Text

Forwards Proposed Amend to OL Changing Tech Specs Re Inservice Insp Program
	ML19276F895
Person / Time
Site:	Quad Cities
Issue date:	04/18/1979
From:	Reed C; COMMONWEALTH EDISON CO.
To:	; Office of Nuclear Reactor Regulation
References
	NUDOCS 7904240437
	Download: ML19276F895 (65)
	v • d • e

Commonwealth Edison One First Noow Pfila CNcap

[ l' h n o 's Address Reply to Post Off.ce Box 767 Chicago, IM.nois 60690 April 18, 1979 THIS DOCUMENT CONTAINS Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission POOR QUAUTY PAGES Washington, D.C. 20555

Subject:

Quad-Cities Station Units 1 and 2 Proposed Amendment to Facility Operating Licenses Nos. DPR-29 and DPR-30 Regarding Inservice Inspection (ISI)

NRC Docket Nos. 50-254 and 50-265 References (a) : C. Reed letter to E. G. Case dated May 22, 1979 (b) : D. L. Ziemann letter to R. L. Bolger dated November 24, 1979

Dear Sir:

Pursuant to 10 CFR 50.59, commonwealth Edison proposes to make amendments to Quad-Cities 1 & 2 Technical Specifications regarding the Inservice Inspection (ISI) Program. These changes modify the Technical Specifications as required by 10 CFR 50, Section 50.55a(g). The changes bring the Technical Specifications into conformance with ASME Boiler and Pressure Vessel Code Section XI requirements so that no conflicts exist between the two documents. In all cases, the Section XI requirements were determined to be sufficient in malntaining an adequate margin of safety. The existing Technical Specifications were also reviewed to identify those surveillance requirements that could potentially place the plant in an unsafe mode of operation should a single failure occur in the process of testing. This resulted J in the elimination of certain existing requirements, primarily, the testing of redundant components when the related subsystems are determined to be inoperable. The elimination of this testing has been reviewed with 'NRC Staff Guidance for Complying with Certain Provisions of 10 CFR 50.55a(g) ' Inservice Inspection Requirements'" transmitted by Reference (b) and has been 0

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Commonwealth Edison NRC Docket Nos. 50-254/265 Director of Nuclear Reactor Regulation April 18, 1979 Page 2 determined to be within the intent of the Staff Guidelines.

With the additional testing requirements of Section XI, there is added assurance that the varit essential components will have the required operational reae.aess when required.

Reference (a) previously transmitted a proposed Technical Specification change which is related to the proposed ISI Technical Specification changes and should be reviawed on a comparable schedule-Enclosure 1 contains amended Technical Specification pagas for Quad-Cities Unit 1 and Enclosure 2 contains amended pages for Quad-Cities Unit 2. These proposed changes have received on-site and off-site review and approval.

Pursuant to 10 CFR 170, Commonwealth Edison has determined that these proposed snendments are one (1) class III and one (1) class I Amendments. As such, Commonwealth Edison has enclosed a fee remittance in the amount of $4,400.00 for the proposed amendments.

Please direct any questions concerning this matter to this office.

Three (3) signed originals and thirty-seven (37) copics of this transmittal are provided for your use.

Very truly yours, On -

Cordell Reed Assistant Vice-President to SUBSCRIBED andIAqQg{ , day before of l/dil me.[this , 1979.

N/) 11 0 / L/ ddfI>L>O

, Notary Public L)

ENCLOSURE 1 Amended Technical Specification Pages Quad Cities Unit 1, DPR-29 The following pages have been revised:

1.0-4 3.4/4.4-1 3.4/4.4-2 3.4/4.4-3 3.5/4.5-1 3.5/4.5-2 3.5/4.5-3 3.5/4.5-4 3.5/4.5-5 3.5/4.5-6 3.5/4.5-7 3.5/4.5-8 3.5/4.5-9 3.5/4.5-10 3.5/4.5-11 3.5/4.5-12 3.5/4.5-13 3.5/4.5-14 3.5/4.5-16 3.6/4.6-4 3.6/4.6-11 3.6/4.6-12 3.6/4.6-16 3.6/4.6-17 3.7/4.7-9 3.7/4.7-10 3.7/4.7-18 3.9/4.9-3 The following pages have been added:

1.0-5 3.4/4.4-2a The following pages have been deleted:

3.6/4.6-18 3.6/4.6-19 3.6/4.6-20 3.6/4.6-21

QUAD-CITIES DPR-29 -

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Y. Shutdown - The reactor is in a shutdown condhion vehen the reactor mode switch is in the Shutdown po,itian and no core a!!crations are being performed.

1. !!o: Shutdown incans conditions as above, with reactor ccotant temperature greater than 212

F.

2. Cohl Shutdown means condi' ion

as above, with reactor coofant temperature equal to or less than 212 ' F. ,

Z. Simulated Auto:natic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question _ ,

AA. Total Peakin;: Factor - The total ,$eaking factor (TPi') is the highest prcduct of radial, axial, and local peaking facters simultaneously opuative at any segment of fue! red. ,

BB. Transition Boi?ing - Transition boiling means the boiling regime betizeen nuc!cate and n!m boilina.

Traie hion, boiling is the regime in which botn nucicate and lilm boitin3 cccur mtermittently, with neithEr .

type bein3' completely stable. .- 3 .

...A.,.. .

CC. Critical Power P.atio (CMt) - The critical power ratio is the ratio of that assemb!y power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor conditica ofintercst as calculated by app!ication of the GEXL ccrrelation (reference NEDO.10958).

DD. Minimum Critical Power Ratin (M CPP.) - The minimum incore critica! power ratio corresponding to the mest limiting fuel assembly in the ccre. .

EE. Sunciliance intcival - Each surveillance ;equirement shall'be performed within the speciSed sun eil.

lance interval with: .

a. A mn.imum a!!ewable exte:" inn not to exr. -d 2M of tr r smei!hnee ime va!.

. b. A to:a! maximum combined interval time for any 3 consecutive surveillance intervals not to exceed

. 3.25 times the speciGed survei!!ance interval. .

FF . - Inservice Testing Prograra For Purps And-Valverr The testing progra:a developed for inservice testing of pumps and valves is written in accordance with Section p of thc AS:2 Boiler and Pressure vessel Code and applicable Addenda as required b/ lOOPn50, -

Section 50.55 a(g) , except where specific written relief has been granted by the.rmC persuant to 10CPP.50, Section 50.55a(g) (6) .(i) .

Certain clarifications regarding the use of the above raentioned .

progra:a are as follows:

. a. The surveillance intervals specifiell by the Inservice Testing Program shall be perfortaed as specified in Definition "EC.-Surveillance Interval"' with the following clarifications:

Weekly at least once per 7 days Monthly at least once per 31 days Quarterly or every 3 months at least once per 92 days semiannual]y or every 6 no. at least once per 184 days Annually at least once per 3G6. days

b. tiothing in the AS'1" Eoiler and Pressure Vessel Code shall be cannt. rued t.o supercede the require:.qnts of any Technical 1.0-4 ,

QUAD-CITIES DPR-29 Specification, For example the requirements under limiting conditions for operation that specify when certain equipment shal] be operable takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to one week after return to normal operation. Also, the Technical Specification definition of OPEFABLE does not grant a grace period before equipment that is not capable of per-forming its function is declared inoperable. This takes precedence over the ASME Code provision that allows a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before being declared inoperabic,

c. Performance of a Surveillance Requirement within the specified time interval shall constitute compliance with OPERABILITY requirements for a Limiting Condition for Operation and associated action statements unless otherwise required by the specification.

GG. Ir.scrvice Inspection Progran for Components Classified as ASME Code Class 1, 2 and 3.

The inspection program developed for inservice inspection of components classified as ASME Code Class 1, Class 2, and Clars 3 is written in accordance wi th Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10CFR50, Section 50.55a(g), except where sp;cific written relief has been granted by the NRC pursuant to 10CFR50, Section

50. 55a (g) (6) (i) .

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QUAD-CITIES -

DPlt-29 3.4/4.4 STANDBY LIQUID CONTROL SYSTEM .

I,lMITING CONDITIONS FOR OPERATION SURVEILLIJ:CE REQUIREttENTS .

Applicability:

Applicability: ,

.The operational readiness of the Standby Lig-Applies to the operating status of the standby liquid uid Control System shall .be insured and deraon control system. ~

strated by the Inservice Testing Program for Pumps and Valven defined in Section 1.O(P.F)'

Objective: of these Specificati.ons. Additional require-To assure the availability of an independent reactiv- ments are listed below. ,

ity control mechanisra.

-To verify the operability of th'c standby' ,

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SPECIFICATIONgid control systera.

A. !!oriaal Operation A. Norma! Operation The operability of the standby liquid centrol During periods when fuel is in-th system shall be verified by performance of th:

, reactor the standby. liquid control pggg,,; m, i system shall be operable except when s the reactor is in the Cold, S'autdown ' l. At least once per quarter. l Condition and all control rods ar Demineralized water shall be recycled fully inserted and Specification' to the test ' tank. Pump minimum flow 3.3. A is raet. or as speciiled -in rate of 39 gpm shall be verified anainst

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3.4.B below.

a sy. stem head of 1275 psig.

B. Operation with Inoferable Components 2. At least once during each operating ,

From and after the date that a redundant component is made or found to be inoper- Manually initiate the system, except able, Specificati.on 3.4.A shall be con- the explosion valves and pump solu-sidered fulfilled, and continued. opera- tion in the tecirculation path. to dem.

tion permitted provided that the compo- onstrate that the pump suction line nont. is returned to an operable condition from the storage tank is not plugged.

within 7 days. .

E plode two of six charges or two of four charges manufactured in the same batch using the permanent system wir-ing to verify prcper fonction. Then install the untested charges in the ex-plosion valves.

Dennneiatu.ed watti siiail be inycted via a test wnncction into the reactor

- vessel to test that valves (enept egio-sion valves) not checked by the recir-r culation test are not clogge'd.'

a 3.4 / 4. 4 - 1

QUAD-CITIES DPR-29 Test that the setting of the system pressure relief valves'is between 1400 and_1490 psig.

3. Disassemble and inspect one explosion valve so that it can be established that the valve is not clogged. Both valves shall be inspected in the course of two operating cycles.

C. 1.iquid Poison Tank-Doron Concentration C. Liquid Poison Tank-Boron Concentration The liquid poison tank shall contain a boron-The availability of the proper boron-bearing bearing solutien that satis 6es the volume- solution shall be veri 6ed by performance of the concentration requirements of Figure 3.4-1 and following tests:

at all times when the standby liquid control system is required to be operab!c and the solu- 1. At least once per month t tempeutur sha!! not be less then the Boron concentration shall be datcr-

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' .iperature presented in Figure 3.4-2. mined. In addition, the boron ccncen-

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tration shall be determined any time D. If Speci6 cations 3.4.A through C are not met, water or boron arc ~ added or if the an orderly shutdown shall be initiated and the solution temperature drops below the reactor shall be in the co!d shutdown condition limits speci6c.i by Figure 3.4-2.

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

2. At least once per day Solution volune shall be checked.

3. At least once per day The solution temperature shall be checked.

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QUAD-cl~11ES OPR29 .

3.4 LIMITING CONDITIONS FOP OPEllATION BASES ,

A. The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the veithdrawn control rods can be inserted. To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentrat!'n of C00 ppm of boron in the reactor core in approximately 90 to 120 minutes with imperfect inixing. A baron concentration of 600 ppm in the reactor core is required to bring the reactor from full power to a 3Eo ak suberitical condition considering the hot to cold reactivity swing, xenon poisoning and an additional margin of 150 ppm in the reactor core for imperfect mixing of the chemical solution in the reactor water. A normal quantity of 3470 gallons of solution having a 13.47c sodium pentaborate cor. centration is required to meet this shutdown requirement.

%c time requirement (90 to 120 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown 'of the reactor fo!!owing the xenon poison peak. For a required pumping rate of 39 gpm, the maximum storage volume of the boron solution is established as 4S75 gallons (195 gallons are contained below the pump suction and, therefore, cannot be inserted).

Boron concentration, solution temperature, and volume are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Performing pump tests in accordance with the Inservice Test Program provides adequate assurance that the pump will be operable.

The only practical time to test the standby liquid control system is during a refueling outage and by initiation from local stations. Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refueling outage r unnecess?.ry. A test of exptmive charges from one rmnufacturing batch is made to assure that the charges arc sausfac: cry. A contin =1 check cf t!.: firing ciremt continu:ty is r svid:d by pilm !!; hts ir. In: ccmrul Toom.

B. Only one c.f the two standby !iquid control pumping circuits is needed for proper operation of the system.

If one pumping circuit is fo.md to be inoperable, there is no immediate threat to shutdown capahifity, and reactor cperation may continue while repairs are being made. Assurance that the remaining system will perforra its intended function and that the reliability of the system is good is obtained by completing the tests outlined in the Inservice Test Program for pumps and valves.

I C. The solution saturation temperature of 139- sodium pentaborate, by weight,is 59' F.The solution shall be kept at least 10" F above the saturation temperature to guard against baron precipitation.The 10

F margi.n is included in Figure 3.3-1. Temperature and liquid level alarms for the system are annunciated in the control room.

Once the solution has been made up, boron concentration will not vary unless more boron or more water is added. I.cvel indication and alarm indicate whether the solut. ion volume has changed, which mieht indicate a possible solution concer.tration change. Considering these factors, the test int established.

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3.4 / 4.4 -3

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. DPR-29 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SunVEILLANCE'nEQUIREMENTS

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Applicabil'ty:

Applicability:

. Applies to the operational status of the emergency cooling subsystems. . The operational readiness of the following s, subsystems shall be' demonstrated in accord-Objective:

ance with the Inservicle Testing Program for Pumps and valves defined.in

'To assure adequate cooling capability for heat re- Section,1.0 (F.F.) of these Technical moval in the event of a less f l i Sp cifications. Additional requirements ant.for each subsystem are listea below:

1 solation from the n.ormal.

heat smk. -o -coo acc dent or reactor .

Objective:

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To verify the operability of the core and e

containment cooling subsystems.

SPECIFICATIONS. - . -

A. Core Spray Subsystems A. . Core Spray Subsystems h

W -

. Surveiliance of the core spray subsynems shall be performed las follows:

1. Both core spray subsystems shall be 1. Core Spray Subsystem Testing operab!: whenever irradia ted fuel is in the reactcr vessel and prior to reactor licm INguency startup frem a cold condition except as specified in 3.5.A.2 and -S mu ated auto-. Each -

3.5.G.2. matic actuation s efcelinae

2. From and after the date that one " ' Y of,the core spray subsystems is . . . core spray pumps made or found to be inoperable for shall deliver at

' any reason, reactor operation is least 4500 gpm .

permissible on1.y during the . "E "5

" 7"

succeeding seven days.unless such head correspond-subsystem i_s sooner made operable. Ing to a reador provided th' at during such seven vessel pressurc' days all active components of the OI 99 PS10 other core spray subsystem and the LPCI subsystem and the diesel gen-erators required for operation of such components if no external source of power were available shall be operable.

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QUAD-CITIES.

DPR-29 '

If the requircraents of 3.5. A cannot be Sj 3.

met, an orderly shutdown of the reactor

c. Core spray '

licider Ap shall'be initiated and the reactor shall instrutnentation be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . check Once/ day Once/3 calibrate

. . months '

-test Once/3 months

d. Logic systern Each refueling l

functional test outag::

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QUAD-CITIES DPR-29 ,

1 B. LPCI t!cde of PllR D. Surveillance of LPCI !sode of IEIR

1. The LPCI mode of t,he RHR system 1. Once ench quarter it shall be verified simll be operable wikn ver irradiated that three IUIR pumps deliver at least fuel is in the reactor vesse! and prior to 14,500 gpt against a systera head corre-reattor st.t: tup from a cold condition, sponding to a reactor vessel pressure except as specified in 3.5.H.2, of 20 psig. , ,

3.5.U.3 and 3.5.G.2.

