Text

Proposed Tech Specs 3.2/4.2 Re Protective Instrumentation & 3.9/4.9 Concerning Auxiliary Electrical Sys
	ML20079G104
Person / Time
Site:	Quad Cities
Issue date:	09/30/1991
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20079G094	List: ML20079G092; ML20079G104;
References
	NUDOCS 9110080410
	Download: ML20079G104 (156)
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PROPOSED TECH SPEC TS 3.2/4.2

' PROTECTIVE INSTRUMENTATION" -

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 3.2/4.2 PROTECTIVE INSTRUMENTATION SPECIFICATIONS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS A. Primary Containment Isolation A. Primary Containment Isolation Funotions Functions The isolation actuation instru- 1. Instrumentation and logic mentation shown in Table 3.2-1 systems shall be functio-shall be OPERABLE with their nallj tested, calibrated, setpoints set consistent with the and checked as indicated in values shown in the Trip Level Table 4.2-1.

Settings coluntn.

2. LOGIC SYSTEM FUNCTIONAL APPLIQARILITY: TESTS and simulated automa-tic operation of all chan-As shown in Table 3.2-1. nels shall be performed at least once each REFUELING ACTION: OUTAGE.

1. With an isolation actuation instrumentation- setpoint less conservative than the value shown in the Trip Level Settings column of Table 3.2-1, declare the CHANNEL inoperable until the CHANNEL is restored to OPERABLE status with its trip- setpoint adjusted i consistent with the Trip 1 Level Settings value.

2. With the number of OPERABLE CHANNELS less than required by the Minimum OPERABLE CHANNELS per TRIP SYSTEM requirement for ona TRIP SYSTEM:

a. If placing the in-operable CHANNEL (s) in the tripped condition would cause an isola-tion, the inoperable CHANNEL (s) shall be restored to OPERABLE 3.2/4.2-1

- . . - .- - . .- - - - . . ~ . . - . . . - - . - - - . - . - . - . - . - - - - - . - . . - -

QUAD CITIES UNITS 1 & 2 -

DPR-29 & DPR-30 status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or the ACTION required by Table 3.2-1 for the affected Trip Function shall be taken.

-b," If placing the in-in operable the tripped CHANNEL nondit (s) ion

, would- not cause- an isolation, the inop-erable CHANNEL (s) and/

Jor that TRIP SYSTEM ,

shall be pit:ed in the tripped condition withint

1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common to RPS Instrumen-tation; snd

2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not i connon to RPS i Instrumentation.

3. With the number of OPERABLE CHANNELS lasa than required by the Minimum OPERABLE CHANNELS per' TRIP SYSTEM requirement for both TRIP -

SYSTEMS,- place one - TRIP '

SYSTEM (with the most' inoperable CHANNELS)lin the i tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and take tha ACTION ,

required by Table 3.2-1. ,

The TRIP SYSTEM need not be placed in .the tripped condition if this would cause the ' isolation to occur.

3.2/4.2-2 b-wr- , * - w w - c , v- =+,,ep .we-

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

B. Core and Containment Cooling B. Core and Containment Cooling systems Initiation and Control Systems Initiation and Control j The core and containment cooling 1. Instrumentation and logic systems actuation instrumentation systems shall be functio-shown in Table 3.2-2 shall be nally tested, calibrated, l OPERABLE with their setpoints set and checked as indicated in i consistent with the values shown Table 4.2-1.

in the Trip Level Settings column. 2. IhGIC SYSTEM FUNCTIONAL TESTS and simulated automa-APPLICABILITYt tic operation of all chan-nels shall be performed at As shown in Table 3.2-2. least once each REFUELING OUTAGE.

ACTIONr

1. With a core or containment cooling actuation instru-mentation CHANNEL setpoint less conservative than the value shown in the Trip Level Settings column of Tabla 3.2-2, declare the CHANNEL inoperable until the CHANNEL is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Level Settings value.

2. With one or more core or containment cooling actuation instrumentation CHANNELS inoperable, take the ACTION required by Table 3.2-2.

3. With either Automatic Pressure Relief TRIP SYSTEM inoperable:

a. Continued reactor operation is permis-sible only durtng the succeeding 7 days, provided that during such 7 days the HPCI system is OPERABLE, or 3.2/4.2-3

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

b. With HPCI inoperable, continued reactor operation is permis-sible only dur;.ng the succeeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

c.' If the requirements of ACTIONS 3.2.B.3.a or 3.2.B.3.b cannot be met, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce REAC-TOR VESSEL PRESSURE to 5 150 poig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. Control Rod Block Actuation C. Control Rod Block Actuation The control rod block instrumen- Instrumentation and logic systems tation shown in Table 3.2-3 shall shall be functionally tested, be OPERABLE with their setpoints calibrated, and checked as set consistent with the values indicated in Tabla 4.2-1.

shown in the Trip Level settings

. column.

APPLICABILITYt As shown in Table 3.2-3.

A.CTIOlu

1. With a control rod block instrumentation CHANNEL setpoint less conservative than the value shown in the Trip Level Settings column of Table 3.2-3, declare the CHANNEL inoperable until the CHANNEL is restored to OPERABLE status wit.h its trip setpoint L1)usted consistent with the Trip Level Setting value.

2. With the number of OPERABLE CHANNELS less than required by the Minimum OPERABLE CHANNELS per Trip Function 3.2/4.2-4

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 requirement, take the ACTION required by Table 3.2-3.

D. Refueling Floor Radiation D. Refueling Floor Radiation Monitors Monitors Two refueling floor radiation The two refueling floor radiation monitors shall be OPERABLE with a monitors shall be functionally setpoint of 5 100 mR/hr. tested, calibrated, and checked as indicated in Table 4.2-1.

AP.P11CABILITY1 Reactor building ventilation isolation and standby gas When irradiated fuel or the fuel treatment system initiation shall cask is being handled in the be performed at least cach reactor building, during CORP OPERATING CYCLE.

ALTERATIONS, and during opera-tions with a potential for draining the reactor vessel.

ACTIONI

1. One of tho two refuelina floor radiation monitors may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If the inoperable monitor is not restored to service in 24 hours, the reactor building ventilation system shall ba isolated and the standby gas treatment system operated until repairs are conplate.

2. Upon loss of both refueling floor radiation monitors while in use, the reactor building ventilation system shall be isolated and the standby gas treatment system operated.

3. The provisions of Specifica-tion 3.0.C are not applicable. ,

3.2/4.2-5

k 1

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 E. Postaccident Instrumentation E. Postaccident Instrumentation '

The postaccident -monitoring Postaccident instrumentation instrumentation CHANNELS shown in shall be functionally tested, Table 3.2-4 shall be OPERABLE. calibrated, and checked as indicated in Table 4.2-2.

APPLICABILITY 1

,.; shown in Table 3.2-4.

ACTIONt ;

hith one or more postuccident monitoring J i..St rusonta tion i CHANNELS inortrable, take the ACTION required by Table 3.2-4.

F. Control Room Ventilation System F. Control Room Ventilation System Isolation Isolation ,

Control room ventilation system 1. Surveillance for instru-isolation -instrumentation .nhall mentation whirh ir.itiates be-OPERABLE as follows: isolation of control room ventilation shall be as

1. LIMITING CONDITIONS FOR specified in Table 4.2-1. ,

OPERATION for drywell high prorsure, reactor low water 2. Manual isolation of the level, dryvell high radia- control room ventilation tion and main steamline high system shall be demonstrated flow are as shown in Table once every REFUELING OUTAGE.

-3.2-1.

2. LIMITING CONDITIONS FOR OPERATION for the refueling

- floor radiation monitors are as shown in specifica tion 3.2.D.

