Text

Proposed TS Sections 3.5/4.5 Re Core & Containment Cooling sys,3.6/4.6 Re Primary Sys Boundary & 3.7/4.7 Re Containment Sys
	ML20085C760
Person / Time
Site:	Quad Cities
Issue date:	10/02/1991
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20085C759	List: ML20085C757; ML20085C760;
References
	NUDOCS 9110110110
	Download: ML20085C760 (111)
	v • d • e

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s PROPOSED TECH SPEC o

TS 3.5/4.5 g 'UORE & CONTAINMENT COOLING SYSTEMS" O

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS SPECIFICATIONS LIM 7 TING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS A. Core Spray subsystems - A. Core Spray Subsystems -

Plant Operating Plant Operating Two core spray subsystems shall Core spray subsystem testing be OPERABLE.

Item Frequency APPLICABILITY

' 1. Valve Position M OPERATIONAL MODES 1, 2, and 3. Verification ACTIONSt 2. Flow Rate Test- Per core spray pumps Specifi-

1. From and after the date that shall deliver at cation I one of the core spray sub- least 4500 gpm 4.0.E systems is made or found to againct a system be inoperable, continued head correspond-reactor operation is permis- ing to a reactor sible only during the succe- vessel pressure eding 7 days unless such of 90 psig.

, subsystem is sooner made OPERABLE, provided that 3. Core Spray Header Ap during such 7 days all Instrumentation active components of the other core spray subsystem a) CHANNEL CHECK D and the LPCI mode of the RHR system and the diesel b) CHANNEL R

> generators required for CALIBRATION operation of such components if no external source of c) CHANNEL Q power were available shall FUNCTIONAL TEST be OPERABLE.

4. SIMULATED PUTO- R

> 2. With a core spray subsystem MATIC ACTUATION header Ap instrumentation CHANNEL inoperable, restore $. LOGIC SYSTEM R the inoperable CHANNEL to FUNCTIONAL OPERABLE status within 72 TEST hours or determine the header Ap locally at least 6. Doors of Pump W p

once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; other- Compartment Closed wise, declare the associated core spray subsystem inoperable.

3.5/4.5-1

O QUAD CITIES UNITS 1 & 2 DPR-2E & DPR-30 9 3. If the requirements of ACTION 3.5.A.1 cannot be met, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 4 hours.

B. LPCI Mode of the RJ{R System - B. LPCI Mode of the RIP System -

Plant Operating Plant Operating The LPCI mode of the RHR system LPCI mode of the RalR system shall be OPERABLE. testing shall be as specified in O Specifications 4.5.A.1, 2, 4, 5 hPPLICABILITY: and 6 for the Core Spray subsys-tem except that each LPCI divi-OPERATIONAL MODES 1, 2, and 3. sion (two RHR pumps per division) shall deliver at least 9000 gpm ACTIONSt aghinst a system head correspond-

ing to a reactor vessel pressure

1. From and af ter the da?.e that of 20 psig, with a minimum flov one of the RHR pumps As made valve open.

er found to be inoperable, continued reactor operation is permissible only during g the succeeding 30 days unless such pump is cooner made OPERABLE, provided that during such 30 days the remaining active components of the LPCI mode of the RHR,

, containment cooling mode of the RHR, all active components of both core spray subsystems, and the diesel generators required for operation of such components if no external 4 source of power were available shall be OPERABLE.

2. From and after the date that the LPCI mode of the RHR system is made or found to 3 be inoperable, continued reactor operation is permis-sible only during the succe-eding 7 days unless it is sooner made OPERABLE, provi-9 3.5/4.5-2 S

9 QUAu CITIES UNITS 1 & 2

- - DPR-29 & DPR-30 O ded that during such 7 days all active components of both core spray subsystems, the containment cooling mode of the AltR (including any two Ri!R pumps), and the Q diesel generators required for operation of such com-ponents if no external source of power were available n>4ll be OPERABLE.

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3. If the reqairements ACTION 3.5.B.i or 3.5.B.2 of cannot be met, be in at least ilOT S!!UTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SilUTDOWN within tha fo3 lowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. Containment Cooling Mode of the C. Containment Cooling Mode of the RHR System - Plant Operating RHR System - Plant Operating Two loops of the containment RHR service water subsystem cooling mode of the RHR system testingt p shall be OPERABLE.

Item Frequency APPLICABILITY:

1. Valve Position H OPERATIONAL HODES 1, 2, and 3. Verification

2. Flow Rate Test - Per g ACTIONS 1 each RHR service Specif1-

1. From and after the date that water pump shall cation one of the RJ1R service water deliver at least 4.0.E pumps is made or found to be 3500 gpm against inoperable, continued a pressure of reactor operation is 198 psig.

O permissible only during the ,

succeeding 30 days unless 3. A LOGIC SYSTEM R such pump is sooner made FUNCTIONAL TEST OPERABLE, provided that during such 30 days all other active components of the containment cooling mode

] of the RHR system are OPERABLE.

2. From and af ter the date that g 3.5/4.5-3 0

" - " ' " " ' ~ ' N' dI , , . , , , , . , , , - . . . . . , ,, , , , , . , , .

L QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 O one loop of the containment cooling mode of the RHR system is made er found to be inoperable, continued reactor operation is permis-sible only during the succe-3 eding 7 days unless such subsystem is sooner made OPERABLE, provided that all active components of the other loop of the contain-ment cooling mode of the RHR system,. b " core spray sub-

) systems, .d both diese; generat't- required for operat' of such components if no texternal source of

> power were available shall be OPERABLE.

~3

3. If the requirements of

' ACTION 3.5.C.1 or 3.5.C.2 cannot be met, be in at least HOT SHUTMWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in 3 COLD SHUTDO'fN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D. Containment Cooling Spray Loops D. Containment Cooling Sprey Loops

> Containment cooling spray loops Surveillance of the containment

" consisting of two drywell loops cooling spray loops shall consist

, and onn suppression chamber loop of performing sn air test on the shall be OPEAABLE. drywell spray headers and nozzles and a water spray test on the APPLICABILTTY? suppress 37n chamber spray header and nozzlsa during each 5-year OPERATIONAL MODES 1, 2, and 3. period.

ACTION:

1. With a maximum of one dry-wel) spray loop inoperabic, reactor operation may con-g tinue for up to 30 days.

2. If the requirements of ACTION 3.5.D.1 cannot be met or if the containment 3.5/4.5-4 i

S c_-. - _ . _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _

)

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o QUAD CITIES UNITS 1 & 2

-DPR-29 & DPR-30 3 cooling spray loops, are otherwise nnoperable, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

9 HPCI Subsystem - Plant Operating E. HPCI Subsystem - Plant Operating E.

The HPCI subsystem shall be Surveillance of HPCI subsystem OPERABLE. shall be performed as specified below with the following limi-APPLICABILITY: tations. For . Ttem 4. 5. E. 3, the g provisions of Specification 4.0.D OPERATIONAL MODE 1 and are not applicable provided the OPERATIONAL MODES 2 and 3 test is perf ormed vichin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever REACTOR VESSEL PRESSURE after REACTOR VESSEL PRESSURE is is greater than 1b0 psig, adequate to perform the test. In addition, the testing required by J ACTION: Item 4.5.E.3.a shall be completed prior to exceeding 325 psig REAC-

1. During startup following a TOR VESSEL PRESSURE. If HPCI is REFUELING OUTAGE, if ehe made inoperable to perform over-flow rate testing require- speed testing, 24 hours is ments of 4.5.E.3 cannot be allowed to com.nlete the tests 3 met, continued reactor before exceeding 325 psig.

operation is not permitted.

The HPCI subsystem shall be Item Frequency declared inoperable, and the provisions of ACTION 3.5.E.3 1. Valve Position M shall be implemented. Verification 3 2. Except for the limitations 2. Flow Rate Test - Per of ACTION 3.5.E.1, if the HPCI pump shall de- Specifi-HPCI subsystem is made or liver at least 5000 cation-found to be inoperable, con- gpm against a sys- 4.0.E tinued reactor operation is tem head corres-permissible only during the ponding to a reac -

O succeading 14 days unless tor vessel pres-such subrystem is sooner sure of 2 1150 psig made OPERABLE, provided that when steam is being during such 14 days tne supplied to the automatic pressure relief turbine at 920 to subsystems, the core spray 1005 psig, g subsystems, LPCI mode of the RHP, system and the RCIC 3. Flow Rate Test - During system are OPERABLE. HPCI pump shall de- startup liver at least 5000 following

3. If the requirements of gpm against a sys- a REFUEL-g 3.5/4.5-5 0

O QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 0 ACTION 3.5.E.1 or 3.5.E.2 tem head corres- ING OUT-cannot be met, be in at ponding to a REAC- AGE lesst HOT SHUTDOWN within TOR VESSEL PRES-the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce SURE of:

REACTOR VESSEL PRESSURE to s 150 psig within the next a. t 300 psig when 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . steam is being O supplied to the turbine at 250 to 325 psig, and

b. h 1 50 psig when steam is being rU supplied to the turbine at 920 to 1005 psig.

4. SIMULATED AUTOMATIC R ACTUATION Test O

5. LOGIC SYSTEM R FUNCTIONAL TEST F. Automatic Pressure Relief F. Automatic Pressure Relief Subsystems - Plant Operating Subsystems - Plant Operating C' pressure relief The automatic pressure relief Automatic subsystem including at least five subsystem testing:

relief valves shall be OPERABLE.

1. At least once per 18 months APPLICABILITY: by manually opening each automatic relief valve when O OPPRATIONAL MODE 1 and REACTOR VEScEL PRESSURE is 2 OP'; %TIONAL MODES 2 and 3 150 psig r t observing that:

whe. sever REACTOR VESSEL PRESSURE is greater than 150 psig. a. T' m .1R ol valve or b,t .a salve position ACTIONS: respmda accordingly, O or

1. With one of the above re-quired automatic pressure b. There is a correspond-relief valvea inoperable, ing change in the provided the HPCI subsystem, measured steam flow, all active components of g both core spray subsystems, The provisions of Specifi-and the LPCI mode of the RHR cations 4.0.D are not ap-system are OPERABLE, restore plicable provided the sur-the inoperable relief valve veillance is performed to OPERABLE status within la within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after REAC-3 3.5/4.5-6 I

I D

O QUAD CITIES UNITS 1 & 2

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DPR-29 & DPR-30 O days or de in at least HOT TOR VESSEL PRESSURE is ade-SHUTDOWN within the next 12 quate to perform the test.

hours and reduce REACTOR VESSEL PRESSURE to 5 150 2. A LOGIC SYSTEM FUNCTIONAL psig within the next 24 TEST shall be performed each hours. REFUELING OUTAGE.

O A SIMULATED AUTOMATIC With two or more of the 3.

automatic pressure relief ACTUATION which opens all valves inoperable, be in at pilot valves shall be least HOT SHUTDOWN within performed each REFUELING the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce OUTAGE.

REACTOR VESSEL PRESSURE to O s 150 psig within the next 24 hourr.

G. Reactor Core Isolation Cooling G. Reactor Core Isolation Cooling System - Plant operating System - Plant Operating O The BCIC system shall be RCIC system testing shall be as OPERABLE. specified in Specification 4.5.E for the HPCI subsystem, except APPLICABILITY: that the RCIC pump shall deliver at least 400 gpm for the system OPERATIONAL MODE 1 and flow tests and the frequency of D OPERATIONAL MODES 2 and 3 the flow rate test for the RCIC whenever REACTOR VESSEL PRESSURE pump in Specification 4.5.E.2 is greater than 150 psig. shall be quarterly.

ACTIONSi m 1. During startup following a

REFUELING. OUTAGE, if the flow rate testing require-ments of 4.5.E.3 cannot be met, continued reactor operation is not permittcd.

The RCIC system shall be J declared inoperable, and the provisions of ACTION 3.5.G.3 shall be implemented.

2. Except for the limitations of ACTION 3.5.G.1, if the 4 RCIC system is made or found to be inoperable, continued reactor operation is permis-sible only during the succe-eding 14 days unless such g 3.5/4.5-7 l

3

O l QUAD CITIES UNITS 1 & 2

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DPR-29 & DPR-30 O system is sooner made OPERABLE, provided that during such 14 days the HPCI subsystem is OPERABLE.

3. If, the requirements of O ACTION 3.5.G.1 or 3.5.G.2 cannot be met, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce REACTOR VESSEL PRESSURE to s 150 psig within the next 24 h "'**

O H. Minimum Core and Containment H. Minimum Core and Contaimnent cooling System Availability - Cooling System Availability -

Plant Shutdown Plant Shutdown The core and containment cooling Surveillance requirements to O systems shall be OPERABLE as assure that minimum core and follows: containment cooling systems are available have been specified in

1. At least the following Specifications 4.5.A and 4.5.B.

pumps, with each having an OPERABLE flowpath capable of O taking suction from the suppression pool or the condensate storage tank and transferring the water to the reactor vessel, shall be OPERABLE:

O a. Two core spray pumps, or

b. Two RHR pumps, or

c. One core spray pump O and one RHR pump.

2. All low pressure core and containment cooling systems may be inoperable in OPERATIONAL MODE 5 provided O- that:

a. The reactor vessel head is removed, and g 3.5/4.5-8 3

O QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 O b. The reactor cavity is flooded, and

c. The spent fuel pool gates are removed, and O d. The spent fuel pool water level is main-tained above the low level alarm point, and

e. The reactor cavity O water temperature is below 140 *F.

APPLICABILITYt OPERATIONAL MODES 4 and 5.

I ACTIONS:

1. With one of the required pumps and/or associated flow paths 1. sperable, restore at least two pumps and

,O associated flow paths to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or suspend all operations with a potential for draining the reactor vessel.

O With both of the required 2.

pumps and/or associated flow paths inoperable and/or the provisions of LCO 3.5.H.2 not met, suspend CORE ALTER-ATI Ns and all operations O with a potential for draining the reactor vessel.

Restore at least one pump and associated flow path to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or establish SECONDARY O CONTAINMENT INTEGRITY within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Q 3.5/4.5-9 O l

)

QUAD CITIES UNITS 1 & 2

- ~ DPR-29 & DPR-30

) I. Suppression Chamber Requirements I. Suppression Chamber Requirements for Core and Containment Cooling for Core and Conuinment Cooling Systems Systems The suppression chamber shall be Surveillance of the suppression OPERABLE. chamber shall be as follows:

C The suppression pool water

1. In OPERATIONAL MODES 1, 2, 1.

and 3 with a contained water level shall be checked at volume corresponding to at least once per day, least a water level of 14 '1" above the bottom of the 2. In OrERATIONAL MODES 4 or 5 g suppression chamber, with the suppression chamber level less than the limit or

2. In OPERATIONAL MODES 4 or 5 drained, at least once per with a contained water 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify the condi -

volume corresponding to at tions of LCO 3.5.I.2 or least a water level of 7 3.5.I.3 are met, as feet above the bottom of the applicable.

@ suppression chamber, except that the suppression chamber level may be less than the limit or drained provided that:

D a. No operations are performed that have a potential for draining the reactor vessel,

b. The reactor mode g switch is locked in the SHUTDOWN or REFUEL position,

c. The condensate storage tank contains at least 230,000 available O gallons of water, equivalent to a level of 9.5 feet, and

d. The core cooling systems shall be G OPERABLE per Specification 3.5.H.1 requirements.

3. In addition to the g 3.5/4.5-10 j.

O.

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 O requirements of Lc0 3.5.I.2, in OPERATIONAL MODE 5, the suppression chamber is net required to be OPERABLE provided that:

O a.' The reactor vessel head is removed, and

b. The reactor cavity is flooded, and O c. The fuel storage pool gates are removed, and

d. The fuel storage pool water level is maintained above the low level alarm point, 9 and

e. The reactor cavity unter temperaturo is below 140 'F.

Q APPLICABILITY:

OPERATIONAL MODES 1, 2, 3, 4, and 5.

ACTIONS:

0 In OPERATIONAL MODES 1, 2 or 1.

3 when the suppression chen-ber water volume is less than that corresponding to at least a water level of g 14'1" above the bottom of ,

the suppression chamber, restore the water level to within the limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN 9 within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. In OPERATIONAL MODES 4 or 5 with the suppression chamber g 3.5/4.5-11 6

l QUAD CITIES UNITS 1 & 2

- DPR-29 & DPR-30 I

N water level less than the above limit, drained, or with the suppression chamber otherwise inoperable and the above required conditions not satinfied, suspend CORE D ALTERATIONS and e11 op ara-tions that have a potential for draining the reactor vessel and lock the reactor mode switch in the SHUTDOWN position. Establish SECON-g DARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

J. Maintenance of Filled Discharge J. Maintenance of Filled Discharge Pipes for Core Spray, LPCI Mode Pipes for Core Spray, LPCI Mode of RHR and the RHR Loops Used to of RHR, and the RHR Loops Used to Satisfy Specification 3.5.H Satisfy Specification 3.5.H For Core Spray or the LPCI mode The following surveillance of RHR or the RHR loops used to requirements shall be followed to satisfy Specification 3.5.H, the assure that the discharge piping discharge piping from the pump of the core spray or the LPCI discharge to the system motor mode of RHR or the RHR loops used 5 operated isolation valve shall be to satisfy Specification 3.5.H filled and discharge pipe pres- are filled.

sure maintained at greater than 40 psig. 1. At least once per month verify by venting and APPLICABILITY: observing water flow at the y high point vents, that the Whenever core. spray or the LPCI system piping from the pump mode of RHR or the RHR loops used discharge valve to the to satisfy Specification 3.5.H system isolation valve is are required to be OPERABLE. filled with water.

ACTIONS: 2. The pressure switches which monitor the discharge lines

1. With a discharge line keep and the discharge of the filled pressure alarm ins- f311 system pump to ensure trumentation CHANNEL in- that they are full shall operable, perform Surveil- have a CHANNEL FUNCTIONAL lance Requirement 4.5.J.1 at TEST at least once per month-least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . and a CHANNEL CALIBRATION at least once per REFUELING

2. With the pressure in any of OUTAGE. The prescure thase systems less than 40 switches shall be set to 3.5/4.5-12

l l

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30

) psig, initiate corrective alarm at a decreasing action within one hour, pressure of 2 40 psig.

Restore the discharge pipe pressure to within limits within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or declare the affected systems

$ inoperable and follow the appropriate system ACTION requirements.

K. Maintenance of Filled Discharge K. Maintenance of Filled Discharge Pipes for RCIC and HPCI Pipes for RCIC and HPCI O The discharge pipe pressure for The following surveillance the HPCI and RCIC systems shall requirements shall be followed to be ensured by maintaining the assure that the discharge piping level in the Contaminated Conden- of the HPCI and RCIC are filled.

sate Storage Tanks (CCSTs) at or above 9.5 feet. 1. Whenever the HPCI and/or S RCIC system (s) are lined up APPLICABILITY: to take suction from the CCST's, the water level in Whenever the HPCI or RCIC systems the CCST's shall be verified are required to be OPERABLE. to be 2 9.5 feet at least once per day.

$ E TIONSt.

1. If the CCST level falls 2. Whenever the HPCI and/or below 9.5 feet, restore the RCIC system (s) are lined up level within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or to take suction from the line up both HPCI and RCIC suppression chamber, the to take a suction from the discharge piping of the HPCI g suppression chamber. and/or RCIC shall be vented from the high point of the

2. If the requirements of system and water flow ACTION 3.5.K.1 cannot be observed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

met, declare the affected systems inoperable and 3. Every month the HPCI and follow the appropriate RCIC discharge piping shall

system ACTION requirements. be vented from the high point and water flow observed.

L. Condensate Pump Room Flood L. Condensate Pump Room Flood.

O Protection Protection The systems installed to prevent 1. The piping and electrical or mitigate the consequences of penetrations, bulkhead flooding in the condensate pump doors, and submarine doors 9 3.5/4.5-13 i

O

I l

b QUAD CITIES UNITS 1 & 2

- - DPR-29 & DPR-30

) room shall be OPERABLE as for the vaults containing l follows: the RHR service water pumps l and diesel generetor cooling

1. Vaults containing RHR pumps shall be checked service water pumps and during each OPERATING CYCLE diesel generator cooling by pressurizing to 15 1 2 O pumps shall be sealed and watertight. Door operation psig and checking for leaks using a soap bubble solu-for passage is permitted. tion. The criteria for I acceptance shall be no l 2. Tne condenser pit visible leakage through the l

level switches shal' soap bubble solution.

g the condencer circt.' t water pumps and alarm i: '

At least once per OPERATING control room if water le CYCLE, the following flood in the condenser pit exct yroteccion level switches a level of 5 feet above u. dall be functionally tested pit floor.

give the following h APPLICABILITY:

a. Turbine building Whenever the RHR service water or equipment drain sump diesel generator cooling pumps high level, are required to be OPERABLE.

b. Vault high level.

) ACTIONS:

3. At least once per OPERATING

1. With one or more of the CYCLE, the RHR service water vaults not sealed and vault sump pump discharge leaktight, restore the vault check valves outside the to OPERABLE status within 12 vault shall be tested for g hours. Otherwise, declare integrity, using clean the affected RHR service demineralized Vater.

water or diesel generater cooling pumps inoperable and 4. The condenser pit 5-foot follow the appropriate trip circuits for each system ACTION requirements. channel. shall be tested as followr r 2. With one of the trip circuits inoperable, action Item Frequency shall be initiated within one hour and completed a. CHANNEL M within four hours to place FUNCTIOFAL TEST the failed trip circuit in a

!$ trip condition. Restore the b. LOGIC SYSTEM R inoperable trip circuit to FUNCTIONAL TEST OPERABLE status within seven days. Otherwise, declare the affected RHR service

g 3.5/4.5-14 9

.O l-QUAD CITIES UNITS 1 & 2 I -

DPR-29 & DPR-30 O vater or diesel generator cooling pumps inoperable and follow the appropriate system ACTION requirements.

3. With all of the alarm cir-O ' cuits inoperable for more than 30 days, in lieu of a LER, prepara and submit a Special Report to the Com-mission within the next 10 days outlining the cause of O th* **1'"" ti " *"* th*

plana for restoring the instrumentation to OPERABLE status.

M. Average Planar LHGR M. Average Planar LHGR O All AVERAGE PLANAR LINEAR HEAT Daily during steady-state opera-GENERATION RATES (APLHGRs) for tion above 25% RATED THERMAL all the rods in any fuel assem- POWER, the average planar LHGR bly, as a function of AVERAGE shall be determined.

PLANAR EXPOSURE, at any axial location, shall not exceed the Q maximum APLHGR specified in the CORE OPERATING LIMT.TS REPOAT.

APPLICABILITY:

OPERATIONAL MODE 1, when Thermal

""*" 1* 9"****" th*" "

9"

1

O 25% of RATED THERMAL POWER.