. . . .2. A Siraulated Automatic A'ctuation Test sha3

- be completed each refuelin' outa;e.

2. From and after the date that one of the RIIR pumps is made or found to be 3. . A Logic system functional Test shall be inoperable fbr any reason. continued completed each refueling outage.

reactor operation is permissible only -

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during the succeeding 30 (fays unless ; .

such pump i< sooner made operable, provided that during such 30 days the ,

remaining active components of the .

LPCI : node of the RHR, containment cooling mode of the RHR. all active -

components of both core spray subsys-tems, and the diesel venerators re- L quired for operation of such compo- (

nents if no external source df power were avai!ah!c shall be operah:e.

3. From arid after the date that the Li'CI - -

mode of the RllR system is rnade or -

I found to be inoperab:e for any reason. -

continued reactor operation is perrais- - t sible only during the succeeding 7 days :

unless it is sooner made operable, pro- -

l vided th:t during such 7 days all active -

compoace.ts of both core r. pray subsys- -

tems, the containment cooling mode of - -

the RHR (including two RHR .

pumps), and the diesel generators re- '

quired for operation of such compo- - -

nents if no c7.tctnal source of power ,

I were available sha!! be operab!c.

4- If the requirements of Specification ,

3.5.D cannot be met, an ordedy shut-

. .down of the reactor shall be initiated, nnd the reactor shall be in the cold shutdown conditloa withi: 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.5/4.5-7a

QUAD-CITIES I DPR-2()

c.* Containarent Ceoling Moda or the niin~

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C. Coi.tair..a: nt Ccoling I'. lode "of the RIIR Sysic:n ,

' System Surveithnce of the containtr. cat coo!ing nic.d3 of the RllR systera sha!! be perfermed cs -

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. ,- . follows: ,

' 1. RiiR service water subsp; era testing:

1. iiotitloops of the containinent cooling

inode of the RfIR systein. as defined in -

Item ~ Ficquency ,

the bases for Specification'3.Sc . shall ,'

1,>e operable whenever irraiia tail fuel is in'the reactor vessel and prior to reac- , l'

. tor startup from a co!d candition "

'except as specified in 3.5.c.2, a. , Fion rate' test - Quarterly 3.5.c.3, 3.5.c.4, and 3.5.G.2. cach Ri!R service water ptimp shall l-deliver .?.t least

~ 3500' ppra against I

a pressur'e 'of 19S -

- psig ;

A logib system Refueling

b. .

l functi.onal test . ;

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2. From and : frer the d.ite th:?t one of the i

ltilR service water pumps is rda<!c cr '

found to be inoperable for any reason. l continued seatict oper.itian is termis-sib!c en!y durine the succeeding 3t) l days unh a such para;iis sr.oner inade '

operat,!c. prr.vh!cd that duri:' t v.ti: 30 ,

days all ether active com; on enh of the contaiament cooling mor.'e of the l'i!R ,

system arc operab!c. ,

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QUAD-CETirS DPR-29

3. From and after the date that one loop .

of the containment cooling mode of -

the RilR sy> tem is made or found to be inoperable for any reason, contin- ,

ued reactor operation is permissible only during the succeeding 7 days un-less such subsystem is sooner made

. operable. provided that all active com-

.ponents of the other loop of the con-tainment cooling mode of the RllR system, both core spray subsystems, and both diesel generators required ,

for operation of such components if no external source of power were availa-ble, shall be operable.

4. Containment cooling spray loops are 2. During each 5-year period, an air test required to be op:rable when the reac- shall be performed on the drywell tor water temperature is greater than spray headers and nonles and a water 212* F and prior to reactor startup spray test performed on the torus from a cold condition. Centinued rea:- spray header and nonles.

tor. operation is permitted provided , ,

,,- - . that a maximum of one drywell spray .

Inop may be inoperable tbr 30 days when the reactor water *:mperature is - o -

greater than 212

F.

5. If the requirements of 3.5.c cannot be met, an orderly shutdown shall be ini-

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,tiated, and the reactor shall be in a cold shutdown condition within 24 .

hours. .-

I o. ireCr subsystem o. nect subsystem Surveillance of !!PCI subsystem c. hall b: per-formed as follows:

1. The !!PCI subsystrm sha!! be operab!c 1. Once per quarter, it shall be verified whenever the reactor pressure is that the HPCI pump delivers at least greater than 90 psig. irradiat:d fuel is 5000 gpm against a system head corre -

in the reac'er vessel and prior to reac- sponding to a reactor vessel pressure oz tor startup from a cold condition, 1150 psig to 150 psig.

except as specified in 3.5.o.2. !

. 2. A Simulated Automatic Actuation Test shal '

be completed each refueling outace.

3. A Logic System Functional. Test shall be

2. From and after the dare that the ;. !PCI completed each refueling outage.

subsystern is made or lound to h inop-crable for any revson. continued reac-tor opera: Ion is permi.sihte only dur-ing the succ:cdint. 7 days unless such subsystem is sooner made operable.

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3.5/ 4.5- I

QUAD-CITIES DPR 29 1

provided that during such 7 days all active components of the automatic pressure relief subsystems, the core spray subsystems. LPCI mode of the RIIR system, and the RCIC system are operable. -

3. If the requirements of S,-ecification g 3.5.D cannot be met, an orderly shut-down shall be initiated, and the reac- i tor pressure shall be reduced to 90 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

E. Automatic Pressure Relief Subsystems E. Automatic Pressure Relief Subsystems i

Surveillance of the automatic pressure relief sub-systems shall be performed as follows:

1. The automatic pressure relief subsys- 1. The following surveillance shall be carried tem shall be operable whenever the out on a 6 month surrveillance interval:

reactor pressure is greater than 90 psig. irradiated fuel is in the reactor a. A simulated automatic initiation which opens all pilot valves.

vessel and prior to reactor startup from a cold condition, except as b. With the reactor at pressure each specified in 3.5.E.2. relief valve shall be manually opened.

Relief valve opening shall be verified by a compensating turbine bypus valve or control valve closure.

2. A logic system functional test shall be performed each refueling outage.

3. When it is determined that one rehef

2. From and after the date that one of the valve of the automatic pressure relief five relief valves of the automatic pres- subsystem is inoperable, the IIPCI shall i sure relief subsystem is made or found be operable.

to be inoperable.when the reactor is '

pressurized about 90 psig with irradi-ated fuel in the reactor vessel, reactor operation is permissible only during the

, succeeding 7 days unless repairs are made and provided that during such time the IIPCI subsystem is operable.

3. . If the requirements of Specification l 3.5.E cannot be met, an orderly shut-down shall be initiated and the reactor pressure shall be reduced to 90 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G 3.5/4.5-5

QUAD-CIT] F.S Dl'R-29 Reactor Core isolation Cooling System F. Reactor Core Isolation Cooling System F.

Surveillance of the RCIC system shal: be per-formed as follows: '

l. The RCIC system will be operab!: 1. once per quarter, it shall be

" whenever the reactor pressurc is verified that the RCIC pump delivers greater than 150 psig. irradiated fuel at least 400 gpm against systen is in the reactor vessel, and pnor to head corresponding to a Reactor startup from a cold condition Vessel Pressure of 1150 psig i , except as specified in to 150 psig.

3.5.F.2.

t

2. A Iogic, Systera functional

. test shall be completed each r fu ling utage.

2. From and after the date that the RCIC system is made or found to be inopera- 3. A Simulated Automatic Actuation b'le for any reason, continued reactor Test shall be completed each operation is permissible only durina refueling outage.

the succeed.ing 7 days unless such sys-tem is sooner made operab!e, provided that during such 7 days all active com-ponents of the IIPCI systen are operable.

3 If the requirements of Specification 3.5.F. land 3.5F.2 cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be re-duced to 90 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . _

. During the period November ^6 through November 20, 1976 the unit rnay be ,

{

' started up with the RCIC inoperable pro-vided that (1) the facility is not more than 7 days with the RCIC inoperable >

and (2) all the liPCI system actisc com-ponents are demonstrated to be operable inenediately after startup and daily therea fter.

.\linimum Core and Containment Cooling Sys- G. Minimum Core and Containment Cooling Sys-l G. tem Availability tem Asailability

1. Any combination of inoperable com- Surveillance requirements to assure that mini-mum core and containment cooling systems are ponents in the core and containment coolin;; systems shall not defeat the available have been specified in Specincation capability of the remaining operable 4.2.B.

components to fulnll the core and con- ,

tainment cooling functions.

2. When irradiated fuel is in the reactor vessel and the reactor is in the cold shutdown conj? tion, all low pressure

- core and containment ecoling systems "

may be inoperable provided no work .

3.5 / .l.5 - 6 ,

QUAD-CITIES DPR-29 cA is being, done which has the potential for draining tl:e reactor vessel.

3. When iriadiated fuel is in the reactor ~

and the ves.e! head is removed, the suppression chamber may be drained comp!ctely 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 poo!

water level is maintained at a !cvel of

greater than 33 feet above tlie 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 blanted following removal of the con-trol rod drive, and a special flange 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

nd the ve:.1 Lead is remeced, "enrt that has the potential fi..- drainine the -

vessel may be carried on with less than i 12,700 !P of water in the suppression -

pool, provided that: (1) the total vol-ume of water in the suppression pool, refueling cavity, and the ft:cl storage pool above the bottom of the fuel pool gate is g cater than 112.200 fth (2) the fuel storage pool gate is rc-moved; (3) the !<'w-pressure core and cor.tainment cooling systems are oper- ,

ab!c; and (4) the automatic mode of the dryv, ell sump pumps is disabled.

ID miatem. nee of Filled Ibeharge Pipe II . Maintenance of Filled Dischar,a,e Pipe f

(

The follov.ing surveillance requirements shall be adhered to to assure that the discherre piping of the core spray, LPCI mode of the RilR, llPCI, and RCIC are filled:

1. Whenewr enre snray. I.PCI mocie of 1. Every quarter prior to the the Rll!'. IIPCI. or RCIC are required testing of the LPCI mode of the to be operabic the diwharge piping RIIR and core spray ECCS, the l discharge piping of these systems from ti.e pump di charpe of thcsc sys-

, term to the I.c.t chaL VMees shall be shall be vented from the high

\ s, g;treg, point and water flow observed.

3.5 / 4.5-7

QUAD-CITIES DPR 29

2. The discharge pipe pressure for the 2. Following any period where the LPCI systems in Specification 3.5.11.1 shall mode of the RHR or core spray ECCS be maintained at greater than 40 psig have not been required to be opera 6te, and less than 74 psig. If pressere in the discharge piping of the inoperable any of these systems is less than 40 system shall be vented from the high psig or greater than 74 psig, this con- point prior to the return of the system dition shall be alarmed in the control to service.

room and immediate corrective action

3. WM ver the llPCI or RCIC system ~

taken. If the discharge pipy pressure is is lined up to take suction frorn the not withm these hmits in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months torus, the discharge pipinn of the after the cccurrence, an orderly shut- ~

down shall be imtiated, and the reac- IIPCI and RCIC shall be vented from the high point of the system and water-tor shall be m a cold shutdown condi-6 okrved on a monthly basis.

tion within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after mittation.

4. The pressure switches which monitor the discharge lines to ensure that they are full shall be functionally tested every month and calibrated every 3 months. The pressure switches shall be set to alarm at a decreasing pressure of 240 psig and an increasing pressure of

s74 psig.

I. Condensate Pump Room Flood Proicction I. Condensate Pump Room Flood Protection k_ l. The systems installed to prevent or 1. The following surveillance require-mitigate the consequences of flooding ments shall be observed to assure that of the condensate pump room shall be the condensate pump room flood pro-operable prior to startup of the tcction is operable.

a. The piping and electrical penetra-

2. The condenser pit water level switches tions and bulkhead doors for the shall trip the condenser circulating vaults containing the RHR 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 1 2 psig and these trip and alarm circuits. the failed checking for leaks using a soap circuit shall be immediately placed in bubb!c 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 floor drains from the vaults

3. If Specification 3.5 I 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 as<uring 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 RilR service water pump and withm 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

diesel generator cooling water pump bed plate drains shall be k- checked during each operating 3.5 / 4.5-8

QUAD-CITIES DPR-29 cycle run throughby the assuring drain linesthat w. ste 3,,

.~ttuating the air-operated v%

by operal;on Of the follow;n',,

xnsors;

.I) loss of air

2) eqttipment drain sump high Icvel 3)1 vault high level .

, d.' The condenser pit 5-foot trip cir. -

, . cuits for each channel shall tx-

' checked once a month. A lo2ic system functional test sha!! b: p':r. i

' form: 1 during each refu: ling -

.ouugt ,

J. Average Planar Linear Heat Generation Rate J. Average Planar Linear }Icat Generation Rage (APL11GR) . (APLHGR)

During steady-state power operation, the aver- The APLliGR fo'r each type of fu:I as a fune-age linear heat generation rate ( APLiiGR) of lion of average planar exposure shall be deter-all the rods in any fuel assembly, as a function mined daily ducing reactor operation n cf average planar exposure. at any axial loca- 2: 25% rated thermai power.

tion. shall not exceed the maximum averate planar Ll!G R shown in Figure 3.5-1 (3 she:ts). ,

If at any time during operation it is det:rmined ' ,

by normal surveillance that the limiting value for APLHGR is being exceeded, acdon shall be initiated within 15 minutes to restore operation .

to within the prescribed limits. If the APLIIGR is not rettimed in within the prescribed limits '

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the reactor shall be broucht to the cold shutdown condition within 36 iiours.

- 5ttrveillance and corresponding action shall

' continue until reactor operation is within the . .. ,

pres:ribed limits. .

K. !.n:al 1.H G R K. Incal Ll!GR - '

During steady-state power operation. the linear Daily during steady-state power operation

.. he.0 generation rate ( LHG R ) of any rod in any _above 25% of rated therrnal power. th: local

-fuel assembly at any axial location shall not LilGR shall be checked.

eweed the maximum allowable LHGR as cal-culated by the following equation. If. at any ,

time during operation it is deterrained by nar-raa! surveillance that the limiting va!ue for LHG R is being exceeded, action shall.be initi-tied witnin 15 minutes ta restore operation to wiMun the prewribed limits. If the LHG R is not returned to within the prescribed limits within.

3.5 / .l.5 -9 u - : :,-- m.. =. ,. _ _ . - , . _ . . - . . - . . _ . - . . , . _ . . . . . _ . _ _ _ __ __

QUAD-CITIES DPR-29 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Surveit-lance and corresponding action shall continue until reactor operation is within the prescribed limits.

LilG R,, LliGR,1 -( AP/P),,(L/l y) 1.IIG R, = design LilGR where:

= 17.5 kW/ft, 7 x 7 fuel assemblies

= 13.4 kW/ft, 8x8 8 x SR fuel essemblies

( AP/P),, = maximum power spiking penalty

= .035 initial core fuel

= .029 reload 1, 7 x'7 fue!

= .022 reload, 8 x 8 fuel

= .028 reload I, mixed oxide fuel

= .000 reload, 8 x 8R fuel assemblies L, -

= total core length

= 12 feet L -

= Axial distance from bottom of core l L. Minimum Critical Power Ratio (MCPR) L. Minimum Critical Power Ratio (MCPR) l During steady-state operation MCPR shall be The MCPR shall be determined daily during greater than or equal to steady-state power operation above 25'" of rated thermal power.

1.29 (8 x 8 fuel) 1.32 (8 x 8 BLTA) at rated power and flow. If at any time during operation it is determined by normal surveil-lance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the pre-scribed limits. If the steady. state MC.'R is not returned to within the prescribed limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Surveil-lance and corresponding action shall continue until reactor operation is within_th.e prescribed ~~ ~~ ~' ~-

limits. For core flows other than rated, these nominal values of MCPR shall be increased by a factor of L,, where L, is as shown in Figure 3.5 -2.