3. The toxic gas detection instrumentation shall consist of a chlorine, ammonia, and sulphur dioxide analyzer with each trip i setpoint set att

a. Chlorine concentration s 5 ppm 3.2/4.2-6

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

b. Ammonia concentration s 50 ppm

c. Sulphur dioxide concentration 5 3 ppm

4. High radiation detection instrumentation in tha reactor building ventilation exhaust ducts shall km maintained OPERABLE with a trip setpoint ot' s2 rr/br above background.

hPPLICABILITYt

1. The applicable OPEkATIONAL MODES for the drywell high pressure, reactor low water level, drywell nigh radia-tion, and main steamline high flow isolation instru-mentation are as shown in Table 3.2-1,

2. The applicability for the refueling floor radiation monitors is as shown in Specification 3.2.D.

3. The applicability for toxic gas detection instrumenta-tion is at all times.

4. The applicability for reactor building ventilation exhaust high radiation instrumentation is all OPERATIONAL MODES and while handling irradiated fuel in the reactor building.

ACTION _t_

1. The ACTIONS for the drywell high pressure, reactor low water level, drywell high radiation, and main steam-line high flow isolation 3.2/4.2-7

l QUAD CITIES UNITS 1 & 2 j DPR-29 & DPR-30 instrumentation are as shown ;

in Table 3.2-1.

2. The ACTIONS for the refuelf.ng floor radiation 1 monitors are as shown in specification 3.2.D.

3. The ACTIONS for the reactor building ventilation exhaust high tradiation and toxic gas detection instrumentation are au follows

a. With one CHANNEL of the reactor building ventilation exhaust high radiation detec- +

tion instrumentation inoperable, restore the inoperable CHANNEL t to . OPERABLE status ,

within 7 days or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , manually iso- .

late the control room ventilation system.

b. With more than one CHANNEL of the reactor building ventilation exhaust high radiation detection instrumenta- -

tion inoprrable, e with-in i honr restore at least ona CHANNEL to OPERABLE status, or manvally. isolate the control room ventila-tion system.

c. With any toxic gas monitor CHANNEL inop- '

erable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore all CHANNELS to OPERABLE status or manually isolate the control room ventila-tion system.

3.2/4.2-8 I

L

n. . .- - . ..- - - _.._. _ _ . - - -.--_,,_.-

QUAD CITIES UNITS 1 4 2 DPR-29 & DPR-30

4. The provisions of Specifica-tion 3.0.C are not applicable.

G. Steam Jet Air Ejector (SJAE) G. Steam Jet Air Ejector (SJAE)

Radiation Monitors Radiation Monitors At least one SJAE radiation 1. The SJAE radiation monitors monitor shall be operable with shall be functionally alarm / trip setpoints determined tested, calibrated, and in accordance with the ODCM. checked as indicated in Table 4.2-1.

APPL 7CABILITYi

2. The SJAE radiation monitors During SJAE vparation, shall be tietermined OPERABLE by perfotmance of a SOURCE ACTIONt CHECK at least each REFUELING OUTAGE.

With no SJAE radiation monitor OPERABLE, -gases from the main condenser off gas system may be released to the environment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least one chimney monitor is OPERABLE; otherwise, be in at least HOT STANDBY in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.2/4.2-9

QUAD CITIES UNITE 1&2 DPR-29 & DPR-30 TABLE 3.2-1 INSTRUKENTATIOH Il{hI JE1TI ATE $ PRIMARY CQFJAINHENT ISOLATION f,VJ[C1]QEE Valve Minimum Croups OPE RAHLE Operat ed CHANNdLS Applicable Trip By Trip per TRIP Trip Level OPERATIONAL Tunction Function (1) SYS*EM (a)(h) Setting MODES ACTION

1. Reactor Low 2, 3 2 > 144 inches above 1, 2, 3 11 Water Level (d) top of active fuel (g)

2. Reactor Low Low 1 2 2 84 inches above 1, 2, 3 11 Water Level top of active fuel (g)

3. Drywell High 2 2 5 2.5 poig (c) 1, 2 , 3 11 Pressure (d)

4. Drywell High 2 1 < 100 R/hr 1, 2, 3 16 Radiation (d)

5. Main. Steam Line 1 8 5 140% of rated 1, 2 (j) 12 High riow (d) steam flow 1 (j) 12 0

6. Main Steam Line 1 8 $ 200 F 1, 2 (j) 12 Tunnel High 3 (j) 12 Temperature

7. Main Steam Line 1 2 s 15 X normal rated 1, 2 (j) 12 Tunnel High power background 3 (j) 12 Radiation (e) (witbout hydrogen addition)

8. Main Steam Line 1 2 2 825 poig 1 15 Low Pressure (b)

9. RCIC Steam Line 5 1 s 300% of rated 1, 2, 3 13 High Flow steam flow (f)

10. RCIC Turbino Area 5 2 5 1700 r 1, 2, 3 13 High Temperature

11. RCIC Steam Supply 5 2 2 60 poig 1, 2, 3 13 Line Low Pressure

12. HPCI Steam Line 4 1 $ 300% of rated 1, 2 , 3 14 High Flow steam flow (f) 3.2/4.2-10

QUAD CITIES UNITS 1 & 2 DPR.29 & DPR.30 TABLE 3.2 1 (continued)

INSTRUKENTATION IBhT INITIATES PRIKARY QONTAINM[HI ISOLATION FUNCTION 1 Valve Minimum Groups OPERABLE Operated CllANNELS Applicable Trip by T*4m per TRIP Trip Level OPERATIONAL Function F6 net bn (1) SYSTEM (a)(h) setting MODES ACTION

13. HPCI Area High 4 2 s 1700F 1,2,3 14 Temperatura

14. HPCI Steam 4 2 2 100 psig 1,2,3 14 supply Line .

Low Pressure 3.2/4.2 11

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

i QUAD CITIES UNITS 1 & 2 DPR'-29 & DPR-30 TABLE 3.2-1 (Continued) '

INSTRU}iENTATION TilAT INITIATES PRIMARY CONTAINEENT ISOLATION FUNCTIONS ACTIONS ACTION 11 - De in at least il0T SilOTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SilUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 12 - Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least il0T SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SilUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 13 - Close the isolation valves in the RCIC system within one hour.

ACTION 14 - Close the isolation valves in the IIPCI subsystem within one hour.

ACTION 15 - Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 16 - Close the affected system isolation valves within one hour and declare the affected system inoperabic.

4 l 3.2/4.2-12 l

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l QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 3.2-1 (Continued)

IEEIRW fliTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION FUNCTIONS IABLE NOTATIQHS (a) Two TRIP SYSTEMS shall bo OPERABLE in the applicablo OPERATIONAL MODES for the specified trip function unless otherwise allowed by the ACTION provisions.

(b) The isolation trip signal is bypassed unen the modo switch is not in RUN.

(c) Need not be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(d) The instrumentation also isolates the control room ventilation system.

(e) This signal also trips the mechanical vacuum pump and closes the SJAE suction valves.

(f) Includes a time delay of 3 5 t 5 9 seconds.

(g) Top of active fuoi is defined as 360" abovo vessel zero for all water lovels used in the LOCA analysos. (Joe Bases 3.2)

(h) A CHANNEL may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required survoillance without placing the TRIP SYSTEM in the tripped condition provided at least one other OPERABLE CHANNEL in the same TRIP SYSTEM is monitoring that paramotor.

(i) The valves associated with each valve group are designated in Table 3.7-1.

(j) Required OPERABLE with any main steam lino not isolated.

3.2/4.2-13 b

)

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 3.2-2 INSTRUMENTATION I1162 INITI ATFS OR CONTRQ11 Illg ppf4 N{f4 CONTAINMENT COOLING H MItig Minimum OPERABLE CHANNELS Applicable Per TRIP Trip Level OPERATIONAL Trip Function SYSTEM (a)(f) Setting MODES ACTION

1. Reactor Low Low 2 2 84* above top of 1, 2, 3, 4(g), 5(g) 20 Water Level active fuel (e)

2. Drywell High Pressure 4 5 2.5 poig 1, 2, 3 20 ,

(b), (c), (d) '

3. Reactor Low Pressure 1 2 300 poig and 1, 2, 3 21 5 350 poig 4 (g), 5 (g) 22

4. Containment Spray Interlock

a. 2/3 core height 1 (d) 2 2/3 core height 1, 2, 3 21

b. Containment High 2 (d) 2 0.5 peig and 1, 2, 3 21 Pressure 5 1.6 pelg

5. Timer Auto Blowdown' 1 5 120 Seconds 1, 2, 3 21

6. Low Pressure Core 2 2 100 peig and 1, 2, 3 21 Cooling Pump _ s 150 psig Discharge Pressure

7. _4 KV Emergency Buses 2/ Bus 3045 volts 1 54 1, 2, 3, 4(h), 5(h) 23 Under Voltage

8. 4 KV Emergency Buses 2/ Bus 3840 volta 1 24 1, 2, 3, 4(h), 5(h) 23 ,

Degraded voltage with 5 minute 1 5%

time delay and 7 second i 20%

inherent time delay.

