ML20128K470

Proposed Tech Specs,Extending Allowed Period for Operation W/Rhr Pump Out of Svc from 30 to 60 Days

ML20128K470
Person / Time
Site:	Quad Cities
Issue date:	07/11/1985
From:	COMMONWEALTH EDISON CO.
To:
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ML20128K468	List: ML20128K461 ML20128K470
References
NUDOCS 8507240147
Download: ML20128K470 (5)

Text

QUAD CITIES DPR-29

e. Core spray header Ap instrumentation check Once/ day

^

calibrate Once/3 months 3

test Once/3 months

f. Logic system &ch efueling functional outage test

2. From and after the date that one of the 2. When it is determined that one core. spray subsystems is made or found core spray subsystem is inoperable, to be inoperable for any reason, continued the operable core spray subsystem, reactor operation is permissible only the LPCI mode of the RHR system, during the succeeding 7 days unless such and the diesel generators required subsystem is sooner made operable, provided for operation of such components if that during such 7 days all active components no external source of power were of the other core spray suosystem and the available shall be, demonstrated to LPCI mode of the RHR system and the diesel be operable immediately. The opera-generators required for operation of such ble core spray subsystem shall be components if no external source of power demonstrated to be operable daily were available shall be operable. thereafter.

3. The LPCI mode of the RHR system shall be 3. LPCI mode of the RHR system testing operable whenever irradiated fuel is in the shall be as specified in Specifica-reactor vessel and prior to reactor startup tions 4.5.A.l.a, b, c, d, and f from a cold condition. except that three RHR pumps shall oeliver at least 14500 gpm against a system head corresponding to a reactor vessel pressure of 20 psig.

42. From and after the date that one of the RHR l 4. When it is determined that one of pumps is maoe or found to be inoperable for the RHR pumps is inoperable, the any reason, continued reactor operation is remaining active components of the permissible only during the succeeding 30 -

LPCI mode of the RHR, containment days unless such pump is sooner made operaole cooling mode of the RHR, both core provided that during such 30 days the spray subsystems, and the diesel remaining active components of the LPCI mode generators required for operation of of the RHR, containment cooling mode of the such components if no external RHR, all active components of both core source of power were available spray subsystems and the diesel generators shall be demonstrated to be operable required for operation of yh components if immediately and the operable RHR no external source of power were available pumps daily thereafter.

shall be operable.

S M72;o g g q g .

P 3.5/4.5-2 Amendment 26 6312N L

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QUAD CITIES DPR-29 Ab. From the effective date of this amendment until Septenber.1,1985, one RHR pump may be inoperable and continued reactor opera-tionLis nermissible during the succeeding 60 days with~the same restilLEicas 02 3.5.A.4.a above.

5. From and after that the date the LPCI mode 5. When it is determined that the LPCI of the RHR system is made or found to be mode of the RHR system is inoperable, inopc:able for any reason, both core spray subsystems, the

?

E 3.5/4.5-2a l 6312N u

. . i QUAli. CITIES

• * *

• Dieg.29 8

O 1 , ,

45 LDitTING CONDITION FOR Ol'ERATION DASES A. Core Sprac and LPCI Mode of the ItllR S 3stem '

This specification assures that adequate ernergency cooling capability is availab!c whenever irradiated fuel ss in the rescior vessel. -

Based on the lessof. coolant ansJytiesi methods described in General Electric* Topical Report NEDO.20566 and 15e specific analysis in Reference 1 core cooling systems provide sufficient cochng to

.,,,,d.e core to dissipste the energv associated with thelossef<oolant acedent.tolmt calcu!ated f perature to less thsn 2:00'F. to avure that core geometry rems:ns intset.to limit c!sdding m actico to less tna 1%,and to lunit the calculated local metal. water scaction to less than .. 173. .:

The limiting conditions of cperation in Specifications 3.5.A.I through db 3.5.A.6 No single specify the co of operable subsystems to assure she availsbihty of the minimum coolin; systems note a os c. .

- . failure of ECCS equipment occurring during a less.of coolant accident under these limitin ,

of opers on will result in inadequate cooling of the reactor core.