3.5/4.5-10 Amendm-nt No. 50

QUAD-CITIES DPR-29

(

3.51.lMITING COND! TION FOR OPERATION HASES A &F1 Core Spray and LPCI Mode of the ItllR S 3stem l This specification assures tnat adequate emergency cooling capability is available whenever irradiated fuel is m the reactor vessel.

Ba:cd on the loss-of-coolant analytical methods descrPxd in General Electric Topical Report NEDO 20566 and the specifie analysis in NFDOO4146. " Loss-of-Coobnt Analysis Report for Dresden Units 2,3 and Quad.

Uties Units 1,2 Nudear Power Stations, S ptember 1978, core coobng systems provide sufficient cooling to th? core to diaipate the eneyy associated with tht loss-of-coolant accident,tolimit calculated fuelcladdingtem-per ature to less than 220f F, to assure that core geometry remains intact, to limit cladding metal water re-action to less than 1%, and to limit the calculated local metal-water reaction to less than 17%.

The limiting conditions of operation in Specifications 3.5.A.1 through 3.5. A.3. and 3.5.B.1 through 3.5.B.4 specify the combinations of opor5ble subsystems to assure the availability of the niniraum cooling systema noted above. 1:o single failure of Eccs equipment occuring during a loss-of-coolant accident under those liraiting conditions of operation will recult in inadequate cooling of the reactor core.

Core spray distribution has been shown, in full-scale tests of systems similar in design to that of Quad-Citics I and 2. to exceed the minimum requirements by at least 25% in addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods ao duplicate the decay heat characteristics of irradiated fuel. The accident analysis is additionally conservative in that no credit is taken for spray cooling of the reactor core before the internal pressure has fah?n to 90 pig.

{

The LPCI mode of the RHR system is designed to provide emergency cooling to the core by flooding in the event of a los-of-coolant accident. This system functions in combination with the core spray system to prevent excessis e fuel cladding te nperature. The LPCI mode of the RHR system in combination with the core spray subsystem provide.. .idequate cooling for break areas of approximately 0.2 ft' up to and including 4.18 ft', the 1.itter being the double ended recirculation line break with the equalver line between the recirculation loops closed without assistance frem the high-pressure emerEency core cochng su bsptems.

The allowable repair tiraes are established so that the average risk rate Ihr repair would be no greater than the basic risk rate. The riethod and concept are described in Reference 1. Using the results developed in this reference the repair period is found to be less than half the test interval. This assumes that the core spray subsptems and LPCI constitute a one-out-of-two system; however, the combined etreet of the two systems to limit excessive cladding temperature must also be considered. The test interval specified in Specification 4.5 was 3 mo,ths. Therefore, an allowable repair period which maintains the basic risk considering single failures saould be less than 30 days, and this specification is within this period.

Although it is recognized that the information given in Reference 1 provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach prevents conplete acceptance of this method at this time. Therefore, the times stated in the specific items were established with due regard to judgment.

Should one core spray subsystem become inoperabic, the re-maining core spray subrystem and the entire LPCI mode of the RHR system are available should the need for core cooling arise. Based on judgments of the reliability of the remain-ing systems, i.e., the core spray and LUCI, a 7-day repair period was obtained.

3.5/4.5-11

QUAD-CITIES DPR-29

(

Should the loss of one RIIR pump occur, a nearly full complement of core and containment cooling equipment is available.Three RH R pumps in conjunction with the core spray subsystem will perform the core cooling function. Because of the availability cf the majority of the core cooling equipment, a 30-day repair period is justified. If the LPCI mode of the RilR system is not available, at least two RIIR pumps must be available to fulfill the containment cooling function. The 7-day repair period is set on this basis.

, C. RIIR Senice Water l

The containment cooling mode of the RIIR system is provided to remove heat energy from the containment in the event of a loss-of-coolant accident. For the flow specified, the containment long-term pressure is limited to less than 8 psig and is therefore more than ample to provide the required heat-removal capability (reference SAR Section 5.2.3.2).

The containment coating mode of the RIIR system consists of two loops, each containing two RilR service water pump,, one heat exchanger, two RIIR pumps, and the associated valves, piping, electrical equipment, and instrumentation. Either set cf equipment is capable of performing the containment cooling functmn. Loss of one RilR service water pump does not seriously jeopardize the containment cooling capability, as any one of the remaining three pumps can satisfy the cooling requirements. Since there is some redundancy left, a 30-day repair period is adequate. Loss of one loop of the containment cooling mode of the RIIR system leaves one remaining system to perform the containment cooling function.

l Based on the fact that when one loop of the containment cooling mode of the RiiR system becomes inoperable, only one system remains, a 7-day repair period was specified. l

, D. liigh-Pressure Coolant Injection l

The high-pressure coolar; injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI mode of the RilR system or core spray subsystems can protect the core.

The llPCI meets this requirement without the use of offsite electrical power. For the pipe breaks for which the llPCI is intended to function, the core never uncovers and is can'inuously cooled, thus no cladding damage occurs (reference SAR Section 6.2.5.3). The repair times for the limiting conditions of operation were set considering the use of the llPCI as part of the isolation cooling system.

E. Automatic Preuure Relief Up.on failure of the HPCI to function properly after a small break loss-of-coolant accident, the ADS automatically causes the safety-relief valves to open, depressurizing the reactor so that flow from the low pressure cooling systems can enter the core in time to limit fuel cladding temperature to 1 ,s than 2200 F. ADS is conservatively required to be operabic whenever reactor vessel pressure exceeds 90 psig even though low pressure cooling systems provide adequate core cooling up to'350 psig.

.I

r. RCIC i

The RCIC system is provided to supply continuous makeup water to the reactor core when the reactor

, is isolated from the turbine and when the feedwater system is not available. Under these conditions the pumping capacity of the RCIC system is sufficient to maintain the water level above the core without any other water system in operation. If the water level in the reactor vessel decreases to the RCIC initiation level, the system automatically starts. The system may also he manually initiated at any time.

3.5/4.5-12

QUAD-CITIES DPR-29

(

The HPCI system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable. Therefore, the specification calls for an operability check of the llPCI system should the RCIC system be found to be inoperable.

G. Err.ergency Cooling Availability l

The purpose of Specircation3.5s is to assure a minimum of core cooling equipment is available at all I times. If, for example, one core spray were out of service and the diesel which powered the opposite core spray were out of service, only two RilR pumps would be available. Likewise,if two RHR pumps were out of service and two RHR service water pumps on the opposite side were also out of service no containment cooling would be available. It is during refueling outages that major maintenance is performed and during such time that all low-pressure core cooling systems may be out of service. This specification provides that should this occur, no work will be performed on the primary system which could lead to draining the vessel. This work would ine'ude work on certain control rod drive components and recirculation system. Thus, the specification precludes the events which could require core cooling.

Specification 3.9 must also be consulted to determine other requirements for the diesel generators.

Quad-Cities Units I and 2 share certain process systems such as the makeup demineralizers and the radwaste system and also some safety systems such as the standby gas treatment system, batteries, and diesel generators. All of these systems have been sized to perform their intended function considering the simultaneous operation of both units.

These technical specifications contain only a single reference :o the operability and surveillance requirements for the shared safety-related features of each plant. The level of operability for one unit must be maintained independently of the status of the other. For example, a diesel (1/2 diesel) which is shared between Units I and 2 would have to be operable for continuing Unit 1 operation even if Unit 2 were in a cold shutdown condition and needed no diesel power.

~

Specification 3.5,G3 provides that should this occur, no work will be performed which could preclude [

adequate emergency cooling capability being available. Work is prohibite ! unless it is in accordance with specified procedures which limit the period that the control rod drive housing is open and assures that the worst possible loss of coolant resulting from the work will not result in uncovering the reactor core.

Thus, this specification assures adequate core cooling. Specification 3.9 must be consulted to determine other requirements for the diesel generator.

II. Maintenance of Filled Discharge Pipe i

If the discharge piping of the core spray, LPCI mode of the RHR, HPCI, and RCIC are not filled, a water hammer can develop in this piping, threatening system damage and thus the availability of emergency cooling systems when the pump and/or pumps are started. An analysis has been done which shows that if a water hammer were to occur at the time emergency cooling was required, the systems would still perform their design function. However, to minimize damage to the discharge systems and to ensure added margin in the operation of these systems, this technical specification requires the discharge lines to be filled whenever the system is in an operable condition.

Specification 3.5SA provides assurance that an adequate supply of coolant water is immediately l available to the low-pressure core cooling systems and that the core will remain covered in the event of a loss-of-coolant accident while the reactor is depressurized with the head removed.

3.5/4.5-13

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

QllAI)-CITil?S I)PR-29 I. Condensate Pump Room Flood Protection l

See Specification 3. 5. I .

l J. Aserage Planar LilGR g This specineation assures that the peak cladding temperature following the postulated design-basis lossof<oolant a:cident will not exceed the 2200 F limit speciGed in the 10 CFR 50 Appendix K considering the postulated effects of fuel pel!ct densification.

The peak cladding temperature following a postulated loss-of<oolant accident is primarily a function of the average heat. generation rate of all the rods of a fuel assembly at any axiallocation and is only secondarily dependent on the rod to-roJ power distribution within an .,ssembly. Since expected local variations in power distribution within a fuel assembly affect the calculated peak cladding temperature by less than i20 F relative to the peak temperature for a typical fuel design, the limit on the average planar LilGR is suf-ficient to assure that calculated temperatures are below the limit. The maximum average planar LIIGR's shown in Figure 3.5-1 are based on calculations employing the models described in Reference 2.

K. local LIIGR g This specification asseres that the maximum linear heat-generation rate in any rod is less than the design linear heat-generation rate even if fuel pellet densification is postulated. The power spike penalty specified is based on that presented in Reference 3 and assumes a linearly increasing variation in axial gaps between core bottom and top and assures with a 95% confidence that no more than one fuel rod exceeds the design linear heat-generation rate due to power spiking. An irradiation growth factor of 0.25's was used as the basis for determining A/P in accordance with References 4 and 5.

b1 s/ L. Minimum Critical Power datio (MCPR) l The steady state values for MCPR specified in this specification were selected to provide margin to accommo-date transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical power correlation itself. These values also assure that operation will be such that the initial condition assumed for the IDCA analysis, an MCPR of 1.18,issatisfi:d. For any of the special set of transients or disturbances caused by single operator error or single equipment malfunction,it is required that design analyses initialized at this steady state operating limit yield a MCPR of not less than that specified in Specification 1.1.A at any time during the transient, assuming instrument trip settings given in Specification 2.1. For analysis of the thermal consequences of these transients, the limiting value of MCPR stated in this specification is con-servatively assumed to exist prior to the initiation of the transients. The results apply with increased con.

servatism while operating with MCPR's greater than specified.

The most limiting transients with respect to MCPR are generally:

a) Rod withdrawal error b) Turbine trip without bypass c) Loss of feedwater heater Several factors influence which of these transients results in the largest reduction in critical power ratio such as the specific fuel loading, exposure, and fuel type. The current cycles reload licensing submittal specifies the limiting transients for a given exposure increment for each fuel type. The values specified as the Limiting Condition st Operation are conservatisely chosen as the most restrictive over the entire cycle for each fuel type.

i A

.~

3.5/4.5-14

~

DPle-29 ~

4.5 SunyEILLANCE ImQlIIIEMENTS !!ASES ,

'Ihe testin'o i[t rval for the core and containment coo!in3 syste:ns is based on a quantitative reliability endy judomeni, .kui! practiculi:y. The core cooling systems have not been designed to ba fully testable during o Fo[examp?e..the core spray final admission valves do not open until reactor pressure has fa!!en to 350 psig.T during oper: don, even if high drywell pressure were simulated, the final valves would not open. In the case o .

IIPC[, automatic'initircon during power opuation would result in pumping cohl water into the re2ctor vessd .

whis is not d.esirable. ,

'The system.E6-bFautomatically actuated dering a refuding cutage and this will be done. To increase the avai!2bility or theindividual components of the core and containment cooling systems, the componen3 which snake .up tht.iystem; i.e., instrumentation, pumps, valve' operators, etc., are tested more frequen!!y. The ,

instrar.:enta(orr,iy functionally tmted each month. The pda ps and valves pill be tested in accordance i Vith the Inservice T&; ting Program to. insure operability. This program which is based 5 on ISME Code,Section XI requirements prc,vides adequate assurance that tie core and I contairuaent cooling systcas vill be operable when required. When co aponer.ts or sub-systens are taken out of service, surveillance tests of the rentaining redu! dant -

f conponents or subsysteras is not required because of the increased risk that a single j

' failure occurring during this testing will result ^ in a loss of total system i'metion. d This is justified by the additional testing requircraents of the 76ME Code whic't providen adequate assurance that the pur,ps and. valves vill be operable when called upon Po wer- ^'

form their specified function. The limiting conditions for operation ensure thac

. U redundancy is naintained.at all tines. '

E n: . crift ati r:cf.theimaut.stcamadi,r va1e ep--2bilitY duSe mmed actuation surveithnee testing rnust h made

. a -.

uant .ce pa~.., -- s indicatca by thermorouples do c,.. t . .. o.r relief valv'es. It has been'foun'. thr_t a rteera.a c . -. -

d # dod This is de: to 3 eam beh.g vented t,u:ough tne pdot vdus t -@rature iner:dsunay result.witn th -

,, _ , g . - d * - W r Y

Salve :..m u a tio n; po n,t ~ m;[; ,ti . n ber i: Io i 3 2C l P-n-

the turo.ne contro1 yahe. durin.g relief vaive manu21 actu1tton wouy

~ g.,sec.t., -

syy., ,,s an aderu2te verif::atton :or tne

- i tor pressures greater tn:n 120 pg..

rM)vahi a ophiing This test method may b: perforrmd ove., a wi3e r.,.-% dYd.; o ,

~~ -

TF 5"veillin&re'quirements to ensure that t.ne d.ischarge pipm3 o' rp. co~. spray LPCI rnode of the RHR,HPCL - -

a .

This ensure 3 i9

?nd RC'IC system's is titled lprovides for a visual observation t1 tat . ater nows from a h,t.gn .notn vent ,

that U line is in a tul[ conditton. In drumentation aas Le n provhled r. > incaitor then1 ressurt- oi. ua te;, 2.n t,r. .: !

~ i dinh5r7 "9 P I 'Nibberdeen the quarterly intirvals lt w!.ich the lines ne wnW and ab.m tu conm! mm e , ,,

Tn. -is Instrumentatlon V111 %-- calibrated on the same f re-the pressure is inadequate. .

quency an the safety systen . instrumentation and th'e alarm system tested .no n uly. Anis ter.d i re ensures that, during the interval between the quarterly venting checks, the st1tus ,

of thu discharge piping is nonitored on a. continuous bu .is. .

m An alarm point of 40 psig for the low pressure of the till system has been chosen because, due to elevations of piping within the plant. 39 psig is required to keep the lines full. Ihe shutolThead of the till system pumps is 74 psig and therefore will not defeat the low-pressure cooling pump discharge press interlock of 75 psig as shown in Table 3.2 2.

t 3.5/4.5-36

I

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

Www-cinis s

2. 110th the sump and air sampling sys. 1)PR-29 terns shall be operable during reactor ,

power operation.1;rorn and erger the date that one of these systerr.s is made or found to be inoperable for any rea-son, reactor power operation is per-missibic only during the suceceding 7 days.