3.2/4.2-14 l

l 1

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 3.2-2 (Continued)

INSTRUMENTATION TilAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLIBG SYSTEMS ACTIONS ,

ACTION 20 - With the number of OPERABLE CNANNELS loss than required by the Minimum OPERABLE CllANNELS per TRIP SYSTEM requirements

a. With one CllANNEL inoperable, place the inoperable CllANNEL in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the associated system inoperable,

b. With more than one CilANNEL inoperablo, declaro the associated system inoperable.

ACTION 21 - With the number of OPERABLE CilANNELS loss than required by the Minimum OPERABLE CllANNELS per TRIP SYSTEM requirement, declare the associated Core or containment Cooling system inoperable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 22 - With the number of OPERABLE CilANNELS less than required by the Minimum OPERABLE CllANNELS por TRIP SYSTEM requirement, place the inoperable CilANNEL in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 23 - With the number of OPERABLE CilANNELS one loss than the Minimum OPERABLE CRANNELS per TRIP-SYSTEM requirement, place the inoperable CHANNEL in the tripped condition within one hour; operation may then continue until performanco of the next required CllANNEL FUNCTIONAL TEST. Otherwise, declare the associated emergency diesel generatur inoperable and take the appropriate ACTIONS.

3.2/4.2-15

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 3.2-2 (Continued)

INSTRUMENTATIQ11 THAT IFITIATES QB CONTROLS Tl[E QQBI Al{D CONTAINMENT COOLING SYSTEMS TABLE NOTATIONS (a) Two TRIP SYSTEMS shall be OPERABLE in the required OPERATIONAL MODES for the specified trip function unless otherwise allowed by the ACTION provisions.

(b) Need not be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(c) There are a total of eight drywell high pressure sensors.

Four are used for core spray and LPCI initiation, and four >

are used for HPCI and auto blowdown initiation. This specification applies to each set of four sensors.

(d) If an instrument is inoperable, it shall be placed (or simulated) in the tripped condition so that it will not prevent containment spray.

(c) Top of active fuel is defined as 360" above vessel zero for all water levels used in the LOCA analyses. (see Bases 3.2)

(f) A CHANNEL may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillance without placing the TRIP SYSTEM in the tripped condition provided at least one OPERABLE CHANNEL in the same TRIP SYSTEM'is monitoring that parameter.

(g) When the associated Core and Containment Cooling System is required to be OPERABLE in COLD SHUTDOWN or REFUELING.

(h) Required OPERABLE when equipment relying on essential 4 KV buses are required to be OPERABLE.

3.2/4.2-16

QUAD CITIES UNITS 1 & 2 e DPR-29 & LPR-30 TABLE 3.2-3 1HEIEVt1EHIATIQlf IllAI IN1TI ATES RQQ g(QQE Minimum OPERABLE Applicable CHANNELS per Trip Level OPERATIONAL ,

Trip runction Trip runction (a', Setting HoDES ACTION

1. APRM Upscale 4 5 (0.58 Wn + 50) 1 31 (flow hias) (FRP/MrLPb) (b) ,

2. APRN Upscale 4 5 12/125 full scale 2, 5 31

3. APRM Downscale 4 2 3/125 full scale 1 31

4. APRM Inoperative 4 NA 1,2,5 31

5. Rod Bloek Honitor 2 (j) (i) 1 (f) 30 Upscale (flow bias)

6. Rod Block Monitor 2 (j) 2 3/126 full scale 1 (f) 30 Downscale

7. Rod Block Honitor 2 (j) NA 1 (f) 30 Inoperative

8. IRH Downscale (c) 6 2 3/125 full scalo 2, 5 31

9. IRH Upscale 6 s 108/125 full scale 2, 5 31

10. IRH Inoperative 6 NA 2, 5 31

11. IRH Detector Not In 6 2 2 feet below core 2, 5 31 Startup Position centerlino

12. SRM Detector Not in 3 2 2 feet below core 2 31 startup Position (d) 2 (k) centerline 5 31

13. SRH Upecale 3 (e) $ 1 X 105counts /sec 2 31 2 (k) 5 31

14. SRM Downscale (h) 3 2 100 counts /see 2 31 2 (k) 5 31

15. SRM Inoperative 3 NA 2 31 2 5 31 3.2/4.2-17

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

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 3.2-3 (Continued)

INSTRUMENTATION THAT INITIATES EQQ BLQCL Minimum OPERABLE Applicable CHANNELS ter Trip Level OPERATIONAL Trip Function Trip Function (a) Setting HODES ACTION

16. Scram Discharge 1/ bank 5 25 gallone (per bank) 1, 2, 5(g) 32 Volume (SDV) High Water Level

17. SDV digh Water Level 1 NA 3, 4, 5(g) 32 Scram Trip Bypassed 3.2/4.2-18

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 3.2-3 (Continued)

INSTRUMENTATION TilAT INITIATES ROD BLOCK ACTIONS ACTION 30 - With the number of OPERABLE CllANNELS less than required by the Minimum OPERABLE CllANNELS por Trip Function requiremont, dociare the RBM inoperable and take the ACTION required by Specification 3.3.M.

ACTION 31 - With the number of OPERABLE CilANNELS

a. One less than required by the Minimum OPERABLE CilANNELS por Trip Function requiremont, rostore the inoperable CilANNEL to OPERABLE status within 7 days or place the inoperable CilANNEL in the tripped condition within the next hour.

b. Two or more less than required by the Minimum OPERABLE CHANNELS per Trip Function requirement, place the inoperable CilANNEL in the tripped condition within ono hour.

ACTION 32 - With the number of OPERABLE CilANNELS loss than required by the liinimum OPERABLE CilANNELS por Trip Function requirement, place the inoporable CllANNEL in the tripped condition within one hour.

3.2/4.2-19

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 3.2-3 (Continued)

INSTRUMENTATION THAT IlilTIATEs ROD nLocK TABLE NOTATIONS (a) One TRIP S' STEP. shall be OPERABLE in the required MODES for the specified 'Arip Function unicos otherwise allowed by the ACTION provisions.

(b) Wo is the percent of drive flow required to produce a rated core flow of 98 Hlb /hr. Trip level setting is in percent of rated power (2511 MWt).

(c) IRM downscale may be bypassed when it is on its lowest range.

(d) This function is bypassed when the SRM count rate is 2 100 cps.

(e) This SRM functior sv be bypassed in the higher IRM ranges (ranges 8, 9, and to) when the IRM upscale rod block is OPERABLE.

(f) Both RBM CHANNELS are automatically bypassed at less than 30 percent RATED THERMAL POWER or when a peripheral rod is selected.

(g) With more than one control rod withdrawn. Not applicable to control rode removed per Specification 3.10.D or 3.10.E.

(h) This trip is bypassed when the SRM is fully inserted.

(i) The RBM upscale setpoint shall be established as specified in the CORE OPERATING LIMITS REPORT.

(j) The minimum number of OPERABLE CRANNELS may be reduced by one for maintenance and/or testing, provided that this condition does not last longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in-a 30 day period when in a LIMITING CONTROL ROD PATTERN. If this condition exists for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 30 day period when in a LIMITING CONTROL ROD PATTERN, the provisions of ACTION 30 shall be followed for an inoperable CHANNEL.

