ML20126J629

Amends 66 & 60 to Licenses DPR-29 & DPR-30,respectively, Revising Tech Specs to Change Setpoints for Certain Sys Settings

ML20126J629
Person / Time
Site:	Quad Cities
Issue date:	04/16/1981
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20126J633	List: ML20126J629 ML20126J634 ML20126J639
References
NUDOCS 8105010517
Download: ML20126J629 (35)

Text

,

fR "*0 9'o 0

UNITED STATES I

+85 NUCLEAR REGULATORY COMMISSION l

m

-g,

,3 E WASHINGTON, D. C, 20555

' (,,/

[

s t

COMMONWEALTH EDISON COMPANY i

AND f

IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY l

i DOCKET NO. 50-254 t

00AD CITIES STATION UNIT NO. 1 i

AMENDMENT TO FACILITY OPERATING LICENSE l

Amendment No. 66 l

License No. DPR-29 l

1.

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

~i e

A. The applications for amendment by the Commonwaalth Edison Company (the licensee) dated November 7,1976, February 21, 1978, as supplemented

{

May 31,1978. April 25,1979 and February 14, 1979, comply with the standards and re as amended (the Act)quirements of the Atomic Energy Act of 1954,

, and the Commission's rules and regulations set forth in in CFR Chapter I;

\\

B. The facility will operate in conformity with the application.

l 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 I

and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; j

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; l

and i

E. The issuance of this amendment is in accordance with 10 CFR Part l

51 of the Commission's regulations and all applicable requirements have been satisfied.

i s

2.

Accordingly, the license is amended by changes to the Technical Specifi-i cations as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-29 is hereby amended to read as follows:

i l

l i

I 18105010 N.

l i

i B.

Technical Specifications j

I The Technical Specifications contained in Appendices A and B, as

~

revised through Amendment No. 66, are hereby incorporated in the l

license.. The licensee shall operate the facility in accordance i

with the Technical Specifications.

3.

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

FOR THE N LEAR REGULATORY COMMISSION f

-y [

Thomes A. Ippolito, Chief 0perating Reactors Branch #2 i

Livision of Licensing j

Attachment:

Changes to the Technical i

Specifications l

Date of Issuance:

April 16,1981 l

)

i t

1 I

i 1

l l

i l

t I

.,,,,.,_,_.._,_,....,_,,_,..,,._,,,.,,.._..1

I l

j i

ATTACHMENT TO LICENSE AMENDMENT NO. 66 FACILITY OPERATING LICENSE NO. OPR-29 DOCKET NO.

50-254 Revise the Appendix "A" Technical Specifications as follows:

Remove Add 1.0-2 1.0-2 1.1/ 2.1 -2a 1.1/2.1-2a

.f 1.1/2.1-3 1.1/ 2.1 -3 1.1/ 2.1 -8 1.1/ 2.1 -8 1.1/2.1-10 1.1/2.1-10 3.1/4.1 -3 3.1/ 4.1 -3 3.1/4.1-8 3.1/4.1-8 i

3.1/4.1-9 3.1/4.1 -9 3.1/4.1-10 3.1/4.1-10 i

3.2/4/2-6 3.2/4.2-6 l

3.2/4.2-11 3.2/4.2-11 3.2/4.2-12 3.2/4.2-12 3.5/4.5-8 3.5/4.5-8 i

3.5/4.5-16 3.5/4.5-16 i

E S

?

f P

p

=

h a

b b

t i

7, y

QUAD-CITirS

-DPR-29 l

k,,)-

j ify the minimum 1L1 Limiting Conditions for Operation (LCO);. The limising conditions ecteptable levels of system performance necessary to i

be safely 6

controlled.

i i strumenta.

Lirnising Safety Sprem Setting (LSSS) The limiting safety system se l

l tion which initiate the automatic protective action at a leve suc ts margin, with normal j

1.

exceeded. The region between the safety limit and these settings represen' h i opetation ly?ng below these setdnss.The margin has been established the instrumentation, the safety limits will never be execeded.

I i

d t cts of 1

K. Logic Sprem runctinn$1 Test A logie system functional test me

-l i

a logie circuit from sensor to activated device to ens d

i r

Modes of Operation. A reactor mode switch selects the properl lk ided:

f shutdown condition of the plant. Following are the modes and inter oc s prov 1.

i i

ved.

1. Shutdown. In this position, a reactor scram is initiated, power d

ior to petminive ibr ~

l and the reactor protection trip systems have been deenergized for 10 secon l

l

[.

manual reset.

b ithdrawn

2. Refuel. In this position, interlocks are established so that one col l

l when flux' amplifiers are set at the proper sensitivity eve an lves. main steam isolation l

i reactor. Abo. the trips froia the turbine control valves, turb ne stop va i

h ctor, all rods must i

valves, and condenser vacuum are bypassed. If the refeeling crane is over f

be fully inserted and none can be withdrawn.

tw

3. Startup/Ilot. tandby. Jn thh position,the reactor protection main steamhd j

vacuum and main steamline isolation valve closure, are bypassed, the lo i d ith IRM and isolation valve closure trip is hypassed, and the i

vie l

ti n l

l Run In this position the reactor system pressure is at or above l

i l ding the 159 high system is energized, with APRM protecdon and RMB inicilocLs in 4.

l flun seram).

l f

f ing its M. Operable A syuem or component shall be considered operab l

5 intended function in its required manner.

f i

in its N. Operating Operating means that a system or component is pe i l it and the end of l

required manner.

O. Operating Cycle. Interval between the end of one refueling outage l

i

[

the next subsequent refueling outage for the same un t.

!! and pressure Prhnery Conininment Integrity Primary containment integrity means 1

ifd suppression chamber are intact and all of the following conditions are P.

l

. C.,)

or

l.. All manual containment isn!ation valves on lines connecting to l

d l

centainment which are not required to be open during accident condidon f

Amendment No. 66 g,0 2 I

r

UUAD-CITIES OPP-29 The definitions used above for the ArnN scre.m trip apply. In the event of oper-(

ation with a maximu:n fraction limiting power dens ity (l'f!.PD) greater than the fraction of rated power (TRP), the scttsng shall be r:odified as follows:

rRP

?

S $ (.65Wp + 43) r.FLPD The definitions used above for the APRM ocra n trip apply.

The ' ratio of TRP to MFLPo shall be set 1

oqual to 1.0 unless the actual operating

.e value is icss than 1.0, in which case the actual operating value will be used.

This rnay also be performed by it.cr' easing the APRM gain by the inverse ratio, MFLPD/PRP, which accornplishes the same degree of pro-tection as rc'ducing the trip setting by FRP/MFLPD.

C. Reactor v water level scram set: ting j

shall i.144 inches above the top of the fuel

• at nor:nal operating condi-activ tions.

icvel ECCS initiation I

D. Reactor Icrw water inches /-0 inch) shall bc 84 inches (+4 above the top of the retive fuel

• at nor sal operating conditions.

'(

valve E. Turbine stop valve scram shall be s 10u l

i closure from full open.

l F. Turbine control valve fast closure scram shall Initiate upon actuation of the fasi closure sofe.

noid valves which, trip, the turbine control vatves.

Q. Main stesmline isofation valve closure scrarn shall be ::: 10% valve closure from full open.

, H. Main stearnline tow. pres,sure initiation of main l

steam,ine isolation' valve closure shall be l

j 2 82s' psig.

of active fuel is defined to

• Top

• 60 inches above vessel ::ero be 3 (See Bases 3 2)

\\

Q Arnendment No. 66 1.1/2.1 ?a 9

6 4

e O

en e 9

9 ii i O

QUAD-CITIES DPR-29 1, Turbine E11C control fluid low. pressure scram t

on loss of control oil pressure shall be set at greater than or equal to 900 psig.

J. Condenser low vacuum scram shall be set at l

2 n inches lig vacuum.

i i

f (l

t i

i i

I t

I l

i i

n..

l s.

[

I Amendment No. 66 1.1/ 2.1 -3 t

g A1)- C l'11 L'.i DPa-29 in present before the in tho MM scram trip setting would decrease the tsargThe Arm scram trip integrity safety limit. is reached.

for maneuvering during

, An i nc r e s s o ge t arns,

l by an analysis of marytne required to provide a reasonable ran fcJucing thts operating margtn would increase the freq fuel cleJ11nij

,f her.ai stresses.

l on reactor safety becavur of the resultir.g tat provtrio ade uate ma t

h ope r a tion.

reduces the posssbil-effect Thus, the Arm scr.w trip sett sng was sc1ceted'::ecau:;callows operat ing rearg an tha vhtch have an adverse I

fuel cladJang inteqrs'cy safety limit yet ity of unnecessary scrams.

the LUCR transient peak is not and hat ing power density (Mtt.ro)

The scram trip setticq ' rust be ad justed to ensure t Ircreased for' any coetnation of maxt:num f riction of lisaitThe scram settin ith the formula in Specification 2.1. A.1. when the MTLPD,is greater than the,inof prot the raciprocal, reactor core therral powcc.

by he APM ga Tne adjuatncnt ruy be accomplished by increasing t This provides the same degreehe initial APRM readings the trip cotting by FRP/MFLPD by raising t ld be received at l

o f FilP/MFi.PO.

closer to the trip settings such that a scram wouttings had been re-.

