contro! od withdrawalsequences properly ful611 its function shall be a.

shall be established so that maa.

versAed by the followint cheets:

imum re.etivity that could be 8.

. e**Puiu Mne &ag.

added by dropout of anv~ encre.

aosde ws: shah k succeWuHy ment of any one control b!:de Perfumed.

would be cuch that the rod drcp accident

b. Proper annunciation of the se:ec.

design lirsit of 280 cal /c=..in.not excoedad.

tion error of one out of seovence

b. Whenever the reactor as in the

"# # ""*" Y W'0" Saartup' Hot Stancey or Run

e. The rod block f' unction of the mode belom OM rated thermal RWM shall be ver Red by with.

power. the rod worth minimiter drawing the Arst rod as an ost.

aball be operab!e A second cpera.

of.wquence control rod no more est or qual: Sed technstal pe son i

may be used as a substitute for an than to the block posat.

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inoperable rod worin minimater which fa !s after withcrawal of at least 12 control rods to t. e fully withdrawn pos:tinn. The red worth man: mazer may :lso be bypassed for low power physics i

testing to demonstrate the shut.

down martin requartments of Spec 4Ecatinn 3.3.A ff a rivelcar engineer is present and verifies the seep by.ste,i rod movements of the esse procedure.

l 4 Control rods shall not be withdrawn d.

Prior to control rod withdrawal fer for nattup or re'veling unten at least startup or during refuehng. verify that two source ta.ite channels have an at least two source range charinels observed count rate equal to or greater have an observed count rate of at least than three counts per second and these three counts per second.

SRM's are fully inserted.

5. During operation with hmiting con-S. Whe'n a limiting control tod pattern trol rod passerns. as determined by the exists. sn instrument functional test of nuclear engbert. either:

the RSM shall be perfermed prict.to i

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s. both RDM channels shall be opershle.

and daily theresfier.

b control rnd withdrawat shall be

s. T*wn.artas distNr.w* u*lises. v==

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3.3#43 3 i

Amendment No. E, 84

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b. the delayed neutron fraction chosen for the bounding reactivity curve

c. a ber' inning-of-life Doppler reactivity feedback l

d.. scram times slower than the Technical Specification rod scram insertion j

rate (Section 3 3.c.1)

e. the maximum possible rod drop velocity of 3 11 fps

f. the design accident and scram reactivity shape function, and

g. the moderator temperature at which criticality occurs In most cases the worth of insequence rods or rod seCments in conjunction with the actual values of the other important accident analysis parameters

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described ~above, would most likely result in a peak fuel enthalpy sub-stantially less than 280 cal /g design limit.

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should a control drop accident result in a peak lhel energy content of 280 cal /3. fewer than 660 (7 a *

7) fbel rods are conservatively estamated to perforate. This would result in an offsier done well below the guideline value of 10 CFR 100. For 8 a 8 fuel, fewer than 850 rods are conservatively estimated I

a perforate, with nantly the same consequences as for the 7 a 7 fuel case because of the rod powee differences.

The rod worth minimiser provides aviomatic supervision to assure that out of sequence control rods will not be withdrawn or inserted i.e., it limits operator deviations frorn planned withdrawal sequences (reference SAR Section 7.9). It serves as a backup to procedural control of control red worth. In the event that the rod worth minimiter is out ofservice when required, a licensed opersoor M

or other qualified technical employee can manually fhlAll the control rod pattern conformance W

ikaction of the rod worth minimizer. In this case, the normal procedural controls are backed up by independent procedural controls to assure conformance.

4. The source range monitor ($RM) system performs no automatic safety system ibaction. Le.,it has no scram function. It does provide the operator with a visual indication of neutron level This is needed for knowledgeable and efRcient reactor startup at low neutron levels. The consequences of reactivity accidents are functions of the initial neutron flua.The requirement of at least 3 counts per second assures that any transient should it occur, begins at or above the initialvalue of 108 of rated power used in the analyses of transients from cold conditions. One operable SRM channel would adequate io monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal A minimum of two operable SRM's is provided as an added conservatism.