(k) OPERABLE CHANNELS must be associated with SRMs required OPERABLE per Specification 3.10.B.

3.2/4.2-20

QUAD CITIEJ UNITS 1 & 2 DPR-29 & DPR-30 TABLE 3.2-4 ,

POSTACCIDENT tiQJilipg1RQ 1NSTRUMENTATION REQVJEltil113 Minin urn Applicable CHANNELS OPERATIONAL Instrument OPERABLE MODES ACTION

1. Reactor Pressure 2 1, 2 40

2. Reactor Water Level 2 1, 2 40 P

3. Torus Water Temperature 2 1, 2 40

4. Torus Wat er Level (wide range) 2 1, 2 40 5.- Drywell Pressure (narrow range) 2 1, 2 40 (wide range) 1, 2 40

6. Drywell Hydrogen 2 1, 2 43 concentration

7. Drywell Oxygen 2 1, 2 43 concentration

8. Drywell Radiation Monitot 2 1, 2, 3 42

9. Main Steam RV Position, 2/ valve 1, 2 41 Acoustic Monitor Temperature Monttwr

10. Main St eam RV Pooltion, 2/ valve 1, 2 41 Acoustic Monster Temperature Moniuor l

3.2/4.2-21 l-

l QUAD CITIES UNITS 1 &2 DPR-29 & DPR-30 TABLE 3.2-4 (Continued)

POSTACCIDENI tipH1TORIl{Q INSTRUMENTATION REQUIREliCl{IS ACTIONS ACTION 40 -

a. With the AMAb;* '

>>^ Q8ERABLE postaccident monitoring instrumor.ksii- pt;M NELS one loss than the Minimum CHANNELSOthE$h'kRiogrirement,restoretheinoperable CHANNEL to OPERABLE status in order to moet the Minimum OPERABLE CHANNELS requirement within 30 days or be in at least HOT SHUTDOWN in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. With the number of OPERABLE postaccident monitoring instrumentation CHANNELS two less than the Minimum rostoro at least one CHANNELS CHANNEL to an OPERABLE OPERABLE requirements,ithin status w 7 days or be in HOT SHUTDOWN in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 41 - If the number of position indicators is reduced to one indication on one or more valvos, continued operation is permissible; however, if oho reactor is in a COLD SHUTDOWN not be started condition for longer up until than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , all position ion isit may indicat rostored. In the event that all position indication is lost on one or more valvos and such indication cannot be restored in 30 days, an orderly shutdown shall be initiated, and the reactor shall be depressurized to loss than 150 psig in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 42 - With less than the minimum number of OPERABLE CHANNELS, initiate the pre-planned alternato method of monitoring this paramotor within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , andt

a. oither restore the inoperable CHANNEL (s) to OPERABLE status within 7 days of the event, or

b. preparo and submit a Special Roport to the NRC within 30 outlining the ACTION taken, days following the cause of thethe event,ility, and the plans and inoperab schedule for restoring the system to OPERABLE status.

3.2/4.2-22

i i

QUAD CITIES UNITS 1 &2 DPR-29 & DPR-30 TABLE 3.2-4 (Continued) l EQSTACCIERT liQHITOR11{G 1RSIRUMENTATION EQR}1EMENTS ACTIONS ACTION 43 - From and after the date that one of the drpell hydrogen or one of the drywell oxygen monitors becomes inoperable, continued reactor operation is permissible,

a. If both drywell hydrogen or both drywell oxygen monitorn are inoperable, continued reactor operation is permissible for up to 30 days provided that during this time the HRSS monitoring capability for the inoperable 4

drywell monitoring function is OPERABLE.

!? b. If all drywell hydrogen or all drywell oxygen monitoring capability is lost, continued reactor operations is permissible for up to 7 days.

3.2/4.2-23

l l

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 4.2-1 HINIMUM 1111 M CALIBRATION IN OUENCY EQB EQBK hHQ QQNTAINMEN_T Co0 LING SYSTEMS INSTRURENTATION, B2D AL9C m AHD ISOLATIONS CRANNEL CHANNEL Applicable FUNCTIONAL CALIBRATION CHANNEL OPERATIONAL Trip Function TEST (c)(n) (c)(m) CHECK (c) MODES A. Eggi Instrumentation

1. Reactor Low Low Q Q s 1, 2, 3, 4(k), 5(k)

Water Level

2. Drywell High Q Q N.A. 1, 2, 3 Pressure

3. Reactor Low Q Q F.A. 1, 2, 3, 4(k), $(k)

Pressure

4. Containment Spray Interlock

a. 2/3 Core Height Q (p) R (p) 5 1,2,3

b. Containment High Q Q N.A. 1, 2, 3 Pressure S. Low Pressure Core Q Q N.A. 1, 2, 3 Cooling Pump Discharge Pressure

6. 4 KV Emergency Dunes R R N.A. 1, 2, 3, 4(j), 5(j)

Under Voltage

7. 4 KV Emergency Buses R (e) R S 1, 2, 3, 4(j), 5(j)

Degraded Voltage

B. B2d Blocks

1. APRM Upscale S/U (b)(g), Q (b) Q N.A. 1 (flow bias)

2. APRM Downscale S/U (b)(g), Q (b) Q N.A. 1

3. APRM Inoperative S/U (b)(g), Q (b) N.A. N.A. 1,2,5 l

3.2/4.2-24

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR 30 TABLE 4.2 1 (Continued) g1H]} igg 1152 hEQ CALIBRATION TREOUENCY EQB CQM hBR QQNTAINMEtil CCOLING RIjlDig INSTRUMENTATIOfk EQQ RLOCKS4 . hED 150LATIOgg

.................................................................................-~~..

CHANNEL CHANNEL Applicable PUNCTIONAL CALIBRATION CHANNEL OPERATIONAL Trip Function TEST (c)(n) (c)(m) CHECK (c) MODES B. Eqd glQgkg (Continued)

4. RBH Downscale S/U (b)(g), Q (b) Q N.A. 1(i)

5. RSH Upscale S/U (b)(g), Q (b) Q N.A. 1(i)

6. RBH Inoperative S/U (b)(g), Q (b) N.A. N.A. 1(1)

7. IRM Downscale S/U (a)(b), W Q N.A. 2, 5

8. IRH Upscale S/U (a)(b), i Q N.A. 2, 5

9. IRH Inoperative S/U (a)(b), W N.A. N.A. 2, 5

10. IRH Detector Not S/U (a), W (d) N.A. 2, 5 In Startup Position

11. SAM Detector Not S/U (a), W (d) N.A. 2, 5 In Startup Position

12. SRM Upecale S/U (a)(b), W Q H.A. 2, 5

13. SRM Downscale S/U (a)(b), W Q N.A. 2, 5

14. SRM Inoperative S/U (a)(b), W N.A. N.A. 2, 5

15. Scram Discharge Q Q N.A. 1, 2, 5(h)

Volume (SDV) High Water Level

16. SDV High Water R N.A. N.A. 3, 4, 5(h)

Level Scram Trip Bypassed C. BAiD G.LRAID L1DR Isolation

1. Main Steam Line R R h.A. 1, 2, 3 Tunnel High Temperature 3.2/4.2 25

i QUAD CITIES UNITS 1 & 2 DPR.29 & DPR-30 TABLE 4.2 1 (Continued)

HINIMUM TEST MiQ CALIBRATIQJi FREOUENCY EQB COP,E M!iQ CONTAINMENT COOLING $YSTEti!i INSTRUMENTATION. EQQ Ri&CLL, MiQ ISOLATIONS CHANNEL CHANNEL Applicable FUNCTIONAL CALIBRATION CHANNEL OPERATIONAL Trip Function TEST (c)(n) (c)(m) CHECK (c) HODES

.................................................................--...c --..............