ACITOFS:

1. With an APLHGR exceeding the limiting value, initiate O corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce thermal power to less than 25% of RATED THERE%

O POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

O 3'5/4 5-15 l

l O

O I

QUAD CITIES UNITS 1 & 2

- - *DPR-29 & DPR-30 0 N. Local utGR N. Local LucR All LINEAR HEAT GENERATION RATES Daily during steady-state opera-(LHGRs) for any rod in any fuel tion above 25% of RATED THERMAL assembly at any axial location POWER, the local LHGR shall be shall not exceed the maximum determined.

O allowable ulGR specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY!

OPERATIONAL MODE 1, when Thermal Power is greater than or equal to 9 25% of RATED THERMAL POWER.

ACTIONS:

1. With a LHGR exceeding the limiting value, initiate O corrective action within 15 minutes and restore LHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce thermal power to less than 25% of RATED THERMAL POWER O within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

O. Minimum Critical Power Ratio O. Minimum Critical Power Ratio (MCPR) (MCPR)

The MCPR shall be equal to or The MCPR shall be determined n greater than the MCPR limit daily during steady-state power

specified in the CORE OPERATING operation above 25% of RATED LIMITS REPORT. For core flows THERMAL POWER.

other than rated, these nominal values of MCPR shall be increased by a f actor of K, where K, is as specified in the CORE OPERATING O LIMITS REPORT.

APPLICABILITY;_

OPERATIONAL MODE 1, when Thermal Power is greater than or equal to O 25% of RATED TilERMAL POWER.

O 3 5/4 5-16 l

O

Q QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 0 3ctrow,

1. With MCPR -less than the applicable MCPR limit, initiate- corrective action within 15 minutee -and 0- restore MCPR to withl" .le required limits witnin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce thermal power to less than 25% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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'O 3.S/4.5-17 O-

O QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 3.5 LIMITING CONDITIONS FOR OPERATION BASES A&B Core Spray Subsystem and the LPCI Mode of the RHR System - Plant Operating g' This specification assures that adequate emergency cooling capability is available during plant operation in Modes 1,2, and 3.

Based on the loss-of-coolant analyses included in References 1 and 2 and in accordance with 10 CFR 50.46 n

and Appendix K, core cooling systems provide sufficient cooling to the core to dissipate the energy associated with tha 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 1%;

and, t limit the calculated local metal-water reaction O to less than 17%.

The LIMITING CONDITIONS FOR OPERATION in Specifications 3.5.A and 3.5.B specify the combinations of OPERABLE subsystems to assure the availability of the minimum g cooling systems noted above.

Core spray distribution has been chown, in full-scale tests of systems similar in design to that of Quad Cities 1 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

'O simulated fuel assemblies with heater rods to 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 fallen to 90 psig.

O The LPCI mode of the RHR system is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident. This system functions in combination with the core spray subsystem to prevent excessive fuel cladding temperature. The LPCI mode of p" the RHR system, in combination with the core spray subsystem, provides adequate cooling for break areas of approximately 0.05 ft2 up to and including 4.26 ft, 2 the latter being the double-ended recirculation line break with the equalizer line between the recirculation loops closed and without assistance from the high-pressure m rgency core cooling subsystems.

O B 3.5/4.5-1

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Q QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 O'

If one core spray subsystem becomes inoperable, the remaining core spray subsystem and the entire LPCI mode of the RHR system are available should the need for core cooling arise. Based on judgement for the reliability of the remaining systems, i.e, the core spray and LPCI, a O 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 F.ajority of the core O~

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.

x0 When a core spray or LPCI subsystem (or pump) is inoperable, there is a requirement that other emergency core cooling systems be OPERABLE. The verification of OPERABILITY, as used in this context, for these other systems means to administrative 1y check by examining logs g" or other information to determine if certain components / systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component / system.

q C. Containment Cooling Mode of the RHR System - Plant Operating The containment cooling mode of the RHR system is provided to remove heat energy from the containment in the event of a loss-of-coolant accident. For the flew n 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 Cooling mode of the RHR System consists q of two loops. Each loop consists of 2 RHR Service Water pumps, 1 Heat Exchanger, 2 RHR Pumps, and the associated valves, piping, electrical equipment, and instrumenta-tion. Either sct of equipment is capable of performing the containment cooling function. Loss of c>ne RHR service water pump does not seriously jeopardize the n containment cooling capability, as any one of the remaining three pumps can satisfy the cc,oling B 3.5/4.5-2 i

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O QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 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 RHR. system leaves one remaining system to perform the containment cooling function. Based on the fact that when one loop of the

c. containment cooling mode of the RHR system becomes

" inoperable, only one system remains, a 7-day repair period was specified.

When a containment cooling subsystem (or pump) is inoperable, there is a requirement that other emergency

, core cooling systems be OPERABLE. The verification of

'3' OPERABILITY, as used in this context, for these other systems means to administratively check by examining logs or other information to deternine if certain components / systems are out-of-service for maintenance or other- reasons. It does not mean to perform the surv illan requirements need d t dem nstrate the O OPERABILITY of the component / system.

D. Containment Cooling Spray Loops The containment cooling spray loops consist of two drywell spray loops, each being 100% capacity, and one

.O suppression chamber spray loop. The containment cooling spray loops are an integral part of the containment cooling cubsystem of the RHR system. The spray headers of the RHR system are not intended to be placed in service until the core cooling requirements of the low pressure coolant injection system have been satisfied.

U, When the reactor water level is restored to two-thirds of the core height, level switches.give a permissive which permits opening of the containment spray valves. These requirements may be bypassed by the operator using a keylock switch in the control room. Because of the

, redundancy in-the drywell spray loops, a 30 day repair O period is justified with one spray loop inoperable.

E. High Pressure Coolant Injection Subsystem - plant

, Operating The high pressure coolant injection subsystem is provided U, to adequately cool the core for all pipe breaks smaller than those for which the LPCI mode of the RHR system or core spray subsystems can protect the core.

The HPCI subsystem meets this requirement without the use

. of offsite electrical power. For the pipe breaks for O which the HPCI subsystem is intended to function, the B 3.5/4.5-3 i

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0-QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 core is never uncovered and is continuously cooled, thus no cladding damage occurs (reference SAR Section 6.2.5.3). The repair times for the LIMITING CONDITIONS FOR OPERATION were set considering the use of the HPCI as part of the isolation cooling system.

O When the HPCI subsystem is inoperable, there is a require.mont that other emergency core cooling systems be OPERABLE. The verification of OPERABILITY, as used in this context, for these other systems means to administrative 1y check by examining logs or other I"# ""*ti " t d***""i"" if ""t*I" "P "*"t"/*Y"t"""

O are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the l component / system.

F. Automatic Pressure Relief Subsystem - Plant Operating The relief valves of the automatic pressure relief subsystem are a backup to the HPCI subsystem. They enable the core spray subsystem and LPCI mode of the RHR system to provide protection against the small pipe break in the event of HPCI subsystem failure by O depressurizing tiie reactor vessel rapidly enough to actuate the core spray subsystem and the LPCI mode of the RHR system. The core spray subsystem and/or the LPCI mode of the RHR system provide sufficient flow of coolant to limit fuel cladding temperatures to less than 2200*F, to assure that core geometry remains intact, to limit the O core wide clad metal-water reaction to less than it, and to limit the calculated local metal-water reaction to less than 17%.

Automatic pressure relief automatically controls five safety / relief valves although the safety analysis only lO takes credit for four valves. It is therefore appropriate to permit one valve to be out-of-service for up to 14 days without materially reducing system reliability.

When one of the relief valves is inoperable, there is a 9 requirement that the HPCI subsystem, all active components of both core spray subsystems, and the LPCI

( mode of the RHR system be OPERABLE. The verification of l OPERABILITY, as used in this context, for HPCI means to l administratively check by examining logs or other information to determine if certain components / systems

,O l are out-of-service for maintenance or other reasons. It B 3.5/4.5-4 l0.

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QUAD CITIES UNITS 1 & 2 DPR-29 4 DPR-30 does not mean to perform the surveillance requirements

-needed- to demonstrate the OPERA'BILITY of the component / system.

G. Reactor Core Isolation Cooling (RCIC) System -

Plant y ,

Operating-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 be manually initiated at any time.

) The HPCI subsystem 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 HPCI subsystem- should the RCIC system be found to be inoperable.

The verification of OPERABILITY, as used in this context, for HPCI means to administratively check by examining. logs or other information to determine if certain components / system are out-of-service for maintenance or other reasons. It does not mean to

). perform the surveillance requirements needed to demonstrate the OPERABILITY of the components / system.

H.- Minimum Core and Containment Cooling Availability - Plant Shutdown f The purpose of Specification 3.5.H is to assure a minimum of core cooling equipment is available at all times. If, for example, one core spray subsystem was out-of-service and the diesel generator which . powered the opposite core spray' subsystem was out-of-service,'only two RHR 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. During REFUELING conditions, all low pressure core-and containment cooling systems may be inoperable in order to allow necessary maintenance and repair work to be performed. With all low-pressure core

} and containment cooling systems out of service, necessary B 3.5/4.5-5 Y

lC QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 g

U water inventory requiroments are maintained, to insure an adequate heat sink, by requiring the reactor vessel head to be removed and the reactor cavity to be flooded, the fuel storago pool gates to be removed, the fuel pool water lovel to be maintained above the low level alarm g' point and the reactor cavity water temperature to be bolow 140 degroos F. With these conditions mot, operations that involve a potential to drr.in the reactor vessel and those that involve coro alterations may be performed. Specification 3.9 must also be consulted to determine other requirements for the diesel generators.

g When the unit or shared diosol generator is inoperable, v there is a requirement that other emergency core cooling systems be OPERABLE. The verification of OPERABILITY, as used in this context, for those other systems means to administratively check by examining logs or other information to determino if cortain components / systems are out-of-service for maintenance or other reasons. It C- does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component / system.

Quad Cities Units 1 and 2 share certain process systems

_ such as the makeup domineralizers and the radwasto system O, 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 n only a single reference to the OPERABILITY and a surveillance requirements for the shared safety-related features of each plant. The level of OPERABILITY for une unit must be maintained independently of the status of the other. For example, a diesel generator (1/2 diesel generator) which is shared between Units 1 and 2 would n

have to be OPERABLE for continuing Unit 1 operation even V if Unit 2 were in a COLD S!!UTDOWN condition and needed no diesel power.

I. Suppression Chamber Requirements for Coro and Containment Cooling Systems O The OPERABILITY of the Suppression Chamber is required in OPERATIONAL MODES 1, 2, and 3 as part of the ECCS to ensure that a sufficient supply of water is available in the event of a LOCA. This limit on Suppression Chamber minimum water level ensures that sufficient water is available to permit recirculation cooling flow to the O core. The specified minimum level is defined at zero B 3.5/4.5-6 O l

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QUAD CITIES UNITS 1 & 2 DPR'-29 & DPR-30

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differential pressure between the Drywell and Suppression Chamber.

Repair work 'might require making the. Suppression Chamber-L inoperable. This specification will permit those repairs p '

to be made:and_at..the same time give assurance that the irradiated fuel has-an adequate cooling water supply when '

the~ Suppression Chamber must be- made inoperable, including draining, in OPERATIONAL MODES 4 or 5. ,

J&K. Maintenance of Filled Discharge Pipe

$. If. the discharge piping of the core spray subsystem, LPCI mode of the RHR system, HPCI subsystem, and RCIC system are not- fi'lled, 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 i 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 3.5.I provides assurance that an adequate supply;of coolant water.is immediately available to the low-pressure core cooling systems'and that the core will y- remain covered in the event of-a loss-of-coolant accident while the reactor is depressurized with the head removed.

L. Condensate Pump Room Flood Protection j See Bases 4.5 M. AVERAGE PLANAR LHGR

-This specification assures that the peak cladding temperature following the -postulated _ design-basis loss-of-coolant accident will not eFCeed the-22000F limit specified in 10 CFR 50, Appendix K considering- the postulated effects.of fuel pellet'densification.

The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat-generation rate of all the rods of a fuel b= assembly at any axial location and is only secondarily B 3.5/4.5-7

O QUAD CITIES UNITS 1 &2 DPR-29 & DPR-30 n" dependent on the rod-to-rod power distribution within a fuel assembly. Since expected local variations in power distribution within a fuel assembly af fect the calcut e,ted peak cladding temperature by less than i 20'F relative to the peak temperature for a typical fuel design, the limit n

on the AVERAGE PLANAR UlGR is sufficient to assure that calculated temperatures are below the 10 CFR 50 Appendix K limit. The maximum AVERAGE PLANAR L11GR specified in the CORE OPERATING LIMITS REPORT are based on calculations which employ the models described _in Reference 2. Power operation with UIGRs at or below those specified in the e, CORE OPERATING LIMITS REPORT 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 Appendix K only if they are more limiting than other design parameters, n#

The maximum AVERAGE PLANAR UlGRs specified in the CORE OPERATING LIMITS REPORT at higher exposures result in a peak cladding temperature of less than 2200'F. Ilowever, the maximum AVERAGE PLANAR LHGRs are specified in the CORE OPERATING LIMITS REPORT as limits because n conformance calculations have not been performed to V

justify operation at LHGRs in excess of those shown.

The AVERAGE PLANAR LHGR (APLHGR) also serves a secondary function which is to assure fuel rod mechanical integrity.

O N. Local LucR This specification assures that the maximum LINEAR HEAT GENERATION RATE in any rod is less than the design LINEAR HEAT GENERATION RATE specified in the CORE OPERATING LIMITS REPORT, even if fuel pellet densification is O~ postulated. The power spike penalty is discussed in Reference 2 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 LHGR due to power spiking. No penalty is required in Specification 3.5.N because it has been

,J accounted for in the reload transient analyses by increasing the calculated peak LHGR by 2.2%.

O. MINIMUM CRITICAL POWER RATIO (MCPR)

Thu steady state values for MCPR specified in the CORE O OPERATING LIMITS REPORT were selected to provide margin B 3.5/4.5-8 l

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 0- to accommodate transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical power correlation itself. These values also assure that the operation will be such that the initial condition assumed for the LOCA analysis plus two percent p for uncerta$nty is satisfied. For any of the special set d 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 p transient, assuming instrument trip settings given in d Specification 2.1. For analysis of the thermal consequences of these transients, the value of the MCPR stated in the CORE OPERATING LIMITS REPORT for the LIMITING CONDITION FOR OPERATION boonds the initial value of MCPR assumed to exict prior to the initiation of n the transients. This initial condition, which is used in

the transient analyues, will preclude violation of the fuel cladding integrity SAFETY LIMIT. Assumptions and methods used in calculating the required steady-state MCPR limit for each reload cycle are documented in References 2 and 4. The results apply with increased n conservatism while operating with MCPRs greater than U specified.

The most limiting transients with respect to MCPR are generally:

O

)

d "it*"*""1 *"" "'

b) Load rejection or turbine trip without bypass; or, c) Loss of feedwater heater.

n The MCPR Operating Limit reflects an increase of 0.03 V over the most limiting transient to allow continued operation with one feedwater heater out-of-service.

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

,U type. The current cycle's reload licensing analyses specifies the limiting transients for a given exposure increment for each fuel type. The values specified as the LIMITING CONDITION FOR OPERATION are conservatively chosen to bound the most restrictive over the entire a cycle for each fuel type.

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B 3.5/4.5-9 O )

O-QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 n.

~

The need to adjust the MCPR operating limit as a function of scram time arises from the statistical approach used in the implementation of the ODYN computer code for analyzing rapid pressurization events. Generic statistical analyses were performed for plant groupings n' of similar design which considered the statistical

'~

variation in several paramotors (initial power level, CRD scram insertion time, and model uncertainty). These analyses (which are described further in Reference 4) produced generic Statistical Adjustment Factors which have been applied to plant and cycle specific ODYN g' results to yield operating limits which provide a 95%

probability with 95% confidence that the limiting pressurization event will not cause MCPR to fall b. low the fuel cladding integrity SAFETY LIMIT.

For core flow rates less than rated, the steady-state n MCPR is increased by a factor of k where k' is as r) " specified in the CORE OPERATING LIMIhS REPORT. 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 coatroller causes the scoop tube positioner for the fluid coupler to move to this maximum speed position.

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B 3.5/4.5-10 I

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O QUAD CITIES UNITS 1&2 DPR-29 & DPR-30 O

References

1. " Quad Cities Nuclear Power Station Unita 1 & 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis",

NEDC-3134SP.*

O " Generic Reload Fuel Applications", NEDE-2 4 011-P-A**

2.

3. Deleted

4. " Qualification of the One-Dimensional Core Transient Q Model for Boiling Water Reactors", General Electric Co.

Licensing Topical Report NEDO 24154 Vols. I and II and NEDE-24154 Vol.III as supplemented by letter dated September 5, 1980 from R.H. Buchholz (GIL) to P.S. Check (NRC).

O'

Approved revision at the time of plant operation.

Approved revision number at time reload fuel analyses are performed.

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e B 3.5/4.5-11 e

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QUAD _ CITIES UNITS-11 & 2-DPR-29.& DPR-30 4.5 - SURVEILIANCE REQUIREMENTS BASES The testing interval for the core and containment cooling systems - is based on a quantitative reliability . analysis,

-)udgement , and practicality. The core cooling systems have O no,t.been designed to be fully testable during-operation. For example,- the corr spray final admission valves do not open until reactor pressure has decreased to 350 psig. Thus, during operation, even if high drywell pressure were-simulated,- the final valves would not open. In the case of the HPCI subsystem, automatic initiation -during power-O - operation would result in pumping cold water into the reactor vessel which is'not desirable.

The surveillance requirements bases described _ in this

-paragraph apply to all core and containment cooling systems except - HPCI and RCIC. The systems can be automatically O actuated during a REFUELING OUTAGE and this will be done. .To increase .the availability of the individual components of the core and containment cooling systems, the components which make up the - system, i.e.,' instrumentation, pumps, valve operators, etc., are tested more frequently. The

instrumentation is functionally tested in accordance with the provisions of Specification 4.2.B. Likewise the pumps _and O motor-operated-valves are tested quarterly in accordance with the Inservice Testing Program provisions to assure their OPERABILITY. The combination of a refueling . interval simulated automatic actuation test and quarterly tests of: the

_ pumps and valve operators is deemed to be adequate testing of these systems. In-addition, monthly checks are made on the O position of.each manual, power operated or automatic valve _

-installed _ in the - direct flowpath of the suction or discharge of each pump that -is not locked, sealed, or otherwise- secured in position. With components or subsystems out of' service,

- overall core and containment cooling reliability is maintained by- verifying the OPERABILITY of the remaining cooling O' equipment. The verification of OPERABILITY, as used in this context, for these other systems means to -administratively check by examining logs or other information to determine-if certain components / systems are out-of-service for maintenance or other reasons. It does not mean to perform the g' surveillance _ requirements needed to demonstrate the OPERABILITY of the component / system. However, if a-failure, design ~ deficiency, etc., causes the out-of-service pericd, then the verification of OPERABILITY should be thorough enough to assure that a similar problem does not exist on the remaining components. For example, if an out-of-service peri d is caused by failure of a pump to deliver rated O capacity due to a-design deficiency, the other pumps of this B 3.5/4.5-12 O

O-QUAD CITIES UNITS 1 & 2 DPR'-29 & DPR-30 C

typo might be subjected to a flow rate test in addition to the OPERABILITY checks.

The survoillance raquirements bases described in this paragraph apply only to the RCIC and llPCI systems. With a o cooling system out-of-norvico, overall core and containment cooling reliability is maintained by verifying the OPERABILITY of the remaining cooling systems. The verification of OPERABILITY , as used in this context, for the remaining cooling systems means to administratively check by examining logs or other information to verify that the remaining systems are not out-of-service for maintenaneci or other reasons. It O does not mean to perform the surveillance requinments nooded to demonstrato the OPERABILITY of the romai..ing systems.

Ilowever, if a failuro, design deficiency, etc., causes the out-of-service period, then the verification of OPERABILITY should be thorough enough to assure that a similar problem does not exist on the remaining components. For examplo, if 9 an out-of-service period is caused by failuro of a pump to deliver rated capacity duo to a design deficiency, the other pumps of this type might be subjected to a flow rate test in addition to the OPERABILITY chocks. Following a REFUELING OUTAGE or an outage in which work was performed that directly affects system operability, the llPCI and RCIC pumps are flow 9 rato testod prior to oxceeding 325 poig and again at rated reactor steam pressure. This combination of testing provides adequato assurance of pump performance throughout the range of reactor pressures at which it is required to operato. The low prossure limit is selected to allow tonting at a point of g stable plant operation and also to provido overlap with low pressure core cooling systems. A time limit is provided in which to perform the required tests during start up. This timo limit is considered adequato to allow stablo plant conditions to be achieved and the required tests to be performed. Flow rato testing of the llPCI and RCIC pumps is

, also conducted ovary 92 days at rated reactor pressure to demonstrate system OPERABILITY in accordance with the LCO provisions and to mes.t inservice testing requirements for the HPCI subsystem. Applicablo valves are tested in accordance with the provisions of the Inservice Testing program. In addition, monthly checks are made on the position of each g manual, power operated or automatic valvo installed in the direct flowpath of the suction or discharge of the pump or turbine that is not locked, Lealed, or otherwise secured in position. At each REFUELING OUTAGE, a LOGIC SYSTEM FUNCTIONAL TEST and a SIMULATED AUTOMATIC ACTUATION TEST is performed on the HPCI and RCIC systems. The tests and checks described

, above are considered adequate to assure system OPERABILITY.

B 3.5/4.5-13 9

QUAD CITIES UN3TS 1 & 2 DPR-29 & DIR-30 D

The verification of the m.ain ateam relief valve OPERABILITY during manu.11 actuation atr',eillance testing must be made independent of temperetus os indicated by thermocouples downstream of the relief valves. It has been found that a temperature increase may result with the valve still closed.

g This is due to steam being vented through the pilot valves during the surveillance test. By first opening a turoine bypass valve, and then observing its closure responso during relief valve actuation, positive verification can be made for the relief valve opening and passing ateam flow. Closure response of the turbine control valves during relief valve g manual actuation would '.ikewise serve as an adequate verification for the relief valve opening. This test method may be performed over a wide range of reactor pressures greator than 150 psig. Valve operation below 150 psig is limited by the spring tension exhibited be the relief valves.

p The surveillance requirements to ensure that the discharge piping of the core spray, LPCI mode of the RHR, HPCI, and RCIC systems is filled provides for a visual observation that water flows from a high point vent. This ensures that the line is in a full condition, y Instrumentation has boon prov:.ded on core spray and LPCI mode of RHR to monitor the pressure of water in the discharge piping between the monthly intervals at which the lines are i vented and alarm the control room if the pressure is inadequate. This instrumentation will be calibrated on the same frequency as the safety system instrumentation .and the g alarm system tested monthly. This testing ensures that, during the interval between the monthly venting checks, the status of the discharge piping is monitored on a continuous i basis. An alarm point of greater than or equal to 40 psig for the low pressure of the fill system has been chos^n because, due to elevations of piping within the plant, 39 psig is p

required to keep the lines full.