~

i the came point in a transient as if the tr p se i

duccd by F_RP_

MFLPD*

l Arm riux scram Trip setting (Refuel or Startup/ Mot Standby Mode)at low 2.

j,,-

Tor operation in the Startup rnode while the reactor is safety between the setpoint and thea s iloc kst e d of 15% of ratcd pcwer provides adequate thermal rnarginThe margin is ade v nd centent art press :re at rrro or icw lcc:.Jy '.n *:14 limit. 25:: of raced.

Ef fects of increasinct available.surict; abst;up ta nut muca colder u:.2:s Gt uare constr2ined with ocwor plant startup.

l red patterns Of all possibic source ninor, coll wt.:r rr :Nru:mm atwr.:.c., c l a r ient s are small, and contro h csinimizer.

unifonn by operating proceJure's barud up by the rod wort system, tu probable cause of significan:

the t.ost uniterm control rod withdrawal is d withdrawals does not 1

Because the flux di tri'eution associated wtth uniform rond beca of reactivity input, bv a siself:

l power rise.

is very slow.'

1=veiv hi,* io<at xs*

,0 cant percentaos of rated power, the rate of power riseIn an assumed uniform rod sichdtcwal apptc and la in near equiltbrien with the fission rate.

than 5*1 of rated power per minute, to the scram level, the rate of power risc, is no taore a scram before the power could excced I ;

che ; node switch is placed in the f f.,

the APM cyster. would be more than adequate to assureThe 15% ATM scram the safety limit.This switch occurs when reactor pressurri Run position.

2M Flux Scram Trip Setting rotection system log l 3,

The IM systeci consists of eight charsbers, four in each of 'the reactor p l between th i a 5-decade instrument which covers the range of pow i

h channels. The IM is I

covered by the SM and the APM.

f For example one-half a decode in size.

in each range of the IM.

The IM scram trip setting of 120 divisions is activeif the instrumen 120 divisions for that ranger range. j instrument were on Range 5, the increase in power level', the scram trap se' t

likewise, if the l

the

{ *j Thus, as the IM to ranc.ed up to accommodate ting is also ranged up.

due to cent sources of reactivity chango during the power increase are

)

i t the The enest a tqntf tcantIn order to ensure that the IM provides adequate protection ag

' i s analyzed. Thts analys ;

single rod withJrawal error, a range of rod withdrawal accidents wa rod witt!rawl.

involves an The most severe case included starting the accident at vartous power icvels. initial con IM system is not yet or he IM, channel C105c! f ecale.

Additional connervatism was taken in this analys te by asru:ning that tof tats -nalys ts fuc1 cladding the reector to scr..:

trnoved. Ths rhsult:

to 1/. of rated power, thus ma tnta tnAng MCPR above theIM pro I

I wt'.% :r wr. t e t 5

{

4. the

• • r"r t ua n teJ Dased on the above analysts, the in sequence arid 44 local contrul ro.1 wittv! raw.61 errors and continuous wAth inteyt tty naf ety 11t".it.

l l

(')

provides backup psuteetton for the APM.

L ll H

1 1.1/2.1-8 bendment No. 66 l

l

4

.e i

QUAD-CillRS l

{"

DPR-29 Volve Closure G. Henr:.,r Conlant low Prewure Initiates Mahi Steam isntr.tlon l

825 psig was provided to give protection against fast reactor de fh am feature which The inw pressure isolation at

. suritation and the resrEnyapid cooldown of the vessel. Advantage was taken o t e scr f

j oscurs in the Run mode when the main steamline isolation valies ele closed to li h l

shutdown so that opervion at pressures lower than th l

l j

constitue an unsafe condition.

l-f H. Main Steamline Isolation to Yahe Clusure Scrant The low pressure isolation of the' main mamlines at.02) psig was pro f

l rapid reactor depressurization and the resuhing rapid cooldown of the i li l es are closed to l

the scram feature in the Run mode which occurs when the main steamli l

provide for reactor shutdown so that high power operation at low providing protection for the fuel cladding integrity safety lir. tit. Opera t

{

lower than 825 p<ig requires that the reactor mode swhch he in the Star-l Sux scrams.

of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutl

. Thus, the combination of main steamline low. pressure isolatio i

h Run mode of the fuel cladding inie;rity safety limit. In addition, the isolation valve closure scra isolation valve j

anticipaies the pressure and nux transients which occur during normal or inadv no increase in neutron l

clnsure. With the scrams set at 10% valve closure in the Run mode, there is-6 t

flux.

]

Turbine EHC Control Fluid Low Preuure Scram He turbine EHC control system operates using high pre:sure oil.There a

!.~

This fast j

system where a lost of oil pressure could result in a fast closure of the tur i

,j closure of the turbine control vahes is not protected by the turbine control vah; failure of the oil system would not resuk in the fast closure solenoid vahe control valve fast closure. the core would he protected by the APRM and high l

fthe l

}!awever, to provide the same margins as provided for the genesator inad rejectij turbine control valves, a te' ram has been added to the reactor protection system w i

j control oil preuure to the turbine control system. This i il s

i f 900 psig is set to that resulting from the turbine control valve fast closure seram.The scram setpo nt o ii i the number of j

high enough to provide the necessary anticipatory function and low enough to m l l ill

[

spurious scrams. Not.nal operating pressure for this system is 1250 psi l

l sure l

not start until the fluid pressure is 600 psig. Therefore, the scram occurs well beI begins.

i l

Condenser Low Yacuum Scram Loss of condenser vacuum occurs when the condenser can no longer L

l hich condenser vacuum initiates a closure nf the turbine stof i

transient. neutron flux rise, and an increase in surrec

)

dding safety limit.from.

j turbine stop valve closure scram function alonc is adequate to prevent the cla being execeded in the event of a turbine trip transient with bypass closure.

-i The condemer low vacuum scram is anticipatory to t

{-)

l t 7. inch g

l

/' ~

rnode al.217 nch lig vacuum stop valve closure occurs at 20 inch Hg vacuum l

i 1

lig vacuum.

l Arnendrnent' fio. 66 1.1/2.1-10 l

I

..:nn u !

QU AD-CITIES DPR-29 t.

(- /-

d h

ram withoet impairment of the scram times or amount ofinscrtion of the control ro

'h h

actor dowe,. hile sumcient tion in which a scram would be d'h it vo!ume remains to accommodate the discharged water and preclu e t e s ua required but not be able to perform its function adequately.

f d

r vacuum Loss orcondenser vacuum occurs when the condenser can no longer hen i

h heat input to the initiates a closure of the turbine stop valves and ttirbine bypass valves, which condenser. Closue of the turbine stop and bypass valves c if this occurs, a reactor scram f

occurs on turbine stop vahe closure. The turbine stop valve dosure scram uncbine trip transien the cladding safety limit from being enceded in the event of a turThe cond l

s at n inches Ifg l

scram bercre the stop vahes are closed, thus the resukin6 transient is less soc 7 inches Hg vacuum.

vacuum, stop valve closure occurs at 20 inches Hg vacuum, and bypass closure at dioactivity High radiation levels in the main steamline tunnel above that due to G ii l

are an indication of feaking fuel. A scram is initiated when to prevent the site environs is prevented ii excessive turbine contamination. Discharge of excessiv ided the limit

+

specified in Specification 3.3 is exceeded.

10% cksed from The main steamline isolation valve closure scram is l

lose. By scramming at this settinC. the resuhant transient is in>i niGeant, 6

iate to the i

{j A reactor mode switch is provided which actuales or hypasses the various sc ode switch is in the h

particular plant eperating status (reference SAR Section i

li e isoladon 1

d i yahc closure scram arc bypassed.This bypass has been provided Tot nexibility u to be made to the turbine condenser. While this bypass is in c i his mode.

i line If the reactor were brought to a hot standby condition for repairs to the tu d f r don isolation valves would be closed. No hypothesized single failure or sir.gk o can result in an unreviewed radiological release.

i tol The manual scram function is active in all modes, thus providing for a m rods during all modes of reactor operation.

r d

The IRM system provides protection against excessive power !cvels a i

(SRM) sy stem intermediate power ranges (reference SAR Sections f

i ns H Standby modn. In Oeference SAR Section 7.4.3.2). Thus the IRM is required in the Refuel and Start addition. protection is provided in this range by the APRM 15% scram as 52 Thus, the 2.l. In the power range.ihe APRM system providn h IRM's provide adequate coverage in the startup and intermediase range.

h highfevel The high reactor pressure, high.drywell pressure, reactor low water level, an i

scrams are required for the Startup/ Hot Standby and Run modes of plant operat on.

to be operational for these modes of reactor operation.

d to start The turbine condenser low vacuuan scram is required only during power o 3

up the unit.