C. Main Steam Line Isolation igentinued)

2. Main Steam Line Q Q S 1, 2, 3 High Flow

3. Main ; team Line Q Q N.A. 1 Low Pressure

4. Main Steam Line Q (b) R S 1, 2, 3 Tunnel High Radiation (q)

5. Reactor Low Low Q (p) R (p) S 1, 2, 3 Water Level (r)

D. RCIC igolatioD

1. RCIC Steam Line Q (f) Q (f) N.A. 1, 2, 3 High Flow

2. RCIC Turbine Area R R N.A. 1, 2, 3 High Temperature

3. RCIC Steam Supply Q Q N.A. 1,2,3 Line Low Pressure E. ILEgl Isolation

1. HPCI Steam Line Q (f)(p) R (f)(p) N.A. 1, 2, 3 High Flow

2. HPCI Area High R R N.A. 1, 2, 3 Temperature

3. HPCI Steam Supply Q (p) R (p) N.A. 1,2,3 Line Low Pressure 3.2/4.2-26

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 4.2-1 (Continued)

MIRDillti IllI NiD CALIBRATION FREOllit(Cf EQB CQRE htig CONTAINMYRI SQOLING SYSTEMS INSTRUMENTATION. SQQ BLOCKS, htiQ ISOLATIONS CHANNEL CilANNEL Applicable FUNCTIONAL CALIBRATION CHANNEL OPERATIONAL Trip Function TEST (c)(n) (c)(m) CHECK (c) MODES F. Egaetor guildino Ventilation Rystem Isolation End ADCTS InitiatiED

1. Refueling Floor O Q s (1)

Radiation Monitors

2. Reactor Building Q R S (u)

Ventilation Exhaust Duct Radiation Monitors G. Steam 231 hir Eiector Off-cas isolation

1. SJAE Radiation Q (b) R S (t)

Honitors H. Control Room yentilation System isolation

1. Reactor Low Water 0 Q s 1, 2 , 3 Level (q)

2. Drywell High Q Q N.A. 1, 2, 3 Pressure (s)

3. Main Steam Line Q Q S 1,2,3 High Flow

4. Toxic Gas Analyzers M R S (o)

(Chlorine, Ammonia, sulphur Dioxide)

I. Drywell Hiah Radiation

1. Drywell High Q h S 1,2,3 Radiation 3.2/4.2-27

{

L 1 l

. _. _ _ . _ . _ _ _ _ _ _ _ _ _ _ __ _ __ . _ ____. _ _ _ _ _ ___.m _

i QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 4. 2-1 (Continued)

MINIMUM TEST AND CALIBRATION FREQUENCY FOR CORE AND CONTAINMENT COOLING SYSTEMS INSTRUMENTATION. ROD BLOCKS. AND ISOLATIONS l TABLE NOTATIONS !

1 (a) Functional tests shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup if not performed within the previous 7 days, but are not required to be performed more than once per week.

(b) This instrumentation is excepted from the functional test definition. The functional test shall consist of injecting a simulated electrical signal into the ueasurement CHANNEL.

(c) Functional tests, calibrations, and checks are not required when these instruments are not required to be OPERABLE or are tripped.

(d) The positioning mechanism sh'111 be calibrated every REFUELING OUTAGE.

(e) Functional tests shall include verification of operation of the degraded voltage 5-minute timer and 7-second inherent timer.

(f) Verification of time delay setting of 3 5 t s 9 seconds shall also be performed.

(g) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

(h) Required with more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.10.D or 3.10.E.

(i) With thermal power greater than or equal to 30 percent of RATED THERMAL POWER.

(j)

~

Required OPERABLE when equipment relying on essential 4 KV buses are required to be OPERABLE.

l (k) When the associated Core er Containment Cooling System is required to be OPERABLE in COLD SHUTDOWN or REFUELING.

(1) Required OPERABLE whenever SECONDARY CONTAINMENT INTEGRITY is required or when required OPERABLE by Specification 3.2.D.

3.2/4.2-28

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

. - ~ . .

i OUAD CITIES UNITS 1 & 2 -

DPR-29 & DPR-30 .

TABLE 4.2-1 (Continued) '

M1HIMUli_T19T AND CALIBRATION GEQUENCY FOR_ CORE AND CONTAINMENT ,

COOLING SYSTEMS INSTRUMENTATION. P.OD BLOCKS. AN G QLATIONE i

TARLE NOTATIONS (m) Neutron detectors may be excluded from calibration.

(n) LOGIC SYSTEM FUNCTIONAL TESTS are performed as specified in the applicable section for these systems. i (o) Required OPERABLE at all times.

(p) Trip units are calibrated at least once per 92 days. A functional test of the trip unit is performed concurrently with the calibration.- Transmitters are calibrated once per OPERATING CYCLE.

(q) These Trip Functions are common to the RPS Trip Functions.

(r) These Trip Functions are common to the Core and containmsnt Cooling Actuation Trip Functions.

(s) These Trip Functions are common to the RPS and the Core and Containment Cooling Actuation Trip Functions.

(t) Required OPERABLE during SJAE operation.

(u) Required OPERABLE whenever SECONDARY CONTAINMENT INTEGRITY is required or when required OPERABLE by Specification 3.2.F.

8 i

3.2/4.2-29

.a-...-_ _ _ _ - . - - - - . - . _ . - . . - - . - _ . . - . . - . - . -

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 Table 4.2-2 E

POSTAccIDENT MONITORIF2 ll!STRUMENTATION SURVEILLANCE REQJl.11E M Applicable CHANNEL CHANNEL OPERATIONAL Instrument CALIBRATION CHECK HODES

1. Reactor Prosauro R H 1, 2

'u

) 2. Reactor Pater Level R M 1, 2

3. Torus Water Tempe r'.ture T H 1, 2

4. Totus Water L*.

(wide rangel R H 1, 2 r

5. Drywell Pressure R H 1, 2

6. Drywell Hydrogen Q M 1, 2 concentration

7. Drywell Oxygen Q M 1, 2 concontration

8. Drywell Radiation Morfter R (b) M 1,2,3

9. Main Steam RV Position, Acoustic Monttor (a) M 1, 2 Temperature Monitor R H 1, 2

10. Mt.in Steam SY Position, Acoustic Monitor (a) M 1, 2

. Temperature Monitor R H 1, 2 3.2/4.2-30 E

QUAD CITIES UNITS 1 &2 DPR-29 & DPR-30 TABLE 4.2-2 (Continued)

POSTACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REOJIMEMENTS TABLE NOTATIONS

~

(a) Functional tests will be conducted befole startup at the and of each REFUELING OUTAGE.

(b) Calibration shall consist of an electronic calibration of the CRANNEL, not including the detector, for range decados above 10 R/hr; and a one-point calibration check of the detector below 10 R/hr with an installed or portable gamma source.

3.2/4.2-31

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 3.2 LIMITING CONDITIONS FOR OPERATION BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidants which are beyond the operator's ability to control, or terminace, operator errors before they result in serious consequences. This set of specifications provides the LIMITING CONDITIONS FOR OPERATION for the - primary system isolation function, initiation of the emergency core cooling system, control rod block and standby gas treatment systems.

The-objectives of the specifications are (1) to assure the effectiveness of the protective instrumentation when required by preservint its cEpability to tolerate a single failure of any component of such systems even during periode when portions of such systems are out-of-service for maintenance and (2) to prescribe the trip settings required to assure adequate performance. When necessary, one CHANNEL may be made inoperable for brief intervals to conduct required CHANNEL FUNCTIONAL TESTS and CHANNEL CALIBRATIONS. Some of the settings on the instrumentation that initiate or control coro and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. It should be noted that the setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvartent actuation of the safety system involved and exposure to abnormal situations.

Isolation valves are installed in those lines that panetrate tha primary containment and must be isolated during ss-of-coorant accident so-that the radiation dose livis> . t c- not-exceeded during an accident condition. Actuat eo et 1.hese valves is initiated by the protective instrumen.

uon which serves the condition for which isolation is requned (this instrumentation is shown in Table 3.2-1). Such instrumentation must be available for the Applicable OPERATIONAL MODES shown in Table 3.2-1. The objective is to isolate the primary containment so that the guidelines of 10 CFR 100 are not exceeded during an accident.

The instrumentation which initiates prir.ary system isolation is connected in a dual bus arrangement. Thus the discussion provided in the bases for Specification 3.1 is applicable here.