HPCI and RCIC systems normally take suction from the Contaminated Condensate Storage Tanks (CCSTs). The level in the CCSTs is maintained at or above 9.5 feet. This level corresponds to an elevation which is greater than the elevation of the last check valves in the discharge pipes of I either the HPCI or RCIC systems. Therefore, filled discharge piping of HPCI or RCIC systems is ensured when lined up to the CCST and tank level is at or above 9.5 feet.

The watertight bulkhead and submarine doors and the penetration seals for pipes and cables penetrating the vault I walls and ceilings have be'n designed to withstand the maximum B 3.5/4.5-14 E

O QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 Q 31ood conditions. To assure that their installation is l

adequate for maximum flood conditions, a method of testing each seal has boon devised. I l

In order to tout an electrical penetration or pipo seal, o"

compressed air is supplied to a test connection and the spaco betwoon the tittings is pressurized to approximately 15 psig.

The outer f acer are then tested for leaks using a soap bubble solution. In order to test the submarino doors, a test framo must be installed around each door. The fremo is then pumped i to a pressure of approximately 15 psig and hold to test for n leaktightness. The atortight bulkhead doors are tested by 1 pressuriring the volumo betwoon the double-gasket seals to approximately 15 psig. The gaskot non1 area is inspected using a soap bubble solution. Each RHR service water vault contains a sump, which will collect any floor or equipment leakago in insido the vault. A sump pump will automatically p' start on high level in the sump, and will pu';ap the vator out of the vault, v'2 two dischargo chcek valves outside tho vault to the servic' ater dischargo pipo. A composito sampler is located on the sump dischargo line. A radiation monitor is a~ located on the service water discharge. The sump

d. .arge water is not expected to be contaminated, and any in o. age to the vault is provented by two check valves.

O Sun tillarco of those check valvos is performed each OPERATING CYCLE to ausure their integrity. Tao previously installed bodplato drains to the turbine buildir.g equipment drain sump have bacn capped off permanently.

A level svitch set at a wt er lavel of 6 inches is located

.O insido nach vault. Upon actuation, the switch alarms in the control room to notify the operator of trouble in the vault.

The operator will also be aware of croblems in the vad ts/ condensate pump room if the high-level alarm on the eqt.ipment drain sump is not terminated in a reasonable amount of time, n

wJ A system of level switches has been installed in the condenser pjt to indicato and control flooding of the condenser area.

The following switches are installed:

Lovel Function

O.

a. 1 foot (one alarm, low water switch) level

b. 3 feet (ono alarm, high water switch) level B ').5/4.5-15 l

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i QUAD CITIES UNITS 1 & 2 DPR'-29 & DPR-30

c. 5 feet (two alarm and cir-redundant culating water ,

switch pairs) pump trip _

Level (a) indicates water in the condonner pit from either the O hotwell or the circulating water system. Lovel (b) is above the hotwell capacity and indicates a probable circulating water failure.

Should the switches at levels (a) and (b) fail or the operator fails to trip the circulating water pumps on alarms at level (b), the actuation of either level switch pair at level (c) 9 shall trip the circulating water pumps autom6 *;ically and alarm in the control room. These radiandant icvol switch pairs at level (c) are designed and installed to IEEE-W79, " Criteria for Nuclear Power Plant Protection Systems". As the circulating water pumps are trippud, either manually or aut mati ally at level (c) f 5 fe t, the maximum water lavel O reached in the condensor pit due to pumping will be at elevation 568 feet 6 inches (10 foot abovo condenser pit floor elevation 558 foot 6 inches; 5 foot plus an additional 5 feet attributed to pump coastdown).

g In order to prevent the RHR service water pump motors and diesel generator cooling water pump motors from overheating,

a vault cooler is supplied for each pump. Each vault cooler is designed to maintain the vault at a maximum of 105'F temperature during operation of its respective pump. For example, if diesel generator cooling water pump 1/2-3903

.O - sta.rts, its cooler alto starts' and maintains the vault at 1o5 F by removing heat r.upplied to the vault by the motor of pump 1/2-3903. If, at the same timo that pump 1/2-3903 is in operation, RHR service water pump 1C starts, its cooler will also start and compc:isate for the added heat supplied to the vault by tha 1C pump motor, keeping the vault at 105'F.

'O rach of the coolers is supplied with cooling water from its respecti/o pump's dischargo line. After the water has been passed through the cooler, it returns to its respective pump's suction line. The cooling water quantity needea for each coolor is approximately 1% to 5% of the design flow of the pumps so that the recirculution of this small amount of heated O water will not affect pump or cooler operation, operation of the fans and coolers is required during shutdown and thus additional surveillance is not required.

Verification that access doors to each vault are closed O following entrance by personnel is covered by station B 3.5/4.5-16 O.

l l

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 D operating procedures.

Average Planar LHGR At core thermal power levels loss than or equal to 25%,

p - operating plant experience and thermal hydraulic analyses indicate that the resulting AVERAGE PLANAR LHGR is below the !

l maximum AVERAGE PLANAR LHGR by a considerable margini l

! therefore, evaluation of the AVERAGE PLANAR LHGR below this ;

power level' is not necessary. The daily requirement for '

calculating AVERAGE PLANAR LHGR above 25% rated thermal power <

is sufficient, since power distribution shifts are slow when I p there have not been significant power or control rod chsnges.

L Local LINEAR-HEAT GENERATION RATE (LHGR) l The LHGR as a function of core height shall be checked daily juring reactor operation at greater than or equal to 25% power ,

to determine if fuel burnup or control rod movement has caused i changes in power distribution. A limiting LHGR valae is l precluded by considerable margin when employing any l permissible control rod pattern below 25% rated thermal power.

Since changes due to burnup are slow and only a few control :

g rods are moved daily, a daily check of power distribution is l i- - adequate.

! Minimum Critical Power Ratio (MCPR)'

At core thermal powers less than or equal to 25%, the reactor ,

will be operating at minimum recirculation pump speed and the 3 moderator void content will be very small. For all designated control rod patterns which may be employed at-this point, operating plant exp.rience and thermal hydraulic analysis L indicate that the resulting .MCPR value is in excess of l requirements by a considerable margin. With this low void .

j content, any inadvertent core flow increase would only place

operation in a more conservative mode relative to MCPR.

l The daily tequirement for calculating MCPR above 25% rated thermal power is sufficient, since power distribution shifts are very slow when there have not been-significant power or control rod changes. In addition, the K, correction, as specified-in the CORE OFERATING LIMITS REPORT, applied to the LCO provides margin for flow increases from low flows.

l-B 3.5/4.5-17 o

, ..- m., m-_ _ ;, ,_ ,i... _ ,,2 . ,,.---;..-_._f_,,_,..._.,_J.__.;,..- ._..-.n,,, _ _ , _

D EXISTING TECH SPEC TS 3,5/4.5

, ' CORE & CONTAINMENT COOLING SYSTEMS" O

4 3

3 3

e O

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

O QUAD-CITIES DPR 29 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS O

LlHITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS ,

Applicability: Applicability:

~ Applies to-the operational status of the Applies to periodic testing of the emer-

.O emergency cooling subsystems.

l gency cooling subsystems.

Objective: Objective: ,

To assure adequate cooling capability for To verify the operability of the core and '

heat removal in the event of a loss of- containment cooling subsystems O coolant accident or isolation from-the normal reactor heat sink.

SPECIFICATIONS O A. Core Spray Subsystems and the LPCI A. Core Spray Subsystems and the LPCI Mode of the RHR System Mode of the RHR System Surveillance of the core spray sub-systems and the LPCI mode of the RHR '

system shall be performed as follows:

O

1. Both core spray subsystems shall 1. Core Spray Subsystem Testing be operable whenever irradiated fuel is in the reactor vessel Item Frequency and prior to reactor startup from a cold condition, a. Simulated Each O automatic refueling actuation

b. Flow rate After pump test - core maintenance o- spray pumps and every shall 3 months LO deliver at least 4300 gpm against a system head corres-ponding to a

.O reactor vessel pressure of ,

90 psig

c. Pump opera- Once/ month bility lO d.. M t r- Once/ month operated valve 3.5/4.5-1 Amtndment No. 114 O

.. .. . . = -

=- - . - - -. - ; -_. . ._ - -_- -

O QUADCITIES DPR 29

e. Core spray O header Ap instrumentation check Once/ day calibrate Once/3 months test Once/3 0 " "th5

f. Logic system Once/Each functional refueling test outage

2. From and after the date that one 2. When it is determined that one O of the core spray subsystems is cota spray subsystem is inoper-

, made or found to be inoparable able, the operable core spray for any reason, continued reac- subsystem and the LPCI mode of tor operation is permissible the RHR system shall be only during the succeedin,g 7 demonstrated to be operable days unless such subsysten is immediately. The operable core O sooner made operable, provided spray subsystem shall be that during such 7 days all ac- demonstrated to be operable tive components of the other daily thereafter, core spray subsystem and the LPCI mode of the RHR system and the diesel generators required Q for operation of such compenents if no external source of power were available shall be operable. '

6

3. The LPCI mode of the RHR system 3. LPCI mode of the RHR system shall be operable whenever testing shall be as specified in g irradisted fuel is in the Specifications 4.5.A.I.a. b, c, reactor vessel and prior to d, and f, except that each LPCI reactor startup f roni a cold division (two RHR puinps per condition, division) shall deliver at least 9000 gpm against a system head

, corresponding to a reactor ves-sel pressure of 20 psig, with a O minimum flow valve open.

4. From and after the date that one 4. When it is determined that one of the RHR pumps is made or of the RHR pumps is inoperable, found to be inoperable for any the remaining active components reason, continued reactor opera- of the LPCI mode of the RHR, O tion is permissible only during containment cooling mode of the the succeeding 30 days unless RHR, and both core spray such pump is sooner made oper- subsystems shall be demonstrated able, provided that during such to be operable immediately and 30 days the remaining active the operable RHR pumps daily components of the LPCI mode of thereafter.

O the RHR, containment cooling 3.5/4.5-2 Amendment No. 114 0

O QUAD-CITIES i DPR-29 !

niode of the RHR, all active i O c mp nents f both core spray l subsystems, and the diesel generators required for operation of such components if no external source of power were available shall be operable.

O 5. iromandafterthedatethatthe 5. When it is determined that the LPCi mode of the RHR system is LPCI mode of the RHR systet.i is made or found to be inoperable inoperable, both core spray sub-for any reason continued reac- systems, the containment cooling toroperationIspermissible mode of the RHR shall be only during the succeeding 7 demonstrated to be operable O days unless it is sooner made immediately and daily thereafter.

operable, provided that during such 7 days all active compo-nents of both core spray sub-systems, the containment cooling mode of the RHR (including two '

O RHR pumps), and the diesel gen-erators required for operation of such components if no exter-nal source of power were avail-able shall be operable, g 6. If the requirements of Specifi-cation 3.5.A cannot be met, an orderly shutdown of the reactor shall be initiated, and the re-actor shall be in the cold shut-down condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

O B. Containment Cooling Mode of the RHR B. Containment Cooling Mode of the RHR System System Surveillance of the containment 4 .

cooling mode of the RHR system shall be pertormed as f0110ws:

1. a. Both loops of the 1. RHR service water subsystem containment cooling mode of testing:

the RHR system, as defined in the bases for Spe. Item Frequency cification 3.5.B. shall be O operable whenever irradiated a. Pump and valve Once/3 fuel is in the reactor operability months vessel and prior to reactor startup from a cold condition.

O 3.5/4.5-3 Amendment No. 114 lO l

I

O QUAD-CITIES DPR 29

1. b. From the effective date of b. Flow rate test - After pur this amendment until Novem- each RHR service maintenan:

0 ber 1, 1989, the "B" loop of water pump shall and every .

the containment cooling mode deliver at least 3 months of the RHR system for each 3500 gpm against reactormaysharetheUnit1 a pressure of 198 "C" and "D ' RHR servi:e psig I water pumps using" cross tie O . line 1/2-10509-16 D. c. A logic system Each Consequently, the require- functional test refueling-ments of Specifications outage 3.5 B.2 and 3.5.B.3 will impose the correspohding surveillance testing of equipment associated with O both reactors if the shared RHR tervice water pump or oumos, or the cross tie line, are made or found to be inoperable.

O 2. From and after the date that one 2. When it is determined that one of the RHR service water pumps RHR service water pump is inop-is made or found to be inoper- erable, the remaining components able for any reast.n, continued of that loop and the other con-reactor operation is permissible tainment cooiing loop of the RHR only during the succeeding 30 system shall be demonstrated to O days unless such pump is sooner be operable immediately and made operable, provided that daily thereafter.

during such 30 days all other active components of the con-tainment cooling mode of the RHR system are operable.

O

3. From and after the date that one 3. When one loop of the containment loop of the containment cooling cooling mode of the RHR system mode of the RHR system is made becomes inoperable, the operable or found to be inoperable for loop shall be demonstrated to be any reason, continued reactor operable immediately, and daily g operation is permissible only during the succeeding 7 days un-thereafter.

less such subsystem is sooner made operable, provided that all j active components of the other l -loop of the containment cooling mode of the RHR system, both O core spray subsystems, and both diesel generators required for operation of such components if no external source of power were I

available, shali be operable.

\

lO 1

l 3.5/4.5-4 Amendment No. 119 l

'O

'O QUAD-CITIES DPR-29

4. Containment cooling spray loops 4 During each 5 year period, an 0 are required to be operable when air test shall be performed on the reactor water temperature is the drywell spray headers and greater than 212'F and prior to nozzles and a water spray test reactor startup from a cold con- performed on the torus spray dition. Continued reactor oper- header and nozzles.

ation is permitted provided that g a maximum of one drywell spray loop may be inoperable for 30 days when the reactor water tem-perature is greater than 212'F.

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

C. HPCI Subsystem C. HPCI Subsystem O 1. The HPCI subsystem shall be Surveillance of HPCI subsystem shall operable whenever the reactor be performed as specified below with pressure is greater than 150 the following limitations. For item psig and fuel is in the 4.5.C.3, the plant is allowed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months reactor vessel, in which to successfully complete the test once reactor vessel pressure is O Y 2. During startup following a refuel adequate to perform each test. In outage or an outage in which work addition, the testing required by item was performed that directly affects 4.5.C.3.a shall be completed prior to HPCI system operability, if the exceeding 325 psig reactor vessel testing requirements of 4.5.C.3 pressure. If HPCI is made inoperable cannot be met, continued reactor to perform overspeed testing, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months O startup is not pemitted. The is allowed to complete the tests before HPCI subsystem shall be declared exceeding 325 psig, inoperable, and the provisions of Specification 3.5.C.4 shall be Item Frequency implemented.

1. Valve Position Every 31 days O 3. Except for the limitations of 3.5.C.2, if the HPCI subsystem 2. Flow Rate Test- Every 92 days is made or found to be inoperable, HPCI Pump shall continued reactor operation is deliver at least permissible only during the suc- 5000 gpm against ceeding 14 days unless such sub- I a system head cor-system is sooner made operabic, respunding to a O provided that during such 14 days I reactor vessel the automatic pressure relief pressure of > 1150 subsystems, the core spray sub- psig when steam is systems, LPCI mode of the RHR being supplied to system, and the RCIC system are the turbine at 920 operable. Otherwise, the pro- to 1005 prig.

O visions of Specification 3.5.C.4 shall be implemented.

3.5/4.5-5 Amendment No. 130 0

)

QUAD-CITIES OPR-29

3. Flow Rate Test- During startup p HPCI pump shall following a deliver at least refuel outage 5000 gpm against or an outage it a system head which work was corresponding to performed that a reactor vessel directly affect g

pressure of: HPCI system operability,

~

a. > 300 psig when steam is being supplied to

> the turbine at 250 to 325 psig, and

b. > 1150 psig when steam

> is being sup-plied to the turbine at 920 to 1005 psig.

4. If the requirements of Specifica- 4. Simulated Auto- Each refueling p tion 3. 5. C.1, 3. 5. C. 2 or 3. 5. C. 3 matic Actuation outage cannot be met, an orderly shutdown Test shall be initiated, and the reactor pressure shall be reduced to < 150 l 5. Logic System Each refueling psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Functional Test outage.

g D. Automatic Pressure Relief Subsystems D. Auto.zatic Pressure Relief Subsystems Surveillat.:e of the automatic pressure relief subsystem shall be performed as follows:

, 1. The automatic pressure relief 1. The following surveillanc.e shall subsystem shall be operable te carried out on a six-month whenever the reactor pressure is surveillance interval:

greater than 90 psig, irradiated fuel is in the reactor vessel a. With the reactor at pressure and prior to reactor startup each Velief valve shall be from a cold condition. manually opened. Relief I valve opening shall be verified by a compensating turbine bypass valve or control valve closure.

2. From and after the date that two 2. A logic system functional test sha'

> of the five relief valves of the be performed each refueling outage, automatic pressure relief subsystem are mede or found to be inoperable ,

3.5/4.5-6 Amendment No.130

O ,

QUAD CITIES OPR-29 when the reactor is pressurized O above 90 psig with irradiated 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 HPCI subsystem is operable.

3. If the requirements of Specifi- 3. A simulated automatic initiation cation 3.5.D cannot be met, an which opens all pilot valves orderly shutdown shall be initi- shall be performed each re-ated and the reactor pressurt fueling outage.

shall be reduced to 90 psig O within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 4. When it is determined that two valves of the automatic pressure relief subsystem are inoperable, the HPCI shall be demonstrated to be operable insnediately.

O E. Reactor Core Isolation Cooling System E. Reactor Core Isolation Cooling System

1. The RCIC system will be operable Surveillance of the RCIC system shall whenever the reactor pressure is be performed as specified below with greater than 150 psig and fuel is the following limitations. For item O in the reactor vessel. l 4.5 E.3, the plant is allowed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in which to successfully complete the

2. During startup following a refuel test once reactor vessel pressure is outage or an outage in which work adequate to perform each test. In was performed that directly affects addition, the testing required by item the RCIC system operability, if the 4.5 E.3.a shall be completed prior to O testing requirements of 4.5.E.3 exceeding 325 psig reactor vessel cannot be met, continur.d reactor pressure. If RCIC is made inoperable startup is not permitted. The to perform overspeed testing, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months RCIC system shall be declared is allowed to complete the tests before inoperable, and the provisions of exceeding 325 psig.

Specification 3.5.E.4 shall be O implemented. Item Frequency

3. Except for the limitations of 1. Valve Position Every 31 dayi 3.5.E.2, if the RCIC system is made or found to be inoperable, 2. Flow Rate Test - Every 92 days continued reactor operation is RCIC Pump shall o' permissible only during the suc- deliver at least ceeding 14 days unless such sys- I 400 gpa against tem h sooner made operable, a system head provided that during such 14 days l corresponding to the HPCI system is operable. a reactor vessel Otherwise, the provisions of pressure of > 1150 g Specification 3.5.E.4 shall be psig when steam is implemented.

3.5/4.5-7 Amendment No. 130 0

e QUAD-CITIES DPR-29 i

i being supplied to the turbine at 920 O to 1005 psig.

3. Flow Rate Test-RCIC During start pump shall deliver following a at least 400 gpm refuel outag against a system or an outage g head corresponding in which wor to a reactor vessel was performe pressure of: that direct) affects RCIC system perability.

O

a. > 300 psig when steam is being supplied to the turbine at 250 O to 325 psig, and

b. > 1150 psig when steam is being supplied to the turbine at 920 0 4. If the requirements of Specification to 1005 psig.

3.5.E.1, 3.5.E.2, or 3 E.E.3 cannot l be met, an orderly shutdown shall be 4. Simulated Automatic Each refuelii initiated and the reactor pressure shall Actuation Test outage be reduced to < 150 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5. Logic System Each refuelii F. Minimum Core and Containment Cooling Functional Test outage O

System Availability F. Minimum Core and Containment Cooling

1. Any combination of inoperable System Availability components in the core and containment cooling systems Surveillance requirements to assure shall not defeat the capability that minimum core and containment O of the remaining operable cooling systems are available have components to fulfill the core been specified in Specification 4.2.B.

and containment cooling functions.

2. When irradiated fuel is in the O reactor vessel and the reactor is in the cold shutdown condition, all low pressure core and contain-ment cooling systems may be in-operable provided no work is being done which has the potential for O draining the reactor vessel.

3.5/4.5-8 Amendment No.130 0

QUAD-CITIES DPR-29

3. When irradieted fuel is in the D reactor and the vessel head is removed, the suppression chamber may be drained completely and no more than one contesi rod drive housing opened at any one time provided that the spent fuel D pool gate is open and the fuel pool water icvel is maintained at a level of greater than 33 feet above the bottom of the pool. Additionally, a minimum condensate storage reserve of

, 230,000 gallons shall be maintained, no work shall be performed in the reactor vessel while a control rod drive housing is blanked following removal of the control 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 and the vessel nead is G removed, work that has the

potential fer draining the vessel may be carried on with 3

less than 112,200 ft of water in the suppression pool, provided that: (1) the total

volume of water in the suppression pool, refueling cavity, and the fuel storage pool above the bottom of the

. fuel pool gate is greater than 3

g 112,200 ft ; (2) the fuel storage pool gate is removed; (3) the low-pressure core and containment cooling systems are operable; and (4) the automatic mode of the drywell sump pumps

, is disabled.

P 3.5/4.5-9 Amendment No. 114 I

t

d lC }

4

'UAD- CITIES

'PR-29 G. Maintenance of Filled Discharge Pipe Maintenance of Filled Discharge Pipe O

The following surveillance require-ments shall be adhered to to assure that the discharge piping of the core spray, LPCI mode of the RHR, HPCI, and RCIC are filled: l O 1..

Whenever core spray, LPCI mode l

1. Every month the LPCI mode of the '

of the RHR, HPCI, or RCIC are RHR, core spray ECCS, HPCI and RCli required to be operable, the discharge piping shall be vented discharge piping from the pump from the high point and water flow discharge of these systems to observed.

O the last check valves shall be filled.

2. Following any period where HPCI, ,

. 2. The di6 charge pipe pressure for RCIC, LPCI mode of the RHR or

f. ore Spray and LPCI mode of RHR core spray have been out of

shall be maintained at greater service and drained for w than 40 psig and less than 90 maintenance, the discharge V psig. If pressure in any of piping of the inoperable system

! these systems is less than 40 shall be vented from the high

psig or greater than 90 psig, point prior to the return of tt.

j this condition shall be alarmed system to service.

in the control room and immediate corrective action

O taken. If the discharge pipe pressure is not witnin these -

limits in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the occurrence, an ords.rly shutdown shall be initiated, and the reactor shall be in a cold '

,0 shutdown conditicn within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after initiation.