Amendment No. 66 3.U4.1-3 t

f i

QL' AI)-Cl l lF.S '

DPR-29 i

i i

f:'}

IABLE 3.11 REACTOR PROTECil0!! SYSTEl.1(SCRAl.t)l!!STRUMENT Ail 0H REQUIREl. EllTS RE

. Itinimum flumber cl Op rat?e or Tilpped Instrument Ch:nn:Is per Actlan:t1

' Trip Spte #1 Tilp Functica kip level Setting A

1 Mode switch in shutdown A

1 Manual scram lRM -

s120/125 of full sca'4 A

3' High flux 3

Inoperative APRMt33 A

2 High flux (15% scram)

Specification 2.1.A.2 A

2 Inoperative A

2 High water Icvelin scram s50 gallons discharge volume # -

t 2

High reactor pressure s1060 psig A

A 2

High drywell pressuret5) s2 psig bI Reactor few water level 28 inches

• A 2

2 Turbine condenser low 2211nches Hg vacuum '

A l

m vacuum 2

Main steamline high s1 X normal full power A

radiation (121 background 4

Main steamline isolation s10% vane closure A

valve c!csure*

I

. i, l

r Amendment No. 66 3.1/ 4.1-8

QU A D-Cl l'IES DPR-29

,.n...

_)

TABLE 3.12 REACTOR PROTECil0!! STSTEM (SCRAM) tilSTRUl1Elli ATIO!! REQUIREl1ElliS STARI MODE l.'inimum fiumter of Op?ratte or j

idpp:d Instrument Chann:Is per Trip Sptem Trip Functica idp level Setting Actico43 l

d3 A

1 Mode saitch in shutdosn A

I Manual scram 1RM 3

High flux s120/125 of full scate A

A 3

beperative APRM'33 2

High llux (15% scram)

Specification 2.1. A.2 A

A 2

Inoperative 2

High reactor pressure s1060 psig A

2 High drywell pressure (53 s2 psig A

2 Reactor !cw water fevel 28 inches (s)

A

(.,,

1 3

2 High nater fevelin scram discharge volume (43 550 ga!fons A

2 Turbine condenser low a n inches Hg vacuum A

\\

vacuum )

d 2

Main steamline high s1 X normal full povier A

tadiation )

background a2 4

Main steamline isolation

$10% valve closure A

vafve closwe03 S

Amendment No. 66 3,g f 4,g.9

1 l

i Ql!AD-CITIES i

DPR-29 i

/

l

~

TABLE 3.14

- ~

REACTOR PROTECIl0N SVSTEM(SCRAM)INSTRUMENTAil0H REQUIREMEl:

i k!!nimum Number of Operatte or Trippcd ' :'re. ment -

l Cfiannefs per Trip level !ctting Action'D l

ul Trip Function t

Trip sptorm A

l 1

Mode switch in shutdown A

i 1

Manual scram i

AFT #

2 High flux (Coiv biased)

Speci5 cation 2.1.A.1 A or 8 A or B l

2 looperative A or B 2

DownscateUD 23/125 of full scale

_f 2

High-reactor pressure

<1060 psig A

2 High :i: /well pressure s2 psig A

2-Reactor low water tevel 28 inches (s)

A, 2

H;gh. water level in scram discharge volume sSOgaltons A

2 Turbine condenser low 221 inches Hg vacuum A or C l

vacuum Main steamline high s7 X normat full

~

2 radiationun power background A or C f

4 Main steamline isolation Valve CICSure )

510% vatve closure A or C 5

2 Turbine control valve fast 2.40% turbine / generator A or C closure #

toad mismatchno t

2 Turbine stop valve s10% valve closure A or C closurecn 2

Turbine EHC control fluid 2900 psig A or C lo.: pressure *

)

t l

l 1

1 Amendment No. 66 3.1/41-10 j

i

'I

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

4

- QU AD CITIF.S I

j f measuring stran) Dow and aho limiting the

]

li e break sccident. In addith.n to monitoring s

\\ emori tubes are proviled in the main steamlines as a means o l

Group i holation valve >. The pdmary funct l

o mau inventory from the sessel during a steam n l Group i valves ase closed. For thel imtrumentation i> provided which c..utes a trip.ofimta

[

r w

l is not ll this trip setting of 140%

lve dosure. limhs the mass invento d

accident, main steamfine breal nuside the rywe.

l li wi b the flow limbers and main steam ne va uncovered, fuel temperatures remain less than 1500'1 l

d 14,2,3.10).

in steamline tunnel to detecs leaks il ded cause closure oPGroup ! isolation va h

Temperature. monitoring instrumentation is provided in t e ma i

i of 5 io 10 3pm; thus it is capabic of eovering l Trips are provided on this instrumentation and when acc d for f 10 Cl R 100 -

l d

setting of 200' F is low encugh to detect leaks of the or erti ity L l i

specitum of breals. For large breats, it is asmall breaks with i

i have been provided to detect l

are exceeded.

i liigh. radiation rnonhors in the main steamline tunne d t d and main steamline isolation valve dosure, l

Section e no: exceeded for this a(cident (referl cstablished setting of 7 times normal backgroun i

release is limited so that 10 CFR 100 guidel nes ar line prenure Jrops hdow 825 psig. A t

)

l de-Pressure instrumentation is provided which trips when ma n steam 12.2.1.7 ).

i l

e sure rept nur j

i vided primarily to provide protection.:ga nst a pr s 1

{

this instrumentation results in dosure of Group bypass valve to open. With the trip sl i

l i

this trip functionis bypassed.This function s pro matrimetion which would cause the control and/orloss is i

' (,

l j

l are no fission products available for release ot e break in their j

d ternperature instrumentation are provided to l

11.2.3 ).

line isntation valves. thus all semorsl

.The RCIC and the 11PCI high flow an l i respective piping. Tripping of this instrumentation resu ts n a f

i gfe. failure critena. Ihe uip i

Tripping logic for this functiun is the same as that for the ma n >

l h

required to be operable or un a tripped condition to m t of.two taken twice logie circuit. Unli ef k

3t:

ip l

stem associated with each func l

is whhin limits.

The instrumentation which initiates ECCS action is artanged in a ol i

the reactor scram circuits, however, there is one tr p sy tic blowdown and high pressure coolans n ec systems in the reactor protection system.The single failure cr bl the.ystem which h activate coohng fanedons are trovided. e.g sprays and automa l

t specilication requires that if a trip system becomes m

3 5 govern.This specitication p d

l

.iut oforrvice specifications orSpecification.

5 itol rod withJiawal so that MCPR l t

i wh!, ropect to the singti faihere criter a even i

The comrol rod block functions are providcJ to prevent excess ve coniI go below the MCPR Fuel Claddipa IntegrThe trip !opic fo el requirements amure suAient l

i-h four SRM's will resuh in a rod block. The minimum f

The minimum imtrument chaj l

instrumentation to assure that the single failure criteria are met.

i This time period is only-3% of the operatin ll i

i I

I preventing an inadvertent control rod withdrawal.

l l

i L

Amendment No. 66 3.2 M.2 6 i

)

QUAD-CITIES DPR-29 TASLE 3.21 6

INSTRllMEHT Ail 0H THAT INiilATES PRIMARY CONTAlHMENT ISOL Ail 0il FllllCil0llS Mr.fmum INmtw of Opratfe or Tr4;d in:trument Chcrxisul Instruments Trip level Setting Actica(2) i 4

Reactor low waterm

>144 inches above top of A

f active fuel' I

4 Reactor few tow water 284 inches above top of A

active fuel" m

A m

gg p3;g 4

High drywell pressure 16 H!gh !!ow main steamline$

s14cg of rated steam f!aw B

\\

i 16 High temperature maili s200*f 8,

steamline tunnel 4

High radiation main s1 x normal rated power B

steamline tunnef5F background 4

tow main steam pressure

• 2825psig 8_

l' l

f 4

High tiow RCIC steamline 5300% of rated steam flow C

{;

f 16 RCIC turbine area high s200* F C

temperature t

4 High flow HPCI steamline

$300% of rated steam flow D

[6 HPCI area high temperature

$200*f D

htes 1 whenever primary containment integrity is required. there shall be (no operable or tripped systems for each function, except for tsw pressure mam steam!me which only need be avalfahre in the Run position.

7.

Aten. If the riist column cannot be inet for one of the trip systems, that trip system shall be tripped.

If the first column cannot be met for both trip systems. the appropriate actions listed below shall be taken:

l A, Initiate an orderly shutdown and have the reactst in Cold Shutdown condition in 24 hourt t

l B. Initiate an orderly load reduction and have reactor in Hot Standby within I t.ours.

[. Close isolation ialves m 8CIC systes f

D. Close isolation valves m hPCI subsysten

)

3 4eed not be operable when pnmary containment integrity is not required.

a 4 The moraten inp signalis bypassed when the mode switch is in Refuel or Startup/ Hot Shutdwn.