3. If the conditions in I or 2 above can-not be inet, an erderly shutdown shall be initiated and the reactor shall ue in

. a co!d shutdown condition within 24

. . . . hours. -

E. Saf:ety and Relief Valves ,- .

E. safety and nelief Valves

1. The pressure set points of the

1. Prior to reactor startup for power op- safety and relief valves shall be

i. ration, during reactor power operat- . checked in accordance.with the- '

ing conditions, and whenever the reac- . requirements of the Innervice -

tor coolant pressure is greater than 90 Testing Program for Pumps and

. psig and temperature greater than Valves defined in Section 1.0(P.F)

, 320* F, all nine of the safety valves of these Technical Specifications.

shall be operable. The solenoid-activated pressure valves shall be oper- 2. The set points for the safety valves able as required by Specificatica shall be as follows:

. _- l 3.5. E . i

-2r If Specification 3.6.E.1 is not met, the Utmtber of Valves Set Point (psig) E reactor shall remain shut down until 1 1115* I

. the condition is corrected or, if sn 2 1240 l

- operat .n, an orderly shutdown shall 2 1250 be initiated and the reactor coolant 4 1260 pressure and temperature shall be below 90 psig and 320* F within 24 The allo.rable set point error for each

. hours. valve shall bc ,t it.

3. The set points for the relief valves shall be as follows:

Humber of valves set Point (nsig)

J

'g 1 1115*

) 2 41130 2 (1135

Target rock co;abination safety / relief valve.

F. Structural Integrity F. Structural Integrity The structural integrity of the primary sys:cm The nondestructive examinations specified '

boundary s* tall be .naintained at the level re- in the Innervice Inspection Progran quired by the AS'.!E 1:oi!:r ca.! Pressure Vesel defined in Section 1.0 (G.G) shall be Code,Section XI, "Pt.!cs for laservire laspectica conducted in accordance uith applicable of Nuclear . Pcwcr Plant Com; onents". - Edition and Addenda of Section XI of the ASM2 noiler and Pressure Ver ael Code .",

required by 10CFR50, section 50.55a (g) .

In addition, t he welds listed in Tabic 4.6.1 nhall be e>:amined at tlu-frequency unecified.

3 f./ 4.6-4 I.

QUAD-CI"h11'S IJPR-2

-.I the boundaries of the reactor, are exceeded. Methods availab!c to the operator for correcting the offstandard condition include operation of the reactor cleanup systeia, reducing the input of impurities, and placing the reactor in the cold shutdown condition.The major f.enefit of cold shutdown is to reduce, the teraperature-dependent corrosion rates and provide time for the cleanup system to reestablish (Fe purity of the reactor coolant. During startup periods, which are in the category ofless than 100,000 lb/hr.

conductivity may exceed 2 pmho/cm because of the initial evolution of gases and the initial addition of dissolved metals. During this period of time when the conductivity exceeds 2 pmho (other t_han short-term spikes), samp!cs will be taken to assure the chloride concentration is less than 0.1 ppm. -

The conductivity of the reactor coolant is continuously monitored.The sampics of the coetant which are taken every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months will serve as a reference for calibration of these monitors and are considered adequate to assure accurate readings of the monitors. If conductivity is within its normal range, chlorides end other impurities .ci!! at;c ba wnhin their normal ranges.The reactor coolant samples will also be used to determine the chlorides. Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content. Isotopic analyses required by Specification 4.6.C.I.b may be performed by a gamma scan.

- . . t; ..

D. Coolant Leakage Allowable leakage rates ofcoolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes and on the ability to make up coolant system leakage in the event of loss of offsite a-c power. The normally expected background leakage due to equipment design and the detection capability for determining coolant system leakage were Ao considered in establishing the limits.The behavior of cracks in piping systems has been experimentally and analytically investigated as part cf the USAEC sponsored Reactor Primary Coclant System Rupture Study (rhe Pipe

N , Rupture Study). Work utilizing the data obtained in thk study indic2!:s th2!leahage f cm a ;7;ck can d 'be detected before the crack grows to a dangerous or critical size bf'inechanically or ther'm' ally in:!uced cyclic loading, stress corrosion cracking, or some other mechanism characterized by gr: deal crack growth. This evidence suggests that for leakage somewhat greater than the limit specified far unMentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly. . . . , . .. m, llowever, the establishment of allowable unidentified leakage greater th'an that given in Specification 3.6.D, on the basis of the data presently available would b~e premature because of uncertainties associated with the data. For Icakage of the order of 5 gpm as specified in Spectrication 3.6.D, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation. Leakap less than the magnitude specified can be detected reasonably in a matter of few hours utilizing the availab!ileaKage detection schemes, and if the orgin cannot be determined in a reasonably short time, the plant should be shut down to allow further investigation and corrective action. ,

The capacity of the drywell sump is 100 rpm, and the capacity of the drywell equipment drain tank pumps is also 100 gpm.

Removal of 50 rpm from either of these sumps can be accomplished with considerah!e margin.

The performance of the reactor coolant leakage detection system will be evaluated during the tirst 2 vears of station operation, and the conclusions of this evaluation will be reported to the NRC.

E. Safety and llelief Yahes The frequency and testing requirements for the safety and relief valves are specified in the Inservice Testing Program which is based on Section XI of the AStiE Itoiler and Pressure Vessel Code. Adh,rence to these code requirements provides adequate assurance as to the proper vperational readiness of these valves. The tolerance value in specified in Section III of the ACME P, oiler and Pressur e Vessel Code as +,1% of design pressure. An analysis has been performd which shows that with all safety valves set It higher the reactor coolant preasure safety limit of 1375 psig is not exceeded. The safety valves are required to be operable above the design pressure (90 psig) at which the core spray subsysteras are not designed to deliver full flow.

3.6/4.6-11

QUAD-CITlES DPn-29 P. Structural Integrity A pre-service inspection of the coraponents in the primary coolant pressure boundary will be conducted after site erection to assure the system is free of gross defects and as a reference base for later inspections. Prior to operation, the reactor primary system will be free of gross defects. In addition, the facilit.y has been designed such that gross defects should not occur throughout life.

Innervice Inspection of ASE Code Class 1, 2 and 3 components vill be perforraed in accordance with the applicable version of Section XI of the ASME Eoiler and Pressure Vessel ~ Code. Relief from any of'the above require-ments has been provided in writing by the Commission. The Innervice.Inspec-tion program and the written relief do not f;orm

a part of these Technical Specifications.

The speb.!' inspection of the main feed and steamlines is to provide added protection against pip-e whip,in addition 1to lhe protective energy absorbin; system to be insta!!ed insid: the d:yv,cil as described in Amend.

ment 27 to the.SAR. The Group I welds are selec:ed on the basis of an analysis that shows these w:!ds era the hi;Eest stress welds and that due to their physicallocation, a break would result in the lent interference and inaximurn energy upon impact with the drywe!!. These weids are the on!y ones which offer any si;nifi-cant riskand.are thereforc inspected 4 times as often as the other welds within the drywell.

Group I,Ijkelds'are selected because without regard for the operating stress Icvels and interferin;cquipment, they hafe.,s.fdcient theoretical energy to penetrate and would propel the pipa .oward the containment. They are thereforeincluded in first inapectica. Upon consideration ofimpact ange,inte:fering equipment, and the distance pipe travels, no substantial risk is involved and no c'xtra inspection is needed.

In addition, extens've visual inspection for 12:ks will be made periodically on critical systems. The inspection pro; ram specified encornpasses the major areas of the vessel and piping systems within the drywell. The inspection period is based on the observed rate of growth of defe:ts from fatigue studi:s sponsored by the NRC. 'Ihese studi:s show that it requires thousands of stress cycles at stress oeyond any e.spected to occur '

in a reactor sys!em to propagate a crack. The test frequency establiAed is at intervals such that in com-parison to study results only a sma'l number of stress cyc!cs at val:es'clawlimits will occur. On t!ds basis, itis considered that the test frequencias are adequate.

o 3.6 /.t.6 - 12

UPit-19 i

1ABLE 4.6-1 Itain Stearnfine . Creup I Wclds Group fl We:ds -

Weld IJentifi. Weld l$enti'i. -

Ur.e catica Ca;t 1 Uae

, catica Unit 1 3301A-20-h. 30A-522

e .

30A F23 3001A-20 h. 30A-511 304 F24 , '

3001B 20 h. 308-510 30010-23-h. 308 525 3001C-20 in. 300-510 308-F26

, 30010 20.h. 300-510 308-F21 3TilC-23-in. 30C 521 30C-F22 3]C F23

30310-23 in.

300-520 3CD-F21 300-F22 Feedrater Ua: Crcup i We!ds Creep 11 We'ds Weld Identifi. WelJ llantifi.

tiae catica Unit 1 Uce .

catica Unit 1 32XA-18-b. 32A-55 32MA-13-b- 32A-SI 32MS la-in. 328-34 32A F3 32A F~/

32M3-IS-h. 323 51 328-16 32MC-12 in. 32C-54

. 32MD 12-in. 32D S4

, 320-53 .

32D-F9 3204E-12" 32E-S2 3204F-12" 32F-S2 32F-F6 Frequencu of Inspection:

a. Group-I Welds All welds will be inspected each 10 year interval with 25% i completed at approximately 2.5 year intervals.

B. Group II Welds All welds will be inspected during the first 10 year interval with 25% completed at approximately 2.5 year intervals.

b

}

3.6/4.6-16

. ~ - - - - . , - -

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

Quad-Cities DPR-29 DELETE PAGES 3.6/4.6-18 TIIRU 21 3.6/4.6-17 ,

DPR-29 -

O r

b. The tcactor water temperature is b. Additional. tests shall be per-below 212' F and the reactor ' formed during the nrst oper; ting coolant systems are vented. cycle ender an adequate number of dilTerent environraental wind conditions to enab!c valid er.trapo-lation of the test re;cits.

c. No activity is being performed c. Secoadary centainment capability which can reduce the shutdown to maintaire an average 1/4 inch of margin below that speciGed in w::er vacuum urJer caint w nid -

Spccification 3.3.A. -

(2<ii<5 mph) conditions with a

d. The fuel cask or irradiated fuel is E *' l' I" 0 * '"'# C I " ! ""'i' not being moved in the reactor than 4000 cfm sh;di be demonstrated

- building. at each refucHng outar.e prict to refuclirc.

2. The doors of the core spray and RilR 2. Wheneve'r the LPCI mode of the RHR pump compartments shal! be closed at and core spray subsystems are re-all times except during passage in or- quired to be operable, the doors of the der to consider the core spray system core spray and RHR pump compart-and LPCI mcde of the RIIR system ments shall be verified to be closed operable, weekly.

3. Ir Specircation 3.7.C.I canner he n-t procedurn shnll be inirinted to estab-lish conditions listed in SpeciScations D. Primary Containment Isolation Valves 3.7.C.1 a through d.

1. The operational readiness of the D. I rirnary Containment Isc,lation Va!yes primary containment isolation valves shall be demonstrated in

1. During reactor power operating con- accordance with the Inservice ditions, all isolation va!ves listed in Te, ting Prograra for pumps and Table 3.7-1 and all instrument line valves defined in Section 1.O (F.F) flow check valves ivhich contact the of these specifications. Addi-primary coc!:nt sprem shall be opera- tional requirements are as follows:

b!c cxcept as specined in Specincation.

3.7.D.2. a. The operable isolation valves that are power operated and automatically initiated shall be tested for simulated l auton.atic initiation, at least, once per operating cycle.

b. The instrument line ficw check valves shall be tested for proper operation, at least once per operating cycle.

O

3.7 / 4.7 -9 -

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

s t .

2. In the event any isolati> n valve speci-

c. When an isolation valvc listed in Ta-lied in Table 3.7-1 becomes inopera- ble 3.7-1 is inoperable, the position of

, ble,' reactor power operation may con- at least one other v t!ve in each line tinue provided at least one valve in .having an inoperable valve shall be cach line having an inoperable valve is recorded daily.

in the m'ode corresponding to the ito-lated condition.

3. If Specifications 3.7.D.I and 3.7.D.2 '

cannot be met, an orderly shutdown shall be initiated and the reactor shall

. be ;n the cold shutdown condition wthin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

4. The temperature of the main steam- ,

line air pilot valves shall be less than 170 F cxcept as specified in Specifi- _

cations 3.7.D.5 and 3.7.D.6 below. - ,

3. From and after the date that the tem-perature of any main steamlin,e air pilot valve is found to be greater than 170* F. reactor operation is permissi- '

ble only during the succeeding 7 days unless the temperatu:c of such valve i:. '

sooner reduced to less than 170 F, '

provided the main steamline isolation ,

valves are operable.

~

6. If Specification 3.7.D.5 cannot be me't. .

the main steamline isolation valve '

shall be considered inoperable' and action taken in accordance with Speci-fication 3.7.D.2.

3.7/4.7-10 ,

8*

I U6' WW T9.m _ WeW gwweguggy, _ _,e y , _, _, (

- 3 QUAD-CITIES DPit-29 will la rep! aced with blters qualifi:d fursu:nt to regulatory guh!c position C.3.d of Regulatory Guide 1.52 Revision I (June l'176). Once per operating cycle dem:mstration of !!FPA fdter pressure drop,op:rabiHty ef intet heaters ::t rated power, air distribution to each IIEPA filter,and automatic initiation of each standby gas treatment system circuit is necessary to assure system perform:nce cagubility). Note: bases within parenti. ems will not bc rpp;icab!c until about D:ccmber 31,1976, when equipment modifications are completed to allow increased testin;. ,

. D. Primary Cuntainment holation Vahes

Those large pipes comprising a portion of the reactor coolant system, whose failure could result in -

uncovering the reactor core, arc supplied with automatic isolation valves (except those lines needed for emergency core coating system operation or containment cooling ). The closure times speci ied herein are adequate to prevent loss of more coolant from the circumferential rupture of any of these lines eutside the containment than from a steamline rupture. Therefore, this isolation valve c!osure time is sufHcient

.to prevent uncovering the core.

In order to assure l' at the doses that m :y result from a steamline break do not exceed the 10 CFR 100 guidelines,it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steamline isolation valves. Analyses indicate that fuel rod cladding perforations would be avoided for main steam valve closure times, including instrument delay. as long as 10.5 seconds.

, However, for added margin, the technical specifications require a valve close time of not greater than 5 seconds.

For reactor coolant system temperatures less than 212

F, the containment could not become pressurized due to a loss-of-coolant accident. The 212' F Jimit is based on preventing pressurization of the reactor

. building and rupture of the blowout panels. These valves are highly reliable. have low service requirement, and most are normally closed. The initiating sensors and associated trip channels ar: u%

checked to demonstrate the capaidlity for automatic isolation (reference S/sR Section 5.2.2 and Taum 5.2 A ). -

The test interval at once per operating cycle for automatic initiation results in a failure probability of 1.1 x 104 that a line will not isolate. More frequent testing for valve operability resuhs in a more reliable sy stem..

The containment is penetrated by a large number of small diameter instioment lines which contact the primary coolant system. A program for periodic testing and eumination of the flow check valves in these lines is performed by blosying down the ins: ument line during a vessel hydro and observing the fo!!owing two conditions, which will verify thu ..e flow check valve is operable:

3.7/4.7-18

QUAD-CITIES 1

D PR-29, unless an additional line is sooner placed in scivice, providing both the Unit and Unit 1/2 cmergency diesel generators are operable.

2. From and artcr the date the incoming power is not available from any line, continued reactor operation is permis-

'sible providing both the Unit and Unit i

I/2 emergency diesel generators are -

l operating, all core and containment

, cooling systems are operable, reactor -

i power level is re'duced to 40% of rated. .

and the NRC is notified within 24 -

hours of the situation, the precautions to be taken during this period, and the .

plans for prompt restoration ofincora-ing power -

3. From and efter the date that one of the two 125/250-volt banery systems is made or found to be inoperable for any reason, cer.tinued reactor opera-tion is permissible only during the succeedinn 3 days unless such batterv

~

~ '

system is saoner made operable. .

D. Diesel Fuel . D. Diesel Fuel There shall be a minimum of 10.000 galfor.. af Once a month the quantity of diesel fuel availa-diesel fuel supply on site for each diesel ble shall be logged. ,

E'"*' I I"

,Once a month a sample of diesel fuci shall b:

checked tor quality.