3. Whenever the HPCI or RCIC system

'3. Filled discharge piping for HPCI is lined up to take suction from and RCIC systems is ensured by the torus, the-discharge piping.

maintaining the level in the of the HPCI and RCIC shall be Q- Contaminated Condensate Storage vented from the high point of Tanks (CCST's) at or above 9.5 the system and water flow ob-feet. If the CCST level falls served every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

below 9.5 feet, restore the level within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or line up both HPCI and RCIC to take a O suction from the torus per 4.5.G.3.

[ ,

i O

i

~

3.5/4.5-10 Amendment No. YW,130-

-O

-#o - , , ,,w-+-- - . --en - .-,..,,,m.,.-mr , r,.,ry.%wey....--y , -,.aw+, , . - - - + - - -

QUAD-CITIES DPR-29

4. The pressure switches which non-p itor the discharge lines and the discharge of the fill system pump to ensure that they are full shall be functionally tested every month and cali-brated every 3 months. The p . pressure switches shall be set to alarm at a decreasing pres-sure of > 40 psig and an in-creasing pressure of 5 90 psig.

H. Condensate Pump Room Flood Protection H. Condensate Pump Room Flood Protection

> 1. The systems installed to prevent 1. The following surveillance re-or mitigate the consequences of quirements shall be observed to flooding of the condensate pump assure that the condensate pump room shall be operable prior to room flood protection is oper-startup of the reactor. able.

> a. The piping and electrical penetrations, bulkhead doors, and submarine doors for the vaults containing the RHR service water pumps and diesel generator cooling

> pumps shall be checked during each operating cycle by pressurir.ing to 15 + 2 psig and checking for leaks using a soap bubble solution. The criteria for y acceptance shall be no visible leakage through the soap bubble solution.

b. During each operating cycle, the following flood

, protection level switches shall be functionally tested to give the following control room alarms:

1) turbine building equipment drain sump i high level.

2) vault high level 3.5/4.5-11 Amendment No. 114 1

)

QUAD-CITIES DPR-29

c. The RHR service water vault

) ,

sump pump discharge check valves outside the vault shall be tested for integrity using clean deminerallzedwater,at least once per operating

, cycle.

d. The condenser pit 5-foot trip circuits for each channel shall be checked once a month. A logic sys-D tem functional test shall be performed during each refueling outage.

2. The condenser pit water level switches shall trip the conden-ser circulating water pumps and p alaru in the control room if wa-ter level in the condenser pit exceeds a level of 5 feet above

'the pit floor. If a failure oc-curs in one of these trip and alarm circuits, the failed cir-g cuit stall be immediately placed in a trip condition and reactor operation shall be permissible for the following 7 days unless the circuit is sooner made oper-able.

I

3. If Specification 3.5.11.1 and 2 cannot be met, reactor startup shall not connence or if operat-ing an orderly shutdown shall be

.- initiated and the reac*.or shall be in a cold shutdown con-D dition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.5/4.5-12 Amendment No. 114

O QUAD-CITIES DPR-29 I. Average Planar LHGR I. Average Planar LHGR O

During steady-state power operation, Daily during steady-state operation the average linear heat generation above 25% rated thermal power, the rate (APLHGR) of all the rods in any average planar LHGR shall be deter-fuel assembly, as a function of aver- mined.

age planar exposure, at any axial lo-O cation, shall not exceed the maximum average planar LHGR specified in the ,

-CORE OPERATING LIMITS PEPORT. If at I any time during operation it is deter-mined by normal surveillance that the limiting value for APLHGR is being p" exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR 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 D cold shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Surveillance and corresponding action shell continue until reactor operation is within the prescribed limits.

J. Local LHGR J. Local LHGR O During steady-state power operation, Daily during steady-state power the linear heat generation rate operation above 25% of rated thermal (LHGR) of any rod in any fuel assem- power, the local LHGR shall be bly at any axial location shall not determined, exceed the maximum allowable LHGR specified in the CORE OPERATING LIMITS O REPORT. If at any time during opera- l tion it is deter.iined by normal sur-veillance that the limiting value for LHGR is being exceeded, actien shall be initiated within 15 minutes to restore operation to within the o prescribed limits. If the LHGR is r.ot 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 shill be brought te the cold shutdown condition within ? 50urs.

Surveillance and corresponding action shall continue until reactor O operation is within the prescribed limits.

O 3.5/4.5-13 Amendment No. 120 O.

QUAD-CITIES OPR-29 9 K. Minimum Critical Power Ratio (MCPR) K. Minimum Critical Power Ratio (MCPR)

During steady-state operation at The MCPR shall be determined daily rated core flow, MCPR shall be during steady state power operation equal to or greater than the MCPR above 25% of rated thermal power.

limit specified in the CORE OPERATING e LIMITS REPORT.

For core flows other then rated, these nominal values of MCPR shall be increasedbvafactorofk[wherekOPERATibG is as specified in the COR

, LIMITS REPORT. If any time during l operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore opera', ion to within the

, prescribed limits, if the steady-state MCFR is not returned to within the prest.ribed 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 shttdown cundition 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 operatio') is within the prescribed limits.

O O

S S

3.5/4.5-14 Amendment ho. 120

. 1

O QUAD-CITIES DPR-29 1

3.5 LIMITING CONDITIONS FOR OPERATION BASES i O , Neo A. a S )(, Core Sprayjand LPCI Mode of the RHR System ~ Planf Opemting This specificat on assures that adequate emergency cooling capability is available whenever irr diatwHuel is in the reacter vessel. during piant e,pmeoes in Modes I, L and a.

QM miMude '

ased on the loss-of-coolant :n:1ytt:+1-methed: de: r4 bed in C:-m1

'lectric Tcpical hpert CO-HM6P core cooling systems provide in Re erences Imf J. enc, in m.ordance sufficient cooling to the e e to dis :utte the energy _ associated wit cith fo uR.so.% the loss-of-coolant acciden limi Falculated f4MW cladding k.

ord 4pndh 5., temperature to less than 22 to assure that core geometry mains intac O to limit, cladding meta -water reaction to less than OGhe corewy limit The calculated local metal-water reat. tion to less than 1 ango 8' itte?A-6 3.5 specify the combinations of ///fA5J/ subsystems to availability of the minimum cooling systems noted above. -Under the3+

He+t4ng conditien: ef Oper:t4a r-4*cres:ed-serve!'!ance-t+st4*g-et-the.

O +cmaining ECCS-*y*4**s-tr:vides essurance that edeavete cooling of the ccre will be provided duriftg a%5rviwvviani au. i de ni.

Core spray distribution has been shpwn, in full-scale tests of systems similar in desige to that of QuacECities I and 2, to exceed the minimum requirements by at least 25%. In addition, cooling effectiveness has O been demonstrated at les.5 than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel. Theaccidentanalysisisadditionallyconservativg in that no credit is taken for spray cooling of the reactor core before the internal pressure has fallen to 90 psig.

O The LPCI mode of the RHR system is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant a:cideat. This system functions in combination with the core spray xdJystem to prevent excessive fuel cladding temperature. The LPCI mode oftheRHRsystenmincombinationwiththecorespraysubsystegprovides adequate cooling Jor break areas of approximately 0.05 ft up .o and

Q- -

including 4.26 f t , the latter being the double-ended recirculation line bre6k with the equalizer line between the recirculation loops closed ithout assistance from the high pressure emergency core cooling subsystems. .

The all weble repa4+-t4me; re 0;t:b11:hed te that -the average risk l0 rate-fee-r+pa4r e eld be ne greater than the bas 4<-- ri: A rate. The method-end-eeneept Ort de:Cribed in lleferen c 3. 'Jeing- the re50%s developed in thi: reference, the repair p&fcd i:- found te be less-then O

l

! 3.5/4.5-15 Amendment No. 114 0

l _ . - ___ _ - , - -- . - - - - -

O!

QUAD-CITIES ,

DPR-29

. i O-half the teet int;rv:1. Thh :::=:; th:t the core ; pray subsystems erd tPc: c ae44:et: : car aut ef t.o syst ., a m ,,r, in, symd;we effect ef-the te cyst::: to Ibit excessive cladding i..mperature must ales be ceneidered. Th. test intuivei apecified in 5pecification 4.5

^^^ ' - ^ he. Iherefeie, GG-EIIeWebli iersii ywiiOd which mainisius the D&iis siik Cen;idising Linpli fiileiei ihewld bc Iraa ihan 36.de'ys, O and thh :;::!fS: tion b with.n thh p;ried. fer rsM4 pit-feilwree, --a ,

th:-t:- t-t: :1 h :p=tried; t 1 p ::: te :::e :n:: th:t te ;

remaining systee; will functkn, e deity test ie celled for. Mthewgh ,

it is recegnized th;t the inf; ratha giv:n_ h Ref:r:ra; i provid ; :

qwantitet4v; ::th:d t: 2:th:te albu;;.le 7;p:fr ti;;.;;, the 1;ck ei operet4ng d:t: t: ;;ppert the enelyticei e yieech v pie..nte compitte Therefere, th+ tie .. 16ted- in

.c;;ptance of thit ;,;thed at this the.

O th: :pecif+c ite;; -e. 5t.LIi ;,.d .Iin du. regard- to judgunt.

If Shev44 one core spray subsystem becom inoperable, the remaining core spray subsystem and the entire LPCI mo e of the RHR system are available should the need for core cooling arise. T :::gr: th:t the rema4&g core Spray and the LDCleede f the ""D syste- are ava!!ab!e, O- (n:y 27,.g;;;;;t7:::e t 3 per:b; ,_4 ,44:::1y, Twg e:: net,.44,n.

Mc!rde: :e-"-! M4tiatien ef the pu p: :nd-:::=!:ted v:h::.- Based on judgmentKYtbe reliability of the remaining systems, i.e., the core ipray and LPC1, a 7-day repair period was obtained.

Should the loss of one RHR pump occur, a nearly full complement of core O. '

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 ore cooling equipment, which will h d==:trated te be eperab! a 30-day repair period is. justified. If the LPCI mode of the RHR system is not available, at least two RHR purcps must be available to fulfill the O Insed Mache) containment cooling function. The 7-day repair period is set on this asis.-

t, f. RHR Stryit: N:ter toniainmeni toch'ng Mode. of 4he KHR.Splem - Plani

. trernhn

. .The conta nment cooling mode of the RHR system is provided to remove '

.O P 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).

OL O

3.5/4.5-16 Amendment No. 114

O

~

10 l

O INSERT TO TECHNICAL SPECIFICATION PAGE 3.5/4.5-16 When a core spray or LPCI subsystem (or pump) is inoperable, thore le a requirement that other emergency core cooling systems bo OPERABLE. The verification of OPERADILITY, as used in this context for these other systems means to administrative 1y check O by examlning logs or other information to determine if certain components / systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component / system.

O O

O-QUAD-CITlES OPR-29 A. w, service Wakv p mps.

The Containment Cooltr/g mode of the RHR System consists of two loops.

Each loop consists offl Heat Exchanger, 2 RHR .%mps, and the associated O electrical equipment, and instrumentation. The uAu valves, piping,it-containtr4-RHR4ervice-Watee-Pumpsr-4nt44-evembee-1, 1eep-en-techwn 1989;-the "B" leep-ett-each-unit-eay-ut4Hae-the "C" :nd "0'LRHR4ee-v4ee-Wa te e-Pomps-f eom4ni t-1-v i a-aa ross 4i e-14 ner-- A f te r4o v embe >1 3 19897each "B"-loop-wM4-contain-2-RHR4ervice-WaterJumps. Either set of equipment is capable of performing the containment cooling O,

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 requirements. Since there is some redundancy lef t, a 30-day repair period is adequate.

Loss of one loop of the containment cooling mode of the RHR system leaves one remaining system to perform the containment cooling func-O tion. The-o pe rabl e-sy s tem-4 +-demons tra ted-to-be-opeceble-ea c h-d ay ,

when-the-above-condition-eecues. Based on the fact that when one loop of the containment cooling mode of the RHR system becomes inoperable, only one system remains, which-45-tested-daHy, a 7-day repair period was specified.(4dd msed)

p. (.onfainrnent tochhg Spray Loors ( Add affar.hed) ,

O E f. HigtQtessure Coolant Injection Subsys/em - Plad OremNng The higt M ssure coolant injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI mode of the RHR system or core spray subsystems can protect the core.

I sospkm Subsystem O The HPCI meets this requirement without the use of offsite electrical power. For the pipe bjr aks for which the HPCI is intended to function, thecore7neveruncoverTandiscontinuouslycooled,thusnocladding occurs (referenc jARSection6.2.5.3). The repair times for damayIMU/dl/M4fph the / / #pf#t we.'e set coM1dering the use of the HPCI as part of the isoi tion cooling system. (Md mscrf)

F g. Automatic Pressure Relief su.bsysfem - Plani OfemHg

-,abspfem The relief valves of the automatic pressure relief subsystem are a backup to the HPCI subsystem, hey enable the core spray subsystem and n LPCI mooe of the RHR system to rovide protection against the small v pipe break in the event of HPCI failure by depressurizing the reactor vessel rapidly enough to actuate the core spray subsystem and LPCI mode we of the RHR system. The core spray subsystem and'the LPCI mode of the RHR system provide sufficient flow of coolant to limit fuel c h $ ting temperatures to less than 2200'F, to assure that core geometry remains intact, to limit the core wide clad metal-water reaction to less than 0 1%, and to limit the calculated local metal-water reaction to less than 17%.

l l

lO l 3.5/4.5-17 Amendment No.119 l

l l

l C) ;

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O INSERT TO TECHNICAL SPECIFICATION PAGE 3.5/4.5-17 c SECTION C When a containment cooling subsystem (or pump) is inoperable,

.A3 there is a requirement that other emergency core cooling systems be OPERABLE. The verification of OPEMBILITY as used in this context fortheseothersystemsmeanstoadaknistrativelycheck by examining logs or other information to determine if certain components / systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the

() - component /nystem.

SECTION D Containment Cooling Spray Loops O The containment cooling spray loops consist of two drywell spray loops, each being 100% capacity, and one suppression chamber spray loop. The containment cooling spray loops are an integral part of the containment cooling subsystem of the RHR system. The spray headers of the RNR system are not intended to M placed in service until the core cooling requirements of the low pressure coolant C) injection system have been satisfied. When the reactor water level is restored to two-thirds of the core height, level switches give a permissive which permits opening of the containment spray valves. These requirements may be bypassed by the operator using a keylock switch in the control room. Because of the redundancy in the drywell spray loops a 30 day repair period is justified

{3 with one spray loop inopera,ble.

SECTION E When the HPCI subsystem is inoperable, there is a requirement that other emergency core cooling systems be in OPERABLE. The

,0- verification of OPERABILITY as used this context, for these other systems means to administrative 1y check by examining logs or other information to determine if certain components /nystems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component / system.

O i

!O l

c

D QUAD-CITIES DPR-29 Analyses-ha ve-shown-t ha t-only-f our-of-the-f 4 ve-va l ves-in-the-automa t i c dep re s s u r4 r a tion-sy s t em-a re-requ i red-to-ope ra ter---Los s-o f-ene-o f-the 9 relief-valves-does-not-signi'4cantly-af feet-the-pressure-relieving c ap a b H i tyr- t he re f o re-c ont i nued-ope ra t4 o n-i s-acc ep t a bl e- p ro v i ded4he-app ro pria te-MPLHGR-red uc ti on-f a c to n-45-e ppl i ed-to-a s s u re -compl i a nc e Add wi th4he-2200% PCT--14 m i t,--Lo s s-o f-mo re-tha n-one-reli e f-v alve.

nHacheci ei gni f i cent 4y-reduces-the-press ure-rel4 ef-ce >ab H 4 ty-o f-the-ADSt---thus ,

l a-7-day-repair-period-is-specified-with-the-HP01 available;-end-e-24 4 (-__,, hou r-repa t e-period-wi th- t he-HPC-I-un a v a i l ab l e .

twc) 4f. RGIC- Reader tcre isolaticn Cooling ysteny- Plan t Of>croh.mj lhe RCIC system is provided to supply continuous makeup water to the reactor core when the reactor is isolated from the turbine and when the 9 . 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 be manually initiated at any time.

O The HPCINstem provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable.

AM Therefore, the specification calls fo 'n 8/##f7Hf check of the HPCI aHache) TsystemshouldtheRCICsystembefound- io inoperable.

Minimum Core andlonluinment 9 9 f. emergency 3 Cooling Availability - Plane snufdown genemier ThepurposeofSpecification3.5.fistoassureaminimumofcore cool _ing equipment is available at all Limis. If, for example, one core f~ spray *were ou of service and the diesel which /

powered the opposite M Tcore sprayJls}re-outcofcservice, only two RHR pumps would be available.

9 Likewise, if two RHR pumps were out6ofG,ervice and two RHR service .

water pumps on the opposite side were also outeofeservice no containment cooling would be available. It-is-ducing-refuel 4ng-outages-that-majoe-maintenance-is-performed-and ducing wh Hae_that_All Replue low pressure-core-cooling-systems-may-be-out-of-service. This gj , sp?cification-provides-that-should-this-occurr-no-wortwi3Lbe 9 "gg performed on-the primary-system-which-could-lead-to-draining-the v e s s elv-Thi s-wo rk-would-i nclude-wo r4-on-c e rta4 n-cont rol-rod-d r iv e losed I compone nt s-and-reci rcul a t ion-sys t em. Thusrthe-specification-pr eclude s-t he-e ven t s-wh ic h-coul d-requi re-core-cooling. Specification 3.9 must also be consulted to determine other requirements for the diesel generators. Md aHachec) insert L O -

Quadi> Cities Units 1 and 2 share certain process systems such as the makeup demineralizers and the radwaste system and also some safety system nsuch as the standby gas treatment system, batteries, and diesel genera &rs. All of these systems have been sized to perform their intended function considering the simultaneous operation of both units.

9 3.5/4.5-18 Amendment No. 114 9

l D

1

() ,

INSERT FOR TECHNICAL SPECIFICATION PAGE 3.5/4.5-18 l t

SECTION F

() Automatic pressure relief automatically controls five safetv/r. lief valves although the safety analysis only takes credil for four valves. It is therefore appropriate to permit valva to be out-of-service for up to 14 days without materially one reducing system reliability.

there is a

() When one ofthat requirement the the relief valves HPCI is inoperable,ive subsystem, all act components of both core spray subsystems, and the LPCI mode of the RHR system be OPERABLE. The verification of OPERABILITY, as used in this context, for HPCI means to administrative 1y check by examining logs or other information to determine if certain components / systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance b) requirements needed to demonstrate the OPERABILITY of the component / system.

SECTION G The verification of OPERABILITY, as used in this context, for

() HPCI means to administratively check by examining logs or other information to determine it certain components / system are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the components / system.

C) SECTION H Insert 1 During REFUELING conditions, all low pressure core and containment cooling systems may be inoperable in order to allow necessary mafrtenance i.nd repair work to be perforned. With all low-pr- ure core and containment cooling systems out of service, g neu ssary water inventory requirements are maintained to insure an adequate heat sink,'by requiring the reactor vessel head to be removed and the reactor cavity to be flooded, the fuel storage pool gates to be removed, the fuel pool water level to be maintained above the low level alarm point and the reactor cavity water temperature to be below 140 degrees F. With these conditions met, operations that involve a potential to drain the

() reactor vessel and those that involve core alterations may be performed.

SECTION H Insert 2 When the unit or shared diesel generator is inoperable, there is a requirement that other emergency core cooling systems be OPERABLE.

as used in this context, for O The theseverification of OPERABILITY,inistrative1y other systems means to adm check by examining logs or other information to determine ,1f certain O

I i

O components / systems are out-of-service for maintenance or other O reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the cor.ponent/ system.

O O

O

~O

-O O

O O

O

O QUAD-CITIES DPR-29 These technical specifications contain only a single reference to the

-O ////$$///// and survaillance requirements fo the shared safety-related features of each plant. Thelevelof//g f%y'foroneunitmustbe maintained indeDendently of the status of the other. For example, a C F * *J diese diesef) which is shared between Units I and 2 would have to be for continuing Unit 1 operation even if Unit 2 were in a

// condition and needed no diesel power.

O 3 pre ation 3.5.fW provides that should this occur, no work MF performe icit.could cool a 111ty being available. % preclude . prohibited*

adequate unlessemer ene acc.,rdance with '

specified procedures whic

he ut the control rod drive housing is open and , assures t es nossible loss of coolant

'O resulting from the work 4M tresulti(nuncoveingthereactor core. Thus, th c fication assures adequate (core W . i Specifip .9 must be consulted to determine other requ org y -61esei nonerntne- - - - - ~

LSfte r&n thahber Requitenients for Core and Contamenent ccolor &dem5(MlUN}tC{

f

)L Maintenance of Filled Discharge Pipe O x4x *b51 * * * 'N*

arge piping of the core spray L.PCI mode of the RHR(, HPClf"Y

Ifthedfebwns 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 whit.h shows thatmif a water hammer were to occur at the O time emergency cooling was redifired, the systems would still perform their design function. However9to minimize damage to the discharge systems and to ensure added mar (gin in the operation of these systems, ibis technical specification re 2 whenever the system is in anf/M// // quires condition. the discharge linos to be filled O specification 3.5.-f-4 provides assuranc.e that an adequate supply of coolant water is immeritately 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.

O .,. . L. )(. Condensate Pump Room Flood Protection See Spec!'icatir 3,5,H sases 4.5 M /. A/#/f/ P///// LHGR

'O This specification assures that the peak clado ng temperature following the postylated design-basis loss-of-coolant accident will not exceed the 2200 F limit specified in %e 10 CFR 50, Appendix K considering the postulated effects of fuel pellet densification.

l

'O 3.5/4.5-19 Amendment No. 114 lO l

-- -. s.,,,em---. , - - , , ---.,,.<m,,,,.,,,,w.,,w,..s-- -

t_

9 h) INSERT TO TECHNICAL SPECIFICATION PAGE 3.5/4.5-19 SECTION I

1. Suppression Chamber Requirements for Core and Containment Cooling Systems 0 The OPERABILITY of the Suppression Chamber in required in OPERATIONAL HODES 1, 2, and 3 as part of the ECCS to ensure ,

that a sufficient supply of water is availabic in the event of a LOCA. This limit on Suppression Chamber minimum water level ensures that sufficient water is available to permit recirculation cooling flow to the core. The specified minimum e level is defined at zero differential pressure between the Drywell and Suppression Chamber.

1 Repair work might r6 quire making the Suppression Chamber l inoperable. This specific tion wi'i permit those repairs to l i

be made and at the same time give assurance that the !