1 This estrumentation also isolates the control room ventitation systes f

6.

This signal also automatcally closes the mechanical vacuum pump discharge fine isolation valves.

I

• Top of active fuel is defined as 360" above vessel zero for all water levels used in the j

LOCA analysis hoe Bases 3.2).

f l

Amendment No. 66

-i 3.2/4.2-11

~

?

=

5

~

1 QUAD-CITIF.S DPR-29 TABLE 3.2 2 I!!STRuljEliTAil0tl THAT llllilATES OR C0tliROLS THE CORE A!!D C0llIAll!!'Elli C00Lil:G SYSTEl.iS I

Minimum Rurr.bar.

of Op:r;tJe or Trip;;J l nim:nt Channic.<

trip Function Trip level Setting Remarks,

4 ReJctor !0w low 284 inches ( + 4 inches /-0 inch)

1. In conjunction with farcactor pressure.

water level above top of active 'uel*

initiates core spray and LPCI.

2. In conjunction with high.drywell pressure 120 second time defay and low. pressure l

i core cooling interlock initiates auto

~!

blowdown.

3. Initiates HPCI and RCIC.

4.

Initiates starting of diesel generators.

f 4*

High.drywell s2 psig

1. Initiates core spray, LPCI, HPCl, and f

l

SGTS, pressure,m m

l

2. In conjunction with low low water !evel, 120-second time delay, and tow. pressure i

I core coofing interfock initiates auto blowdown.

3. Initiates starting of diesel generators.

f 4.

Initiates isolation of control roorn j

vcatilation.

f 2

Reactor low 300 psigsps350 psig

1. Permissive for opening core spray and LPCI l

admission valves.

l pressure

2. In conjunction with low low reactor water j

level initiates core spray and LPCI.

l

, Prevents inadvertent operation of containment f

t Containment spray spray during accident conditions.

j interlock i

25 2/3 core height 22/3 core height 4m containment 0.5 psigspsl.5 psig high pressure l

2 Timer auto s120 seconds in conjunction with low low reactor water j

level, high-drywell pressure, and low pressure blowdown core cooling interlock initiates auto blow-down.

t 4

low. pressure core ice psigspsisopsig Defers APR actuation pending confirmation of l

l Iow. pressure core cooling system operation.

l cooling pump dis.

charge pressure

.1 2

(Jndervoltage on N/A

1. Initiates starting of diesel generators.

l

2. Permissive for starting ECCS pumps.

emergency buses

3. Removes nonessential loads from buses.

l

' Top of setive fuel ls defited as 360" above vessel aero for all water levels used in the LOCA anelysis.

i Amendment No. 66 3.2/4.2-12

^

1 I

I QUAD-CITIES DPR j 1

I 2, The discharge pipe pressure for the 2.

rollcuing any pericd whore,the t'XI

(

rede of the rJR or core npray frCS systems in Specification 3.5.G.1 shall h3V' ' " "' f S*'VIC 3"d d' l"*d I

l be maintained at greater than 40 ps'S for raintenance, the discharge piping l

and less than 74 ps.ig, if pressure.'

in of the inopirable system shall te e

vented from the high point prior to any of these' systems is less than 40 th '

t"'"

f th* SY*to" to 56'VIC

I psig or greater than 74 psig, this con-dition shall be alarmed in the control room and immediate corrective action 3'

Dcnever the llPCI or RCIC system-

-l taken. If the discharge pipe pressure is is h.ned up to take sucn,on from the l

not within these limits in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> torus, the discharge piping of the aner the occurrence, an orderly shut.

IIPCI and RCIC shall be vented from l

down shall be initiated, and the reac-the high point of the system and water i

tor shall be in a cold shutdown condi-O w observed on a monthly basis.

tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiation.

_ l 4

The pressure switches which ronitor the dischstge lines aid the discharge 1

of tho fill system p p to en:Nre that l

r they are full shall be functionally tested every rnonth and calibrated every i

3 nonths. The pressure switches shall be set to alarm at a decreasing pressure of 240 poig and an increasing pressure g

of 6 90 psig, g

j

f r,i H.

Condensare Pump Hoom Flood Protection H. Condensate Pump Room Flood Protection j

1. The systems installed to prevent or

1. The following surveillance require-l mitigate the consequences of flooding -

ments shall be observed to assure that j

of the condensate pump room shall be the condensate pump room flood pro-operable prior to startup of the tection is operable.

l reactor.

The piping and electrical penetra-l a.

2. "Ihe condenser pit water level switches tions and bulkhead doors for the j

shall ' rip the condenser circulating vaults containing the RHR service l

water pumps and alarm in the control water pumps and diesel generator l

room if water level in the condenser cooling pum.ps shall be checked pit exceeds a level of 5 feet above the during each operating cycle by pit floor. If a failure occurs in one of pressurizing to 15 i 2 psig and these trip and alarm circuits, the failed checking for leaks using a soap circuit shall be immediately placed in bubble solution. The criteria for i

a trip condition and reactor operation acceptance shall be no visible leak-l shall be permissible for the following age through the soap bubble j

7 days unless the circuit is sooner solution.

i made operable.

b. The floor drains from the vaults

3. If Specification 3.5.H.I and 2 cannot shall be checked during each oper-be met, reactor startup shall not com-ating cycle by removing the end

[

mence or if operating, an orderly shut-cap and assuring that water can be i

down shall be initiated and the reactor run through the drain lines.

-l shall be in a cold shutdown condition D " * tor cooh.** **' pu mp and

1. ~

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

diesel genera ng water 4

[

pump bed plate drains shall be i

checked during each operating j

Amendment No. 66 3.5/ 4.5-8 j

QUAD-CITIES G

DPR-29 e

( =a,)

4.5 SURVEILL\\NCE REQUIREMENTS BASES The testing interval for the core and containment cooling systems is based on a quantitative rel judgment, and practicality. The core cooling systems have not been designed to be fully tes For example, the core spray final admission valves do not open until reactor pressure has fallen during operation, even if high drydell pressure were simulated, the final valves would not open.

r HPCI, automatic initiation during power operation would result in pumping cold water into the react which is not desirable.

The systems can be automatically actuated during a refueling outage and this will be done. To availability of the individual components of the core and containment cooling systems, the com make.op the system, i e., instrumentation, pumps, valve operators, etc., are tested more freq instrumentation is functionally tested each month. Likewise the pumps and motor-operated valves a each month to assure their operability. The combination of a yearly simulated automatic actuation tes monthly tests of the pumps and valve operators is deemed to be adequate testing of these systems.

With components or subsystems out of service, overall core and containment cooling reliability demonstrating the operability of the remaining cooling equipment. The degree of operability to b depends on the nature of the reason for the out-of service equipment. For routine out of-preventative maintenance, etc., the pump and valve operability checks will be performed to demo operability of the remaining components. However,if a failure, design deSciency, etc., cause period, then the demonstration of operability should be thorough enough to assure that a sim not exist on the remaining components. For example,if an out-of service period caused by failure of a pump to deliver rated capacity due to a design denciency, the other pumps of this type might be subjected to a n, (s

test in addition to the operability checks.

The verification of the main steam relief valve operability during manual actuation surveillance testing must b It has been found that a independent of temperatures indicated by thermocouples downstream of the relief valves.

tempersture lacrease may result with the valve still closed. This is due to steam being vented throug By first opening a turbine bypass valve, and then observing its closure response during re during the surveillance. test.

valve actuation, positive verification can be made for the relief valve opening and passing steam flow. C the turbine control valves during relief valve manual actuation would likewise serve as an adequate verificatio relief valve opening. This test method may be performed over a wide range of reactor pressures greater than Valve operation below 150 psig is limited, y the spring tension exhibited by the relief valves.

b The surveillance requirements to ensure that the discharge piping of the core spray, LPCI mode of the RH and RCIC systems is filled provides for a visual observation that water Hows from a high point vent. This that the line is in a full condition. Instrumentatic.n has he:n previ&d to mnitor the prescure of 9;ater in the discharge piping between the monthis/ intervals at unien the lines are vented and al:rm the control reem if the preFcure is inadequate.

his irstru c.tation will be calibrated on the same frequenc/ as the cafety system instrumentation and the alarm system tested ncnthly. This testing ensures that, dur ing the interval between the

onthly venting checks, the status of the diccharge piping is trnitored on a continuous basis.

An alarm point of 40 psig for the low pressure of the fill system has been chosen because, due to elevations of piping within the plant, 39 psig is required to keep the lines full. The shutoff head l

of the fill system pumos is less than 90 psig and therefore vill not defeat the icw-2 pressure cooling p.:mp diccharge press interlock 100 psig as shown in Table 3.2-2.

A margin of 10 psig is prcvided by the high pressure alarm paint of 90 psig.

J Y

Amendment No. 66 I

3.5/4.5-16

i pA Cf 0 S}

[0,j -

+1 UNITED STATES NUCLEAR REGULATORY COMMISSION j

~l; ( '

E WASHINGTON, D. C. 20555 t

o o

9 g

COMMONWEALTH EDISON COMPANY AND IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY

[

t DOCKET NO. 50-265 OVAD CITIES UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 60 l

License No. DPR-30 t

1.