5. The rod block monitor (RBM) is designed to automaticaDy prevent fbei damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided, and one of these may be bypassed from the console for maintenance and/or testing Tripping of one of the channels will block erroneou.s rod withdrawal soon enough to prevent fbel damage. This system hacks up the operator, who withdraws control rnds according to a written sequence. The speciAed restrictions with one channel out of service conservatively assure ena damage will not occut due to rod withdrawal errors when this condition exists. Dunng reactor operation with certain limiting control rod patterns, the withdrawal of i

a designated single control rod could result in one or more fuel rnds wi th -

MCPh's less than the MCPR fuel cladding integrity safety limit.Duringuseorsuchpatier

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it is hdged that testing of the RBM system to assure its operaMlity prior to withdrawal of will assure that improper withdrawal does not occur. It is the responsibility of the Nucicar Engin io identify these limiting patierns and the designated rods either when the patterns are an 0

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Amendment No. $, S4 3.3/4.3-9 e

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pa aseg'o UMTED STATES i

,n NUCLEAR REGULATORY COMMISSION g

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COPEDNWEALTH EDIS0N COWANY

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AND IOWA ILLINOIS GASTD ELECTRIC COWANY DOCKET NO. 50-265 QUAD CITIES STATION UNIT NO. 2, AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 77 License No. DPR-30 1.

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

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated October 14,1980, as supplemented October 22, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regu-1ations set forth in 10 DFR Chapter I; B.

The facility will operate in confonnity with the application, I

the provisions of the Act, and the rules and regulations of the Commission; 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 conducted in compliance with the Commission's regulations; 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.

2.

Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to'this license amendment e

and paragraph 3.8 of Facility License No. DPR-30 is hereby amended to read as follows B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.77, are hereby incorporated in the license. The licensee shall operate the facility in 1

accordance with the Technical Specifications.

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3.

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

l FOR THE NUCLEAR REGULATORY COMNISSION

/

l Domenic B. Vassallo, Chief l

Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specificctions Date of Issuance:

Decerber 23,1982 t

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t ATTACHMENT TO LICENSE AMENDMENT NO. 77 FACILITY OPERATING LICENSE NO. DPR-30 DOCKET NO. 50-265 RehisetheAppendix"A"TechnicalSpecificationsasfollows:

Remohe Insert

-3.2/4.2-14 3.2/4.2-14 3.2/4.2-16 3.2/4.2-16 3.3/4.3-3 3.3/4.3-3 3.3/4.3-9 3.3/4.3-9 l

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DPR-30 188LE3.!(

315filullDITAT!ON TNAT 11@.4T:3 It3D BLCCI

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$12/12$ hg seat

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2" 3158 dowscsW" 230' costs /sse a

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571 hich water level scram FA trip bypassed llotes

1. For the startur/ Mot stander and.itun positions of the reactor mode scleeter switch, there ar.all 3e two ope atle or tripped trio systess fer eat:1 fur.c-tion execpt the 3RM red bloet:s. 711M upscale and IR.v dcwnscale need riot bc opersele in the Run petition, AP tM down=cale, AFEM upscale (flow ti.ised),

and RT4 doernceale ncc*$ not be operable in the Startup/ Hot standuy emu.

Tha M.it! vpccule cod riot be orerable at loss than 3C% reLad ther=al powcr.

Orne chganol may be bypassed e'.:-evo 305 rated themal power provided that a limitirac control rod pattere Joos not exist. For systees with m re than one channel per trip systen, 2.f the first column cannot to not far one of the two trip sy?te s, this conditinn.ay exist far up to 7 days provided that during that time the aperanin systes is fur.ctionally tested in-mediately and daily thereaf ter; if this condition lasta longer tlat 7 dayn the systes shall be tripped. If the first column cannot te met for laoth trip systems, the systems shall be tripped.

2. V is the percent of drive flow required to produce a rated core flow of 793 sillion 13/hr. Trip level setting is in parcent of rated power 4

(2511 E t).

1 sur maman se, as w m e s a as ni n== rase.

& as an.aan a humanad som the amme use e 2:100 CPS.

s om e w in srm.een== = gr===e.

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Amendment No. '/J, 77 w- -- a-

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InsLE421 IlamICII TEST AND CAUSEATION FRE15ENCT FOR CORE ANO CCNTAINMENT C00UNG ST37Di$ INSTileMENTAfl0N.