D irradiated fuel has an adequate cooling water supply when the l suppression Chamber must be made inoperable, including i draining, in OPERATIONAL HODES 4 or 5. '

J

O QUAD-CITIES OPR-29 The' peak cladding temperature following a postulated loss-of-coolant 3

accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only secondarily depenoent on the rod-to-rod power distribution within axCAlel assembly. Since expected local variations in power distribution within a fuel assecly affect the calculated peak cladding temperature by less than 120'F rejetM to the peak temperaturt for a typical fuel design, O the limit on thc /////# #fMf/ LHGR is sufficient to assure that calculated tempenAures are aelow the411mit. Themaximum/////#

"$g ;g (

'o///// LHGTF s specified in the CORE OPERATING LIMITS REPORT are ha n calculations,Mchemploy w k

%e-AW p PT49f Lftw,' U#4 Ogimmhp 47.g ( APtttGR) aisu >vrved a o srecendary f acoon Acn is to assure fudmtxt-mectrmiral integri ty.

/.N.LocalLHGR This specification assures that the maxinium /(( /JI /

inanyrodislessthanthedesign///////(iAs , #ff

[/////d specified 3 in the CORE OPERATING LIMITS REPORh)even if fuel pellet densi ication is postulated. The power spike penalty is discussed in Reference 2 and assumes a linearly increasing variation in axial gaps betweea core bottom and top and assures with 95% confidence that no more than one fuel rod exceeds the design LH due to pcwer spikirg. No penalty is required in Specification 3.5. ecause it has been accounted for in

$ the reload transient analyses by increasing the calculated peak LHGR by 2.2%.

f.o. M/p'////i C/////M P//// R//// (MCPR)

The steady state values for NCPR specified in tha CORT 9PERATING LIMITS O REPORT were selected to provide margia tt ,cconcodate transients and uncertainties in monitoring the core cpe . ting state as well as uncer-tainties in the critical power correlation itself. These salues also assure tnat operation will be such that the initial condition assumed for the LOCA analysis plus two perc(nt far uncertainty is satisfied.

For any of the special set of transients oc disturbances aused by single O operator error or single equipment malfunction, ;t 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 value of MCM stated in the CORE OPERATING LIMITS REPORT l 0 for the 7///ffff ///////h'/ T////Mf//

o bounds the initial value of MCPR assumed to exist prior to the initiation of the transients. This initial condition, which is used in the trhnsient anal will precludeviolationofthefuelcladdingintegrity/

Assumptions and methods used in calculating the requir d steady state

///

es,$(f.

MCPR limit for each '.e'ioad cycle are documented in References 2 and 4.

O The results apply with increased conservatism while operating with MCPRs greater than specified.

3.5/4.5-20 Amendment No. 120 l

'O

ge INSERT FOR TEC"NICAL SPECIFICATION PAGE 3.5/4.5-20 1

0; SecTION M Power operation with LHGRs at or below those specified in the CORE OPERATING LIMITS REPORT assures that the-peak cladding temperature-following--a postulated loss-of-coolant accident

-will not exceed :the 2200'F. limit. These values represent ,

^

limits fr peration to ensure conformance with 10 CFR 50 ,

O. Appendix K only if they are more limiting than other' design ;

parameters.

The- maximum AVERAGE PLANAR LHGRs specifled in the CORE OPERATING LIMITS REPORT at higher exposures result in a peak. '

-cladding temperature of- less than 2 2 00'F. However, the 0- maximum AVERAGE PLANAR LHGRs are specified in the CORE OPERhTING LIMITS REPORT as- -limits ')ecause conformance ,

calculations have not been performed to justify. operation at LHGRs in e:tcess :of those shown.

O .

O

'O :

01 LO -

O 1

1 0

i QUAD-CITIES DPR-29 The most limiting transients with respect to MCPR are generally:

S a) Rod withdrawal erro b) Load rejection or turbine trip without bypas c) LossoffeedwaterheateQ The MCPR Operating Limit reflects an increase of 0.03 over the most limiting transient to allow continued operation with one feedwater heater outeof3 ervice.

nce which of these transients results in the 9 Severalfactorsinflug///

largest reduction in 4 ff ///// f$/ such as the specific fuel loading, exposure, and fue type. The current cycle's reload licensing analysts specifies the limiting transients for a given exposure incrementJgreachfueltype. Thevaluesspecifiedasthel/////[g C////ff wOf(fif/// bre conservatively chosen to bound the most restric ve over the entire cycle for each fuel type.

9 ihe need to adjust the MCPR operating limit as a function of scram time arises from the statistical approach used in the implementation of the ODYN computer code for analyzinc rapid pressurization events. Generic statistical analyses were performed for plant groupings of similar design which considered the statistical variation in several parameters 4 (initial power level, CRD scram insertion time, and model uncertainty). These analyses (which are described further in Reference

4) produced generic Statistical Adjustment Factors which have been applied to plant and cycle specific ODYN results to yield operating limits which provide a 95% probability with 95% confidence that the limiting pressurization event will not cause MCPR to fall below the 9 fuel cladding integrity ////f/ /////.

For core flow rates less than rated, the stead Astate MCPR is increased by the formula given in the specification. This ensures that the MCPR willbemaintainedgreaterthanthatspecifiedinSpecification1.1.%

even in the event that the motor generator set speed controller causer 9 - the scoop tube positioner for the fluid coupler to move to the maximur-speed position.

O O

3.5/4.5-2' Amendment No. 114 l

O QUAD-CITIES DPR-29 References

... O

1. 7"5AFER/6ESTR LOCA Loss of-Coohnt-Analysts-f+r-QuadC4LiesJucIsar Pner mt4en-Units 1 12" NEDC-31345P.*

2. "GenericReloadFuelApplicatiohNEDE-24011-P-A**

O 3. I. ". 'acob;; and i'. W. "arriett, GE Tupicai RepurdPEW36, "Guiuvitimr fee-Determining Safe Test Intervals and-Repair-%es for Engineered Safeguaeds," April,1050r--

4. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors" General Electric Co. Licensing Topical Report NEDO 24154 Vols. I O and 11 an NEDE-24154 Vol. III as supplemented by letter dated September 5, -

1980 from R.H. Buchholz (GE) to P. S. Check (NRC),

t

Approved revision at time of plant operation.

- Approved revision number at time reload fuel analyses are performed.

9 cuad ti+Its NarJear ftwer Station Units t 4 L sArenlqcsrn- LCeA Loss of- Coolanf Accideni Analpis *,

6 g

e 9

3.5/4.5-22 Amendment No. 114 6

C:

QUAD-CITIES DPR-29 4.5 SURVEILLANCE REQUIREMENTS BASES 0, 2

$ The testing interval for the core and containment cooling systeas is based on a 3 quantitative reliability analysis, judgment, and practicality. The core cooling

$ systems have not been designed to be fully testable during operation. For n

v f'r(

u E example,ge_gresprayfinaladmissionvalvesdonotopenuntilreactorpressure has fe" .~ fb"350 psig. Thus, during operation, even if high drywell pressure VE were simulated, the final valves would not open. In the case of the HPCIC *

M y[ automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable.

.T C

(Ln Atterdente with the prw$ looms of SPs*LifNofion 'b A 0 The surveillance requirements bases descrioca in this paragraph apply to all core g

n and containment cooling syst s except HPCI and RCIC. The systems can be V automatically actuated durin a////(///g$$/#(andthiswillbedone.

F[O

g. increase the availability of the individual components of the core and containment coolin To instrumentation, tg *j pumps,g systems, valve the compon operators, et , arents which tested make more up the system, frequently. i.e.,

The instrumentation is c- I functionally tested ca:h n .th. Likewiseithe pumps and motor-operated valves e 0- -alse testecTN:h = nth to assure their #/#/$f/ff/. The combination of a year:y g*ny simulated automatic actuation test and = ntAly tests of the oumos and valve operators is deemed to be adequate testing of these systems.4 With components oD""N subsystems out of service, overall core and containment cooling reliabi nty is (verihn{[The maintained by 6:en:tr:teting d q;r:: ef Operabi'ity the ///f($f7/f/depend be &renstrated of theerremaining the naturecooling equipme of the reasen hjg ifer-the-out-of-:er*4ce eq"!? rent. For reutine cut-cf-cervi:.: period: ::e::d by ,.

U, "%gce th preventative maintenance, etc. , the ? p and valv Operab+11ty checks-will bc

<@f d Lpeefemed te &renstrate Operabi'ity Of the ree:ining ^ penent:. However, if a

'A failure, design deficiency, etc. , causes the out-of-snvice period, then the ng ,-denr.stration of ($g/Mfffff should be thorough enough to assure that a similar problem does not exist on the remaining components. For example, if an out-of-a

+

n Oervice period is caused by failure of a pump to deliver rated capacity due to a

's design deficiency, the other pumps of this type might be subjected to a flow rate test in addition to the /,4([$$Ufff checks.

The surveillance requirements bases described in this paragraph apply only to the RCIC and HPCI systems. With a cooling system outGofGservice, overall core and containment cooling reliability is maintained by verifying the $)f///$fffff of the O remaining cooling systems. The verification of ////#fffff, as used in this context, for the remaining cooling systems means to administrative 1y check by examining logs or other information to verify that the remaining systems are not out-of-service for maintenance or other reasons. It does not mean to perform the surveillancerequirementsneededtodemonstratethe///////////oftheremaining systems. However, if a failure, design deficien O service period, then the verification of (////$f[c , etc. , causes the out-of-//f should be to assure that a similar problem does not exist on the remaining systems. For example, if an out-of-service period is caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test. Followinga////////g or an outage in which work was performed that directly affects system ///jff, the ..PCI and O RCIC pumps are flu rate tested prior to exceedir g 32 ps g and again at rated reactor steam pressure. This combination of testing provides adequate assurance of pump performance throughout the range of reactor pressures at which it is g 3.5/4.5-23 Amendment No. 130

Ot 0- INSERT "A" To TECHNICAL SPECIFICATION PAGE 3.5/4.5-23 The verification of OPERABILITY, as used in this context, for

-these other systems means to administratively check by examining logs or other information to determine if certain components / systems are out-of-service for maintenance or-other

() reasons. It does not usan to perform the surveillance rnquirements needed to demonstrate the OPERABILITY of the component / system.

INSERT "B" TO TECHNICAL SPECIFICATION PAGE 3.5/4.5-23 O

In addition, monthly: checks are-made on-the position of each manual, power operated or automatic valve. installed in the direct flowpath of the suction or discharge of each pump that is not locked, sealed, or otherwise secured in position.

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O QUAD-CITIES DPR-29 gre.

ewlins o required to operate. The low pressure limit is selected to allow testing at a V

point of stable plant operation and also to provide overlap with lo4 pressure ECG system 1., A time limit is provided in which to perform the required tests during startGap. This time limit is considered adequate to allow stable plant conditions to be achieved and the required tests to be performed. Flow rate testing of the HPCI and RCIC pumps is also conducted every 92 days at rated reactor pressure to q' demonstrate system ////gpff[ff in accordancegith the LCO provisions and to meet inservice testing requirements for the HPCI, system. Applicable valves are tested in accordance with the provisions of the inservice testing program. In addition, monthly checks are made on the position of each manual, power operated or automatic valve installed in the direct flowpath of the suction or discharge of the pump or turbine that is not locked, sealed or otherwise secured in sition. At each n fffffff/$ jfffy, a ffjff')f)(%$ ffffanda //t'f/fffffffff-f.ffftestisperformedontheHPCIan RCIC systems. The t s and checks

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described above are considered adequate to assure systemf///gff///[.

The verification of the main steam relief valve /////////// during manual actuation surveillance testing must be made independent of temperatures indicated n"

by thermocouples downstream of the relief valves. It has been found that a temperature increase may result with the valve still closed. This is due to steam being vented through the pilot valves during the surveillance test. By first opening a turbine bypass valve, and then observing its closure response during relief valve actuation, positive verification can be made for the relief valve opening and passing steam flow. Closure response of the turbine control g valves during relief valve manual actuation would likewise serve as an adequate verification for the relief valve opening. This test method may be performed over a wide range of reector pressures greater than 150 psig. Valve operation below 150 psig is limited by the spring tension exhibited by the relief valves.

The surveillance requirements to ensure that the distnarge piping of the core o' spray, LPCI mode of the RHR, HPCI, and RCIC systems is filled provides for a visual observation that water flows from a high point vent. This ensures that the line is in a full condition.

Instrumentation has been provided on core spray and LPCI mode of RHR to monitor the pressure of water in the discharge piping between the monthly intervals at O

which the lines are vented and alarm the control room if the pressure is inade-quate. This instrumentation will be calibrated on the same frequency as the safety system instrumentation and the alarm system tested monthly. This testing 9rdr than cc ensures that, during the interval between the monthly venting checks, the status a ual e of the discharge piping is monitored on a continuous basis. An alarm point of 0 psig for the low pressure of the fill system has been chosen because, due to n"

elevations of piping within the plant, 39 psig is required to keep the lines full.

j The-shutoff heed of the fill cycte pump: ic lecc than 90 pcig ad-therefore j wi44-net-defeat-the icw peenure cochg-pump-ducharge-preccure inter 4oct4f l 100 p:ig-es-shown in-Tab 1c 3,2-2. f ~.aegie-of-14-ps4 9-4+-prww+4ed by the-high-

! prwsure alarm point of 90 psig-l 5 HPCI and RCIC systems normally take a suction f om the Contaminated Condensate Storage Tanks (CCSTs). The level in the CCS '., is maintained at or above I,

3.5/4.5-24 Amendment No. 130 g

O QUAD-CITIES DPR-29 9.5 feet. This level corresponds to an elevation which is greater then the 0 elevation of the last check velves in the discharge pipes of either the HPCI or RCIC systems. Therefore, filled discharge piping of HPCI or RCIC systems is ensured when lined up to the CCST and tank level is at or above 9.5 feet.

The watertight bulkhead and submarine doors and the penetration seals for pipes and cables penetrating the vault walls and ceilings have been designed to U,

withstand the maximun flood conditions. To assure that their installation is adequate for maximum flood conditions, a method of testing each seal has been devised.

In order to test an electrical pei.atration or pipe seal, compressed air is q supplied to a test connection and the space between the fittings is pressurized to approxinately 15 psig. The outer faces are then tested for leaks using a soap bubble solution.

In order to test the submarine doors, a test frame must be installed around each door. The frame is then pumped to a pressure of approximately 15 psig and held g to test for leaktightness. The watertight bulkhead doors are tested by

' pressurizing the volume between the double gasket seals to approximately 15 psig. The gasket seal area is inspected using a soap bubble solction. Each RHR service water vault contains a sump, which will collect any floor or equipment leakage inside the vault. A sump pump will automatically start on high level in the sump, and will pump the water out of the vault, via 2 discharge check valves outsids'the vault to the service water discharge pipe. A composite sampler is g located on the sump discharge line. A radiation monitor is also located on the service water discharge. The sump discharge water is not expected to be contaminated, and any in-leakage to the vault is prevented by 2 check valves.

Surveillance of these check valves is performed each operating cycle to assure their integrity. The previously installed bedplate drains to the turbine building equipment drain sump have been capped off permanently.

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A level switch set at a water level of 6 inches is located inside each vault.

Upon actuation, the switch alarms in the control room to notify the operator of trouble in the vault. The operator will also be aware of problems in the vaults / condensate pump room if the high-level alarm on the equipment drain sump 1

0 is not terminated in a reasonable amount of time.

A system of level switches has been installed in the condenser pit to indicate and control flooding of the condenser area. The following switches are installed:

I Level Function

~O a. I foot (one alarm, icw water switch) level

b. 3 feet (one alarm, high water j switch) level

c. 5 feet (two alarm and cir-redundant culating water switch pairs) pump trip 3.5/4.5-25 Amendment No. 130

,O

O QUAD-CITIES ,

DPR-29 Level (a) indicates water in the condenser pit from either the hotwell or the

, O circulating water system. Level (b) is above the hotwell capacity and indicates a probable circulating water failure.

Should the switches at levels (a) and (b) fail or the operator fails to trip the circulating water pumps on alarm at level (b), the actuation of either level switch pair at level (c) shall trip the circulating water pumps automatically and O alarm in the control room. These redundant level switch pairs at level (c) are designed and installed to IEEE-279, " Criteria for Nuclear Power Plant Protection Systems." As the circulating water pumps are tripped, either manually or automatically at level (c) of 5 feet, the maximum water level reached in the condenser pit due to pumping will be at elevation 568 feet 6 inches clemier. (10 feet above condenser pit floor elevation 558 feet 6 inches; 5 feet plus an 0 additional 5 feet attributed to pump coastdown).

In order to prevent the RHR service water pump motors and diesel generator cooling water pump motors from overheatingig vault cooler is supplied for each pump. Each vault cooler is designed to maTntain the vault at a maximum *105'F temperature during operation of its respective pump. For example, if diesel O generator cooling water pump 1/2-3903 starts, its cooler also starts and maintains the vault at 105'F by removing heat supplied to the vault by the motor of pump 1/2-3903. If, at the same time that pump 1/2-3903 is in operation, RHR service water pump 1C starts, its cooler will also start and compensate for the added heat supplied to the vault by the 1C pump motogkeeping the vault at 105'F.

O Each of the coolers is supplied with cooling water from its respective pump's discharge line. After the water has been passed through the cooler it returns to its respective pump's suction line. The cooling water quantity needed for each cooler is approximately 1% to 5% of the design flow of the pumps so that the recirculation of this small amount of heated water will not affect pump or cooler operation.

Doeration of the fans and coolers is required during shutdown and thus additional surveillance is not required.

Verification that access doors to each vault are closed following entrance by personnel is covered by station operating procedures.

to lhe LHGR shall be checked daily to determine if fuel burnup or control rod move-4 ment has caused changes in power distribution. Since changes due to burnup are Islow and only a few control rods are moved daily, a daily check of power distri--

(@hutionisadequate.

A////gif P///,8f tER./. men. usar 4cucRnM wc 6 M At core thermal power levels less than or equal to 25%, operating plant experienceandthermalhydraulicanalysesir.dicatethattheresulting/////fl ff//f/ LHGR is below the maximum $ff// ///LHGRbyaconsiderablemargin; therefore,evaluationofthe/f(//(ff LHGR below this er level is not O nacessary. ThedailyrequirementforcalculatingJ////// LHGR above 25%

rated thermal power is sufficient, since power distribution shi ts are slow when there have not been significant. power or control rod changes.

3.5/4.5-26 Amendment ko. 130

O QUAD-CITIES DPR-29 O Local t+1GR- unana. usaTr c ewed.nried cars Mc, A)

}.

Add The LHGR as a function of core height shall be checked dcily during reactor PM operation at greater than or equal to 25% power to determine if fuel burnup or Ps control rod movement has caused changes in power distribution. A limiting LHGR M value is precluded by considerable margin when employing any permissible control O rod pattern below 25% rated thermal power.

- M///gfjfp C/fff#/ P//// R/// (MCPR)

At core them power levels less than or equal to 25%, the reactor will be O operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicate that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.

The daily requirement for calculating Mr.FR above 25% rated thermal power is sufficient, since power distribution shifts are very slow when there have not been significant power or control rod changes. In addition, the Kf correction, as specified in the CORE OPERATING LIMITS REPORT, applied to the LCO provides margi.1 for flow increases from low flows.

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> en SIGNIFICANT- HAZARDS: CONSIDERATIONS 3: - -

AND ENVIRONMENTAL ASSESSMENT EVALUATION Q;

PROPOSED TS 3-5/4.5 i:

"COREE& CONTAINMENT: COOLING SYSTEMS" y

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U EVALUATION EQR SIGNIFICANT RAZARDS E9NSIDERATION PROPOSED SPECIFICATION 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS The proposed changes provided in this amendment request are made in order to provide a more user friendly document, incorporate desired technical improvements, and to incorporate some improvements from later operating BWRs. These changes have been reviewed by Commonwealth Edison and we believe that they do not present a Significant Hazards Consideration. The basis for our O determination is documented as follows:

BASIS IQB EQ SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards consideration.

r3 In accordance with the criteria of 10 CFR 50.92[c) a proposed v amendment to an operating license involves no s:.gnificant hazarna consideration if operation of the facility, in accordance with tm.

proposed amendment, would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated, because:

O~

a. The Generic Changes to the technical specifications involve administrative changes to format and arrangement of the material. As such, these changes cannot involve a significant increase in the probability or consequences of an accident previously evaluated.

O b. Proposed Changes to Core Spray Subsystems, LPCI Subsystems and Containment Cooling Mode of the RHR System The proposed change to the Applicability replaces the present operability requirements of whenever irradiated fuel is in the reactor vessel and prior to reactor startup from a cold condition with Operational Modes 1, 2, and 3.

.r)'-

Operational Modee 1, 2, and 3 as defined in Table 1-2 provide equivalent conditions for operability as present requirements. The remaining changes to these sections involve reordering of present requirements and addition of some STS provisions whlch have been evaluated and found acceptable for use at Quad Cities. The deletion of the O one time exception in Specification 3.5.B.1.b is an i

administrative change. Considering the nature of these changes, they cannot result in a significant increase in the probability or consequences of an accident previously evaluated.

Proposed changes to the testing requirements include

^m > deletion of the testing of other systems when one system 1

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-, O is inoperable. The present testing requirements for ECCS J when one system (or component) is inoperable represent requirements .beyond those necessary to adequately demonstrate system operability. Other testing requirements that are not affected by this propor,ed amendment provide assurance that remaining ECCS systems O are operable and capable of performing their design intent. The proposed deletion of multiple system testing will bring Quad Cities Units 1 and 2 in line with current BWR plant operating practices. Since ECCS systems perform accident mitigation functions, this proposed amendment does not affect the probability of an accident previously evaluated. The proposed amendment does not involve a O significant increase in the consequences of an accidant previously evaluated because testing, other than multiple system testing, ensures that the present level of operability for the ECCS systems is maintained. The remaining changes to the testing requirements involve deletion of the monthly pump and valve operability tests, the addition of monthly valve position verification O checks, the deletion of post-maintenance testing caquirements, and changing the channel calibration quency from quarterly to refueling tor the core spray

. er dp instrumentation. The deletion of the monthly p and valve operability tests is based on later plant usating methods and reliance on the quarterly Inservice o Testing Program to determine operability. Additionally, the newly proposed monchly valve position verification checks have been demonstrated through use at other plants to provide assurance of system alignment. Post-maintenance testing requirements are contained in maintenance procedures and are not needed in the Technical Specifications. The change to the channel calibration 9 frequency has been evaluated, considering instrument drift, and found to be acceptable for implementation.

Proposed SR 4.5.A.6 on the doors for the core spray and RHR pump compartments moves this requirement from present Specification 3.7.C.2/4.7.C.2 while retaining present intent.

O The proposed changes do not affect any accident precursors and therefore, do not increase the probability of an acc1 dent previously evaluated. The changes to the Applicability replace present requirements with equivalent operability conditions. The changes in testing requirements represent deletion of outdaced provisions

$ while implementing provisions that will provide assurance of system operability. Therefore, the changes do not involve a significant increase in the consequences of an accident previously evaluated.

c. Proposed Changes to HPCI, Automatic Pressure Relief Subsystemt and RCIC 9

0-C) The present Applicability for the HPCI and RCIC systems is whenever reactor pressure is greater than 150 psig, irradiated fuel is in the reactor vessel and prior to reactor startup from a cold condition. The proposed change for these systoms to Operational Mode 1 and Operational Modes 2 and 3 whenever reactor pressure is o' greater than 150 psig provides equivalent operability requirements to present provisions and allows implementation of the standard Operational Mode terminology defined in Table 1-2.