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

)

A.

The applications for amendment by the Commonwealth Edison Company i

(the licensee) dated November 7,1976, February 21, 1978, as supplemented i

May 31,1978, April 25,1979 and February 14, 1979, comply with the standards and re j

as amended (the Act)quirements of the Atomic Energy Act of 1954,

, and the Commission's rules and regulations

~

set forth in 10 CFR Chapter I; j

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 f

by this amendment can be conducted without endangering the health i

and safety of the public, and (ii) that such activities will be i

conducted in compliance with the Commission's regulations-D.

The issuance of this amendment will not be inimical to the f

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

)

~

i E.

The issuance of this amendment is in accordance with 10 CFR Part i

51 of the Commission's regulations and all applicable requirements i

have been satisfied.

l i

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 3.B of Facility License No. DPR-30 is hereby amended to read as follows:

i

~ !

l I

. ~ ~,,.

,,,-,.,,.,,,,.~.e-

.r.,,

..v..

.v,-,-v.-._,,,

ym -

2 3.B Technical Specifications The Technical Specifications contained in Appendices A and B,

-as revised through Amendment No. 60, are hereby incorporated i

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

.3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

']

er,-e Thomas A. Ippo it. Chief Operating Reactors Branch #2 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: April 16,-1981 1

I i

6 r

a i

t I

i f

I t

1

-, ~ -

E k

ATTACHMENT TO LICENSE AMENDMENT NO. 60 l

i FACILITY OPERATING LICENSE NO. DPR-30 1

DOCKET N0. 50-265 Revise the Appendix "A" Technical Specifications as follows:

Remove.

Add

'1.0-2 1.0-2 I

1.1/ 2.1 -2a l

1.1/2.1-3 1.1/ 2.1 -3 1.1/2.1 1.1/ 2.1 -8 l

1.1/2.1-10 1.1/2.1-10 i

3.1/4.1 -3 3.1/4.1 -3 3.1/4.1-8 3.1/4.1 -8

-l 3.1/4.1 -9 3.1/ 4.1 -9 i

3.1/4.1-10 3.1/4.1-10 3.2/4.2-6' 3.2/4.2-6 3.2/4.2-11 3.2/4.2-11 l

3.2/4.2-12 3.2/4.2-12 3.5/4.5-7 3.5/4.5-7 3.5/4.5-8 3.5/4.5-8 r

3.5/4.5-15 3.5/4.5-15 i

l i

i I

i I

I l

l l

l

. _ ~.

r QUAD-CITil3 oen-3o j

O-

-l ditions for operation specify the minimum i ii i n of the facility.

. It Umiting Conditions for Operation (LCO),.The l m t ng con neceptable !cvels of system performance necessary b

fl l

l contro!!cd.

i t

I. ~ Lhaiting Safety Sprem Setting (LSSS) The limiting safety system j

he safety limits will not he tion which initiate the automatic protective action at a level such that t argin, with normal l

exceeded. The region between the safety limit and these settings represents md th(

operation lying bdow these settings. The margin has been cuablished s l

the instrumentation, the safety limits will never be execeded.

_j 6

l l

D K. !Agic System Functional Test. A logic system functional tes ble per design intent.

-l a logic circuit from sensor to activated device to ensu

?

i i

Modes of Operation A reactor mode switch selects the proper inter!oe I

ided-i shutdown condition of the plant. Following are the modes and interloci,s prov L

d l

1. Shutdown In this position, a reactor scram is initiated. power to th

' Wye (or '

[

and the reactor protection trip systems have been deenergized for 10 set l

f l

-(,

2. Refuel In this position. interlocks are established so that one con manual reset.

l h

f when Dux ampliners are set at the proper sensitivity leve i

isolation l1 rods muu valves, and condenser vacuum are bypassed. If the sefeeling crane is over thl be fully inserted and none can be withdrawn.

f

3. Startup/ Hot Standby In this position,the reactor protection scr in weambos j

vacuum and main steamline isolation valve closure, are bypassed the low preu IRM and l

isolation valve closure trip is bypassed.and the l

i l

f

4. Run In this position the reactor system pressure is at or above p2$ p di h 15% high system is energized, with APRM protection and RMB interlocks in se[

i nux scram).

f M. Operable. A syuem or component shall be considered operal i

intended function in its required manner.

N. Operating Operating means that a system or component is pe l

j required snanner.

f l

Operating Cycle Interval between the end of one refueling outaj O..the next subsequent refueling outage for the same unit.

i l

P. ' Primary Containment Integrity Primary containment integrity mean

_l suppression chamber are intact and all of the following conditions are sa j

-[

1. All manual containment isolation valves on lines connecting to the rea j

eentainment which are not required to be open during accident conditionsI i

r Amendment No. 60 1.0-2 l

l

-._ ~.-. - -..

i l

QUAD-CITNS i

DPR-30 l

(.

1

. The definitions used above for the. APRM f

ocrean trip apply. In the event of oper7 stion with a maxiraum fraction limiting i

power density (MFLPD) greater than the l

frnetion of rated power (FRP), tha r.ctr ing trill be modified as follows:

PRP s 6 (.65Wo + 43)

MFLPD I

' he definitions used above for the APR.4 i

T cerem trip apply.

J t

The ratio of FRP to MTLPD shall be set equal to 1.0 unless the actual operating l

value is less than 1.0, in which case l

the actual operating value will be used.

i C. Reactor icw water leval scram setting chall be 144 inches above the top of the i

activo fuel

• at normal operating condi-I tions.

s D. Reactor low water level ECCS initiation shall bo 04 inches (44 inches /-0 inch) i abovo the top of the active fuel

• at i

normel oporating conditions.

j l

E. Turbine stop valve scram shall be s 10% valve closure from full op:n.

[

F. Turbine control valve fast closure scram shall f

Initiate upon actuation of the fast closure sole-noid valves which trip the turbine centrol l

valves.

G. Main steamline isolation valve closure scram shall be s 10% valve closure frorn full open.

i H. Main steamline low. pressure initiation ormain j

steamline isolation valve closure shall be l

2 825 psig.

g l

i

• Top 60 inches abcVe vessel zero I

of active fuel is defined to i

be 3 (See Bases 3 2)

?

t I

Amendment No.-60 1.1/2.1-2a l-t I

I

- -., +

,,.,,.._n,

._m.,

y._,_...

.,,,,,.._w..,...

,,,.....,w...,.,

,,.,,,.,,,,.m.

.......m.,m._,r

,%.,,__,....,,,,,y.I

0 8

QUAD-CITIES DPR-30 1, Turbine EllC control fluid low-pressure scram on loss of control oil pressure shall be set at greater than or equal to 900 psig.

J. Condenser low vacuum scram shall be set at l

2 21 inches !!g vacuum.

i

. l p

1 i

Amendment No. 60 1.1/ 2.1 -3

e i

OUAD-CITIUS l

DPR.

• ~

l An increase in the APM scram trip setting would decrease the margin present befor is reached.

by an analysis of margins required to provide a ransonable ran!

fuel cladding integrity safety limit

'h".

t 1

operation.

hermal stresses.

which have an adverse ef fect on reactor safety because of the resulting t i for the f

Thus,. the APM sera:n trip setting was selected bJesuse it provid f

I ity of unnecessary screras.

Tha ocrem trip setting nust be adjusted to 'oncure that the IER transient peak is not.

Increased for any c 3mbination of maximwn fraction of li=iting peser de

[

reactor core thermal power.

)

{

in Specification 2.1. A.1, when the MTLPD is groctor than the fraction of rated power (:

t

. ArRH Plux Scrc n Trip Sctting (Refuel or Startup/det Stendby Mode)

-l i

2.

the APM scram detting l

For operation in the Startup mcde while the reactor is at low pressure,

+

of 15% of rated power g rovides adequate thermal nergin batween the t

limit, 25% of rated.

Ef fects of increacing r.rcsoure at zoro or low void content are j

minor, cold water from ocurces available during ecsttep is r.ot much colder than that aircad with peer plant startup.

oyatem, tenparature, coef ficiones sre small, and'E6h661 rod patterns are constrained to be Of all possible soerces uniform by operating procedures backed up by the red worth cinimirer. uniform contrj of reactivity input,Decause the flux distribution associated with uniform red withdrawals does not involve high local peaks, and because several rods must be moved to change power by a signifi power rise.

is very slow, cenere11y,,the heat flux cant percentage of rated power, the rate of pcwer riseIn an assumed uniform rod withdrcwal approach to the scram level, the rate of power rise. is no core than 5% of rated power per minute, and

'j is in near equilibrium with the fission rate.

the APRM system would be more than adoquat's to ascure e scram before the power could exceed

[

The 15% AFM ocrea remains cctivd until the mode switch is placed in the l

l c.,i the safety limit.