E. Diesel-Generator Cperability '

E. Diesel-Generator Operability

l. Whenever th- reactor is in the Star- V/ hen it is determined that either the unit or tup /Ilot Standby or Run mode and shared diesel generator is inopecable. all low-the unit or shared diesel generators pressure core coo!ing systems ar.d al! loops of and/or their respective asociciated the containment coating modes of the RIIR bus:'s are made or found to be inopera- ystem associated with the operab!: diesei gen-ble for any reason. except as specifwd crator shall be operabl6.

in Specification 3.9 E.2 below, contin-ued reactor operation is permissible ,

only during the succeediny 7 days pro- ,

vided that all of the low-prewure core cooling and all loops of the conta;n- '

ment cooiing mode of the itiiR sys:em associa:cd with the operab!c diesel f,ener.itor shall be operable. and two ofRite lias are asadable. If this re-quirement cannot be inet. an orderly shutdown stall be initiated and the g

3.9 / J.9 .i I

~ __

ENCLOSURE 2 Amended Technical specification Pages Quad Cities Unit 2, DPR-30 The following pages have been revised:

1.0-4 3.4/4.4-1 3.4/4.4-2 3.4/4.4-3 3.5/4.5-1 3.5/4.5-2 3.5/4.5-3 3.5/4.5-4 3.5/4.5-4a 3.5/4.5-5 3.5/4.5-6 3.5/4.5-7 3.5/4.5-8 3.5/4.5-9 '

3.5/4.5-10 3.5/4.5-11 3.5/4.5-12 3.5/4.5-13 3.5/4.5-14 3.5/4.5-15 3.6/4.6-4 3.6/4.6-11 3.6/4.6-12 3.6/4.6-16 3.6/4.6-17 3.7/4.7-9 3.7/4.7-10 3.7/4.7-18 3.9/4.9-3 The following pages have been added:

1.0-5 3.4/4.4-2a The following pages have been deleted:

3.6/4.6-18 3.6/4.6-19 3.6/4.6-20 3.6/4.6-21

DPR 30

~

Y. Shutdown - The reactor is in a shutdown condition vehen the reactor mode switch is in the Shutdown position and no core at:crations arc being perfaimcd.

1. Ilot Shutdown means conJitions as above. with reactor ccofant temperature greater than 21P T.

2. Cold Shutdown means condidons as above, with reactor coa! ant temperature equal to or ! css than 212 ' F. .

7_ Simulated Automatic Actuationy Simulated automatic actuation means applying a simulated signa! to the senior to actuate the cucu:t in que,tton ..

AA. Tot.'l P a' sin;: Factor - The total p$ating factor (TPi-) is the highest prcduct of radial, axial, and local peaking facters simultaneously operative at any segment of fuel rod.

'EB. Transition Sniling - Transition boiling raeans the boiling regime betiteen nuc!cate and film boilinc.'

Tratoition, boiling is the regime in which both nucica:e and lilin boiling cccur intermittently, with neithEr .

type being completely stable. .- .s .

. . .): . -

CC. Critical Power Ratio (CPR) - The critical power rado is the ratio of that assemb!y power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition

. ofintercs: as calculated by application of the GF.XL correlation (reference NEDO-1095S).

4

- DD. Minimum Critical Power Rctio (MCI'P.) - The rninimuin incore critica! power ratio correspondiag ro ihe mest limidng fuel assembly in the cere. .

EE. Suncil:ance Intcival - Each surveillance :equirement shall be performed within the specined surveli- '

Ian'ec interval with: .

a. A runimum allow 2bie exten<im[not to em-ed 25% of th- sm eilhnee imcrv-1.

'. . b. A total maximtim cen.bined in:erval time for any 3 consecutive surveillance intervah not to excee

. . . . . . 3.25 times the spe:iGed surveilhace interval. .

FF.. gservice Testing Prograra For Purps And-Valves ,

- The testing program developed for insewice testing of pumps and valves is written in accordance with section ,XT of the As:c soiler and Pressure vessel Code and applicable Addenda as required by 3 0:'FR50, -

Section 50.55 a(g), except where specific written relief has been granted by the. tmC persuant to lOCFF30, Section 50.55a (g) (6) .(i) .

Certain clarifications regarding the use of the above raentioned .

program are as follows:

. a. The surveillance intervals specified by the Inservice Testing Program shall be performed as specified in Definition -

EC -Surveillance Interval"' with the following clarifications :

. at least once per 7 days Ueckly Monthly at least once per 31 days Quarterly or cvery'3 months at least'once per 92 days semiannually or every 6 no. at least once per 184 days Annually at least once per 366. days

b. !!othing in the AS:tF. Doiler and Pressure Vessel Code shall be construed to superced'e the require:ac,nts of any Technical 1.0-4 .

DPR ,30 Specification. For exemple the requirements under limiting conditions for operation that specify when certain equipment shall be operable takes precedence over the ASMC Boiler and Pressure Vessel Code provision which allows pumps to be tested up to one week after return to normal operation. Also, the Technical Specification definition of OPERABLE does not grant a grace period before equipment that is not capable of per-forming its function is declared inoperable. This takes precedence over the ASME Code provision that allows a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before being declared inoperable.

c. Performance of a Surveillance Requirement within the specified time interval shall constitute compliance with OPERABILITY requirements for a Limiting Condition for Operation and associated action statements unless otherwise required by the specification.

GG. Inservice Inspection Program for Components

Classi fied as ASME Code Class 1, 2 and 3.

The inspection program developed for inservice inspec' tion of components classified as ASME Code Class 1, Class 2, and Class 3 is written in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10CFR50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to lOCTR50, Section 50.55a (g) (6) (i) .

[W

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1.0-5

QUAD-CITIES .

Dl'It-3 0 3.4/4.4 STANDBY LIQUID CONTROL SYSTEM ,

SURVEILLANCE REQUIREMENTS ,

LIMITING CONDITIONS FOR OPERATION Applicability:

Applicability: ,

.The operational readiness of the Standby Lig-Applies to the operating status of the standby liquid uid Control System shall.bc_ insured and demon control system. strated by'the Inservice Testing Program for

. Pumps and Valves defined in Section 1.O(P.F)'

Objective of these Specifications. Additional require-To assure the availability of an independent reactiv- ments are listed below. '

ity control mechanism. ,

To verify the operability of th'e standby' -

licuid control system. - - - - -

SPECIFICATIONS .

A. Normal Operation -

A. Normal Operation ,

The operability of the standby liquid centrol During periods when fuel is in th system shall be verified by performar.cc of the

- reactor the standby.11guld control pg);3, g system shall be operable except when

l. Atleast once per quarter.

l C- the reactor is in the Cold, Shutdown '

Condition and all control rods are

~

Dem.merah. zed water shall be recycled fully inserted and Specification- 4

- to the test ' tank. Pump minimum tTow 3.3.A is met or as specified in rate of 39 ppm shall be verified against -

3.4.B below. a system head of 1275 psig.

B. Operation with Inoperable Components 2. At least or.ce during each operating ,

From and after the date that a redundant Manually initiate the system, excep:

component is made or found to be inoper- the explosion valves and pump solu-able, Specification 3.4.A shall be con- tion in the recirculation path, to dem-sidered fulfilled, and continued opera-onstrate that the pump suction line tion permitted provided that the compo- from the storage tank is not plugged.

nent is returned to an operable condition within 7 days. . Explode two of six charges or two of

- - four charges manufactured in the same batch using the permanent system wir-ing to verify prcper function. Then install the untested charges in the ex-p!osion valves.

Deinineiatiwd water shall be injected

- - via a test connectic.n ir.to the reactor vessel to test that v'a!ves (except espio-sion valves) not checked by ,the recir-culation test are not clogged.

a 3.4 N.4- 1

QUAD-CITIES DPR-30 Test that the setting of the system prosaure relief valves'is between 1400 and.1490 psig.

3. Disassemble and inspect one explosion valve so that it can be established that the valvo is not clogged. Both valves shall be inspected in the course of two operating cycles.

Liquid Poison Tank-Iloron Concentration C. Liquid Poison Tank-Boron Concentration The liquid poison tank shall contain a boron- The availability of the proper boron-bearing bearing solution that satis 6es the volume- solution shall be verified by performance of the concentration requirements of Figure 3.4-1 and fo!!owing tests:

. at c!! times when the standby liquid control system is required tv be operabic and the solu- 1. At least once per month O. ,' ter,peraure sh2!! net be less than the Boron wncentration shaii be deter-

a;'erature pr sented in Figure 3.4-2. mined. In addition, the boron concen-tration shall be determined any time

. If Speci6 cations 3.4.A through C are not met, water or boron are'added or if the an orderly shutdown shall be initiated and the solution temperature drops below the reactor shall be in the cold shutdown condition limits specified by Figure 3.4-2.

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . *

2. At least once per day

. Solution volume shall be checked. -

3. At least once per day The solution temperature shall be checked.

0 .

e*

O 3.4/4.4-2

QUAD-Crt lES wn-30 .

d*"

e 3.4 LIMITING CONDITIONS FO!1 OPEllATION llASES A. The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon. free shutdown assuming that none of the veithdrawn control rods can be inserted.To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of 600 ppm of boron in the reactor core in appro'.imately 90 to 120

, rninutes with imperfect mixing. A boron concentration of 600 ppm in the reactor core is required to bring the reactor from full power to a 3Fo ak suberitical condition considering the hot to cold reactivity swing.

xenon poisoning and an additional margin of 150 ppm in the reactor core for imperfect mixing of the chemical solution in the reactor water. A normal quantity of 3470 gallons of solution having a 13.47o sodium pentaborate concentration is required to meet this shutdown requirement.

. The time requirement (90 to 120 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor fol!owing the xenon poison peak. For a required pumping rate of 39 gpm. the maximum storage volume of the boron solution is established as 4875 gallons (195 gallons are contained below the pump suction and, thercior , cannot be inserted).

Boron concentration, solution temperature, and volume are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Performing pump tests iri accordance with the Inservice Test Progran provides c aquate assurance that the pump will be operable.

The only practical time to test the standby liquid control system is ; ring a refueling outage and by inidation from local stations. Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refuding outage

.c t'nnecess?.ry. A test of explo<ive charp from one tronof3cruting batch is made to assure that the charge

. src satisfactcry. A continua! unk cf the f, ring ciremt continmty as p;cvided by pilot lights ir. tx cc,mrvl room. - -

B. Only one cf the two standbyliquid centrol pumping circuits is needed for proper operation of the system.

If one pumping circuit is foand to be inoperable, there is no immediate threat to shutdown capability, and teactor operation may continue while repairs are being made. Assurance that the remaining system will perform its intended function and that the reliability of the system is good is obtained by completing the tests outlined in the Inservice Test- Program for pumps and valves.

C. The solutinn saturation temperature of 137c sodium pentaborate, by weight, is 59; F. The solution shall be Lept at ! cast 10' F nbove the saturation temperature to guard against baron precipitation.The 10

F margin is induded in Figure 3.3-1. Temperature and liquid level alarms for the system are annunciated in the control room. ,

Once the solution has been made up, boron concentration will not vary unless more boron or more water is added. Level indicatian and alarm indicate whether the solution volume has changed. which might indicate a possible solution concer.tration change. Considering these factors, the test interval ha, been established.

h g

/ .

O -

s k

\ *

\ . 3.4 / 4.4 -3

\

. J 'n.30 ~

' 3.5/4.5 COltE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SunvEILLANCp: IREQUIREMEttTS Applicability: *

' Applicability:

. Applies to the op$ rational status of the emergency cooling subsystems. . The operational readiness of the following' subsystems shall be' demonstrated in accord-

'

s' Objective:

'ance with the Inservic'c Testing Program for Pumps and valves defined in

'To assure adequate coe!ing capability for heat re- Section,1.0 (P.F.) of these 'i'echnical moval in the event of a loss-of-cooiant accident or ^

Specifications. -Additional requirements isolation frorn the' normal'rcactor heat sink. f r each subsystem are listed below: '

Objective: ,

~

To verify the operability o'f the core and containment cooling subsystems.

SPECIFICATIOliS. -

A.- Core Spray Subsystems A. . Core Spray Subsystems g

4:./ --

Surveillance of the core spray sosre. ms shall be performed,'as follows: ~ '

l. Both core spray subsystems shall be 1. Core Spray Subsystem Testing operable whenever irradiated fuel is in

the reactor vcssel and prior to reactor liem INguency startup frem a cold condition except as specified in 3.5.A.2 and *

a. Simu ated auto- Each . -

3.5.c.2. -- rnatic actuation ~refuelin"e .

2. From and after the date that one *

^ 'Y '

of.the core spray subsystems 1 .

. core spray pumps made or found to be inoperable for shall deliver at

.any reason, reactor operation is . .

least 4500 rpm .

permissible only during the #E" "5 * *E*

succeeding seven days.unless such . head correspond-subsystem is sooner made operable. Ing to a reactor provided th'at during such seven vessel pressuM days all active components of the -

OI N P5IS other core spray subsystem and the '

LPCI subsystem and the diesel gen-erators required for operation of such components if no external source of power were available shall be operable.

m.

e 3.5 / 4.5- 1

~ ~ - . -

n-- ,e. ~..e... . w w- .. 4.-..

..-,n..r...~....,...,.....,o.~v.mmw-r-*~~~~~'-"~~'

.. . . - . . . . - - . - ~ 5.4 =.3 s

QUAD-CITIL'S .

DPR-30 .

$ 3. If the.requircraents of 3.5.A cannot be j zac t , an orderly t.hutdown of the reactor

c. Core spray

hedder Ap '

shall'bc initiated and the reactor chall instrumentation be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . . check Once/ day .

calibrate Once/3

. . months * '

, . test Once/3 months

d. Logic system Each refueling l

~ .

functional test outage -

' d, S

9 .

I es em "

I 3.5/4.5-2 .

~

. . . . . . . - . . - . - ~ - ~ - ~--+' '-

QUAD-CITIES .

nen- 30 -

B. INCI flode of PltR D. Surveillance of ISCI l' ode of IUIR.

l1". 'The LPCI mode of the RIIR system 1. Once cAch quarter it shall be verified shall he operable whenever irradiated that three luta pumps deliver at least fuelis in the reactor vessd and prior to 14,500 gpa against a syst c:a head corre-spond_ing to a reactor vessel pressure

. except reactor as startup from ain specified cold condition,2, 3.5.B.

of 20 psig. . .

3.5.B.3 and 3.5.G.2. ~

' ~

.2. A Siraulated Automatic Actuation Test sha

. , .be completed each refueling outcce.

2. From and after the date that one of th- '

Ri!R pumps is made or found to be ,3. . A Logic systent functional Test shall be inoperabic for any r,cason, continued completed each refueling outaese.

reactor operation i; perinissible only - -

during the succeeding 30 days unless j -

such pump is sooner made operah!e, ,

provided that during such 30 days the , .

remaining active cornponents of the -

~

. LPCI rnode of the RHR. containment .

cooling ruode of the RHR. all active

component: of both core spray sub7.ys- .

~

tems. and the diesel generators re- ,

quired for operation of such compo- -

nents if no external source of power (

r v,ere avai!ah!e sha!! be operab!e. -

3. From arid after the date that the LPCI -

- - i rnode of the RiiR system is rn.ide or -

i found to be inoperab'e for any reason. -

continued reactor operation is permis , .

sible only duting the succeeding 7 days l

unless it is sooner made operable, pre- -

l vided th:t during such 7 days aliactive .

componer.ts of botit core rpray subsys- - - i tems, the containment cooling mode of '- - l the Ri!R (including two RilR ', .

pumps). cad the diesd generators re- !.

quired for operation of su:h compo-nents if no ex.ternal source of power '. -

~

were availab!e shall be oo cab!c.-

~

. 4- If the requirements of Specification .

3.5.B cannot be met. an or:!ccly shut-(

.. .down of the reactor sh:.!! be initiated. I and the reactar shall be in the cold -

I shutdo.vn condit;on within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

e 3.5/4.5-2a ,

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

QUAD-CITIES DPR- 30 Coteta!raaent Ccolisq Mode "o f the RIIR C'.