The proposed change to the operability for the Automatic Pressure Relief Subsystems will require operability in O Operational Mode 1 and Operational Modes 2 and 3 whenever reactor pressure is greater than 150 psig. The present operability requirement is whenever the reactor pressure is greater than 90 psig, irradiated fuel is in the reactor vessel and prior to startup from a cold condition. This change will retain sufficient pressure overlap between the

,s automatic pressure relief valves and the low pressure ECCS

'd systems to assure that operation of the low pressure ECCS performs as assumed in the accident analysis. The change to 150 psig for operability of the automatic pressure relief valves is supported by testing methods that are presently described in Technical Specification Bases 4.5.D. This testing method requires the reactor to be o greater than 150 psig before operability of the valves.is verified. Since the proposed change retains sufficient overlap to ensure proper low pressure ECCS operation when required by the accident analysis, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The propoced changes to the LCO and Actions for inoperable O relief valves implements STS guidelines which are more restrictive than present provisions. Proposed LCO for relief valves requires all five valves to be operable.

Proposed Actions will allow only one automatic pressure relief valve to be inopera';1e for 14 days, provided HPCI, both core sprays, anu LPCI are operable.

O The proposed deletion of Surveillance Requirement 4.5.D.4 will delete the demonstration of HPCI operability when two valves of the autanatic pressure relief subsystem are inoperable. The operability testing of a system when another system is lnoperable represents unnecessary requirements that are no longer being implemented at later O operating plants. The verification of HPCI operability is presently required by Specification 3.5.D.2 when two of the valves are inoperable. This verification will be performed when one automatic pressure relief valve is inoperable and will involve administrative checks by examining logs or other information to determine if certain components / systems are out-of-service for O maintenance or other reasons. Since system operability is O

O C) being determined and being retained when required to function, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Proposed Surveillance Requirements for the HPCI, Automatic O Pressure Relief System and RCIC are based on present provisions and STS guidelines which are applicable to Quad Cities. Post-maintenance testing provisions are not necessary in the technical specifications and are being deleted. The extension of the present six-month testing frequency to 18 months for manually opening each relief valve at power, is justified, based on corrective actions O performed in the maintenance of these valves and the reduced number of valve failures since 1984. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

'9 d. Proposed Changes to Minimum core and Containment Cooling System Availability The proposed amendment maintains necessary ECCS pumps and flow paths to provide makeup water to the reactor vessel in the cold shutdown and refuel conditions. The proposed action provisions ensure that operations with a potential O for draining the reactor vessel and core alterations are not performed without necessary ECCS makeup capability.

As an additional restriction, with both of the required ECCS pumps and/or associated flow paths inoperable, secondary containment integrity is established. Present BWR operating philosophy for cold shutdown and refueling conditions demonstrates that the availability of two ECCS 9 pumps with associated flos paths to the reactor provide sufficient assurance that makeup capability for accidents that have a potential to drain the vessel in these operating conditions is available. The deletion of present Specification 3.5.F.1 is justified since the proposed changes provide definitive operability statements g for the Core and Containment Cooling systems. Since necessary Core and Containment Cooling capability is maintained, this proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated,

e. Proposed Changes to Suppression Chamber Requirements for 9 Core and Containment Cooling Systems Suppression Chamber operability is required when systems depending on its water supply are required to be operable.

Operability is determined in Operational Modes 1, 2, and 3 by requiring a contained volume corresponding to minus 2 inches on the level indicator. Water level requirements I in Operational Modes 4 and 5 are less than t' tat required .

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when the plant is operating. The proposed water level of

_. at least 7 feet above the bottom of the suppression chamber provides sufficient margin to ensure necessary pump operability during these operational Modes.

Operability in operational Mode 5 is not required if the conditions of proposed LCO 3.5 I.3 are met including g reactor vessel head removal, the reactor vessel cavity is flooded, the spent fuel pool gates are removed, the spent fuel pool water level is maintained above the low level alarm point and the reactor cavity water temperature is below 140 degrees F. With all the stated conditions of the LCO met, core and containment cooling capability is retained when there is a potential need for this function.

The new surveillance Requirements are added to ensure that the conditions of the LCO are being met in the applicable operational Modes. Thus the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

g f. Proposed Changes to Maintenance of Filled Discharge Pipe o

The separation of the present requirements into two distinct specifications is made for clarification purposes. STS guidelines that are applicable at Quad Cities are followed in the addition of Action provisions with a discharge line keep filled pressure alarm channel 4 inoperable. Present Specification 3.5.G.2 requires that if the discharge pressure for the Core Spray and LPCI mode of RHR is not restored within limits in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , then an orderly shutdown.of the reactor shall be initiated and the reactor shall be in a cold condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after initiation. Present Specification 3.5.G.2 is proposed to be changed to require the affected system to be declared 8 inoperable and appropriate actions taken if the discharge ~

pressure cannot be restored in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Since the discharge pipe pressure affects system operability, any actions as a result of a loss of this pressure should be deferred to the affected system operability requirements.

Reliance on affected system operability in this case g follows established STS guidelines for system and support system operability requirements. Similarly, the same type of provision is proposed for the discharge pipe pressure Action requirements for the HPCI and RCIC systems. For HPCI and RCIC, if the level of the CST falls below 9.5 feet for longer than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and/or the RCIC and HPCI are not lined up to take suction from the suppression chamber, 9 HPCI and RCIC are declared inoperable and appropriated system actions are followed. The proposed changes follow proven STS guidelines for declaring affected systems inoperable when discharge pipe pressures are not being met and thus retains necessary system and plant protective features. Proposed Surveillance Requirement 4.5.K.1 is g added to ensure the provisions of the LCO for CCST level is being met. The proposed deletion of the upper limit of 9

O E) 90 psig for the fill systems for Core Spray and RHR removes an unnecessary setpoint from the Technical Specifications that does not perform any safety function.

Thus, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously O evaluated,

g. Proposed Changes to Condensate Pump Room Flood Protection The proposed changes provide clarification by clearly stating the provisions of the LCO and requiring operability of the Flcod Protection provisions when the C) pumps in the Condensate Pump Rooms are required to be operable. Instead of the present provision to shutdown the reactor after 7 dayo if the trip or alarm circuits are not returned to operable, the proposed Action results in declaring the affected systems inoperable if the trip circuits are not restored and preparing a Special Report if the alarm circuits are inoperable for more than 30

-() days. The proposed changes follow proven STS guidelines to defer to affected system operability requirements if support systems are inoperable. The change also recognizes the greater importance of the trip circuits for the Condensate Pump Room Flood Protection over the alarm circuits for the same function. The proposed change does O not alter trip or alarm setpoints and retains present operability requirements for the trip systems. Thus the proposed changes do not involve a significant increase in the-probability or consequences of an accident previously evaluated.

, h. Proposed Changes to Average Planar LHGR, Local LHGR and U- MCPR The only technical changes proposed in these sections is to require operability in Operational Mode 1 when thermal power is greater than or equal to 25% rated thermal power and to require in the Action provisions a reduction to r)-

below 25% rated thermal power if limits are not. restored.

The proposed changes to tne applicability and actions will provide consistency with present surveillance requirements for determining these parameters daily during steady-state operation above 25% rated thermal power. The proposed changes are consistent with later operating plants' requirements and have been evaluated and found to be 70 appropriate for use at Quad Cities Units 1 and 2. Thus

, the proposed changes do not involve a significant increase in the probability or consequences of an accident l

previously evaluated.

2) Create the possioility of a now or different kind of accident

,3 from any previously evaluated because: 1 l'

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{3 O a. Since the Generic Changes proposed to the technical specifications are administrative in nature, they cannot create the possibility of a new or different kind of accident from any previously evaluated.

b. Proposed Changes to Core Spray Subsystems, LPCI Subsystems O and Containment cooling Mode of the RHR System The proposed changes to the Applicability for the subject systems retains equivalent system operability requirements of the present provisions. The proposed addition of STS Action provisions for the core spray header dp instrumentation adds necessary operational flexibility to O ensure that the system is not declared inoperable unless necessary. The proposed deletion of the one time exception of Specification 3.5.B.1.b is administrative in nature and cannot because of its nature affect the accident analysis. Lie proposed changes to the testing requirements implement provisions that are currently being used n later operating BWRs. The deletion of the O multiplo testing of other systems when one system is excessive testing incperable merely requirements. The deletes proposed unnecessary,ilize changes ut proven Inservice Testing Provisions to demonstrate system operability on a quarterly basis. The changes described in this amendment request do not modify system designs or O actuation setpoirits. Since the proposed changes maintain syctem operability requirements when required by accident and transient analysis, there is no possibility of a new or different kind of accident from any previously evaluated.

q c. Proposed changes to HPCI, Automatic Pressure Relief

- Subsystems and RCIC The proposed changes for these systems will require operability in Operational Mode 1 and Operational Modes 2 and 3 whenever reactor pressure is greater than 150 psig.

This change ensures system operability when required and g retains adequate discharge pressure overlap with the low pressura core cooling systems. The r- 7 posed changes to the LCO and Action provisions for th, Automatic Pressure Relief Subsystem helps to ensure operability of the required number of valves assumed for accident considerations. The proposed deletion of present Surveillance Requirement 4.5.D.4 will remove unnecessary C) multiple system testing when one system is inoperable.

The proposed change from the present six-month testing frequency to 18 mon'.hs for manually opening the Automatic Pressure Rolief Valves at power is justified based on corrective actions performed at Quad Cities. The changes described to these systems do not modify system design or 4q actuation setpoints and thus do not create the possibility of a new or different kind of accident.

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Of

..O d. Proposed Changes to Minimum Core and Containment cooling System Availability ,

The proposed changes follow proven STS guidelines for i minimum core and containment cooling system availability during cold shutdown and refueling. The proposed-() requirements are more prescriptive than present provisions and ensure that operational flexibility as well as adequate protective features are provided. The proposed changes maintain necessary ECCS makeup capability during cold shutdown and refueling conditions. The proposed changes add restrictions for taking all low pressure core :

and containment coolin O- refueling conditions. g These systems out of service restrictions during ensure that sufficient water volume is available in the reactor cavity '

and refueling pools before taking all these systems out of

-service. The proposed changes do not introduce any new modes of-operation and maintain necessary ECCS availability: therefore, there is no possibility of a new or different kind of accident from any previously i C) - evaluated,

e. Proposed Changes to-Suppression Chamber Requirements for Core and Containment Cooling Systems The proposed changes define.the opernbility requirements O for the Suppression Chamber in all reactor modes.

Suppression Chamber water level requirements for Operational Modes 4 and 5 are reduced from present provisions in order to allow operational flexibility while maintaining necessary water volumes for plant shutdown conditions. New Surveillance Requirements are added to ensure-that the conditions of the LCO are being met. No

'O- changes are being made to system design requirements or there is no possibility of cooling a-new or capability, therefore,ident different kind of acc from any previously evaluated.

f. Proposed Changes to Maintenance of' Filled Discharge Pipe C The proposed changes to the Action requirements will ensure that affected systems are declared inoperable as applicable, when discharge pipe pressures are not bekng maintained. New Surveillance Requirement 4.5.K.1 is added to ensure that the CCST-level is maintained in accordance with.LCO requirements. The-proposed changes maintain

() necessary system operability and ensure the provisions of the LCO are being met. The changes do not alter system

' design-requirements or initiation setpoints; therefore, there is no possibility of a new or different kind of accident from any previously evaluated..

O g. Proposed Changes to Condensate Pump Room Flood Protection O

Thep$$osedchangesto'theLCOaremadetoclarifythat operab ity of the Condensate Pump Room Flood Protection is required only when equipment that-is served by the cooling pumps in these rooms is required to be operable.

System design is not modified by the proposed changes and present trip and alarm setpoints are maintained. The

). proposed Actions are based on STS guidelines to-defer to the affected system action requirements if a trip circuit is inoperable. The alarm circuits are separate from the trip circuits and requirements for operability have been reduced to reflect the less significant alarm function.

However, operability of the alarm circuit is retained in the-Technical Specifications in order to assure

). operability of a function that can indicate potential

-problems with flooding in the Condensate Pamp Rooms before the trip setpoint is reacned. Since the design function-of the Condensate Pump Room Flood Protection circuits has not been moaified, there is no possibility of a new or different kind of accident.

h. Proposed Changes-to Average Planar LHGR, Local LHGR and MCPR The changes to require applicability in Operational Mode 1 when thermal power-is greater than or equal to 25% rated thermal power and to the actions to require a reduction in

_) power to < 25% rated thermal power.follows later operating F BWR plant provisions and is consistent with present Surveillance Requirements for these functions. Present Surveillance conditions are being maintained and thus monitoring of these functions is not being modified.

Therefore, there is no possibility of a new or different kind of accident.

.i . Proposed Relocation of Figures 3.5-1 and 3.5-2 to the Core-Operating Limits Report The proposed relocation of these Figuren to the COLR does not change any existing limits. Any changes needed to these figures will be handled by revising the COLR.

31 Therefore, the proposed changes do not create the possibility of a new or different kind of accident.

3) Involve a significant-reduction in the margin of safety because:

) a. The Generic Changes are administrative in nature and as such they cannot involve a significant reduction in the margin of safety.

b. Proposed Changes to Core Spray Subsystems, LPCI Subsystems and Containment Cooling Mode of the RHR System

)- Theproposedchan$etorequireoperabilityofthese systems in Operat onal Modes 1, 2, and 3 provides

)

0-

- O equivalent operability requirements to the present

... provisions and thus the margin of safety is maintained.

The addition of STS Action provisions for the core spray header dp instrumentation provides necessary remedial measures to ensure that system operability is maintained.

The reordering of present requirements and the deletion of O '

the one time exception of Specification 3.5.B.1.b are administrative changes that can have no significant affect on any margin of safety.

The proposed changes to the testing requirements will remove excessive multiple system testing requirements while maintaining system availability. Excessive testing O of systems and components can reduce rather than increase reliability. An acceptable level of testing to demonstrate operability currently being used at later BWR plants does not include multiple testing of other systems when one system is inoperable. The testing that will remain in the technical specificatirns provides adequate assurance of system performance. Otuer changes to the O testing requirements include adoption of proven testing methods at other plants and acceptance of the quarterly IST pump and valve testing in lieu of monthly pump and valve operability demonstrations The proposed changes follow similar provisions that are e implemented at BWR plants with systems similar to those at Quad Cities and maintain necessary system availability.

Therefore, the proposed changes do not involve a significant reduction in the margin of safety.

c. Proposed Changes to HPCI, Automatic Pressure Relief Subsystems and RCIC The proposed change to the Applicability for the HPCI and RCIC systems to Operational Mode 1 and Operational Modes 2 and 3 whenever reactor pressure is greater than 150 psig, provides equivalent conditions to the present provisions and thus the margin of safety is preserved.

g The changes to the LCO and Action provisions for the Automatic Pressure Relief Subsystem will require five valves to be operable instead of the present allowance of three valves. Proposed Actions allow one relief valve to be inoperable and plant operation to continue for 14 days provided HPCI, both core sprays and LPCI are operable.

This change follows STS guidelines and will help to ensure 9 system operability when required.

The proposed changes will increase the operational pressure requirement for the automatic pressure relief valvas from 90 to 150 psig. However, operating pressure overla with the low pressure ECCS subsystems is suffic ently maintained such that operation of these O subsystems is assured in accordance with present accident 9

O.

O analysis assumptions. LPCI and core spray will inject at approximately 325 psig and thus the increase from 90 to 150 psig for relief valve operation does not impact this ECCS function. Therefore, the change does not involve a significant reduction in the margin of safety.

O The proposed deletion of Surveillance Requirement 4.5.D.4 will remove unnecessary testing of the HPCI system when two valves of the Automatic Pressure Relief Subsystem are inoperable. The proposed change maintains the requirement for HPCI to be operable, but does not require a demonstration of that operability. This change follows similar testing provisions at later BWR plants and has O been evaluated and found to be applicable for use at ouad Cities Units 1 and 2. The remaining changes to testing provisions follow STS and later operating plant provisions that have been evaluated and found acceptable for use at Quad Cities. Since the Automatic Pressure Relief Subsystem operability is maintained in accordance with desi n assumptions the changes do not involve a fO sign ficant reduct1on in the margin of safety.

d. Proposed Changes to Minimum Core and Containment Cooling System Availability The proposed change will allow fewer ECCS systems to be O required operable in cold shutdown and refueling conditions than is required by present specifications.

However, the present specifications are overly restrictive and do not reflect the difference in water makeup requirements for reactor power operation conditions and cold shutdown and refueling conditions. With the reactor n not pressurized in cold shutdown and refueling conditions, v the ECCS makeup requirements are less than when in the pressurized condition. Later BWR plants have demonstrated that the availability of two ECCS pumps with associated flow paths to the reactor provide sufficient water makeup capability for the cold shutdown and refueling conditions.

The added action requirements if one or both of the 73 required ECCS pumps and/or associated flow paths are

" inoperable ensures that operations such as core alterations and those with a potential for draining the vessel are not conducted without necessary ECCS makeup capability or adequate heat sink available. In addition, the actions proposed will require establishing secondary containment integrity if the required EC^S pumps and/or

{} associated flow paths are not returned to operable status.

Restrictions are also added to present Specification 3.5.F.2 requirements concerning all low pressure core and containment cooling systems being allowed out of service.

These restrictions require the reactor vessel head to be n removed, the cavity to be flooded, the spent fuel pool v gates to be removed, the fuel pool water level to be O

l J'

'O - maintained above the low level alarm point, and reactor cavity water temperature to be below 140 degrees F The provision te allow r.ll core and containment cooling systems to be inoparable applies only to the refueling condition and add restrictions not presently in the technical specifications. The present margin of safety is rw not significantly reduced because necessary ECCS water makeup capability for the cold shutdown and refueling conditionr is being maintained and additional restrictions are being imposed before taking all low pressure core and containment cooling systems out of service,

e. Proposed Changes to Suppression Chamber Requirements for O Core and Containment Cooling Systems The proposed changes to the Suppression Chamber operability requirements implement present Quad Cities provisions and adopt later operating BWR provisions. The proposed changes maintain aither the suppression chamber 7, operable or a sufficient heat sink to handle cooling V requirements in Operational Modes 4 or 5. New Action provisions provide definitive steps to be taken which will provide adequate plant and modes of reactor operation. personnel protection in allSince adequate Suppres Chamber cooling and makeup capacity is maintained by the proposed changes, there is no significant reduction in the

() margin of safety.

f. Proposed Changes to Maintenance of Filled Discharge Pipe The proposed changes to the Action requirements for loss of discharge pipe pressure for the Core Spray and LPCI

, mode of RHR will require appropriate system actions to be L) taken rather than require a plant shutdown per present Specification 3.5.G.2 and will properly address inoperability of keep filled instrumentation. This change will not modify present Core Spray or LPCI system operability requirements and thus the margin of safety is maintained. Likewise, if the level of the CST falls below q

v 9.5 feet for longer than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and/or the RCIC and HPCI are not lined up to the suppression chamber, the proposed changes will require RCIC and HPCI to be declared inoperable and appropriate actions taken. The addition of Surveillance Requirement 4.5.K.1 will help to ensure that the level of the CST is maintained in accordance with LCO provisions. The deletion of the 90.psig upper limit for C). the fill systems for Core Spray and RHR removes an unnecessary requirement. Since core Cooling System operability is not modified by the proposed changes, there is no significant reduction in the margin of safety.

g. Proposed Changes to Condensate Pump Room Flood Protection O The proposed changes will require operability of tne flood O

O

, O protection when systems served by the pumps in the J Condensate Pump Rooms are required to be operable. This change provides operability of the flood protection devices only when required to perform a protective or alarm function. The separation of the requirements for the alarm and trip functions places primary emphasis on the trip and secondary emphasis on the alarm function.

O However, the alarm function is required to be maintained operable in order to provide early warning and detection capability. Since flood protection is being maintained when required to assure operability of affected safety systems, the changes do not involve a significant reduction in the margin of safety.

O

h. Proposed Changes to Average Planar LHGR, Local LHGR and MCPR The proposed changes to require operability in Mode 1 when thermal power is greater than or equal to 25% rated i thermal power and to the actions to require reduction in

power to < 25% rated thermr?. power, provides consistency with present surveillance provisions and later operating BWR plant specifications. Therefore, the proposed change does not involve a significant reduction in the margin of safety.

G 9

9 O

e

D ENVIRONMENTAL ASSESSMENT EVALUATIOE

{) PROPOSED SPECIt'ICATION SECTION 3.5/4.5 CORE & CONTAINMENT COOLING SYSTEMS Commonwealth Edison has evaluated the proposed amendment in accordance with the requirements of 10 CFR 51.21 and has determined that the amendment meets the requirements for

{) categorical exclusion as specified by 10 CFR 51.22(c)(9).

Commonwealth Edison has determined that the amendment involves no significant hazards consideration, there is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite, and there is no significant increase in individual or cumulative occupational radiation exposure.

The proposed amendm2nt does not modify the existing facility and does not involve any new operation of the plant. As such, the proposed amendment does not involve any change in the type or

) significant increases in effluents, or increase individual or cumulative occupational radiation exposure. The proposed amendment to Section 3.5/4.5, " Core & Containment Cooling Systems", contains administrative changes such as including appropriate applicability statements within the specifications to define the applicability during operating mode and the required actions to be implemented in the event that specification cannot h) ae met. The information is consistent with the Standard Technical I

Specifications or later operating plants. In cddition, some existing requirements have been updated and new requirements added to reflect the Standard Technical Specifications or later operating plant requirements.

D l

3 l

P D

?

D-QC-1 / QC-2 DIFFERENCES 3

TS 3.5/4.5 3

' CORE & CONTAINMENT COOLING SYSTEMS' l

l 1

)

D D

I

!IIh i

n

%> COMPARISON'OF UNIT 1 AND UNIT-2 TECHNICAL-SPECIFICATIONS FOR'THE IDENTIFICATION OF. TECHNICAL DIFFERENCES

-SECTION-3.5/4.5 ,

Q CORE AND CONTAIN 4ENT COOLING SYSTEMS r

' Commonwealth Edison has conducted a comparison review of the Unit 1 and Unit 2 Technical Specifications to identify any technical differences in support of combining the '

Technical Specificaticas into one docum6nt. The intent of 3 ': the review was not:to identify-any dif';erences in presentation style (e.g. table formata, use of capital 11etters, etc. ) or punctuation but rr cher to identify areas which the Technical Specifications'are technically or administratively different.- Due to the volume of differences

. Specificatio.the.

identified- appropriate pages from the Technicalns are marked-up-to assist in the 0;--

The-review of Section 3.5/4.5," Core and Containment

~ Cooling Systems" revealed the following administrative and technical differences:

Page 3.5/4.5-3 0 4.5.A.5 Unit 1:

~

subsystems, the containment Unit 2'. subsystems and the containment Pace 3.5/4.5-4 (Unit 1 and Unit 11

~3.5.B.3 Unit 2 has the following paragraph:- During

I) . L the period from April 17, 1978 through April _ >

-30, 1978-while the 2A Containment Cooling Loop of the RHR System-is made inoperable-for heat .

exchanger repair, continued reactor operation is permisnible beyond the 7-day limitation,-

unless such loop is sooner made operable, provided that during the time the 7-day-limit

O' is exceeded, a visual inspection is performed daily-to assure that proper valve alignment and system integrity is maintained in the B

~

RHR loop.