300 BL2CAS, ANO 130LAT10NSh hutumat humussut Assesus humanes 5

tasmai InstW Cuarsessa gasse8 f

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1.

Reacts low.bw ante levei Q)

.0ms/3 months Once/ day 2.

Drymee h:gh yesswa (1)

Once/3 months None 3.

Reactor be pessee (1)

(hes/3 months lune 4

Contammet sorsy aterlock

a. 2/3 can heet (D

Once/3 months Neu

6. Catamment yesswe (D

Once/3 montes Nue 5.

(sw.teauwe core coolmg pumo (D

Ous/3 maths Nue (adurge 4.

Ihnenmange 4.W essental Nefusing outage Refusing outage Nue aus sexes 1.

AMet downscale UI 0)

Once/3 months Nuo 2.

APIIM flow vanacle

' (D 01 Asfusing outage None

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558 upscale W 0)

(S 03 Nun -

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sat domescale S 01 S 0)

Nue 5.

NBM upscale (D 0)

Refusing outage Nuo 6.

ROM downscale Q) 01 Once/3 months Nas 7.

358 isscale S 0)

(S 0)

Nue 8.

358 detector not a startus (S 0)

S Nee postm 9.

IRA detector not a startus psw=m S

(6)

None

10. Sat desnicale (S 0)

S 0)

Nue

11. Nign water level a scram once/3 nonths Not asomaale Nuo

. sechste voinne (SDV) j

12. SDV high level trip Refueling Not acclicable None

• l bypassed cutage ilms a m ans humous 1.

$tses twasi he temocrature Refusing outage R W outage Mme 2.

Steamine hp Sow (D

Once/3 months Once/ day

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Steamfme bw yesswe U)

Once/3 mutns

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Steamtme he radates (1) (4)

Refusing oute Que/m 5.

Asocts low low ware lived n)

(bce/3 months once/dsy acic inneens 1.

Steamwe he now Ous/3 months once/3 momhs Nue

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Tuttee area hp temocratre Re'using outage Wmq outage M

3-18" 'eactor pesswe Once/3 montas once/3 montns None Amendment No. 77, 77 7

l 3.2/4,2-16

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QUAD-CITIES t

DPR*30

3. The control rod drive housing suppert i The correctness of the control rod syhem shall he in place during reactor power operation and when the reactor withdrawal sequence input to the coolant system is pressurized above RWM computer shall be ver:Aed aftee atmospheric pressure with fuelin the loading the sequence.

reactor vessel, unless all control rods a tk stan of contml rod w'th-are fully inserted and SperaAcanon drawal towards ersticality, the capabil.

3 3.A.i as met ity of the red worth minimizer to properly fulf.!! its funenon shall be Controt rod withdrawal sequences a.

shall be estabbshed so that man.

ver:Aed try the following checLs:

imum reactivity that esuid be

a. The RWM computer online dias.

added by dropout of any incre.

aosne ten s succusfuHy ment of any one control blade P"I*'**

would be cuch that the rod drop accident

b. Proper annunciation orthe se:ee.

design limit of 280 cal /c:n..in.not exceeded.

tion error of one out of secuence

b. Whenever the reactor as in the 5 artup' Hot Standbv or Run t.

The rod block function of the mode belo*w 20'a ra'ted thermal RWM shall be verded by with-power, the rod worth minimizer drawing the Jirst toJ as an out.

aball be operab!e A second opera.

of tequer:ee control rod no more set or qual: Sed techns:a! pe son

,than to the block point.

may be used as a subsntute for an inoperable rod worin minimater which faib after withdrawal of at least 12 control rocs to t. e fully withdrawn pos:tann. The rod worth min:rnizer rnay also be bypassed for low power physics testing to demonstrate the shut.

down marcin requirernents of Specif. cation 3.3.A sf a nuelcar angineer is present and verifies the asep by ste,1 rod movements of the test proerdure.

4 Control rods shall not lie withdrawn d.

Prior to control rod withdrawal for for startup or te ueling unless at least startup or during refueling. verify that r

two source raye channels have an at least two source range channels observed count rate equal to or greater have an observed count rate of at least than three counts per second and these three counts per second.

SRM's are fully inserted.