Cots sprap distribution has been shown, in full semic iests of syster's simi!st in de Quad. Cities I and 2. to enced the mir:imum requiremeni: by at least 25%. In addition

, efeedveness has been demonstrated aileis than hsif the rated flow in sim

~ bester rods to duplicate the deC3Y heat Charseteristles

• of ir preuure has fallen to 90 psis.

i

( The 1.PCI mode of the RIIR system is desi;ned to provide emergency cooling to the c she event of a loss.or.coolsnt accident.This system functions in combination with the core spray sy

' so prevent cseessive fuel c!sdding temperature. The LPCI  ! 0 2 fi up8 mode to andof the R11R s the core spray subsystem provides adequate coolin; for break areas of approximste y .

including 4.18 'ft . the latter being the double. ended rec

? ~

j sutissstettrso The allowable repair times are established so that the average risk rate fortjsing repairthewould be l no The method and concept are described in Reference 3.

greater than the basic rate.

results developed in this reference, the subsystems repair period andis fotnd LPCI to be less constitute than half the test a one-out-of-two interval. This assumes that the core, spray system; however, theThe combined effect test interval of the specified two systems in Specification 4.5 towas limit excessive 3 months. cladding temp must also be considered.

Therefore, an allowable repair period which maintains the basic risk considering single failuresDu should be less than 30 days and this specification is within this period. This For July 1 to Septenter 1,1985, the repair period for one RHR pump multiple failures, a shorter interval is specified. Although it is recognized tnat the systems will function, a daily test is called for.

information given in Reference 1 provides a quantitative method to estimate allowable repair times, the lack of operating data tothe Therefore, support the analytical times stated approach in the specific items were prevents complete accep established 1

of this method at this time.

with due regard to judgment. .

' Should one core spray sWsystem become inoperable, the remaining core spray To sWsystem anl i

.' entire LPCI mode of the RHR systemTnisare available should the ne demonstration includes a are available, they are demonstrated to operable immediately. Based on judgments f j

manual initiation of the pumps and associated valves and diesel generato

,) was obtained.

' 3.5/4.5-11 i l i 6312N i,

Amendment No. 61 l

.I D . 1

.. r. a m a ex pump ocent, a nearly ik!! complement cf core and containment cooling DS~

• egripment is available.Three RNR pumps In conjunctirn with th2 core spray opse cooting fhaction. Because of the availabuity of the majority of the core cooling equipment, which win be demonstrated to be operable, a 30 day repair period isjustified. If the ! PCI mode of the RHR .

systemThe is not available, at Isas two RHR pumps must be available to fu!All the conuinenent coolin

7. day repair period is set on this basis.

RHR Servlee Weser *

• The containssent cooling anode of the RHR systern is provided to remove heat energy ikom the aonesia'anat in the event da less of. coolant accident. For the Sow pa-d. the containment long-term pressure is Emited to 1sss than 8 psig and is therefbes more than ample se provide the required heat removal capabDiry (reference SAR Section 12.3.2). *

-- g

' Th3 Containment Cooling mode of the RWE Systes consista of two looos. W Each loop consists of 1 Heat Eschenger, 2 RMA Pumps, and the associated q valv:s, piping, electrical equipment, ans instrumentation. The "S* loop E

d

~en esin unit contains 2 RHR Service Water Pumps. During the perico from 2 ,,,,

Nov eDer 2a, 1981, to .MF 1,1982, the "A" loop on each unit say * '

utilize the "A** and "S" RNR Service Water Pumps. free Unit 2 vie a eress-tie line. After .rair 1,19s2, each "A" loop will contain 2 RMR [

servise unter Pumps.

' Either set of equipment is cepsble c(perfbnaigthe conmia'n==e l

amoung ihnesion. I.am of one RHR service water pump does not ser:onslyjeoprdize the sonestavan=e g

. amoung capabi5ty, as any one of the rammining three pumps can sadsfy thnooling requirements. Since o

thne fa some redandancy left, a 30alay repair period is adequate. lam done loop of du sone =tarmene sooEng mode of the RHR synsa leaves one remaining system to perfons the containment cooling fhassion. The operable system is demonsarstod to be operable ecch day when the above condition ocents.

y Raamd on the fhst that when one loop of the containment cooling mode of the RNR sysism becomes

• Inoperable, only one symem remains, which is testad daDy, a 7. day repair period was speciand.