This paragraph is specific only to Unit 2 during a limited time frame.

.O Pace 3.5/4.5-7 4.5.D.4 Unit 1: two valves of the Unit 2: two relief valves of the

OL O.

O -

O pgge 3,5f4,3_11 4.5.H.1.a Unit 1: by pressurizing to 15 + 2 Unit 2: by pressurizing to 15 + 2 Unit 2 is correct. Error introduced in Unit 1 Technical Specification re-type (Amendment Q 114)

P_ tan 1 5/4.5_n 3.5.I Unit 1: the APLHGR is not returned to within Unit 2: the APLHGR is not returned in within O page 3.5/4.5-is First Unit 1: is available whenever irradiated fuel is in the reactor vessel.

Unit 2: is available.

second Unit 1: analytical methods described in O General Electric Topical Report NEDC 31345P Unit 2: analyses included in References 1 and 2 and in accc. dance witn 10 CFR 50.46 and Appendix K, Unit 1: accident, to limit calculated fuel

$ Unit 2: accident, to limit the calculated fuel Unit 1: intact, to limit cladding Unit 2: intact, to limit the corevide cladding O The third, fourth and fifth paragraphs which are part of the Unit 1 Technical Specification Bases are not included in the Unit 2 Bases.

This information is applicable to both Units, tnerefore, will be retained in the combination.

O Eane 3.5/4.5-16 First Unit 1: given in Reference 1 provides Unit 2: given in Reference 3 provides Reference 3 is the correct reference since the O content of the paragraph is discussing allowable out-of-service and repair times.

O O

Eagg 1 5/4.5-11 Fourth Unit 1: core spray subsystem and the LPCI Unit 2: core spray subcystem and/or the LPCI The Unit 2 bases is correct. The error in

) Unit 1 was introduced during the Unit i re-type (Amendment 114)

Eagg 3.5/4.5-13 First Unit 1: is acceptable provided the appropriate MAPLHGR reduction factor b is applied to assure compliance with the 2200 F PCT limit.

Unit 2: is acceptable.

Un!.t 1: Loss of more thr.n one relief Unit 2: Loss of two relief

> components and recirculation system.

Fourth Unit li Unit 2: components and recirculation systems.

Eag2 2 5/4.5-19 Eixth Unit 1: the 2200 F limit specified in the 10

> Unit 2: the 2200 F limit specified in 10 Eagg 3.5/4.5-20 First Unit 1: a function of the average heat-

< generation rate of Ur.it 2: a function of the average LHGR g

Unit 1: all the rods of a fuel Unit 2: all the rods in a fuel

. a,,t ' 1: power distribution within an atnembly, p Unit 2: power distribution within a fuel assembly.

Unit 1: temperatures are below the limit Unit 2: temperatures are below the 10 CFR 50 Appendix K limit I

m

J o'

Eagn 2 5/4.5-20 Ignptinued) l 1

Unit 1: The Avorage Planar Linear float Generation Rato (APLilGR) also sorvos a secondary function which is to assure fuel rod 7'

' mechanicalintegrity.

Unit 2 Pago 3.5/4.s 13 and -14: Powur operatlon with LllGRs at or below those specified in the CORE OPERATING LIMITS REPORT assures that the peak cladding temperature following a postulated Johs-of-coolant

, accident will not exceed the 2200 F limit.

'J Those values represent limits for operation to ensure conformance with 10 CFR SO Appendix K only if they are more limiting than other design paramotors.

The maximum average planar LilGRs specified in the CORE OPERATING LIMITS PEPORT at higher q exposures result in a peak cladding temperature of less than 2200 P. Ilowever the maximum averago planar LilGRs are specified in the CORE OPERATING LIMITS REPORT as limits because conformac.ce calculations have not boon performed to justify operation at LilGRs in excess of those shown.

O The Unit 2 Technical Specification Bases will be adopted.

Pane 3.5/4.5-21 3 Second Unit 1: The MCPR Operating Limit reflects an increase of 0.03 over the most limiting

~

transient to allow continued operation with one feedwater heater out of service.

Unit 2 does not have this paragraph. Since

.he paragraph is applicable to both units, the paragraph will be retained.

()

EAlg L J / 4 . 5-14 a LUnit 21 The Unit 1 Technical Specification Bases does not have the information which is contained on m

this Unit 2 page. Commonwealth Edison has

<J evaluated this information and has dotormined that the informa' on will not be retained in the combined Technical Specification bases.

The information was determined not to add value to the bases objective of understanding the limiting condition for operation. The information contained in the bases is provided g in other plant documents, o

f)

%)

O Engn 3.5/4.5-22 References Unit 1: 1. " SAFER /GESTR-LOCA Loss of Coolant Analysis for QuadCities Nuclear Power Station Units 1 & 2" O Unit 2: " Quad Cities Nuclear Power Station Units 1 & 2 SAFER /GESTR-LOCA Lons of Coolant Accident Analysis" g) Enna 12h44.5-23 Third Unit 1: out-of-service period is caused by failuro Unit 2: out-of-service period cauaed by failure g EAns 22 5/4.5-24 Second Unit 3: An alarm point of > 40 poig Unit 2: An alarm point of 40 poig Unit 2 Technical Specification bases is correct. The Unit 1 error was introduced O during the re-type (Amendment 114).

Unit 1: cooling pump discharge pressure Unit 2: cooling pump dischargo press Unit 1: interlock of 100 psig g Unit 2: interlock 100 psig a

Pane 3.5/4.5-26 Fifth Unit 1: at a maximum 105 F Unit 2: at a maximum of 105 F O Fourth Unit 1: planar L11GR is below Unit 2: planar LilGR below Enne 3.5/4.5-22 OcVen Unit 1: The daily requirement Unit 2: The daily required O

O

O QUAD-Cll!ES OPR-29 mode of the RHR, all active O components of both core spray subsystems, and the diesel generators required for operation of such components if no external source of power were available shall be operable.

5. from and after the date that the 5. When it is determined that the LPCI mode of the RHR system is LPCI sade of the RHR system is made or found to be inoperable he'ptTatden th core spray sub-for any reason continued reac- . systems, the ontainment cooling tor operation Is permissible inode(nLthe' R shall be

$ only during the succeeding 7 demonstrated to be operable days unless it is sooner made irrnediately and daily thereaf ter.

operable, provided that during such 7 days all active compo-nents of both core spray sub-systems, the containment cooling g mode of the RHR (including two RHR pumps), and the diesel gen-erators required for operation of such components if no exter-nal source of power were avail-able shall be operable.

O

6. If the requirements of Specifi-cation 3.5.A cannot be met, an orderly shutdown of the reactor shall be initiated, and the re-actor shall be in the cold shut-O down condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. Containment Cooling Mode of the RHR B. Containment Cooling Mode of the RHR System System Surveillance of the containment

'O . cooling mode of the RHR system shall be performed as follows:

1. a. Both loops of the 1. RHR service water subsystem containment cooling mode of testing:

the RHR system, as defined lC l in the bases for Spe- Item frequency cification 3.5 B, shall be j operable whenever irradiated a. Pump and valve Once/3 i fuel is in the reactor operability months l vessel and prior to reactor j startup from a cold Q condition.

l l

l 3.5/4.5-3 Amendment No. D 4 lO l

O '

Q M -C111t$

tr$-M O

3. Frts ared af ter the date that ore lo@ of 3.

the contaltwnt coolir nde o De les taen noir ofore tie10@

m R systars teccevsof Uw contatre systen is ude or fcv to te eg erele incperable, the cgerable locp shall te for any rtason crritira.e4 reactor domris rated to te coerable lemediately cgeration is ge,rmissible ont durIng ud and da ly pertaftar.

Succeeding i days un1tts su( sd systen is sooner made (gerable all active ctrgorents of provised the oDer that lo@

Q of the contaltret coating me of the -

M R syste toth cat spray sesystes, andtothdIrsel rakes trtd for e eration of su .( M ts f no s

e arral aa11215_twete*of

, shall te cgerable.sevef' <ee DurinlhesuDe ilee h Aprterlod frtn 4rl) 11 O isto .i so, lors whlie (te 2A

Contaleren Ccc ,ng text of ife fem Systen is ude r eactan r epair, getable contlovedfor heet reactor cgerat on s gerwis lble teyond tie

,7 atove 7-day 11mitat on, unless such lo@

15 er ude gerab1t p ided that dar rig the* tiee

' a c'+**d the

' 5 5"* 1-day *' "alt

' l a50'< is is O perforred dall to assure that p r valve aligten and s ulatalted in U4 *B*ystem integr IDR 10@. ty is

4. t ce, silt's sprayda@saft 4 Charing 41 required crte deerkble den tte reactor Std)) te each Sqtaron gerforved terlod,dran tre air test 11 spray water te erature is greater itan headers and rottles ar.d a va er spray 2120F rior to reactor start @ test gerforved on tie torvs spray header O , f r<n a coi condit cri tirmed reactor cperation s gero.ited prowl 6H!

and roanies.

that a raninn of cre depell spray loop my te troceratie for 30 days ven De reactor water terterature is greater than 212tf.

$. If the regairerents of 3.$.8 cannot te O set an orderly shutaun shaii te inillated, and the reactor stall te in a cold Stut&wn condition uithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

O O

O ^*"d"*"t N " ' 9' O

)

QUAD-CITIES OPR-29 when the reactor is pressurized

) above 90 psig with irradiated 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 HPCI subsystem is opereble.

3. If the requirements of Specifi- 3. A simulated automatic initiation cation 3.5.0 cannot be met, an which opens all pilot valves orderly shutdown shall be initi- shall be performed each re-ated and the reactor pressure fueling outage, shall be reduced to 90 psig

) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 4. Uti is determined that two valves f the automatic pressure reu ubsystem are inoperable, the HPCI shall be demonstrated to be operable immediately.

)

E. Reactor Core Isolation Cooling System E. Reactor Core Isolation Cooling System

1. The RCIC system will be operable Surveillance of the RCIC system shall ssure is be performed as specified below with

) whenever greater thanthe 150reactor psit r'.id fuel is the following limitations. For item in the reactor vessel. 4.5.E.3, the plant is allowed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in which to successfully complete the

2. During startup following a refuel test once reactor vessel pressure is outage or an outage in which work adequate to perform each test. In was performed that directly affects addition, the testing required by item

' the RCIC system cperability, if the 4.5.E.3.a shall be completed prior to testing requirements of 4.5.E.3 exceeding 325 psig reactor vessel cannot be met, continued reactor pressure. If RCIC is made inoperable startup is not permitted. The to perform overspeed testing, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months RCIC system shall be declared is allowed to complete the tests before inoperable, and the provisions of exceeding 325 psig.

' Specification 3.5.E.4 shall be implemented. Item Frequency

3. Except for the limitations of 1. Ydive Position Every 31 days 3.5.E.2, if the RCIC system is made or found to be inoperable. 2. Flow Rate Test - Every 92 days

' continued reactor operation is RCIC Pump shall permissible only during the suc- deliver at least ceeding 14 days unless such sys- l 400 gpm against tem is sooner made operable, a system head provided that during such 14 days l corresponding to the HPCI system is operable. a reactor sessel

' Otherwise, the provisions of pressure of > 1150 Specification 3.5.E.4 shall be psig when steam is implemented.

3.5/4.5-7 Amendment No. 130 t

O QUAD ClllES DPR 29

4. The pressure switches which mon-O itor the discharge lines and the discharge of the fill system pump to ensure that they are full shall be functionally tested every month and cali-brated every 3 months. The

.O pressure switches shall be set to alarm at a decreasing pres-sure of > 40 psig and an in-creasing ~ pressure of < 90 psig.

H. Condensate Pump Room flood Protection H. Condensete Pump Room flood Protection 0

1. The systems installed to prevent 1. The following surveillance re-or mitigate the consequences of quirements shall be observed to flooding of the condensate pump assure that the condensate pump room shall be operable prior to room flood protection is oper-startup of the reactor, able.

O

a. The piping and electrical penetrations, bulkhead doort, and subinarine doors for the vaults containing the RiiR service water pumps O and diesel generator cooling pumps shall be checked during each operati cycle lbypressurizingto 5+$

psig and checking f lenks using a soap bubble O solution. The criteria for acceptance shall be no visible leakage through the soap bubble solution.

b. During each operating cycle, O the following flood protection level switches shall be functionally tested to give the following control room alarms:

O- 1) turbine building equipment drain sump high level.

2) vault high level

'O l 3.5/4.5-11 Amendment No. 114 lO 1

.-. . - - - - _ - . ._ .- _-- .= _ - - . -- _ _ _ _

O QUAD-CITIES OPR-29 !

1. Average Planar LHGR I. Average Planar LHGR During steady-state power operation. Daily during steady-state operation the average linear heat generation above 25% rated thermal power, the rate (APLHGR) of all the rods in any average planar LHGR shall be deter-fuel assembly, as a function of aver- mined, age planar exposure, at any axial lo-O cation, shall not exceed the maximum average planar LHGR specified in the CORE OPERAilNG LIMITS REPORT. If at any time during operation it is deter-mined by normal surveillance that the limiting vaive for APLHGR is being i O exceeded, action shall be initiated ;

within IE minutes to restore operation i to within the prescribed 4 . If the APLHGR is not return to ithin i the prescribed limits wi d hours, the reactor shall be brought to the O cold shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . i Surveillance and corresponding action shall continue until reactor operation is within the proceribed limits.

J. Local LHGR J. Local LHGR During steady-state power operation, Daily during steady-state power the linear heat generation rate operation above 25% of rated thermal (LHGR) of any rod in any fuel assem- power, the local LHGR shall be bly at any axial location shall not determined, exceed the maximum allowable LHGR O specified in the CORE OPERATING LIMITS REPORT. If at any time during opera- l tion it is determined by normal sur-veillance that the limiting valut for LHGR is being exceeded, action shall be initiated within 15 minutes to O restore operation to within the prescribed limits. If the LHGR is r.ot 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 .

O surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

I o

3.5/4.5-13 Amendment No. 120

'O

O QUAD-CITIES DPR-29 3.5 LIH111NG CONDITIONS FOR OPERATION BASES O

A. Core Spray and LPCI Mode of the RHR System Thi5 specif atWaTsNeM76i@D~ttPemergo6Niiol67 cap bility is availabl whenever irradiated fuel is in the reactor vessel.

vm Based on 90 the I M" N W , W h g ical methods described in Genera Electric Topical Report NEDC-31345P ToTetettifi'f'syttWrTrdvitr tfMhth4ttobitnd-ttetheMoYe'toMrsti ate the energy associated with theloss-of-coolantaccident,tolimLt 1culated fuel cladding temperature t en,than 2200'F, to at re that core geometry remains intact, to lin clq1) ding metal-water reaction to less than 1%, and to n

' limit the calcu del local metal-water reaction to less than 17%.

wr

( The limiting conditions of operation in Specifications 3.5.A.1 through 3.5.A.6 specify the combinations of operable subsystems to assure the availability of the minimum cooling systems noted above. Under these limiting conditions of operation, increased surveillance testing of the e remaining ECCS systems provides assurance that adequate cooling of the O

f core will be provided curing a loss 4of-coolant accident.

Core spray distribution has been shown, in full-scale tests of systems similar in design to that of Quad-Cities 1 and 2, to exceed the minimum requirements by at least 25%. In addition, cooling effectiveness has g been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to 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 fallen to 90 psig, g The LPCI mode of the RHR system is designed to provide emergency accident. This system functions in combination with the core spray

([coolingtothecorebyfloodingintheeventofaloss-ofco system to prevent excessive fuel cladding temperature. The LPCI mode oftheRHRsystemincombinationwiththecorespraysubsystemprovides adequate cooling {or break areas of approximately 0.05 f t up to and k/

O '.

including 4.26 f t , the latter being the double ended recirculation line break with the equalizer line between the recirculation loops closed without assistance from the high pressure emergency core cooling 11> systems.

.A__-t_A,g - h t W V L(_A.L'b The allowable repair times are established so that the average risk g 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 O

3.5/4.5-15 Amendment No 114 0 -

O QUAD-CITIES DPR-29 -

O' 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 tw; 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 maintair.s the basic risk considering single failures should be less than 30 days, O- and this specification is within this period, for multiple fal. lures, a >

shorter interval is specified; to improve the assurance that the- '

remaining systems will function, a daily tes srcalledvf Although it is recognized that the information given Reference 1 vides a quantitative method to estimate allowable re 4rAlmu@ ack of operating data to support the analytical approach prevents complete g acceptance of-this method at this time. Therefore, the times stated in

-the specific ttems were established with due regard to judgment.

Should one core spray subsystem become inoperable, the remaining core spray sebsystem and the entire LPCI mode of the RHR system are available should the need for core cooling arise. To assure that the O- remaining core spray and the LPCI mode of the RHR system are available, they are demonstrated to be operable immeDately. This demonrstration includes a manual initiation of the pumps and associated valves. Based on judgments of the' reliability of the remaining systems, i.e., the core spray and LPCI, a 7-day repair period was obtained.

O "

Should the loss of one RHR pump occur, a nearly full complement of cure and containment cooling equipment is available. Three 2HR 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, which will be demonstrated to be operable, a 30-day repair period is justified. If the LPCI mode of the RHR system is not g 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.

B. RliR Service Water 9 '

. The containment cooling mode of the RHR system is provided to remove heat energy from tM 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).

O O -

.O 3.5/4.5-16 Amendment No. 114

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

I O i QUAD-CITIES !

DPR 29 The Containment Cooling mode of the RHR System consists of two loops.

Each loop consists of I Heat Exchanger, 2 RHR Pumps, and the associated O valves, piping, electrical equipment, and instrumentation. The "A" loop en each unit contains 2 RHR fervice Water Pumps. Until November 1, 1989, the "B" loop on each unit may utilize the "C ' and "D" RHR Ser-vice Water Pumps from Unit I via a cross-tie line. After November 1, 1989, each "B" loop will contain 2 RHR Service Water Pumps. Either set of equipment is capable of performing the containment cooling O . 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 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 RHR system leaves one remaining system to perform the containment cooling func-O tion. The operable system is demonstrated to be operable each day ,

when the above condition occurs. Based on the fact that when one

loop of the containment cooling mode of the RHR system becomes inoperable, only one system remains, which is tested daily, a 7-day repair period was specified.

O C. High-Pressure Coolant Injection The high pressure coolant injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI mode of the RHR system ar core spray subsystems can protect the core.

O The HPCI meets this requirement without the use of of fsite electrical power. For the pipe breaks for which the HPCI is intended to function, the core never uncovers and is continuously 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 HPCI as part of the isulation cooling system.

O Automatic Pressure Relief The relief valves of the automatic pressure relief subsystem are a backup to the HPCI subsystem. They enable the core spray subsystem and LPCI mode of the RHR system to provide protection against the small O pipe break in the event of HPCI failure by depressurizing the reactor vessel rapidly enough to actuate the core sp t4 stem and LPCI mode of the RHR system. The core spray subsyste and th LPCI mode of the RHR system provide sufficient flow of coola tM4 mit fuel cladding temperatures to less than 2200'F, to assure that core geometry remains intact, to limit the core wide clad metal-water reaction to less than O 1%, and to limit the calculated local metal-water reaction to less than 17L O

3.5/4.5-17 Amendment No.119 O

._. ~. . -_ __ ____ _-_ ._

O QUAD-CITIES i OPR-29 i l

Analyses have shown that only four of the five valves in the automatic 0 depressurization system are required to operate. Loss of one of the relief valves does not significantly apapi)itynthpreforegorttipue affect the pressure plLoging@

-ope {at,1 4%jsgcqcal,aN appropriate MAP!HGR reductio actor is applied to. assur #proQd tcolnri1Taffce

th the 2200'T PCT limit. fpbrPthan'onPrehtf-Tal s*gnWit^antiyWduteHhd ressure Pet-isha9abiRtf-of4fte ADS: thus, O a 7 day repair period is specified with the HPCI available, and a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months repair period with the HPCI unavailable.

E. RCIC The RCIC system is provided to supply continuous makeup wdter to the O 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 suf ficient 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 O initiated at any time.

The HPCI system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable.

Therefore, the specificath . calls for an operability check of the HPCI system should the RCIC system be fcund to be inoperable.

O F. Emergency Cooling Availability The purpose of Specification 3.5.F is ta assure a minimum of core cooling equipment is available at all times. If, for example, one core spray were out of service and the diesel which powered the opposite O core spray were out of service, only two RHR 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 O . . - specification provides that should this occur, no work will be performed on the primary system w ch could lead to draining the vessel. This work would inc e ohqn certain control rod drive components and recirculatio sie . hus, the specification precludes the events which cou e u re core cooling. Specification 3.9 must also be consulted to determine other requirements for the

,O diesel generators.

Quad-Cities Units 1 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 ther

O intended function considering the simultaneous operation of both units.

l- 3.5/4.5-18 Amendment No. 114 LO _

O QUAD CITIES DPR-29 I

These technical specifications contain only a single reference to the O 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 i diesel (1/2 diesel) which is sharet between Units I and 2 would have to be operable for continuing Unit 1 operation even if Unit 2 we're in a .

cold shutdown condition and needed no diesel power. l O l Specification 3.5.F.3 provides that should this occur, no work will be l performed which could preclude 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 O resulting from the work will not result in uncovering the reactor core. Thus, this specification assures adequate core cooling. l Specification 3,9 must be consulted to determine other requirements for the diesel generator.

G. Maintenance of Filled Discharge Pipe 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

'O time emergency cooling was required, the systems would still perform their deb 1gn 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.

O fpecification 3.5.F.4 provides assurance that an adequate supply of coolant water is immediately availab!e to the low pretsure 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.

Oa, H. Condensate Pump Room Flood Protection See Specification 3.5.H I. Average Planar LHGR O This specification assures that the peak cladding temperature following thepostglateddesign-basisIso9-coolantaccidentwillnotexceed the 2200 F limit specified ir the O CFR 50, Appendix K considering the postulated effects of fuel pe ensification.

O 3.5/4.5-19 Amendment No. 114 g

b .