(d This switch occurs when reactor pi'ennute is greater than 825 psi 9 i

Run position.

3.

IRM Flux Scres Trip Setting The IM cystem consists of eight chambers, fcur in occh of the roector protection system Icgic The Irx le a' 5-decede instru.ent which cevors the rango of power icvel beween that j

each being channels.

The 5 decades are brokon down into 10 ranges, covered by the SPM and the APM.

one-half a decade in size.

For example, i

The IRM scram trip setting of 120 divisions is active in each' range of tl range.

if the instrument were on Range 5, the screm would be 120 divisions on that l

Thus, as the IRM is ranged up to accorraodate the increase in power level, the scram trip set-

likewise, i

ting is also ranged up, i

The most significant sourecs of reactivity chango durirn the pow'er increase are due to contf In order to ensure that the IM provides adequate protection against theThis analysis single rod withdraval error, a rarse of rod withdrawal accidents was analyzed.

red withdrawl.

The most severe case involves an included starting the accident at various power levels.

j initial condition in which the reactor is just suberitical and the IRM system is not yet on t

scale.

[

Additional conservatism was taken in t'his analysis by asouming that the IRM channelclos f

The results of this analysis show that the re, actor is scra=ed '

the withdrawn rod is bypassed.

ding and peak pcreer limited to 1% of rated power, thus maintaining MCPR abf

.i I

local control rod withdrawal errors and continuous withdrawal of control rods in acquence integrity safety limit.

i

~

I j

provides backup protection for the APRM.

i

~

()

1.1/2.1-e'

~ Amendment No. 60' -

4 8

L.

i t

i i

i QU AD. CITIES 1

DPR-30.

i

~ G Itercr Coulant Low Pressure initiates Mafn' Steam Isnir.tlon Yahe l

l

- surir.ation and the reseD'ng capid cooldown of f

hich j

i i

l d to provide for reactor l

occurs in the Run mode whers the main stearnline isolation ve.l'es d

li fety limit shuidown so that opervion at pressures lower than th ily l

l

-l z

constitue an unsafe condition.

{

l Main Sicam!ine Isolation to Yabe Clusure Senim 823 psig was provided to give protection against H.

k n of l

The low pressure isolation of the main steamlines atrapid reac i

r thus j

the scram feature in the Run mode which occurs when the ma n steam n provide for reactor shutdown so that high power operation at low re s

providing protection for the fuel cladding integrity s i

j flux scrams.

of the fuel cladding iniegrity safety limit is provided by the IRM and APRM high h

l Thus, the combination of main steamline low. pr li l

i i h Run mode of the fuel c! adding integrity safety limit. in addition, the isolation val <c closure s d

i lation valve f

anticipales the pressure and flux transients which occur during' normal or in i no increase in neutron closure. With the scrams set at 10% valve closure in the Run mode, there s flux.

)

Turbine EHC Control Fluid Low Preuure Scram i hisoil

[

The turbine EHC control system operates using high pre:sure oil. There a l

This fast l

system where a lost of oil pretsure could result in a fast closure of the h e fast closure scram. since closure of the turbine control valves is not protected by the turbine controlva bine failure of the oil system would not result in the fast closure solenoid valves l

ra ms.

control valve fast closure, the core would be protected by the APRM and high However, to provide the same margins as provided for the genesat t m which senses failure of l

i turbine control valves. a se' ram has been added to the reactor protect on sys e d results in reactor l

control oil preuure to the surbine control system. This is i i imilar i

to that resulting from the turbine control valve fast closure scram. The scram setpo n f

l high enough to provide the necessary anticipatory function and l l es will l

spurious scrams. Nor.nal operating pressure for this system is 1250 p ll before valve closure not start until the fluid pressure is 600 psig. Therefore. the scram occurs we l

begns.

j l

t f

Condenser Low Vacuum Scram Loss of condenser vacuum occurs when il$e condenser can no' long t

1 j

i condenser vacuum initiates a closure of the turbine stop valves and turb ne preuvre l

eliminates the heat input to the condenser. Closure of the turbine stop and b transient. neutron flux rise. and an increase in surfa i

h l dding safety limit from l

turbine stop valve closure s: tam function alonc is adequate to prevent t e c a l

1 being cxceeded in the event of a turbine trip tran>ient with bypass closure.

l The condemer low vacuum scram is anticipatory

- { -)

l at 7. inch g

l mode at.21. inch lig vacuum stop valve closure occurs at 20 inch Hg vacu 1

lig vacuum.

. Amendment Na. 60

- [.

1.1/2.1-10

a QUAD-CITIES DPR-30

(>.

ithovt impair ment galions. As indicated above, there is suf5cient volume in the piping to acco i

'h h eactor dowa.. hile surdcient of the scram times or amount ofinsertion of the co'ntrol rods.This funct on volume rernains to accommodate the di> charged water and preclud required but not be able te perform its function adequardy.

d um Loss of condenser vacuum occurs when the condenser can no longer h heat input to the initiaics a dosure of the turblne stop vahes and itirbine hypass vahes. which c t

condenser. Cbsure of the turbine stop and bypass vahes caus d if this occurs.a reactor scram f

i occurs on turbine stop vahe dosure. The turbine stop valve dosure scram unct onbine trip the cladding safety limit from being euceded in the event of a turThe conden es a rs at n inches Ifg l

scram before the stop vahes are dosed thus the resuhing transient is less snere 7 inches Hg vacuum.

vacuum, stop valve closure occurs at 20 inches lig vacuum, and bypass closure at d

radicactivity Ifi h radLiion levels in the main steamline tunnel above that due to the normd

~

i l

l are an indication of!cating fuel. A scram is inhiated when t

to prnent he site environs is prevented ii enessive turbine contamination. Discharge of excessiv ided the limit specified in Speci6 cation 3.3 is exceeded.

'~ dased from The main stean. fine isolation vahe closure scram h

h s close. By scramming at this setting, the resultant transient is insi nincant, riate to the A reactor mode switch is provided which actuates or hypasses the various sc d switch is in the

( j particular plant operadng status (reference SAR Section i

li isolation

~

d to a!!aw repairs i

yahc dosure scram arc a> passed.This bypass has been provided for Rexibi f

to be made to the turbine condenser. While this bypass is in ef i his mode.

i mline If the reactor were brought to a hot standby condition for repairs to the turb i his mode of operation isolation valves would be dosed. No hypothesized single failure or sir.gle operato can result in an unteviewed radiological re! ease.

i l

The manual scram function is active in all modes, thus providing for a rods during all modes of reactor operation.

h and The IRM system provides protection against exces funciions is also provided to supply additional neutron level information during sta i h (reference SAR Section 7.43.2). Thus the IRM is required n t e discussed in the bases for Specification addition. protection is provided in this range by the APRM 15c5 scram as 5 2 ). Thus, the 2.1. In the power range.the APRM system provid the I RM 's provide adequate coverage in the startup and intermediate range.

l high fes el The high reactor pressure. hi;;h drywell pressuse. reactor low water leve d

fl scrams are required for the Stariup/itot Standby and Run mo es o p an to be operational for these modes of reactor operation.

bypassed to start The turbine condenser low-vacuum scram is required only during power op 3

up the unit.

Arnendrnent No. 60 3.1M.1-3 i

A Ql'.\\ D-Cl i IES DPR-30 a

TABLE 3.11 REACTOR PROTECil0H SYSTElf(SCRAli)I!!$TRUMENTAll0H REQUIREl.tElliS REFUEL MO tiinitaum hmbat cl Op;re!.!a of Trlfrd Icstrm:::nl Ch:nn:ts par Actica )

tz ul Trip Function idp level Setting Trip !) stem A

1 Mode sydtth in shuthwn A

1 Manual scram liiM s120/125 cf full scate A

3 High flux 3

lacperative APRMt3) 2 High flux (15% scram)

Specification 2.1.A.2 A

A 2

InoperatNe 2

High water fevelin scram s50 gallons A

discharge volume (U '

I 2

H!gh reactor pressure s1060 psig A

.r A

2 High drywell pressure (5) s2 psig

( l Reactor low water level 28 inches:st A

2 l

2 Turbine condenser low 221 inches Hg vacuum A

m vacuum 2

Main steamline high s1 X normal full power A

radiation )

background nt 4

Main steamline isolation s10% vafve closure A

valve c!csurem e

V Amendment No. 60 3.1/ 4.1-8

i j

.Qll AI)-CITil'.S

[

DPR-30:

i j

g

)

j s

TABLE 3.12 r

REACIOR PROTECil0H SYSTEM (SCRAM)(H$iRUI.tEllidil0H REQUIREMENTS STAR l

MODE 1

l L3nimum !! umber l

'of op:ratte or l

Trlpped instrument.

5 frlp sptr,r/Il trip runction Trlp level $ctting Actiodri l

Chann:Is per A

f 1

Made switch in shutdown A

1-Manual scrari IRM 3.