Conta*carent Ccoling Moda or the nna' C.

Syste:n .

Sys!t.n . .

~

- Surveille.iice of the containment coolhtg nicis of th,e Rl!R system shal! be petrermed cs - '

< .- - -' . follows: . .

1. Riin service water.subsys:en. t: sung:

1. noin scop > or trie to:ttaintnent cooiing ,

rnode or the RflR system. as defined in - "

Item ' Trepency

.the ~b. asis far SpeciGcation 3.5c .~sital! ,

1,)e,op,crabic whenever irradiat. d fue!is In'the rea::or vess::1 and prior to reac. , *

. tor startup fre,m a co!d candition ,

Ouarterly i except as specified in 3.5.C.2, a. . Flow rate" test -

3.5.C.3, 3.5.C.4, and 3.5.G.2. crich RflR service

r water pump shall deliver ;tt least ,

'3500 rpm apinst a pressure 'of 198 - .

F5'8 '

Reft eling

b. A logic system

. i

- functional test .

. i I

2. From'and rter the date th:rt one of t!ie '

RIIR service v.at'er pumps is rdn<!c cr ,

j fou.nd to be inoper:tNe for any reason. !

contint:cd reacter oret 2:ian is permis-sib!: en!y ciuring the sua ceding 30 days ur.I v such parap is Sr.na.:r made operate!c. provided that durine wtt: 30 ,

d.ays all other aetive coni; onents of Llie cont:siamtitt cooling mode o!'llie l'1lR systcm at: op::ub!c. .

t . _ . . _ .

3.5/4.5-3

QUAD-CITIES DPit-30

3. From and after the date that one loop of the containment cooling inode of the RilR system is made or found to be inoperable for any season, contin-ued reactor operation is permissible only during the succeeding ' days un-less such subsystem is so3ner made operable, provided that all active com-ponents of the other loop of the con-tainment cooling mode of the RilR system, both core spray subsystems, and both diesel generators required for operation ofsuch components if no external source of power were availa-ble, shall be operable.

During the time period from April 17, 1978 through April 30, 1978 while the -,

2A Containment Cooling Loop of the RilR System is made inoperable for heat exchanger repair, continued reactor operation is permissible beyond the above 7-day limitation, unless such loop is sooner made operable, provided that during the time the 74ay limit is ex-ceedeJ, a visual inspection is performed daily to assure that proper valve align.

ment and system integrity is maintained in the "B" R!lR loop.

"r 2. During each 5-year period, an air test l

4. Containment cooling spray loops are shall be performed on the drywell required to be operable when the reac- spray headers and nonles and a water for water temperature is greater than spray test performed on the torus 212 ' F and prior to reactor startup spray header and nonles.

from a cold condition. Continued reac-tor operation is permitted provided that a masimum of one drywell spray loop may be inoperable for 30 days when the reactor water temperature is greater than 212 = F.

5. If the requirements of 3.5.B cannot be met. an orderly shutdown shall be ini-tiated, and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 3.5 / 4.5-4

Q U A D-CITIES D Pit.30

\ D, IIPCI Subsystem D. IIPCI Subsystem l

Surveillance of IIPCI subsystem shall be per-formed as follows:

1. The llPCl subsyst:m shafi be operab!c 1. Once per cuarter, it shall be verified whenever the reactor pressure is that the IIPCI pump delivers at least p,reater than 90 psip. irradiat:J fuelis 5000 gpm against a system head corre--

, in the reac'er vessel. and princ to rea . sponding to a reactor vessel pressure of for startup from a cold condition, 1150 peig to 150 psig.

except as specified in 3.5.D.2.

, 2. A Simulated Automatic Actuation Test shal' be completed each refueling outace. '

2. From and after the date that the IIPCI 3. A Logic System Functional. Test shall be subsystem is inade or found to be inop- completed each refueling outage, erable for any reason, continued reac-tor operation is permissible only dur-ing the succeeding 7 days unless such -

subsystem is sooner made operable, 3.5/4.54a

QUAD-CITIES DPR.30 I

provided that during such 7 days all active components of the automatic pressure relief subsystems. the core spray subsystems. LPCI mode of the RilR system.and the RCICsystem are - - -

. operable. -

3. If the requirements of Specification l ED cannot be met, an orderly shut-down shall be initiated. and the reac- i tor pressure shall be reduced to 90 psig within 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months .

. 1 E. Automatic Prewure Relief Subsysicms E. Automatic Pressure Relief Subsystems Surveillance of the automatic pressure relief sub-I systems shall be performed as follows:

1. The automatic pressure relief subsys- 1. The following surveillance shall be carried tem shall be operable whenever the out on a 6 month surrveillance interval:

reactor pressure is greater than 90

a. A simulated automatic initiation psig. irradiated fuel is in the reactor which opens all pilot valves.

vessel and prior to reactor startup -

from a cold condition, except as b. With the reactor at pressure each spccified in 3.5.E.2. relief valve shall be manually opened.

Relief valve opening shall be verified by a compensating turbine bypass valve or control valve closure.

2. A logic system functional test shall be performed each refueling outage.

, 3. When it is determined that one relief

2. From and after the date that one of the valve of the automatic pressure relief five re!!ef valves of the automatic pres- subsystem is inoperable, the liPCI shall 1 sure relief subsystem is made or found be operable.

to be inoperable.when the reactor is pressurized about 90 psig with irradi-ated fuel in the reactor vessel, reactor operation is permissible only during the succeeding 7 days un'ess repairs are made and provided that during such time the IIPCI subsystem is operable.

3. . If the requirements of Specification l 3.5.E cannot be met, an orderly shut-down shall be initiated and the reactor -

pressure shall be reduced to 90 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.5/ l.5-5

DPR 30 .. .

F. Itescror Core Isolation Cooling System F. Reactor Core Isolation Cooling System . l Survei!!ance of the RCIC system shall be p:r-formed as follows:

I. The RCIC system will be operable 1. .

Once per quarter, it shall be whenever the reactor pressure is verif2cd that the RCIC pump delivers greater than 150 ps.ig. irrad.iated fuel at least 400 gpa against systen n m the reactor vessel, and prior to head corresponding to a Reactor startup from a cold condition Vessel Pressure of 1150 psig

. except as specified in to 150 psig.

3.5.F.2.

2. A Logic. System functional test shall be completed each refu ling utage.

2. From and after the dat: that the RCIC system is made or found to b:inopera- 3. A Simulated Automatic Actuation b'le for any reason, continued reactor Test shall be completed each operation is permissible only during # f" li"9 "t^9 -

the succeeding 7 days unless such sys-tem is sooner mad: operable, provid:d that during such 7 days all active com-ponents of the llPCI systern are operable.

3.. If the requirements of Specification i 3.5.F. land 3.5F.2 cannot be met, an

~

orderly shutdown shall b: initiated and the reactor pressure shall be re-duced to 90 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . , . . ,

.\linimum Core and Containment Cooling Sys- G. ' Minimum Core and Containment Cooling Sys.

} G- tem Asailability tem Asaila'oility

1. Any combination of inop:rable com- Surveillance requirements to assure that mini-ponents in the core and containment mum core and containment coo!ing syst .s are cooling sy3tenis shall not defeat the availab!: have been specified in Speci5 cation capability of the remaining operable 4.2.B.

' components to follill the core and con. ,

tainment cooling functions.

2. When in2diated fe:1 is in the rea: tor vessel and the reactor is in the cold .

. shutdown cond? tion all low-pressure

- core and containment cooling sys: cms

~

may b: inoperable provided no work .

3.f.MS-6 ,

-- - - - - . . . , _ _ _ ~ , _ _ L_ _ _ _ __ __

QUAD-CITIES DPR. 30 ,

s .

is being done which has the potential -

for draining the reactor vessel. ,

3. When iriadiated fuel is in the reactor * -

and the ver.cl head is removed, the suppression chamber may be drained comp?ct:ly and no more than one con- .

trol rod drive housir.g opened at any one time provided that the spent fuel ,

pool gate is op:n and the fuel pool , ,

.- ~

water Icrel is snaintained at a level of '

greater than 33 feet abose the bottom of the paal. 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 blanhed following removal of the con-trol rod drive, and a special flange shal: 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 O

__.....-.,i.,,a:,,....,...a...~i.

that h:n the potential f r drainine the - '

vessel may be carried on with less than 112,200 !!' of water in the sup;.rcssion .

pool, provided that: (1) the total vol-ume of water in the .<uppression pool. ,

refueling cavity, and the fuct storage

pool above the bottom of the fuel pool .

gate is greater than i12,200 ft); ,

(2) the fuel storage pool gate is re- '

moved;(3) the low-pressu re core and ,

cont.,inment cooling systems are oper- ,

able; and (4) the automatic niode of the drywell sump pumps is disab!cd.

IR Maintem. nee of Filkd Dhcharge Pipe 11 s Maintenance of Fi!!cd Discharge Pipe

[

The follov.ing surveillance requirements shall be adhered to to assure that the discherge piping of the core spray. LPCI mode of the RilR, llPCI, and 11CIC are filled:

13 Whenever enre spr.iy. T.PCI mode of .. l.. Every quarter prior to the l

' the RllR. IIPCI. or 1* CIC are required testing of the LPCI mode of the 3

to be operabic the disci:arge piping RHR and core spray ECCS, the l discharge piping of these system from ti.e pump t -harpe ofincsc sys-

, tenn to the 1.c.t cnet! valves shai! be shall be vented from the high

\s filled. point. and water flow ob. served.

e 3.5/ 4.5-7

QUAD-CITIES DPR-30 A

1

2. The discharge pipe pressure for the 2. Following any period where the 1.PCI systems in Specification 3.5.11.1 shall mode of the RilR or core spray ECCS be maintained at greater than 40 psig have not been required to be operable, and less than 74 psig. If pressure in the discharge piping of the inoperable any of these systems is less than 40 system shall be vented from the high psig or greater than 74 psig, this con- point prior to the return of the system dition shall be alarmed in the control to service.

room and immediate corrective action

taken. If the discharge pipe pressure is

3. Whenever the llPCI or RCIC system not within these hmits in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is lined up to take suction from the after the occurrence, an orderly shut- torus, the discharge piping of the down shall be mitiated, and the rear- IIPCI and RCIC shall be vented from tor shall be in a cold shutdown condi- the high point of the system and water tion within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after initiation. Ilow observed on a monthly basis.

4. The pressure switches which monitor the discharge lines to ensure that they are full shall be functionally tested every month and calibrated every 3 months. The pressure switches shall be set to alarm at a decreasing pressure of 240 psig and an increasing pressure of s74 psig.

I. Condensate Pump Room Flood Protection I. Condensate Pump Room Flood Protection

( '

( l. The systems installed to prevent or 1. The following surveillance require-mitigate the consequences of floorng ments shall be observed to assure that of the condensate pump room shall be the condensate pump room flood pro-operable prior to startup of the tection is operable.

a. The piping and electrical penetra-

2. The condenser pit water level switches tions and bulkhead deors for the shall trip the condenser circulating vaults containing the RHR 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 alarin 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 opc4able.

b. The floor drains from the vaults

3. If Specification 3.5 I.1 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 operatieg, 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 RIIR service water pump and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

diesel generator cooling water pump bed plate drains shall be s checked during each operating 3.5 / 4.5-8

QUAD-CITIES DPR- 30 cycle by assuring that water can y run through the drain lines y actuating the air-e; e ated s%

by operation of the (nllowing sensors;

.I) loss of air .

2) equipment level drain sump high ,_

3)i vault high level .

d.'

Die condenser pit 5-foot : rip cir. -

, . cuits for each channel shall te

', . checked once a month. /s loeie system fecctional test sha!! be p'er.

formed d ring each refueling -

s outage. t.

c. ..-

J. Average Planar Linear lleur Gener.ition Rate J. Average Planar Linear lleat Generation Rate (APL11GR) .. (API.HGR) ,

Dur? .g steady-state power operation, the aver- - The APLIIGR fo' reach type of fuel as a func.

age' linear heat generation rate ( APLHGR) of tion of average planar exposure sha!! be deter.

all the rods in any fuel assembly, as a function mined daily dudn3 reactor operation 2t of average planar exposure. at any axial laca- 2: 25% rated thermal power. ,

tion. shall not exceed the maximum average , ,

planar LHGR shown in Figure 3.5-1 (3 sheets). ., ,

If at say time during operation it is determined ' * ,

by normal surveillance that the limiting va!ue for APLf!GR is being exceeded. action shall be initiated within 15 minutes to restore operatien ,.

to within the prescribed limits. If the APLHGR .

is not rettirned in within the prescrib'ed limits -

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the reactor shall be brought to .

the cold shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

.- Surveillance and corresponding action shall ,

' continue until reactor operation is within the ., _ . _ , ,

prescribed limits. -

K. La a; 1.11Gn '

K. I n al I.11C R - -

During steady-state power operation the linear Daily during steady-state power operation

.. he.0 generation rate ( LHG R ) of any rod in any above 25'A. of rated thermal power, the local

'Itcl assembly at any axial loc.ition shall not LHGR shall be checked. ,

exceed the maximum a!!owable LHGR as cal-culated by the fo!!owing equation. If, at any ,

time during operation it is determined by nor-rnal surveillance that the limitine value for '

L!!GR is being exceeded. action shall.be initi-Tred wittiin 15 minutes ta restore operation to within the prescribed limits if the LUGR is not ~

returned to within the prescribed limits within.

e -

3.5 / 1.5-9

_ ~ . . _ . . -

~

QUAD-CITIES DPR-30 within the prescribed limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Surveillance and cor- l responding action shall continue until reactor operation is within the prescribed limits. .

LilG R,,,, < LilG R, 1 -( .1P/ P ),,,,,,( L/ L, ) r-where:

LilG R, = design LilGR

= 17.5 LW/ft. 7 x 7 fuel assemblies

= 13.4 kW/ft. 8 x 8 fuel auemblies ,

( AP/ P),, ,

= maximum power spiking penalty

= .035 initial core fuel

= .029 reload I . 7 x 7 fuel

= .022 reload. 8 x 8 fuel 31 = .028 reloao I . mixed oxide fuel 1., = total core length

= 12 feet L = Axial distance from bottom of core L. Minimum Critical Power Ratio (MCPR) L. Minimum Critical Pmur Ratio (N1CPR)

The MCPR shall be determined daily during During steady-state operation MCPR shall be greater than or equal ta steady-state power operation above 25'i of 1.35 (7 x 7 fuel) rated thermal power.

1.35 (8 x 8 fuel) at rated power and flow. If at any time durina operation it is determined by normal sunetllance

, that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed hmits.

If the steady state MCPR is not returned to within the prescribed lunits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the reactor ,

shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Surveillance and corresponding action shall continue untd reactor operation is within the prescribed limits. For core 00ws other than rated. these nominal values of MCPR shall be incicazd by a factor of kg where kg is as show n in Figure 3.5.2.

3.5/4.5-10

QUAD-CITIES DPR-30 3.5 LIMITING CONDITIONS FOR OPERATION DASES A. &Il Core Spray and LPCI Mode of the RilR System l This specification assures that adequate emergency cooling capability is availab!c. .- .

Based on the loss-of coolant analyscs included in References I and 2 and in accordance with 10 CFR 50A6 and Appendix K, core coo!ing systems provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident, to limit the calculated peak cladding temperature to less than 2200' F, to assure that core geometry remains intact, to limit the corewide cladding metal-water reaction to less than I"c, and to limit the calculated local metal-water reaction to less than 177o.

The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in Reference 3. Using the results developed in this reference, the repair period is found to be less than half the test interval. This assumes that the core spray subsystems and LPCI constitute a one-out-of-two system: however, the combined effect of the two systems to limit excessive cladding temperature must also be considered. The test interval specified in Specification 4.5 was 3 months. Therefore, an allowable repair period which maintains the basic risk considering single failures should be less than 30 days, and this specification is within this period.