QUAD-ClTIES DPR-29 The peak cladding temperature following a post sted-1,op wficDolan !

y accident is pdm ly a function of the averag heat generation rate f '

alltherodfofa uel assembly at any axial 1 d.tinunNHnl , i egender. 6egm8ntontherod-to-rodpowerdistributionwithi[an3 l sembiv. Sinceexpectedlocalvariationsinpowerdistribution#1 thin '

a Get-il sembly affect the calculated peak cladding temperature by less l than 120'F relative to the peak temperature for a typical fuel. design, ;

y the limit on the average planar LHGR sqic qt 4 e a s,nr e n g a

on calculations employing the models described in Reference 2.

5 GL 1 R ased l' The Average Planar Linear Heat Generation Rate (APLHGR) also serves a ;

) econdary function which is to assure fuel rod mechanical integrity.

NM ,

J. Local LHGR This specification assures that the maximum linear heat generation rate in any rod is less than the design ~1inear heat generation rate specified in the CORE OPERATING LIMITS REPORT even if fuel pellet densification 3 is postulated. The powet spike penalty is discussed in Reference 2 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 LHGR due to power spiking. No~ penalty is required in Specification 3.5.L because it has been accounted for in y the reload traasient analyses by increasing the calculated peak LHGR by 2.2%.

K. Minimum Critical Power Ratio (MCPR) '

The steady state values for MCPR specified in the CORE OPERATING LIMITS n" REPORT were selected to provide margin to accommodate transients and uncertainties in monitoring the core operating state as well as uncer-tainties in the critical power correlation itself. These values also assure that operation will be such that the initial condition assumod for the LOCA analysis plus two percent for uncertaint) is satisfied.

For any of the special set of transients or disturbances caused by single 3' 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 Spacification 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 value of MCPR stated-in the CORE OPERATING LIMITS REPORT l g for the limiting condition of operation bounds the initial value of MCPR assumed to exist prior to the initiation of the transients. This-initial condition, which is used in the transient analyses, will preclude violation of the fuel cladding integrity safety limit.

Assumptions and methods used in calculating the required steady state MCPR limit for each reload cycle are documented in References ? and 4, g' The results apply with increased conservatism while operating with MCPRs greater than specified.

3 5/4.5-20 Amendment No. 120 0-

W O ' dt:se) ponerators. All Cf these systems have been steed to perferis their

'. 1; tended functlen constdortng the simultaneous operation of both units.

These technical spectf testions contain only a single reference to the -

operebility and survet11ance reestrements for the shared safety related ~

foetures'st each p)eht. The level of operablitty for one unit must be maintained ineependently of the status of the other. for esamete 4 O diesel O/3 diesel) i41ch is shared between Units 1 and I would have to be operable for contt% stag tpilt 1 operetton even if Unit I were in a cold utdo.ei consin and needed n. diciei ,o.or.

l-

$pesificatten 1.5.F.3 provides that should this occur, no work will be performed which could preclude adequate peersency coeting capahttity being avattable, teork ts prohibited unless it is in accordance with $pecified procedures estich iteit the period that the contre) red drive houstne is O- open and assures that the worst possible loss of coolant resulting from the work wt11 not result in uncovering the reacter core. Thus, this specification assures adequate core coeling. Specification 3.g ansst be consulted to esterialne other requirements for the dietet generator.

G, . Maintenance of Fl11ed 01scharge Pipe ff the discharge piping of the core spray tPCI mode of the RHe. HPCI.

.O and RCIC are not f111ed, a water hapsner can develop in this piping, threatening system damage and thus the availablitty of emergency cooling systems unen the pump and/or pues are started. An analysts has been does !

antch shows that if a water hanser were to occur at the time emergency I cooling was reeutred, the systems would sit 11 perform their desten j function. However to ministre damage to the discharge systems and to ensure added margin in the operation of these systems, this technical _i l

specification requires the discharge itnes to be f t11ed whenever the

.n

'v system is in an operable conditten.

Specifica' ton 3.5.r.4 provides assurance that-an adeguate supply of coolant wter is inmediately available to the low pressure core cooling i 1

systems <nd that the core ut11 remain covered in the event of a loss-of-coolant accident while the reactor is depressurized with the head l removed.

iO H. Condensate em noo. riood protectiun l see $pecification 3.$.H

1. overage Planar tHGA l-This spectf tcation assue ss that the peak c14dding temperature following a postulated design-basis loss of-coolant acetosat wt)) not esteed the

% v 2290'y limit spectf ted in 10 Cf t $8 appendts K considering the postulated

effects of fuel pellet denstf tcation.

The peak cladding teigerature fc11owing a postulated less-of-coelant accident is primarity a function of the average tHQA of all the rods in a fuel assembly at any antal location and is only setorsarily dependent on >

the rod t'o. rod power distributton within a fuel assembly. Since espected local variations .in power distributton within a fuel assembly af fect the

.n calculated peak cladett:g temperature by less than 120'r relative to the

.V' peak temperature for a typical fuel design. the limit on the average planar UeGA is sufficient to assure that calculated temperatures are below the it CFR $8 Appendts K itett.

MW The maatsnam avet age planar tHGA's'spectf ted in the (Det OPERATING t! NIT $

REPORT are based on calculations eay1oying the models described in ')

hv Reference 2. Power operatton with (HGR's at or below those spectf ted in the CORE OPLRAf teG LIMITS etPORT assures that the peak cladding

}

temperature following a postulated loss-of-coolant accident will not esteed the 2200'r limit, fhese values represent limits for operation to ensure conformance with 10 cfR $0 and appendts K only if they are more 11stling than other design parameters.

The maalaus average pl.nar LHGA's specif ted in the C0tt OPERATING t!MITS REPORT snan 2:00at. higher esposures

r. Howe,er, tne ma i.u. resuit in a peak average cladding pianar (nGR'stangerature are of less O

OY 1.$/4,$.1) amendnent so. II6 1977H

+ .:

- - , ., e,-' . - . . . - - - ~ - .,r- - --.~.~-a-r +.-~~-,~-n---- - - ~ , - , - - - - , - . - n , - - , - ~~,s.----

W M y l n specified 12 the CTE OPERAf!he LIMlf 5 RIPoet as it tts because conformance !

l

"' . calc 31stions have not been performed te j:stify Cperation at LMGA's ta encess cf I these shown.

J. Local this spectittatten assures that the manianae itnear heat-generation rate

in any rod is less tMe the desten linear heat-generatton rate specified in the (Det OPERA 11mG LIMITS REPWT even if fuel pellet denstf tcation is postulated. The powr sptke Q penalty it discussed t's esference I and assumes a 1tnearly *Nreasing variation in asial paps between core trottom and top and assures with 951 tonfidence that no more !

than one fuel rod entsees the design (HGa due to power sputag. no penalty is 1 rewtres in spectrication 3.5.t because it has been accounted for in the reload l transtant analyses by increasing the calculated peak LHCR by 2.rt.

K. Minimum Critical Power latio (MCPt)

O The steady state values for MCra specified in the C0At CPitATING t!M175 introtf were l selected to providta margin to accommodate transtents and uncertainties in nionitoring the core operating state as well at uncertainties in the critical power correlation itself. These values also assure that operation will be such that the initial comettion assumed for the LOCA analysis plus two percent for uncertainty is satisfied. for any of the special set of transients or disturbances caused by single operator error or single equipment malfunction, it is required that design

. O.

analyses intitaltred at this steady-state operating Itatt yield a MCPR of not 1 ell than that spectfied in Spectf tcation 1.1.A at any time during the transtent, assuming instrument trip settings gtven in spectf tcation 2.1. For analysts of the thermal consecuences of these transtents. the value of MCPR stated in the CORE OrtRATING t!MIT5 etPott for the Itatting condition of operation bounds the initial value of HCPR assweed to entst prior to the initiation of the transients. This inittal condition, untch ts used in the transtant analyses, will prec166e violation -

of the fuel cladding integrtty safety limit. assumptions and methods used in O calcuistias the autr** st***r state The"cr* 'iait apply results for **ch withr increated 1**8 cic1* are conservatt's docunented in References 2 and 4.

Witle operating utth MCra's greater than spectf ted.

The most Itatting transtants witu respect to MCPR are generally:

a) Rod withdrama) error

'O' b) toad rejection or turbine trip without bypass c) Loss of feedwater heater Several f actors influence etch of the;e transients results in the 1ergest reduction in critical power ratto such as the specific fuel loading. esposure, and fuel type.

The current cycle's reload Itcensing analyses specifies the 11eittog transients for a given esposure increment for each fuel type. The values specified as the Limiting Condition of Operation are conservatively chosen to bound the most restrictive over Q the onttre cycle for each fuel type. ,

The need to adjust the MCPt operating limit as a function of scram time arises f rom the statistical approach used in the implementation of the 00Yu asputer code for analyzing rapid pressurization events.~ cenerts stattstical analyses were performed for p1snt groupings of stattar oesign which considered the statistical variation in several parameters (initial power level. CRD scram insertton time, and model uncertainty). These analyses (which are described further in Reference el produced Q generic Statistica) adjustment f actors which have been applied to plant and cycle specific 00Yu results to yteld operating Ilmits which provide a 951 probability with 951 conf teente that the Ilmiting pressuritation event will not cause MCPR to f all below the fuel cladding integrity safety limit.

O O

ignH/o6tsz .3. 5 / s . s- i s amenavent no.116 O

e m -, % , - ,,.7-- -. er ---.: . -- _. -. <. w -

O QUAD-ClllES DPR-29 The most limiting transients with respect to MCPR are generally:

O !

a) Rod withdrawal error b) Loadrejectionorturbinetripwithoutbypass Loss of feedwater heater ec) w -evw'"

O The MCPR Operating Limit reflects an increase of 0.03 over the most limiting transient to allow continued operation with one feedwater J. A A svP Several factort influence which of these transients results in the O largest reduction in critical power ratio such as the specific fuel loading, exposure, and fuel type. The corrent cycle's reload licensing analyses specifies the limiting transients for a given exposure iMrement for each fuel type. The values specified as the Limiting Condition of Operation are conservatively chosen to bound the most restrictive over the entire cycle for each fuel type.

The need to adjust the MCPR operating limit as a function of scram time arises from the statistical approach used in the implementation of the ODYN computer code for analyzing rapid pressurization events. Generic statistical analyses were performed for plant roupings of similar design which considered the statistical variat on in several parameters O (initial power level, CRD scram insertion time, and model uncertainty). These analyses (which are described further in Reference

4) produced generic Statistical Adjustment Factors which hcve been applied to plant and cycle specific ODYN results to yield operating limits which provide a 95% probability with 95% confidence that the limiting pressurization event will not cause MCPR to fall below the O- fuel cladding integrity safety limit.

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 0 .

the scoop tube positioner for the fluid coupler to move to the maximum speed position.

O O

3.5/4.5-/1 Amendmtnt No. 114 g

OPR-30 j . .

swym/mrvy,a n, ,y ,

As a result of this 95/95 approach, the average 20% insertten scran time must be honttored to assure compliance with the assumed statistical distributton. If the mean value on a cycle cumulative (running average) basis were to exceed a 5%

significance level compared to the distribution assumed in the 00YN statistical b

analyses, the HCPR limit must be increased linearly (as a function of the van 20%

scram time) to a more conservative value which reflects an NRC determined (

uncertainty penalty of 4.41. This penalty is applied to the plant specific 00YN results (i.e. without statistical adjustment) for the limiting single failure pressuri m lon event occurring at the limiting point in the cycle. It is not applied full until the mean of all current cycle 20% scram tirnes reaches the 0.90 g skcs v- e of Specification 3.3.C.). In practice, however, the requirements of 3.3.C.1 would most likely be reached (i.e. Individual data set average > .90 secs) and the required actions taken (3.3.C.2) well before the running average exceeds 0.90 sect.

The 51 significance level is defined in Reference 4 as: /

9 D

TB=p+1.65(Nj/{Ng)l/2o 11 )

where:

y . Mean value for statistical scram ttine distributton to 20%

inserted f

y o . standard deviation of above distributton N)

- number of rods tested at BOC (all operable rods) n INg total number of operable rods tested in the current cycle 11 D

The value for tg used in Spectf tcation 3.5.k is specified in the CORE OPERATlHG LIMITS REPORT and is conservative for the following reasons:

a) For simplicity in formulating and implementing the L'CO, a n

g conservative value for I Ng of 708 (i.e. 4: 177) was used.

1.)

This represents one full core data set at BOC plus 6 half core data sets. At the maximum frequency allowed by Specification

.- 4.3.C.2 (16 week intervals) this is equivalent to 24 orerating mor.t h s . That is, a cycle length was assumed which is longer than f any past or contemplated refueling interval and the number of rods K D tested was maximized in order to simplify and conservatively f reduce the criteria for the scram time at which HCPR penalization is necessary.

b) The values of p and o were also chosen conservatively based on g the dropout of the position 39 RPIS switch, since pos. 38.4 15 the precise point at which 20% insertion is reached. As a result Specification 3.5.k initiates the linear HCPR penalty at a slightly lower value tave. This also produces the full 4.4%

penalty at 0.86 secs which would occur sooner than the required value of 0.90 secs.

A ,xJ V^V G'Q w I 1976H 3.5/4.5-14a Amendment No.116 B

n -

P QUAD-ClllES DPR-29 l

References '

D ' W V'~V Vv'N W C > N

1. " SAFER /GESTR-LOCA Loss of Coolant Analysis for QuadCities Nuclear Power Station Units 1 & 2" NEDC-31345P.*

wxwen wVWG

2. " Generic Reload fuel Application," NEDE-24011-P-A**

3 3. I. H. Jacobs and P. W. Harriott, GE Topical Report APED 5736, " Guidelines for Determining Safe Test Intervals and Repair Times for Engineered i Safeguards," April,1969.

4. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors" General Electric Co. Licensing Topical Report NEDO 24154 Vols. 1 J and 11 and NEDE-241$4 Vol. III as supplemented by letter dated September 5, 1980 f rom R.H. Buchholz (GE) to P. S. Check (NRC).

Approved revision at time of plant operation.

- Approved revision number at time reload fuel analyses are performed.

J O

J

~

O O

O 3.5/4.5-22 Amendment No. 114 0

o QUAD CITIES DPR-29 4.5 SURVEILLANCE REQUIREMENTS BASES 0

The testing interva5 for the core and containment cooling systems is based on a quantitative reliability analysis, judgment, and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, the core spray final admission valves do not open until reactor pressure

, has fallen to 350 psig. Thus, during operation, even if high drywell pressure O were simulated, the final valves would not open. In the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable.

The surveillance requirements bases described in this paragraph apply to all core and containment cooling systems except HPCI and RCIC. The systems can be O automatically actuated during a refueling outage and this will be done. To increase the availability of the individual components of the core and containment cooling systems, the components which make up the system, i.e. , instrumentation, pumps, valve operators, etc. , are tested more f requently. The inftrumentation is functionally tested each month. Likewise the pumps and motor-operated valves are also tested each month to essure their operability. The combinotion of a yearly O. simulated automatic actuation test and monthly tests of the pumps and valve operators is deemed to be adequate testing of these systems. With components or subsystems out of service, overall core and containment cooling reliability is maintained by demonstrating the operability of the remaining cooling equipment.

The degree of operability to be demonstrated depends on the nature of the reason for the out-of-service equipment, for routine out-of-service periods caused by O preventative maintenance, etc. , the pump and valve operability checks will be performed to demonstrate operability of the remaining components. However, if a failure, design deficiency, etc. , causes the out-of-service period, then the demonstration of operability should be thorough enough to assure that a $1milar problem does obgxist on the remaining compcnents, for example, if an cut-of-service peri is3aused by f ailure of a pump to deliver rated capacity due to a O design defic nty7 the other pumps of this type might be subjected to a flow rate test in addition to the operability checks, The surveillanco requirements bases described in this paragraph apply only to the RCIC and HPCI systems. With a cooling system out of service, overall core and g' containment cooling reliability is maintained by verifying the operability of the remaining cooling systems. The verification of operability, as used in this context, for the remaining cooling systems means to administrative 1y check by examining logs or other information to verify that the remaining systems are not out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the remaining q systems. However, if a failure, design deficiency, etc. , causes the out-of-

' service period, then the verification of operability should be thorough enough to assure that a similar problem does not exist on the remaining systems, for example, if an out-of-:,ervice period is caused by f ailure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test. Following a refueling outage or an outage in g' which work was perforc'ed that directly affects system operability, the HPCI and RCIC pumps are flo.' rate tested prior to exceeding 325 psig and again at rated reactor steam pressure. This combination of testing provides adequate assurance of pump performance throughout the range of reactor pressures at which it is 3.5/4.5-23 Amendment No. 130 )

O

O QUAD ClllES DPR-29 required to operate. The low pressut e limit is selected to allow testing at a O point of stable plant operation and also to provide overlap with low pressure ECC systems. A time limit is provided in which to perform the required tests during start-up. This time limit is considered adequate to allow stable plant conditions to be achieved and the required tests to be performed. Flow rate testing of the HPCI and RCIC pumps is also conducted every 92 days at rated reactor pressure to demonstrate system operability in accordance with the LCO provisions and to meet 0 inservice testing requirements for the HPCI system. Applicable valves are tested in acy rdance with the provisions of the inservice testing program. In addition, monthiy checks are made on the position of each manual, power operated or automatic val' e installed in the direct flowpath of the suction or discharge of the pump or turbine that is not locked, sealed, or otherwise secured in position. At each refueling outage, a logic system functional test and a simulated automatic actua-O tion test is performed on the HPCI and RCIC systems. The tests and checks described above are considered adequate to assure system operability.

The verification of the main steam relief valve operability during manual actuation surveillance testing must be made independent of temperatures indicated by thermocouples downstream of the relief valves. It has been found that a O temperature increase may result with the valve still closed. This is due to steam being vented through the pilot valves during the surveillance test. By first opening a turbine bypass valve, anti then observing its closure response during relief valve actuation, positive verification can be made for the relief valve opening and passing steam flow. Closure response of the turbine control valves during relief valve manual actuation would likewise serve as an adequate O verification for the relief valve opening. This test method may be performed over a wide range of reactor pressures greater than 150 psig. Valve operation below 150 psig is limited by the spring tension exhibited by the relief valves.

The surveillance requirements to ensure that the discharge piping of the core spray, LPCI mode of the RHR, HPCI, and RCIC systems is filled provides for a O visual observation that water flows from a high point vent. This ensures that the line is in a full condition.

Instrumentation has been provided on core spray and LPCI mode of RHR to monitor the pressure of water in the discharge piping between the monthly intervals at which the lines are vented and alarm the control room if the pressure is inade-O quate. This instrumentation will be calibrated on the same frequency as the safety system instrumentation and the alarm system tested monthly. This testing ensures that, during the interval between the monthly venting checks, the statyse of the discharge piping is monitored on a continuous basis. An alarm point of 40 psig for the low pressure of the fill system has been chosen because, due to

g elevations of piping within the plant, 39 psig is required to keep the lines full.

TheshutoffheadofthefillsystempumpsislessthanSD.psig,aedtherefor%j f of will not defeat the low pressure cooling 100 psig as shown in Table 3.2-2. A margin of 10 psig % ptovided by the n pomp discharge pressure Giterloc(igh pressure alarm point of 90 psig.

HPCI and RCIC systems normally take a suction from the Contaminated Condensate 3 Storage Tanks (CCSTs). The level in the CCST'3 is maintained at or 5ove 3.5/4.5-24 Amendment No.130 g

3 QUAD-CITIES DPR-29 level (a) indicates water in the condenser pit from either the hotwell or the c rculating water system. Level (b) is above the hotwell ca racity and indicates

~

S a probable circulating water failure.

Should the switches at levels (a) and (b) fail or the operator fails to trip the circulating water pumps on alarm at level (b), the actuation of either level switch pair at level (c) shall trip the circulating water pumps automatically and 9 alarm in the control room. These redundant level switch pairs at level (c) are designed and installed to IEEE-279, " Criteria for Nuclear Power Plant Protection Systems." As the circulating water pumps are tripped, either manually or automatically at level (c) of 5 feet, the maximum water level reached in the condenser pit due to pumping will be at elevation 568 feet 6 inches elevation (10 feet above condenser pit floor elevation 558 feet 6 inches; 5 feet plus an

.9 additional 5 feet attributed to pump coastdown),

in order to prevent the RHR service water pump motors and diesel generator cooling water pump motors from overheating a vault cooler is supplie( each pump. Each vault cooler is designed to raintain the vault at a maxfgum 5'F temperature during operation of its respective pump. For example, iMit el 9 generator cooling water pump 1/2-3903 starts, its cooler also starts and maintains the vault at 105'F by removing heat supplied to the vault by the motor of pump 1/2-3903 If, at the same time that pump 1/2-3903 is in operation, RHR service water pu p IC starts, its cooler will also start and compensate for the added heat suppi ed to the vault by the IC pump motor keeping the vault at 105 F.

3 Each of the coolers is supplied with cooling water from its respective pump's discharge line. After the water has been passed through the cooler it returns to its respective pump's suction line. The cooling water quantitv needed for each cooler is approximately 1% to 5% of the design flow of the r , a do that the recirculation of this small amount of heated water will not e fect pump or cooler operation.

J Operation of the fans and coolers is required during shutdown and thus additional surveillance is not required.

Verification that access doors to each vault are closed following entrance by personnel is covered by station operating procedures.

U The LHGR shall be checked daily to determine if fuel burnup or control rod move-ment has caused changes in power distribution. Since changes due to burnup are slow and only a few control rods are moved daily, a daily check of power distri-bution is adequate.

3 Average Planar LHGR At core thermal power levels less than or equal to 25%, operating plant experience nti hermal hydraulic analyses indicate that the resulting average planar LHG is elow the maximum average planar LHGR by a considerable margin; therefore, efa ation of the average planar LHGR below this power level is not J necessary. The daily requirement for calculating average planar LHGR above 25%

rated thermal power is sufficient, since power distribution shifts are slow when there have not been significant power or control rod changes.

3.5/4.5-26 Amendment ho. 130

O QUAD-CITIES DPR-29 1 O Local LHGR The LHGR as a function of core height shall be checked daily d uring reactor operation at greater than or equal to 25% gwer to determine if f 3

control rod movement has caused changes in power distributlon uel burnup or value rod is below pattern precluded bythermal 25% rated considereo,e paer. margin when .

employing A limiting LHGR any permi s>l51e control Minimum Critical Power Ratio (MCPR)

I At core thermal power levels less than or equal to 25%

3 operating be very small.at minimum recirculation pump , the reactorand speed will be the moderator void content will this point, operating plant experience and thermal hydrauliFor a at that the margin. resulting MCPR value is in excess of crequirements analysis indicateb With this low void content y a considerable

) only place operation in a more conse,rvative mode relative to MCPRany i The dail utrement sufficidr4AsinceJo r distribution shifts are very 3.or calculating MCPR abov been sigrificant power or control rod changes. In addition, the K< when there have not margir, for flow increases from low flows.specified rovides in the CORE OP

)

i 3.5/4.5-27 Amendment No. 130

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ML20085C760

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