/

5. During operation with limiting con.

1 When a limiting control rod pattern trol rod pas: erns, as determined by the exists an instrument functional test of nuclear engineer, either:

the RBM shall be performed prior to drawal of th dnignard md(s) w

a. both RDM channels shall be

operable, and daily thereafter.

s. 1*wa :cr.us *1ssemr ism instem ww.ssvf etrasn valvee b.

control rod withdrawal shall be stut t ev var iri..e at t -'..e.

,..r it.i.,v bloeLed, or N. ine w*e.s w'in.'ra, iv-,"=."...c'"n*e* 'i ' e

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tw orire ar's %=fisat s r=.nst v.. 'V-i t.am d sr=tir ia i

witwar varit tvus arain vaivae wat t av. *!versistrate t-tOf

a. Close within 1s

• 4:"Pv's after racetfit oc a sanvias for contraa reets to 'icts, arul ti. Ctrn when the eram ucmt i s raget, 13/43*3 Amendment No. U, 77 u

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QUAD CITIES

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DPR-30

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b. the delayed neutron fraction chosen for the bounding reactivity curvu l

c. a beginning-of-life Drappler reactivity feedback

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d. scram times slower than the Technical Specification rod scram insertion rate (Section 3 3.c.1)

a. the maximum possible rod drop velocity of 3 11 fps

f. the desigrt accident and scram reactivity shape function, and

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g. the moderator. temperature at which criticality occurs In most cases the worth of insequsnee rods or red segments in conjunction l

with the a:tual values of the other important accident analysis pars =eters described above, would ::so likely resu'.0 in a peak fuel enthalpy sus-stantially less than 2D cal /g design limit.

j Should a control drop accident resuit in a peak fuel energy content of 280 cal /g. firwer than 660 (7 a *

7) thel rods are conservatively estimated to perforate. This would result in an offsite dose well below l

the guidelinevalue of 10 CFR 100. For 8 a I fuel. fewer than 850 rods are conservatively estima:ed se perforate. with nearly the same consequences sa for the 7 a 7 fuel case because of the rod power 4

differernes.

r The rod worth minimiaer provides automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted: i.e it limits operator deviations from planned withdrawal requences (reference SAR Section 7.9). It serves as a backup to procedural control of control red worth. In the event that the rod worth minimizer is out of service when required, a licensed opera:or or other qualified te:hnical employee can manually ful.H1 the control rod pattern cuformarwe

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knction of the rod worth minimiser. In this case. the normal procedi.ral controls are backed up by independent precedural controis to assure conformar.co.

The source range monitor (SRM) system performs no automatic safety systern knMon. i.e.,it has 4.

no scram funcnon. It does provide the operator with a visual indication of neutron level This is needed for knowledgeable and e'5eient reactor startup at low neutron levels. The consequences of reactivity accidents are functions of the initial neutron us.The res,uirement of at least 3 counts per a

second assures that any transient, should it occur. begins at or above the initial value of 104 of rated power used in the snaiyses of transients from cold conditions. One operable SRM channel would be adequate to morsitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal A minimurn of two operable SRM's is provided as an added conse:vatism.

S. The rod block monitor (R3M) is designed to automatica!!y prevent thei damage in the event of l

erroneous red withdrawal from locations of high power density during high power operation. Two channels are provided, and one of these may be bypassed from the console for maintenance and/or sessing. Tripping or one of the channels will block erroneous rod withdrawal soon enough to prevent thei damage.This sys:em hacks up the operator,who withdraws control rods according to a wettien sequence. The specified restrictions with one chsanel out of r,arwice conservatively assure snat (sel l

damage will not occur due to rod withdrawal errors when this condition esists. During reactor l

operation with certain limiting control, rod' patterns, the withdrawaf ot a designated single control rod could result in one or more fuel rods with MCPh's less than the MCP.C. fuel claddin6 integrity safety limit.During use of tsch pa::ert:-

it isjudged that testir:5 cf the RSM system to assure its oper:bility prior to withd :wal of such rods will usure that impreper withdrawal does not o::t:r. It is the responsikility of the Sue! car Engineer in identify these limiting patterns and the designated rc:s either when the patterns are initia*!y established or as they develop due to the occurrence ofincperabic control rods in other than limiting patterns.

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Amendment No. 57, 77 3 3/A 3-9 l

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