High Oux s120/125 of full scale A

.[

f A

3 Inoperative f

APRit33 2-High flux (15% scram)

Specificatbn 2.1. A.2 A

j A

j 2-Inoperative 2

High reacter pressure

$1000 psig A

t 2

High drywell pressure'5) s2 psig A

2 Reactor low water level 28 inchestsi A

I

-(,,,

)

2 High water level in scram f

discharge volump')

s50 gallons A

2 Turbine condenser low 221 inches Hg vacuum A

l l

vacuumW 2

Mah steamline high s1 X normal full power A

radiation"2) background

{

4 Main steamline isolation

$10% valve closure A

I valve cbsurth I

l f

i 1

]

8

~

l l

Amendment No. 60 3.1/ 4.1 -9

+mm

t

-e QtIAD-CITIES j

DPR-30 1

i i

f

~

1ABLE 3.14 E

i REACTOR PROTECil0N SYSTEl.1(SCRAM)INSTRUf1EllTAil0ll REQUIRE 5

ISnimum Number of Op:rable er Tripp:d lastrument -

P

' Actiod!!

}

Chann:Is per Trip systerdll.

Trip ranctka trfp terel $ctting A

l Mode switch in shutdawn t

A l~

Manual scram APRMt31 '

2 High Cux (Uciv biased)

Specification 2.1.A.1 A or B A or B 2

fnoperative A or 8 i

2 Downscalec ti 23/125 of full scale

,f

'2 High reactor pressure

<1060 psig A

i 2

High dryweH pressure sz psig A

\\

j 2

Reactor low water level 28 inches ts)

A, i

2 High water level in scram A'

l discharge volume s50 galfons.

2 Turbine ccndenser low 2 ninches Hg vacuurn A or C

' l l

vacuum Q~

2 Main steamline high s1 X normal fuH radiatiodirl nwer backga ad A or C l

t

~

4 Mah steamline isolation -

I valve closure'81 s10% vafve closure A or C

-t 2

Turbirvi control valve fast 240% turbine / generator A or C l

closure

• load mismatchd83 i

2 Turbine stop valve s10% valve closure A or C closure

• l 2

Turbine EHC control Duid 2900 psig A or C m

low pressura 4.

I l

-l 3'I' Amendment No. 60 I

.E m

gml).CITif.S DPR-30 l

9 t

and also limiting the loss

()

n Gainsi tubes are prodied in the main steamlines as a means o break accident. In addition to monitoring steam fh i isolation valves. The primary nmction o t e li j

of" mass incentory from the vend during a steam ne Group ! valves are dosed.For the v.orst. case

{

l imtrumentadon is provided.which c..uses u trip pf Grof i

uccident. main sicamline break outside the drywe. tsteamline vahe dosure limbs the mas the environs is well bdew i

i I

with the now IImiters and main than 1500' F. and release of radioactiv ty to i l d 14.2.3.10).

taoveredifueltemperatures rema n ess10 CFR 100 guidelines (re'e t amline tunnel to detect leaks in this d d cause closure cP Group i isolation valve i

Temperature. monitoring instrumentation is provided in the ma n s e i

der of 5 to 10 gpm; thus it is capable of cov Trips are prosided on this instrumentadon and when excee e I

f highoteam flow instrumentation discuoed fh setting of 200' F islow encugh to detect trats o t e or 00 i hy, gives isolation before the guidelinc>

spectrum of breaks. For large bre tks. it is a backup tosma ided to detect gross fud failure. This l

i d to dose for this aseident. With are cuceded.

High. radiation monitors in the main steamline tunnel have be

' instrumentation causes dosure of Group i valves, the only vahes req i

established sening of 7 time > normal background and ma nt excl i

release is limited so that 10 CI'R 100 guiddines are no remure drops bdow 825 psiy. A trip nf l

[

)

12.2.1.7 ).

li t

Pressure instrumentation is provided which trips when main steam n l

i ided primarily to provide protection aga ns this instrumentation resuks in dosure of Group

)

alve to open. With the trip set at 8

, j this trip functiun is hypased. This function is prov l

malfunction which would cause the control and/or hypas l

h nie no fission products available for release ot er t a are provided to detect a break in thei f

actuation of the RCIC 'or of HPCIiso i

The RCIC and the HPCI high now and ternperature instrumentat on i 1.2.3 ).

li e isniation vahes, thus all senwrs are hi respective piping. Tripping of this imirumentation resu s n F and j

f Tripping logic for this function is the same as that for the main 3 te l failure criteria. Ihe trip 3ettinp of.'t.')

required to be operable ur m a tripped condition to meet the s i

f 10% of desi n flow and valve dosure time are suc r two taken 1wice logic circuit. Unlike l

e 3

h he two trip is within limits.

The instrumentation which initiates ECCS action is arranged in a one tem associated with each function rather t l

riteria are met by virtue of the fac,t that redun i

the reactor scram circuits, however, there is one tr p sys i blowdown and high.pressur'e coolant inje f il sy>tems in the reactor protection system.The single a l

d inoperaNe.

l b

Specircation requires that if a trip system ecomFor, exampl This specification preacnes the cafetthenes d

l aut.of6ervice specificadon> of Speci0 cation 3.5 SoYern.during period

. I d withdrawal 3o that MCPR. doe >l with rc3pect to the single.faihue criteria even l

l The comiol rod block functions are provided to preveni excessive co l

ft Limit.i APRM's, eight IRM's go below the MCPR Fuel Claddipo Integrity Sa e yT l

[

l four SRM's will result in a rod block. The minimum in m instrument channd requirements ii intenance, testing, or calibration.

instrumentation to assure that the sing!c. failure I

f This time period is only~3% nf the operating time in a month and a

t

)

preventing an inadvertent control rod withdrawal.

l' Amendment No. 60 i

a 3.1/ 4.2-6 4

t

- ~

. QUAD-ClTIES DPR-30 b

+

w e

iABLE 3.21 INSTRUMElliAT10tl THAT ll:lilATES PRIMARY CONTAINMEHi lSOLAil0!! FUllCil0NS I

l.'Irdmum l'ur.ibar of Oprabia or Tr4;:d lastrum:nt l

Chenrelsm Instruments Trip levet Setting ActicaW 1

4 Reactor fow waterW yl44 inches above top of A

i e

i active fuel

• 1 I

4 Reactor fow low water 284 inches above tcp of A

l active fuel' m

s2 psigm A

i 4'

High drywell pressure m

sidog of rated steam hw B

l

]

5 16 H'gh hw main steamline 16 High temperature main

$200*F B.

1 steamline tunnel l

4 High radiation main s1 x normal rated power B

steamline tunne(H' background 4

Lcw main stearn pressure

• 2825 psig B.

l 4

High hw RCIC,tcamline s300% of rated steam flow C

{;

l 16 RCIC turbine area high s200*f C

temperature l

4 High kw HPCI steamline s300% of rated steam flow 0

16 HPCI area high temperature s200

• F 0

P.'ates 1 Whenever primary ccatainment integrity is required, there shall be two operable or tripped systems for each function. except I

for los pressure main steamline which only need be available in the Run position.

I 2.

Acton. If the Erst column cannot be met for one of the trip systems that trip system shall be tripped.

]

If the fust column cannot be met for both trip systems, the appropriate actions listed belon shall be taken:

A, Initiate an orderty shutdown and have the reactor in Cold $hutdown condition in 24 ' ourt n

B. Initiate an orderly load reduction and have reactor in Hot Standby within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, f

[, Cbse isolaten valves in RCIC systen j

D. Cbse isolaten valves in HPCI subsystea l

) Need act be ope <able when prunary containment integrity is not sequired.

i 4 the aciates trip signalis bypassed when the mode switch is in Refuel or $tartup/ Hot Shutdown.

4

$ Thrs mstrumentation also isolates the control roorn ventilation systes 6 This signal also automaticalty closes the mechanical vacuum pump discharge line isolation valvet

~

• Top of active fuel is defined as 360** above vmsel aer o for all water levels used in the LOCA analysis (see Ba:.es 3.2).

l l

l l

Amendment No. 60 3.2/4.2-11

~

~

k j

QUAD-CITIES DPR-30 l

l TABtE 3.22 liiSTRl!!.iE!! Tail 0!! THAT ll!!TIATES OR CONTROLS THE CORE All0 C0!!TA Minimuni INew of OS:r;t'e or Trippad fastrtm rrt Channsl:43 Trip Function Trip level Sett!rtg Remarks,

i 4

Reactor low bw 234 inches ( + 4 inches /-0 inch)

1. In conjunction with low-reactor pressure _

initiates core spray and LPCI.

water fevel above top of active fuel'

2. In conjunction with high.drywell pressure 120 second time delay and law. pressure core cooling interlock initiates auto blowdown.

3. Initiates HPCI and RCIC.

4 Initiates starting of diesel generators.

1. Initiates core spray, LPCI, HPCI, and 4"3 Egh-drywell s2 psig

scis, pressure ),cs)

2. In conjunctica with low low water fevel, r2 120-second time delay, and low pressure core cooling interlock initiates auto bfowdown.