The reactor :ow lc. vel instrumentation is set to trip at greater than 8 inches on the level instrument (top of active B 3.2/4.2-1

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 fuel is defined to be 360 inches above vessel zero) and after allowing for the full power pressure drop across the steam dryer the low-level trip is at 504 inches above vessel zero, or 144 inches above the top of active fuel. Retrofit 8x8 fuel has an active fuel length 1.24 inches longer than earlier fuel designs. However, present trip setpoints were used in the LOCA analyses (Lons-of-coolant accident analysis for Dresden Units 2 & 3 and Quad Cities Unita 1 & 2, NEDO-24146A, April, 1979). This trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation pumps (reference SAR Section 7.7.2) . For a trip setting of 504 inches above vessel zero and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximum break. The setting is therefore adequate.

The reactor low-low level instrumentation is set to trip when reactor water level is greater than or equal to 444 inches abovc vescal zero (with top of active fuel defined as 360 inches above vessel zero, -59 inches is 84 inches above the top of active fuel). This trip initiates closure of Group 1 primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems, starts the emergency diesel generator, and trips the recirculation pumps.

This trip setting level was chosen to be Icw enough to prevent spurious operation but high enough to initiate ECCS operation nd primary system isolation so that no molting of the fuel

-cladding will occur and so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded. For the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary isolation are initiated in time to meet the above criteria. The instrumentation also covers the full spectrum of breaks and meets the above criteria.

The drywell high pressure instrumentation is a backup to the water level instrumentation and, in addition to initiating ECCS, it causes isolation of Group 2 isolation valves. For the breaks discussed above, this instrumentation will initiate ECCS operation at about the same time as the low-low water level instrumentation; thus the results given above are applicable here also. Group 2 isolution valves include the drywell vent, purge and sump isolation valves. Drywell high pressure activates only these valves because drywell high pressure could occur as the result of non safety-related causes such as not purging the drywell air during start up.

Total system isolation is not desirable for these conditions, and only the Group 2 valves are required to close. The low-low water level instrumentation initiates protection for the B 3.2/4.2-2

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 full spectrum of loss-of-coolant accidents and causes a trip of Group 1 primary system isolation valves.

High radiation in the drywell indicates an abnormal situation due to a line break or other abnormal occurrence. To preclude release of potentially highly contaminated material from the drywall, this high radiation isolation automatically closes the Group 2 isolation valves.

Venturi tubes are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line b i :k accident.

In addition to monitoring steam flow, instrumentation is provided which causes a trip of Group 1 isolation valves. The primary function of the instrumentation is to detect a break in the main steam line, thus only Group 1 valves are closed.

For the worst-case accident, main steam line break outside the drywell, this trip setting of 140% of rated steam flow, in conjunction with the flow limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel temperatures remain less than 15 0 0*F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines (reference SAR Sections 14.2.3.9 and 14.2.3.10).

Temperature-monitoring instrumentation is provided in the main steam line tunnel to detect leaks in this area. Trips are provided on this instrumentation and when exceeded cause closure of Group 1 isolation valves. Its setting of 200'F is low enough to detect leaks in the order of 5 to 10 gpm; thus it is capable of covering the entire spectrum of breaks. For large breaks, it is a backup to high-steam flow instrumentation discussed above, and for small breaks with the resulting small release of radioactivity, provides isolation bafore the guidelines of 10 CFR 100 are exceeded.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure. This instrumentation causes closure of Group 1 valves, the only valves required to close for this accident. With the established setting of 15 times normal background (without hydrogen addition) and main steam line isolation valve closure, fissi7n product release is limited co that 10 CFR 100 caidelines are not exceeded for this accident (reference SAR Section 14.2.1.7).

Pressure instrumentation is provided which trips when main l steam line pressure drops below 825 psig. A trip of this instrumentation results in closure of Group 1 isolation B 3.2/4.2-3 1

i QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 valves. In the REFUEL and STARTUP/ HOT STANDBY OPERATIONAL MODES this trip function is bypassed. This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the control and/or bypass valves to opon. With the trip set at 825 poig, inventory loss is limited so that fuel is not uncovered and peak cladding temperatures are much less than 1500*F; thus, there are no fission products available for release other than those in the reactor water (reference SAR Section 11.2.3).

The RCIC and the HPCI high flow and temperature instrumentation are provided to detect a break in their respective piping. Tripping of this instrumentation results in actuation of the RCIC or of HPCI isolation valves. Tripping logic for this function is the same as that for the main steam line isolation valves, thus all sensors are required to be operable or in a tripped condition to meet single-failure criteria. The trip settings of 170'F and 300% of design flow and valve closure time are such that core uncovery is prevented and fission product release is within limits.

The instrumentation which initiates ECCS action is arranged in a one-out af-two taken twice logic circuit. The single-failure critoria are met by virtue of the fact that redundant core cooling functions are provided, e.g., sprays and automatic blowdown and high pressure coolant injection.

The specification requires that with the number of OPERABLE CHANNELS less than required by the minimum OPERABLE CHANNELS per TRIP SYSTEM requirement, either the CHANNEL (s) is placed in the trip condition or the associated system which it activates is declared inoperable and the ACTION statements of ;

Specification 3.5 govern. This specification preserves the ef fectiveness of the system with respect to the single-f ailure criteria even during periods when maintenance or testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not go below the MCPR fuel cladding integrity SAFETY LIMIT. The trip logic for this function is one-out-of-n; e.g. , any trip on one of the six APRMs, eight IRMs, four SRMs will result in a rod block. The minimum instrument channel requirements assure sufficient instrumentation to assure that the single-failure criteria are met. The minimum instrument channel requirements for the RBM may be reduced by one for a short period of time to allow for maintenance, testing, or calibration. This time period is only 3% of the operating time in a month and does not significantly increase the risk of failing to prevent an inadvertent control rod withdrawal.

B 3.2/4.2-4 I

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 The APRM rod block function is flow-biased and prevents a significant reduction in MCPR, especially during operation at reducod flow. The APRM provides gross core protection, i.e.,

limits the gross withdrawal of control rods in the normal withdrawal sequence.

In the REFUEL and STARTUP/ HOT STANDBY OPERATIONAL MODES, the APRM rod block function is set at 12% of rated power. This control rod block provides the same type of protection in the REFUEL and STARTUP/ HOT STANDBY OPERATIONAL MODES as the APRM flow-biased rod block does in the RUN OPERATIONAL MODE, i.e.,

prevents control rod withdrawal before a scram is reached.

The RBM rod block function provides local protection of the core, i.e., the prevention of TRANSITION BOILING in a local region of the core for a single rod withdrawal error from a LIMITING CONTROL ROD PATTERN. The trip point is flow-biased.

The worst-case single control rod withdrawal error is analyzed for each reload -to assure that, with the specific trip settings, rod withdrawal is blocked before the MCPR reaches the fuel cladding integrity SAFETY LIMIT.

Below 30% power, the worst-case withdrawal of a single control rod without rod block action will not violate the fuel cladding integrity SAFETY LIMIT. Thus - the RBM rod block function is not required below this power level. If a peripheral control rod is selected, the neutron leakage is sufficiently high such that withdrawal of this rod will not violate the fuel cladding integrity SAFETY LIMIT. Thus, the RBM function is not required for withdrawal of peripheral control rods.

The=IRM rod block function provides local as well as gross core protection. The scaling arrangement.is such that the trip setting is less than a factor of 10 above the indicated level. Analysis of the worst-case accident results in rod l

block action before MCPR approaches the MCPR fuel cladding integrity SAFETY LIMIT.

l A downscale indication on-an APRM is en indication that the instrument has failed or~is not sensitive enough. In-either

! case,_the instrument will not respond to changes in control l rod motion, and the control rod motion is thus prevented. The downscale trips are set at 3/125 of full scale.