Although it is recognized that the information given in Reference 3 provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach prevents complete acceptance of this method at this time. Therefore, the times stated in the specific items were established with due reoard to judgment.

\ Should one core spray subsystem become iroperable, the remaining core spray subsystem and the entire LPCI mode of the R!lR system are available should the need for core cooling arise.

Based onjudgments of the reliability af the remaining systems, i.e., the core spray and LPCI, a 7-day repair period was obtained.

Should the loss of one RHR pump occur, a nearly full complement of core and containment cooling equipment is available. Three RHR pumps in conjunction with the core spray subsystem will perform the core cooling function. Because of the availability of the majority of the core cooling equipment, a 30-day repair period is justified. If the LPCI mode of the RHR system is not available, at least two RHR pumps must be available to fulfill the containment cooling function. The 7-day repair period is set on this basis.

C. RHR Senice Water g The containment cooling mode of the RHR system is provided to remove heat energy from the comainment in the event of a loss-of-coolant accident. For the flow specified, the containment long-term pressure is limited to ! css than 8 psig and is therefore more than ample to provide the required heat-removal capability (reference SAR Section 5.2.3.2 ).

The containment cooling mode of the RHR systen consists of two loops, each containing two RHR service water pumps. one heat exchanger, two RHR pumps, and the associated valves, piping, e!ectrical equipment, and instrumentation. Either set of equipment is capable of performing the containment cooling function. Loss of one RHR service water pump does not seriously jeopardize the containment cooling capability, as any one of the remaining three pumps can satisfy the cooling requirc ments. Since the.re is some redundancy left, a 30-day repair period is adequate. Loss of one loop of the containment

(- cooling mode of the RHR system leaves one rema>ing system to perform the containment cooling function. l 3.5/4.5-11

QUAD-CITIES DPR-30 Based on the fact that when one loop of the containment cooling mode of the RilR system becomes inoperable, only one system remains, a 7-day repair period was specified. l D. Iligh.Prewure Coolant Injection g

The high-pressure coolant injection subsystem is provided to adequately cool the core li>r all pipe breaks smaller than those li>r w hich the LPCI mode of the RilR system or core. spray subsystems can protect the core.

The ilPCI meets this requirement without the use ofotTsite electrical power. For the pipe breaks liir which the lli CI is intended to function. the core never uncovers and is continuously cooled. thus no cladding damage occun ( reference SAR Section 6.2.5.3 ). The repair times liir the limiting conditions of operation were set considering the use of the llPCI as part of the isolation cooling system.

E. Automatic Prewure Relief Upon failure of the HPCI to function properly after a small break loss-of-coolant accident, the ADS automatically causes the safety-relief valves to open, depressurizing the reactor so that flow from the low pressure cooling systems can enter the core in time to limit fuel cladding temperature to less than 22000F. ADS is conservatively required to be operable whenever reactor vessel pressure exceeds 90 psig even though low pressure cooling systems provide adequate core

( cooling up t> 350 psig.

F. RCIC g

The RCIC system is provided to supply continuous makeup vater to the reactor core when the reactor is isolated from the turbine and when the feedwater system is not available. Under these conditions the pumping capacity of the RCIC system is sullicient to maintain the water level above the core without any other water system in operation. If the water level in the reactor vessel decreases to the RCIC initiation level, the system automatically starts. The system may also be manually initiated at any time.

The llPCI system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable. Therefi)re. the specification calls for an operability check of the llPCI system should the RCIC system be found to be inoperable.

, G. Emergency Cooling Asailability l

The purpose of Specitication3.5G is to assure a minimum of core cooling equipment is available at all [

times. If. for esample. one core . spray were out of service and the diesel which powered the opposite core spray were out of service, only two RilR pumps would be available. Likewise. if two RilR pumps were out of service and two RilR service water pumps on the opposite side were also out of service no containment cooling would be available. It is during refueling outages that major maintenance is performed and during such time that all low-prewure core cooling systems may be out of service. This specification prosides that should this occur. no work will be perti>rmed on the primary system w hich could lead to draining the vewek This work would include work on certain control rod drive components and recirculation system. Thus, the specification precludes the events which could rcquire core cooling.

Specification 3.9 must also be consulted to determine other requirements tiir the diesel generators.

Quad-Cities Units I and 2 share certain procew systems suth as the makeup demineralizers and the radwaste system and also some safety systems such as the standby pas treatment system. batteries. and i

3.5/.l.5-12 i

~ _ _

QUAD-CITIF.S DPR-30 L,

diesel generators. All of these systems have been sized to perform their intended function considering the simultaneous operation of both units.

These technical specifications contain only a single reference to the operability and surveillance requirements for the shared safety-related features of each plant. The level of operability for one unit must be maintained independently of the status of the other. For example, a diesel (1/2 diesel) which is shared between Units I and 2 would have to be operable for continuing Unit 1 operation even if Unit 2 were in a cold shutdown condition and needed no diesel power.

Specification 3.5E3 provides that should this occur, no work will be performed which could preclude l adequate emergency cooling capability being available. Work is prohibited unless it is in accordance with specified procedures which limit the period that the control rod drive housing is open and assures that the worst possible loss of coolant resulting from the work will not result in uncovering the reactor core.

Thus, this specification assures adequate core cooling. Specification 3.9 must be consulted to determine other requirements for the diesel generator.

H. Maintenance of Filled Discharge Pipe l If the discharge piping of the core spray, LPCI mode of the RilR,11PCI, and RCIC are not filled, a water hammer can develop in this piping, threatening system damage and thus the availability of emergency cooling systems when the pump and/or pumps are started. An analysis has been done which shows that if a water hammer were to occur at the time emergency cooling was required, the systems would still perform their design function. However, to minimize damage to the discharge systems and to ensure added m trgin in the operation of these systems, this technical specification requires the discharge lines p to be filled whenever the system is in an operable condition.

w Specification 3,SLGA I provides assurance that an adequate supply of coolant water is immediately l available to the low-pressure core cooling systems and that the core will remain covered in the event of a loss-of-coolant accident while the reactor is depressurized with the head removed.

I. Condensare Pump Room Flood Protection l

See Specification 3.5.I. j J. merage Planar LIIGR l This specification assures that the peak cladding temperature following a postulated design-basis loss-of-coolant accident will not exceed the 2200* F limit specified in 10 CFR 50 Appendix K considering the postulated etTects of fuel pellet deasification.

The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average LIIGR of all the rods in a fuel assembly at any axial location and is only secondarily dependent on the rod-to-rod power distribution within a fuel assembly. Since expected local variations in power distribution within a fuel assembly arTect the calculated peak cladding temperature by less than i20' F relative to the peak temperature for a typical fuel design, the limit on the average planar LilGR is. sufficient to assure that calculated temperatures are below the 10 CFR 50 Appendix K limit.

The maximum average plan r LilGR's shown in Figure 3.5-1 are based on calculations employing the models described in Reference 1. Power operation with LHGR's at or below those shown in Figure 3.5-1 assures that the peak cladding temperature following a postulated loss-of-coolant accident will not exceed the 2200

F limit. These values represent limits for operation to ensure conformance with 10 CFR 50

- and Appendix K only if they are more limiting than other design parameters.

( '."he maximum average planar Ll!GR's plotted in Figure 3.5-1 at higher exposures result in a calculated psak cladding temperature ofless than 2200* F. Ilowever, the maximum average planar LHGR's are 3.5/4.5-13

QUAO-CITIES DPR-30

?

shown on Figure 3.5-1 as limits because conformance calculations have not been performed to justify operation at LilGR's in excess of those shown.

K. local 1.IIGR g

This specification assures that the maximum linear heat-generation rate in any rod is less than the Asign linear heat-generation rate even if fuel pellet densification is postulated. The power spike penalty specified is based on that presented in Reference 4 and assumes a linearly increasing variation in axial gaps between core bottom and top and assures with 95% confidence that no more than one fuel rod exceeds the design LIIGR due to power spiking. An irradiation growth factor of 0.25% was used as the basis for determining AP/P in accordance with References 5 and 6. -

L. Minimum Critical Power Ratio (MCPR) l The steady state values for MCPR specified in this specification were selected to provide margin to acconuno.

date transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical power correlation itself. These values also assure that operation will be such that the initial condition assumed for the LOCA analysis, an MCPR of 1.18,is satisifed. For any of the special set of transients or disturbances caused by single operator error or single equipment malfunction,it is required that design analyses initialized at this steady. state operating limit yield a MCPR of rot less than that specified in Specification 1.1.A at any time during the transient, assuming instrument trip settings given in Specification 2.1. For analysis of the thermal consequences of these transients, the limiting value of MCPR stated in this specification is con-servatively assumed to exist prict to the initiation of the transients. The results apply with increased con-servatism while operating with MCPR's greater than specified.

1 The most limiting transients with respect to MCPR are generally:

a) Rod withdrawalerror b) . Load rejection or Turbine Trip without bypass c) Ioss of feedwater heater Several factors influence which of these traasients results in the largest reduction in critical power ratio such as the specifa fuelloading, exposure, and fuel type. The current cycles reload licensing submittal specifies the limiting transients for a given exposure increment for each fuel type. The values specified as the limiting Condition of Operation are conservatively chosen as the most restrictive over t! e entire cycle for each fuel type.

For Cycle 4, the operating limit has been increased by 0.04 over the limit based on transient analyses to assure that boiling transition would not occur in a mistoaded fuel bundle during steady state operation.

3.5/4.5-14

.- - . . . - . . - - _ . _ . - - . _ _ - - - - - . - . . - . . . . - - . . . - - . - . . - - - . . - . ..-.-?

Ot)AD-CIT.li:S -

DPR-30 .

k5 SURVEIIbNCE bEQUIREMENTS BASES TIE testine, i:It'irGal for the core imd containment cooling systems is based on a quantitative reliab

~

judgmeni,'enil pr;.#ality. The core coolin; systems have not been designed to be fully testa For exarapfe,-th- core si, ray final admission valves do not open until reactor pressure has fallert .

during op ration.cven if h@ drywell pressure were simulated, the fmal valves would

~

not open.

2ctoe vessel .

In IIPCI, automatic initiadon dudng power operation would result in pumping cold water into the r:

(

I which is not d.esirable., . .' .

' To increase the .

'the'systemsifM-bFdutomatical.y actuated during a refueling cuta3: and this wi!! be dot -

availability of the: individual components of th: core and containment cooling systems, the compon:n's w in. Instrumentation, pumps, valve' operators, etc., are tested more frequently. The make .1:p Jhe.~systerat instrumenta:!or-is fcnctionally tested each rnonth. The pumps and valves will be' tested in accordance Y

'with the Inservice TOsting Progcma to1 insure operabilit'y. This program which is based I on IGS Code,Section XI requircraents prov. ides adequate assurance that 'the core and .

containment. cooling nystems will be operable when required. Uhen components or sub-systens are fteken out of service, surveillance tests of the rentaining redundant components or nubsystens is not required because of the increaned risk that.a single j

' failure occurring during this testing vill result in a loss of total synten function.

This is justified by the additional testing requirements of the ISIE Code which provides I adequate assurance' that 1.he purms and valves will be operable when called upon to per- '

}

forn their specified function. .The limiting conditions for operation ensure that. j l

redundancy in reaintained, at all' tiraes. , ,

. j e

ih: verificatErroEthim2'ersteam I. rcilef valve operability during rninual ectuation surnillance testin; must b: ma 2*uas indicated by thermoccup!:s downstre::n of the relief valves. It has been found that a iadepend ent .cf.1 cape:

temp:rature ine=asen:y result.with the valve sti!.! closed. This is d": to steam being vented throup.h

' durin3 the surb.ll!ance. test. .P,y first opening a tu:bine bypas valve,and then observin;ite closure rcs ulveT:$2tMiikositive v-dfication can be mid: for the relief valve opening end p2ssing steam flow. Closur2 resp the 2:hne contcci'vheiduring relief valve manual actuation wou!d likewis: serve es an adequate veriSection f .

' rebf valviopi:iing This test method may be performed over a wide range of textor pressures greater thm 150

.. Wlve o.OatiadL4hi5150 psi 7 lis liruited by the sprin; tension exhibited by the relief valves.

Thesu eitlinde riquire/aenis to' ensure that the discharge piping of'the corespray,LPCI rnode of t and ljCfC sys:em's is fi!! dlprovides far a visual' observation that water flows frorn a high point v

. that t.h.e.line.d, b. a tu!I condition Intrumentation has been provided t > monitor the pressure of water discharge pip,iSt between the quarterly intervals 3.t which the lines me vented and n!. tm the control 'n the pre:.sure is inadequate. .This instrunentation vill be calibrated on the sana f re--

quency as the safety systen instrunwntation and the alarn systera tested boathly. This testing ensures th 9t, during the interval betwen the quarterly venting checks, the status e of the discharge piptng

.ts monitored on a continuous basis. s An alarm point of 2 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 shutotT head of the till system pumps is 74 psig and therefore will not defeat the low-pressure cooling pump discharge press interlock of 2 75 p>ig as shown in Table 3.2-2.

3.5/4.5- 15 .

sot i a: , o,,... .

. 3 . . . ,, y DPR- O

rns shali be operab!: during rea: tor ,

power operation. From and t ft:r the date that one cf these systerr.s is made or found ta be inoperab!: for any rea-son, reactor power operation is per-inissib!: only during the succeeding

- "I days.

3. If the conditions in I or 2 above can-not be met, an erderly shutdown shall

~

be initiated and the reactor shall be in

. , , . h co!d shutdown condition within 24 *

.. . hours. ,

~

.E. Safety and Relief Valve's' ,- ,

~

E. safety and 1:clier valves , ,

.: ssu s t points of the

1. Prior to reactor startup for power op- safety and relief valves shall be cration, during reactor power operat-checked in accordance.with the-

~- . -

ing conditions.and whenever the res:- . requirements of the Inservice -

tor coolant pressure is greater than 90

. Testa.ng Program for Puraps and

. . psig and ternperattne greater than Valves defined in Section 1.o(P.F) 320* I, all nme of the safety valves of these Technical Specifications.

--- - shall be operable. The 'solenotd-activated pressure valves shal1 be oper- 2.* The set points for the safety valves ab!: as required by Specification shall be as follous:

. 3.5. E . . .

Numb r f Valv s set Point (psig)

-2.- If Specification 3.6.E.1 is not m:t, ti:e reactor shall remain shut down until 1 1115*

. the condition is corrected or, if in

~

- 2 1240

. oneration, 2

an order!"1 shutdown shall 2 1250

-be initiated and the reactor coolant 4 1260 .

pressure and temperature shall be

- . . below 90 psig and 320* F within 24 The allowable set point error for each

- - hours. -

valve shall be + lt. ~ .

t * '

- 3. The set points for the relief valves shall be as follows:

Number of Valves Set Point (ps ig,)

1115*

-

1

- 2 -

41130 2 (1135 I

Target rock co;nbination safety / relief i valve.

F.. Structural Integrity F. Structural latepity ,

i 'Ihe structur2l integrity of the primary syMem The nondestructive exa:ainations specified boundary sha'd b: maintain:3 at the lnel re- in the Inservice Inspectica Prograa quired by the AS'.!E P. oiler an-! hessure Vesel defined in Section 1.0 (G.G) shall be Code, S:etion x!, P.t.!cs for burvice laspectica conducted in accordance with applicable i of h'uclear Power Plant Com; onents". . Edit. ion and Addenda of Section XI of the ASMN Doller and Pressure Vessel Code as required by loci'R50, Section 50.55a (g) .

In addition, the welds listed in Table 4.6.1 shall be c:.:amined at the frequency specified.