j f m sb.Prasm.ca.ia nIss ales ,

U e .The higby . coolant injecdon subsystem is provMed to adequately cool the core for an pipe breaks O anaDer than those for which the I.PCI mode of the RHR system or core spray subsynams can prosect the )

i The EPCI marm this requirement withoes the use ofofhite electrical power. For the pipe breaks fbr which the HPCl is latended to fhaction, the core never uncovers and is continsonsly cooled. thus no cladding Q

u damage occurs (reibrancs SAR Secti on 4.2.13). The repair times for the limiting conditions of operation' were set considering the use of the NPQ as part af the isolation cooling system. Q Assematic Pressure RaBar -

q The relidvalves of the astonistic pressure neuersubsystem are a backup to the HPQ subspens. They I anable the core spray subsystern or LPCI mode of the RHR system to provide protection agains the suas pipe break in the event ofHPQ thilure by d.y. L.isg the reactor venel rapidly enough to actuate the @N -

ease spray subsyssenes or LPC mode of the RNR system.The core spray subsystem,and/or.the L.PCI l nods of the RHR symem provide sufBciset ihrw of coolant to Emit fhs!claddhgtemperaturestalen_than-

2200*F to amare that esse geometry remains haast, to Emit she'sers widIsted metal water rencdon to km then 15, and to Emit the alenistad lemt metal. water Isaction ta less than 175.

l Imst of 1 of tha must vabes afEneas the pcomme seleving esymbDity and, therefore, a 7 day repair period.h specified. Lass of more than ons antist valve sH8---'1y reduces the'presure rallef capabDity,the a 24. hour <

~sepair period is specifInd based on the HFC! system avaitsbGity during this peded.

! ROC The Rcc system is provided to supply continsons makeup water to the reactor core when the reactor is isolated Dom the turbine and when the feedwater system is not available. Under these conditions the jumping capacity of the ROC syssets is sufReient to maimain,ibe water livel above the core withou't any Cher water system in operation. If the water level in the reactor vessel decreases to the RCIC'aitiation

! . vel, the system automaticaDy starts. The system may also be manuaDy initiated at iny time.

~

~

3.5/d.S-12~ ~~~ ~ ~

Amendsent No.. fJ -78

1

• ~ ~ %,

QUA0 CITIES OPR- 29 B '

' For core flow rates less than rated, the steady state MCPR is increased by the formula given in the specification. This ensures that the MCPR will be maintained greater than that specified in Specification 1.1.A even in the event that the motor-generator set speed controller causes the scoop tube positoner position.

for the fluid coupler to move to the maximum speed References

1. " Loss-of-Coolant Analysis Report for Dresden Units 2, 3, and Quan Cities Units 1, 2 Nuclear Power Stations," NE00-24146A*,

April, 1979

2. " Generic Reload Fuel Application," NEDE-2401 MP-A** 4

(

3. I. M. Jacobs and P. W. Marriott, GE Topical Report APED 5736, '

' Guidelines for Determining Safe Test Intervals and Repair Times for Engineereo Safeguards " April, 1969.

4. " Qualification of the One-Dimensional Core Transient Model for ,

Boiling Water Reactors

• General Electric Co. Licensing Topical  ;

- s Report NEDO 24154 Vols. I and II and NEDE-24154 Vol. III as

• supplementeo by letter datee September 5, 1980 from R. M. t.

Buchholz (GE) to P. S. Check (NRC). ,

5

5. Letter, R. H. Suchholz (GE) to P. S. Check (NRC) dated January 't 19, 1981 "0DYN Adjustment Methods For Determination of Operating #

o Limits".

dg Approved revision at time of plant operation. I e

Appro-ved revision number at time reload fuel analyses are ,f performed.

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d 3.5/4.5-15 l

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knendment No. 37 , 83 l

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ew and approval. We have reviewed this amendment [l, ,j -

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significant hazards consideration exists. Our ~ .i ' - -J #

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wm-H. R. Denton July 15, 1985 Please direct any questions you may have concerning this matter to this office.

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

Very truly yours, H. L. Massin Nuclear Licensing Administrator 1m Attachments- A: Background and Discussion 8: Technical Specification Change to NPF-11 C: Evaluation of Significant Hazards Consideration j cc: Region III Inspector - LSCS A. Bournia - NRR G. Wright - State of Ill SUB ' N to befo 0AtoSg1-day this[

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Notary PGblic d dhh Q

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