The SRM rod block of less than or equal to 100 CPS and the detector is not fully inserted assures that the SRMs are not withdrawn from the core prior to commencing rod withdrawal for L startup. The scram discharge volume, high water level rod B 3.2/4.2-5

,y y- --y*ew q+ d+g --

4my p t-'m-- -*-r me-- wvm.* g y

i QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 block provides annunciation for operator action. The alarm setpoint.has been selected to provide adequate time to allow for the determination of the cause for the level increase and corrective action prior to automatic scram initiation.

For effective emergency core cooling during small pipe breaks, ;

the HPCI system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided.as a backup to the HPCI, in the event the HPCI does l not operate. The arrangement of the tripping contacts is such ;

as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria are met (reference SAR Section 6.2.6.3). The specification preserves the-effectiveness of the system during periods of maintenance, testing or calibration and also minimizes the risk of inadvertent operation, i.e., only one instrument channel out-of-service.

' Two radiation monitors are provided on the refueling floor which initiate isolation of the reactor building and operation of the standby gas treatment systems. The trip logic is one-out-of-two. Trip settings of less than or equal to 100 Mr/hr for the_ monitors on the refueling -floor are based upon initiating normal ventilation isolation and standby gas 1 treatment system operation so that none of the activity released during the refueling accident leaves the reactor building.via the normal ventilation stack but that all the ,

activity is processed by the standby gas treatment system. l The instrumentation which is provided to monitor the post accident condition' is listed jn Table 3.2-4. The

-instrumentation listed and the LIMITING CONDITIONS FOR OPERATION-on these systems ensure adequate monitoring of the

. containment following a loss-of-coolant accident. Information from this instrumentation will. provide the operator with a detailed knowledge of the conditions resulting from the accident;-based on this information, the operator can make logical decisions regarding post accident recovery.

Allowable outage-times are based on the fact that several diverse instruments are available for guiding the operator shoulG an accident occur, and on the low probability of an instrument being.out-of-service and an accident occurring.

The normal supply of air for the control room ventilation system trains "A" and "B" is outside the service building. In the event of an accident, this source of air may be required B 3.2/4.2-6

l QUAD CITIES UNITS 1 & 2 DPft-29 & DPR-30 to be isolated to prevent high doses of radiation in the control room. Rather than provide this isolation function on a radiation monitor installed in the intake air duct, signals which indicate an accident, i.e., drywell hioh pressure, low water level, main steam line high flow; drywell high radiation, high radiation at the refueling floor monitors, or high radiation in the reactor building ventilation duct, will cause isolation of the intake air to the control room. The above trip signals result in immediate isolation of the control room ventilation system and thus minimize any radiation dose. Manual isolation capability is also provided.

Isolation from high toxic chemical concentration has been added as a result of the " Control Room Habitability Study" submitted to the NRC in December 1981 in response to NUREG-0737 Item III D.3.4. As explained in Section 3 of this study, ammonia, chlorine, and sulphur dioxide detection capability has been provided. The setpointo chosen for the control room ventilation isolation are based on early detectiog in the outside air supply at the odor threshold, so that the toxic chemical will not achieve toxicity limit concentrations in the Control Room.

The instrumentation that monitors the release of gases from the main condenser offgas system, provides reasonable assurance that the effluents to unrestricted areas will not exceed the guidelines of 10 CFR Part 20. The alarm setpoints for the instruments are provided to ensure that the alarms will occur prior to exceeding the limits of 10 CFR 20.

4.2 SURVEILLANCE REQUIREMENTS BASES Surveillance requirements for the instrumentation in Technical Specification Section 3.2/4.2 are selected in order to demonstrate proper function and operability. The surveillance intervals can be determined by different means, such as, 1) operating experience, 2) good engineering judgement, 3) reliability analyses, or 4) other analyses that are found acceptable to the NRC.

Reactor water level instruments 1(2)-263-73A & B, HPCI high steam flow instruments 1(2)-2352 & 1(2)-2353, and HPCI steam line low pressure instruments 1(2)-2389A-D have been modified to be analog trip systems. The analog trip system consists of an analog sensor (transmitter) and a master / slave trip unit setup which ultimately drives a trip relay. The frequency of calibration and functional testing for instrument loops of the analog trip system has been established in Licensing Topical Reoort NEDO- 21617-A (December 1978) . With the one-out-of-two taken twice logic, NEDO-21617-A states that each trip unit be B 3.2/4.2-7 I

l QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 subjected to a calibration / functional test frequency of one month. This monthly calibration / functional test f requency has been extended to quarterly by the analyses contained in NEDC-31677P for trip units in analog trip systems. An adequate calibration / surveillance test interval for the transmitter is once per operating cycle.

The adiation monitors in the ventilation duct and on the re;% ling floor which initiate reactor building ventilation system isolation, control room ventilation system isolation, and standby gas treatment operation are arranged in two one-out-of-two logic systems. The bases given above for the rod blocks apply here also and were used to arrive at the functional testing frequency.

Based on experience at Dresden Unit 1 with instruments of similar design, a testing interval of once overy 3 months has been found to be adequate.

The automatic pressure relief instrumentation can be considered to be a one-out-of-two logic system, and the discussion above applies to it also.

The instrumentation which is required for the post accident condition will be tested and calibrated at regularly scheduled intervals. The basis for the calibration and terting of this instrumentation is the same as was discussed alove for the reactor protection system and the emergency core cooling systems.

B 3.2/4.2-8 l

I EXISTING TECH SPEC TS ? <,

' PROTECTIVE lh

ATIO N' - - -

QUAD-CITIES DPR-29 3,2/4.2 PROTECTIVE INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS Applicability: Applicability:

Applies to the plant instrumentation Applies to the surveillance requirements which performs a protective function. of the instrumentation that performs a protective function.

Objective: Objective:

To assure the operability of protective To specify the type and frequency of sur-instrumentation, veillance to be applied to protective instrumentation.

SPECIFICATIONS A. Primary Containment Isolation Func- A. Primary Containment Isolation Func-

, tions tions When primary containment integrity is Instrumentation and logic systems required, the limiting conditions of shall be functionally tested and cal-operation for the instrumentation ibrated as indicated in Table 4.2-1.

that initiates primary containment isolation are given in Table 3.2-1.

B. Core and Containment Cooling Systems - B. Core and Containment Cooling Systems -

Initiation and Control Initiation and Control The limiting conditions for operation Instrumentation and logic systems for the instrumentation that shall be functionally tested and initiates or controls the core and calibrated as indicated in Table containnient cooling systems are given 4.2-1.

in Table 3.2-2. This instrumentation must be operable when the system (s) it initiates or controls are required to be operable as specified in Speci-fication 3.5.

C. Control Rod Block Actuation C. Control Rod Block Actuation

1. The limiting conditions of oper- Instrumentation and Logic systems ation for the instrumentation shall be functionally tested and cal-that initiates control rod block ibrated as indicated in Table 4.2-1.

are given in Table 3.2-3.

3.2/4.2-1 Amendment No. 114

QUAD-CITIES DPR-29

2. The minimum number of operable instrument channels specified in Table 3.2-3 for the rod block monitor may be reduced by one in one of the trip systems for maintenance and/or testing, pro-vided that this condition does not last longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in

' any 30-day period. If this condition exists for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 30-day period, the system shall be tripped.

D. Refueling Floor Radiation Monitors D. Refueling Floor Radiation Monitors

1. Except as specified in Specifi. The two refueling floor radiation cation 3.2.D.2, the two refuel- monitors shall be functionally tested ing floor radiation monitors and calibrated as indicated in Table shall be operable whenever irra- 4.2-1. Reactor building ventilation diated fuel or components are isolation and standby gas treatment present in the fuel storage pool system initiation shall be performed and during refueling or fuel at least each operating cycle.

movement operations.

2. One of the two refueling floor radiation monitors may be inop-erable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the

-inoperable monitor is not re-stored to service in this time, the reactor building ventilation system shall be isolated and the standby gas treatment operated until repairs are complete.