(

3. Initiates starting of diesel generators.

4.

Initiates isolation of control roarn ventilation.

2 Reactor low 300 psigsps350 psig

1. Permissive for cpening core spray and LPCI admission vakes.

pressure

2. In conjunction with bw bw reactor water levelinitiates core spray aM LPCI.

, Prevents inadvertent operation of containment Containment spray spray during accidenLconditions.

interlock 2(31 2/3 core height 22/3 core height 4'38 containment 0.5 psigspsl.S psig high pressure In conjunction with low low reactor water 2

Timer auto

$120 seconds tevel, high-drywell pressure, and low-pressure blowdown core cooting interlock initiates auto blow-down.

Defers APR actuation pending confirmation of

\\

4 Low. pressure core 100 psigspsisopst low pressure core cooling system cperation, coofing pump dis.

charge pressure

1. Initiates starting of diesel generators.

2 Undervoltage on N/A

2. Permisske for starting ECCS pumps.

emergency buses

3. Removes nonessentialloads from buses.

• Top of active fuelis defined as 360" above vessel zero for aff water levels used in the t.OCA enelysis.

= - _ _ _ ___Amndm nt No. 60 3.2/4.2-12 l

P QUAD-CITIES DPR-30 C..,J is being done which has the potential for draining the reactor vessel.

3. When irradiated fuel is in the reactor and the vessel head is removed, the suppression chamber may be drained completely and no more than one con-trol rod drive housing opened at any one time provided that the spent fuel pool gate is open and the fuel pool water levelis maintained at a level of greater than 33 feet above the bottom of the pool. Additionally, a minimum condensate storage reserve of 230,000 gallons shall be maintained, no work shall be performed in the reactor vessel while a control rod drive housing is blanked following removal of the con-i trol rod drive, and a special llange shall be available which can be used to blank an open housing in the event of a leak.

4.

When irradiated fuel is in the reactor and the vessel head is removed, work h.

that has the potential for draining the vessel may be carried on with less than i12,200 ft of water in the suppression 8

pool, provided that: (1) the total vol-ume of water in the suppression pool, refueling cavity, and the fuel storage l

pool above the bottom of the fuel pool 5

gate is greater than 112,200 ft ;

l (2) the fuel storage pool gate is re-moved;(3) thelow pressure core and containment cooling systems are oper-t able; and (4) the automatic mode of l

the drywell sump pumps is disabled.

G. Maintenance of Filled Discharge Pipe G. Maintenance of Filled Discharge Pipe The following surveillance requirements shall be adhered to to assure that the discharge i

piping of the core spray, LPCI mode of the RHR, HPCI, and RCIC are filled:

1. Every month prior to the testing of the

1. Whenever core spray, LPCI mode of LPCI mode of the RHR and core spray the RHR, HPCI, or RCIC are required ECCS, the discharge piping of these

]

to be operable, the discharge piping systems shall be vented from the high from the pump discharge of these sys-tems to the last check valves shall be point and water flow observed.

fi!!ed.

w:

Amendment No. 60 3.5/ 4.5-7

QUAD-CITIES DPR-30 g-(...)

2. The discharge pipe pressure for the 2.

rolicwing any period where the t x I rcx!e of the FJct or core roray ars systems in Speci0 cation 3.5.G.1 shall have toen out of.;ervice and drained be ma.mtained at greater than 10 ps.ig for raintenance, the discharge pioing and less than 74 psig. If pressure in of the inepetable system shall ce' vented frca the high toint prior to I

any of these systems is less than 40

'h t"'" f 'h* 573t * '

S*'"I **

psig or greater than 74 psig. this con-f dition shall be alarmed in the conttol room and immediate corrective action

. 'henever the llPCI or RCIC system.

n 1

taken. If the discharge pipe pressure is is k.ned up to take suction from the not within these limits in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> torus, the discharge pipmg of the after the occurrence, an orderly shut-IIPCI and RCIC shall be vented from down shall be initiated, and the reac-the high point of the system and water tor shall be in a cold shutdown condi-Ow bserved on a monthly basis.

tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiation.

4 The pre:sure switches which monitor the discharge lines and the discharge l

of the fill system pep to ensure that j

they are full shall be functionally tested every unth and calibrated every 3 nonths. The pressure switches shall be cet to alarm at a decreasing pressure of 140 psig an3 an increasing pressure of 6 90 psig.

[

H.

Condensate Pump Room Flool Protection H.

Condensate Pump Room Flood Protection

1. The systems innaued to prevent or

1. The following surveillance require-mitigate the consequences of Gooding ments shall be observed to assure that of the condensate pump room shall be the condensate pump room Good pro-operable prior to startup of the tection is operable.

reactor.

The piping and electrical penetra-a.

2. The condenser pit water level switches tions and bulkhead doors for the shall trip the condenser circulating vaults containing the RliR service water pumps and alarm in the control water pumps and diesel generator room if water level in the condenser cooling pumps shall be checked pit exceeds a level of 5 feet above the during each ' operating cycle by pit floor. If a failure occurs in one of pressurizing to 15 i 2 psig and these trip and alarm circuits, the failed checking for leaks using a soap circuit shall be immediately placed in bubble solution. The criteria for a trip condition and reactor operation acceptance shall be no visible leak-shall be permissible for the following age through the soap bubble 7 days unless the circuit is sooner solution.

made operable.

b. The door drains from the vaults

3. If Speci0 cation 3.5.H.! and 2 cannot shall be checked during each oper-be met, reactor startup shall not com-ating cycle by removing the end mence or if operating, an orderly shut-cap and assuring that water can be down shall be initiated and the reactor run through the drain lines.

shall be in a cold shutdown condition c.

The RHR service water pump and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

diesel generator cooling water pump bed plate drains shall be checked during each operating Amendment No. 60

L QUAD-CITIES DPR-30

~

.s 45 SURVEILLANCE REQUIRE 3 TENTS BASES i

The testing interval for the core and containment cooling systems is based on a quantitative reliability a judgment, and practicality. The core cooling systems have not been designed to be fully testable durin For example, the core spray Onal admission valves do not open until reactor pressure has fallen to 350 during operation, even if high dryucll pressure were simulated, the Gnal valves would not open. In the llPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel l

which is not desirable.

The systems can b'e automatically actuated during a refueling outage and this will be done. To increase L

availability of the indisidual components of the core and containment cooling systems, the component make up the system, i.e., instrumentation, pumps, valve operators, etc., are tested more frequently. T instrumentation is functionally tested cach month. I.ikewise the pumps and motor-operated valves are also test each month to assure their operability. The combination of a yearly simulated automatic actuatiert rest and monthly tests of the pumps and valve operators is deemed to be adequate testing of these systems.

~

With components or subsystems out of service, overall core and containment cooling reliability is maint demonstrating the operability of the remaining cooling equipment. The degree of operability to be demon depends on the nature of the reason for the out of service equipment. For routine out-of service peri preventative maintenance, etc., the pump and valve operability checks will be performed to demonstrate operability of the remaining components. However, if a failure, design de6ciency, etc., causes the o period, then the demonstration of operability should be thorough enough to assure that a similar pro not exist on the remaining components. For example,if an out-of service period caused by failure of a pump to deliver rated capacity due to a design denciency, the other pumps of this type might be subjected to a now rate

(,_.I test in addition to the operability checks.

The verification of the main steam relief valve operability during manual actuation surveillance testing must be made It has been found that a independent of temperatures indicated by thennocouples downstream of the reffer valves.

temperature increase may result with the valve still closed. This is due to steam being vented through the By first opening a turbine bypass valve,and then observ ng its closure response during relie during the surveillance test.

valve actuation, positive verification can be made for the relief valve opening and passing steam now. Closu the turbine control valves during relief valve manual actuation would likewise serve as an adequate verification for the relief valve opening. This test method may be performed over a wide range of reactor pressures greater than 1 Valve operation below 150 psig is limited by the spring tension exhibi'ed by the relief valves.

The surveillance requirements to ensure that the discharge piping of the core spray, LPCI mode of the R and RCIC systems is lilled provides for a visual observation that water dows from a high poim vent. Thi that the line is in a full condition. Instrumentation has been provided to mnitor the pressure of water in the discharge piping between the monthly intervals at which the lines This instru.entation are vented and alarm the control ream if the pressure is inadequate.

will be calibrated on the same frequency as the safety system instrumentation and the This testing ensures that, during the interval between the alarm system tested tronthly.

tronthly venting checks, the status of the discharge piping is racnitored on a continuous I

basis.

40 psig for the low pressure of the fill system has been chosen because r

An alarm point of 2:

piping within' the plant, 39 psig is required to keep the lines full. The shutoff head of the fill system pumps is less than 90 psig and therefore will not defeat the icw-pressure cooling pamp discharge press inter 1cck 100 psig as shown in Table 3.2-2.

A margin of 10 psig is provided by the high pressure alarm paint of 90 psig.

O Amendment No. 60 3.5/45-15