3.6M.6-4

$6 *#M* ND NN * .% w g p, 4b e y.gm.-h.gW*'*-..*-.hMg***-***MMee* wy me - **-**e**- = %e e **8""* WM+ ' ' ' ' * - * " *- '

M

- m 2 o DPR-30

\2 the boundaries of the reactor. ::re exceeded. Methods available to the operator for correcting the offstandard condition include operation of the reactor cleanup specu. reducing the input ofimpuritie;.

and placing the reactor in the cold shutdown condition.The nujer benefit of cold shutdown is to reduce.

the temperature. dependent corrosion rates and provide time for the deanup system to reestablish the purity of the reactor coolant. During startup periods. which are in the category ofless than 100,000 lb/h r.

conductivity may exceed 2 pmho/cm because of the initial cvotation of gar.es and the initial addition of dissolved metals. During this period of time when the conductivity exceeds 2 pmho (other than short. term spikes), samples wil! Le taken to assure the chloride concentration is less than 0.1 ppm. -

The conductivity of the reactor coolant is continuously monitored.The sampics of the coolant which are taken every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months will serve as a reference for calibration of these monitors and are considered

, adequate to assure accurate readings of the rnonitors. If conductivity is within its normal range, chlorides end other irnperities vi!! also ha ydthin their normai ranges.The reactor coolant samp!es will also be used

_ to determine the chlorides.Therefore, the samplin'; frequency is considered adequate to detect long-term changes in the chloride ion content. Isotopic analyses required by Specification 4.6.C.I.b may be performed by a gamma scan. , ,

.. . ..q .. .. ..'..

D. Coolant Leakage .. .. - - .

. . Allowab!c leakage rates ofcoolant from the reactor coolant system have been based on the predicted and '

experimentally observed behavior ofcracks in pipes and on the ability to make up coolant system leakace

~ , 'in the event ofloss of offsite a-c power. The norma!!y expected background leakage due to equipment design and the detection capability for determining coolant system leakage were a!so considered in establishing the limits.The behavior of cracks in piping systems has been experimenta!!y and analytica!!y investigated as part cf the USAEC sponsored Reactor Primary Coolant System Rupture Study (rhe oipe f

t Itupture Study). Work utilizing the data obtained iri this study indi;?2t?? th2t eakage frCm .a crack can

'be detected before the crack gro.vs to a dangerous or critical size bf mechanically or ther' mil!y induced cyclic loading, stress corrosion cracking or some other mechanism characterized by gradual crack growth.This evidence suggests that for leakage somewhat greater than the limit speciSed for unZentified Icakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly. . . . , . . , . , . ., y , ,

II'Iwever, o the establishment of allowable unidentified Icakage'greitter th'an that given in Specification 3.6.D, on the basis of the data presently available would b'e premature because of uncertainties associated with the data. For Icakage of the order of 5 gpm as specifud in Specitication 3.6.D, the experim:ntal and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation. Leakagc! css than the magnitude specified can be detected reasonably in a matter of few hours utilizing the availab!ilcakage detection schemes, and if

~

the orgin cannot be deterrained in a reasonably short tin.:. the plant should be shut down to allow further investigation ;md corrective action. ,

The capacity of the drywel! sump is 100 ppm, and the capacity of the drywell equipment drain tank pumps is also 100 gpm.

Itemoval of 50 gpm from either of these sumps can be accomplished with :onsiderah!e margin.

The performance of the reactor coolant leakage detection system will be eva!aated during the first 2 years ~

of station operation, and the conclusions of this evaluation will be reported to the NRC.

. E. Safety and Relief Yuhes The frequency and testing requircraents for the safety and relief valves are specified in the Inservice Testing Program which is based on Section XI of the ASBIE P. oiler and Pressure Vessel Code. Adherence to these codo requirementu provides adequate assurance as to the proper operational readiness of these valves. The tolerance value is specified in Section III of the AS'tE Boiler and Pressure Vessel Code as + 11 of design pressure. An antlysis has been i perforraed which shous that with all safety valves set It higher the reactor I

, coolant pressure safety lirait of 1375 psig is not exceeded. The safe ty valvan are required to be operable abow the design pressure (90 psig) at which the core spray subsysteris are not designed to deliver full flow.

3 (d4 r-

QUAD-CITIES ,

DPn 30 - _ _

P. Structural Integrity A pre-service inspection of the components.in the primary coolant pressure is free boundary will be conducted after site crection to assure the systent Prior to of gross defects and as a ref erence base foe later inspections. In -

operation, the reactor prinary r,ystem will be free of gross defects.

addition, the f acility has been designed such.that gross defects should not occur throughout life.

Inservice Inspection of AS:m Code Class l', 2.and 3 components will bc perforraed in accordance with the applicable version of Section XI of the AstE Iloiler and Pressure Vessel Code. Relief fro:a any ofThe theInservice.Inspec-above require-ments has been provided in writing by the Comraission.

tion program and the written relief do not f;orm a part of these Technical _

Specifications.

'Ibe sp:Livinspection of the snain feed and s*,e:mlines is to provide :dded protection ogsinst pipe whip,in additloai toihe prote tive energy absorbin; system to be instalkdinsid: the dryv.e!!:s des::ibed in A:nend.

raent 27 to the1\R. The Group I wdds are selected on the basis of :n an:Iysis that shows then welds :r i

the hi;1rsytress welds and that due to their physicallocation, a break would result in the I:ast nterferen:e 3rd snariarsa energy upon imp:ct with th: drywell. These welds are th- only cnes whi:h offer any si;nifi.

cant rbkani:re therefore inspected 4 times:s often es the other welds witida th: drywell.

Group I,Ijsilfs'ere sdected b:caun without regard for the c; eratin ; stress 1:v:!s:nd interf::in; equipment, they hafe's0fficient th:oretied energy to pen:trete end would p:opel the pip: toward the centainm:nt.They are theri:forcincluded in first in;pictica. Upon consider tion ofimp::t 2nje,inte:ferin; quipment, d.d the distance pipe travels, no substantial ric is in,ched and no c.v.t:a insp:: tion is needed.

h a

f e#

e e

3.6/.16-12

. . ( vs. . .;1TJ:.S Dl'i:-30

[ .

11.ME 4.6 l

- pl.:. 5lncil:.2 . Ct:;) l C'dh

, ET '4 II \IAh ysf I.'n;in. tlcu IArdifi-py c:.ti:3 tr.: 2 L'sa -

t2a l':.!t 2

,. 3031A 23 h. .NA5?G

... . 3 A T;:

.v/lA 20 .5. 3'A !!' '

YA'f24 3::;.e .;p,. >3 313 300!E.23.b. 303 52%.

' 3:.);C W -h. 33C 5?J 333 f25 -

. l.n.2;.h. M 3 51c .

. 393-F23 3); 5.'!A

.l.'

300!C 20-b. vCm 30C-Fis I

. . 300i9 29-h. E 3 5?!

3:C FN

. * *30J 125 T:cGn;'t L:: Crcn I \?Ch. C.T.9 II W' E3 '

t,'tu !b'.T iltM lb'iii-Ur.: cath Edt 2 OM C2U.~1 Erh I 2:(%E h. 301. 54 3%t5 r'. 3IA-31 U13.:;.b. 323 55 3M I3 3204D-18-in. 32B-S1 22:1-f4 3?i: Fi

. 3 M*C 12 la. 3?C 52

, 320;D.:!'i.1.

U3 51 .

32D 15

. k 32D-Fi #

3204E-12-in.32E-S2 32NF-12.h. 321 51 22f-fi

Prequency of Inspection:

a. Group I welds All welds will be inspected each 10 year interval with 25% completed at approximately 2.5 year intervals. .

b. Group II Welds .

All welds will be inspected during the first 10 year interval with 2%

completed at approximately 2.5 year intervals. ~

6

l g /

g .

g l

e 3.6/4.6-16 4

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

\

l Quad-Cities ppn 30 DELETE PAGES 3. 6/.1. 6-18 'nlRU 21 ,

kN

# e d

4 I

e a

9 1

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1 e

a>% /

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~ r 4

6 h 2 a

j s b

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* j I

l t '

3.6/4.6-17 ,

i

I;PR 30 .

O. -

l .

b. The reactor water temperature is b. Addition:i! tests shall be per-below 212 ' F and tSe reactor formed during the Sr.st oper: ting coolant systems are voned. . cycle ender an adequate m:mber of di!Terent environr.wntal wind

, conditions to enabic valid c..trapo-lation of the test re;ults.

c. No activity is being performed c. Secondary ccatainment capabi!ity which can reduce the shutdown to malataire an average 1/4 inch of margin below that specified in water vacuum ur.&r caha w.ad Specification 3.3.A. -

(2<ii<5 mph) conditions with a

' d. ' Die fuel cash or irmdiated fuel is fdter train flow rate cf not more not being moved in the reactor thm 4000 cfm shaU Ee demo:tstrated building. at each refueling outage prict to refueling.

2. The doors of the core 3. pray and RllR 2. Wheneve'r the LPCI mode of the RHR.

pump compartments shal! be closed at , and core spray subsystems are re-all times except dudag passage in or- quired to be operable. the doors of the der to consider the core spray system core spray and RHR pump compart-and LPCI mcde of the Ri!R system ments shall be verified to be closed operable. weekly.

Q, 3. ITSpeci6 cation 3.7.Cl canner he r et.

procedures shn!! be initiated to estab.

~ ~

9sh condit ons listed in SpeciScations D. Primary contain:acnt Isolation valves

. 7.Cl a throt.gh d.

1. The operational readiness of the D. Primary Containment Isclation Valves priraary containment isolation valves shall be demonstrated in I. During reactor power operating con- accordance with the Inservice ditions, all isolation valves listed iri -

Te, ting Program for pumps and Table 3.7-1 and all instrumcat line valves defined in section 1.o(r.r) flow check valves ivhich contact the of these specifications. Addi-primary coolant sprem shall be opera- , tional requirements are as follows:

ble except as specineJ in Speci6 cation-3.7.D.2, a. The operable isolation valves :

that are power operated and I automatically initiated shall s be tested for simulated automatic initiation, at

- least, once per operating cycle.

b. The instrument line flow check

, valves shall be tested for s proper operation, at least once per operating cycle.

O '

3.7/.t? -9 '

4 QUbD -CITIES DPR-30

~' . .

- 1

2. In the event any isolation valve speci- ._c. . When an isolation valve listed in Ta-fic'd in Table 3.7-1 becomes inopera- ble 3.7-1 is inoperable, the position of a ble,' reactor power operation may con- at least one other vilve in each line

~ tinue provided at least one valve in .having an inoperable valve shall be .

cach line having an inoperable valve is recorded daily. -

-in the mode corresponding to the iso-

' lated condition. .

3. If Specifications 3.7.D.! and 3.7.D.2 ~

cannot be met. an orderly shutdown '

shall be initiated and the reactor shall ,.

. be in the cold shutdown condition -

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . , .

4. The temperatme of the main steam- -

line air pilot valves shall be tess than ,

170* F cxcept as seccir:ed in Specifi-cations 3.7.D.5 and 3.7.D.6 below. ,

~

3. ' From and after the date that the tem-perature of any main steamline air _

pilot valve is found to be greater than - - - - -

170* F. reactor operation is permissi- '

b!d only during the succeeding 7 days .

unless~ the temperature of such valve is .

soo r reduced to less than 170 F, j, . .

provided the main steamline isolation ' '

valves are operable. ,

~

6. If Specification 3.7.D.5 cannot be me't. . ,

the main steamline isolation valve -

shall be considered inoperable und -

action takta in accordance with Speci-

. fication 3.7.D.2. '!

3.7/ 4 7-10 ,

n.

QUAD-CFl !ES DPit- 30 wiH I;e rep!ned with rdt' ers r;ualified pursu:nt to rei;ulatory guide position C.3.d of Regulatory Guide 1.52 l'evision I (June l'176). Once p:r operating cyt!c demonstr: tion of !!FPA filter pressure drop.operabinty er

- inlet heaters :t rated power, air distributiva to each IIEPA filter,and automatie initiation of each standby pas treatmerit sy2 tem circuit is necessary to sssure system performance capability). Note: basas witida parenthes.:s will not be rpp'ieah!c entil about 11:ccmber 31,1976, when equipment modifications are completed to allow increased testin;. , ,

. D. Primary Cuntainment isniation Vahes Those large pipes comprising a portion of the reactor coolant system. whose failure could result in *

. oncovering the reacmr core, are supplied with automatic isolation valves (except those lines needed for emergency core continf, system operation or containment cooling ).The closure times specitied herein are y ,

adequate to p: event loss of more coolant from the circumferential cupture of any of these lines outside

. the containment than from a steamline rupture. Therefore, this isolation valve closure time is sumcient '

.to prevent uncovering the core.

in order to assure that the doses that may result from a steamline break do not exceed the 10 CFR 100 guidelines, it is necessary that no fuct rod perforation resulting from the accident occur prior to closure of the main steamline isolation valves. Analyses indicate that fuel rod cladding perforations wou!d be avoided for main steam valve closure times, including instrument delay, as long as 10.5 seconds.

Ilowever, for added margin, the technical specifications require a vah e close time of not greater than 5

!.econds.

For reactor coolant system temperatures less tha'n 212* F, the containment could not become pressurized due to a loss-of-coolant accident.The 212' Flimit is based on preventing pressurization of the reactor

- . building and rupture of the blowout panels These valves are highly reliable, have low service requirement. ,nd most are normal ly closed. The initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolation (referenr- SAR Section 5.2.2 and Table

. 5.2.4 ).

The test interval at once per operating cycle for automatic initiation results in a failure probability of 1.1 x 104 that a line will not isclate. More frequent testing for valve operability resuhs in a more reliable system.

The containment is penetrated by a large number of small diameter inst ument lines which contact the primary coolant system. A program for periodic testing and examination of the flow check vahes in these lines is performed by blosying down the instrument line during a vessel hydro and observing the following two cor.ditions, which will verify that the flow check valve is operab!e:

A af O

e 3.7/4.7-18

QUAD-CITIES DPR-30

~. ,,

untess an additional line is sooner .

placed in service, providing both the

Unit and Unit 1/2 emergency dies:1 generators are operable. *

2. From and after the date the incoming power is hot available from any line, continued reactor operation is permis-I sible providing both the Unit and Unit

1/2 cmergency diesel generators are

, operating, all core and containment -

- cooling systems are operable, reactor power level is re'duced to 40% of rated.

~

and the NRC is notified within 24 .

. hours of the situation, the precautions to be taken during this period, and the ,

plaas for prompt restoration ofincom- *

- . ing power. ,

. 3. From and after the date rhat one of the '

two 125/250-volt battery systems is made or found to be inoperable for .

~ .

any :cason, continued reactor opera- -

tion is permissible only during the -

succeedin<t 3 days unless such battery' I. system is sooner'made operaMe. ,

D'.. Diesel Fuel D. Diesel Fuel There shall be a minimum of 10.000 gititons of Once a month the quantity of diesel fuel availa-ble shall be logged.

' diesel fuel supply on site for each diesel E*"* # * #~

~

. Once a month a sample of diesel fuel shall be

. . checked for quality.

3 C- - -- -

c - .

~

~ ~ ~-" E. ' Diesel-Generator Operability E.1' Diesei-Generator Cperability

~

i.

. 1. Whenever the reactor is in the Star- When it is determined that either the unit or shared diesel generator is inoperab!e. all low-

- tup /Ilot Standby or Run mode and the unit or shared diesel generators pressure core cooling systems and ai! loops of and/or their respective associated the containment cooling modes of the RHR buses are made or found to be inopera- system associated with the operab!: diesei gen-crator shall be operable.

ble for any reason. except as speeitied in Specification 3.9.E.2 below, contin- .,

' ued reactor operation is permissible ,

~

only during the succeeding 7 days pro- ,

vided that all of the low-pressure core .

. cooling and all loops of the contain-ment coaiing mode of the iniii system ano:iated with the operah!e diesel generator shall be operable, and two ofhite lines are available, if this re-quirement cannot 1 e met. an orderly shutdown shall be initiated and the e

3.9/.19-3

ML19276F895

Text

Subject:

Dear Sir:

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