3. The trip setting for the refuel-ing floor radiation monitors shall be set at a value of < 100

~

l -

mR/hr.

l 4. Upon loss of both refueling i floor radiatio 1 monitors whilt L in use, the reactor building ventilation system shall be iso-lated and the standby gas treat-ment operated.

l l

3.2/4.2-2 Amendment No. 114

QUAD-CITIES DPR-29 E. Postaccident Instrumentation E. Postaccident Instrumentation The limiting conditions for operation Postaccident instrumentation shall be for the instrumentation which is read functionally tested and calibrated as out in the control room, required for indicated in Table 4.2-2.

postaccident monitoring are given in Table 3.2-4.

F. Control Room Ventilation System F. Control Room Ventilation System Iso-Isolation lation

1. The control room ventilation 1. Surveillance for instrumentation systems are isolated from which initiates isolation of outside air on a signal of high control room ventilation shall drywell pressure, low water be as specified in Table 4.2-1.

level, high main steamline flow, high toxic gas concentration, high radiation ir, either of the reactor building ventilation exhaust ducts, or manually.

Limiting conditions for operation shall be as indicated in Table 3.2-1 and Specification 3.2.H. and 3.2.F.2.

2. The toxic gas detection 2. Manual isolation of the control instrumentation shall consist of room ventilation system shall be a chlorine, ammonia, and sulphur demonstrated once every dioxide analyzer with each trip refueling outage, setpoint set at:

a. Chlorine concentration

< 5 ppm.

b. Ammonia concentration

< 50 ppm.

c. 3ulphur dioxide

-o' concentration 5,3 ppm.

The provisions of Specificatinn 3.0.A, are not applicable.

3.2/4.2-3 Amendment No. 114 L

QUAD-CITIES DPR-29 G. Radioactive Liquid Effluent Instru- G. Radioactive Liquid Effluent Instru-mentation mentation The effluent monitoring instrumenta- Each radioactive liquid effluent mon-tion shown in Table 3.2-5 shall be itoring instrument shown in Table operable with alarm setpuints set to 4.2-3 shall be demonstrated operable ensure that the limits of Specifi- by performance of the given source cation 3.8.B are not exceeded. The check, instrument cf.eck, calibration, alarm setpoints shall be determined and_ functional test operations at the in accordance with the ODCM. frequencies shown in Table 4.2-3.

1. With a radioactive liquid ef-fluent monitoring instrument alar'm/ trip setpoint less con-servative than required, without delay suspend the release of ra-dioactive liquid effluents moni-tored by the affected instru-ment, or declare the instrument inoperable, or change the set-point so it is acceptabiy con-servative.

2. With one or more radioactive li-quid effluent monitoring instru-ments inoperable, take the ac-tion shown in Table 3.2-5.

Exert best efforts to return the instrument to operable status ,

within 30 days and, if unsuc-cessful, explain in the next Semi-Annual Radioactive Effluent Release Report why the inoper-ability was not corrected in a timely manner. This is in lieu of an LER.

3. In the event a limiting condi-tion for operation and associ-ated action requirements cannot be satisfied because of circum-stances in excess of those ad-dressed in the specifications, provide a 30-day written report to the NRC, and no changes are required in the operational condition of the plant, and this does not prevent the plant from entry into an operational mode.

3.2/4.2-4 Amendment No. 114 I

l

l QUAD-CITIES DPR-29 H. Radioactive Gaseous Effluent Instru- H. Radioactive Gaseous Effluent Instru-mentation mentation The effluent monitoring instrumenta- Each radioactive gaseous radiation tion shown in Table 3.2-6 shall be monitoring instrument in Table 4.2-4 operable with alarm / trip setpoints shall be demonstrated operable by set to ensure that the limits of Spe- performance of the given source cification 3.8.A are not exceeded, check, instrument check, calibration, The alarm / trip setpoints shall be and functional test operations at the determined in accordance with the frequency shown in Table 4.2-4.

ODCM.

1. With a radioactive gaseous ef-fluent monitoring instrument alarm / trip setpoint less con-servative than required, without delay suspend the release of ra-dioactive gaseous effluents mon-itored by the affected instru-ment, or declare the instrument inoperable, or change the set-point so it is acceptably con-servative.

2. With one or more radioactive gaseous effluent monitoring instruments inoperable, take the action shown in Table 3.2-6.

Exert best efforts to return the instrument to operable status within 30 days and, if unsuc-cessful, explain in the next Semi-Annual Radioactive Effluent Release Report why the inoper-ability was not corrected in a timely manner. This is in lieu of an LER.

3. In the event a limiting condi-tion for operation and associ-ated action requirements cannot be satisfied because of circum-stances in excess of those ad-dressed in the specifications, provide a 30-day written report to the NRC and no changes are required in the operational condition of the plant, and this does not prevent the plant from entry into an operational mode.

3.2/4.2-5 Amendment No. 114

)

QUAD-CITIES l DPR-29 3.2 LIMITING CONDITIONS FOR OPERATION BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in uences. This set of specifications provides the I,//M/d/

rw. serinm

/

enn eq/M////dn for the primary system isolation function, in////f//

itiation of the emergency core cooling system, control rod block and standby gas treatment systems. The objectives of the specifications are (1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are outMD service for maintenance and (2) to prescribe the tri settings required to assure adequate performance. When necessary, one / M/ may be made in_ operable for brief intervals to conduct required 4 /df/47/Hff and _ eme

(/ft1#fitf$#f. Some of the settings on the instrume tation that initiate or control core anu containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. It should be noted that the setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss-of coolant accident so that the radiation dose limits are not exceeded during an accident condition.

Actuation of these valves is initiated by the protective instrumentation which serves the condition for which isohtion is required (this instrumentation is shown in Table 3.2;1). Such instrumentation must be availablwhenever primary contain cwe 'ntcgritj is rcquirci The objective

( is to isolate the primary containment s3 that the guidelines of 10 CFR 100 f are not exceeded during an accident.

The instrumentation which initiates primary system isolation is connected in

{ ' a dual bus arrangement. Thus the discussion ' 'n the bes46 for

~

Specification 3.1 is applicable.hu o f"'

hws tea. greater than The b reactoralevel instrumentation is set to trip at A 8 inches on the level instrument (top of active fuel is defined to be 360 inches above vessel zero) and after allowing for the full power pressure drop across the steam dryer the low-level trip is at 504 inches above vessel zero, or 144 inches above the top of active fuel, Retrofit 8x8 fuel has an active fuel G for the Applica.ble OPEMnDNRL MODES .shown ir) Tabic 3.3.~/

3.2/4.2-6 Amendment No. 114

QUAD-CITIES l DPR-29 length 1.24 inches longer than earlier fuel designs. However, present trip setpoints were used in the LOCA analyses *. This trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation pumps (reference SAR Section 7.7.2). For a trip setting of 504 inches above vessel zero snd a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximumbreaky,./hesettingisthereforeadequate.

toup lou)

The Tec h w reactor 31evel instrumentation is set to trip when reactor water fis greafer MhTeIS)( 444 inches above vessel zero (with top of active fuel defined as I

man or qua/L 360 inches above vessel zero, -59 inches is 84 inches above the top of

' vo '/ active fuel). This trip initiates closure of Group 1 primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems, starts the emergency diesel generator, and trips the recirculation pumps. This trip setting level was chosen to be low enough to prevent spurious operation but high enough to initiate ECCS operation and primary system isolation so that no melting of the fuel cladding will occur andsothatpostpccidentcoolingcanbeaccomplishedandtheguidelinesof 10 CFR 100 will not be exceeded. For the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary isolation are initiated and in time to meet the above criteria. The instrumentation also covers the full spectrum of breaks and

- meets the above criteria.

hiah The M9 h-drywellJressure instrumentation is a backup to the water level instrumentation and, in addition to initiating 2005, it causes isolation of Group 2 isolation valves. For the breaks discussed above, this instrumentation will initiate ECCS operation at about the same time as the lodow water level instrumentation; thus the results given above are applicable here also. s purge and sump isolation Group valves. 2 isolation valve $, ssure

%h-ct'rywell clude activates the drywell vent, only these valvis beccuse Mgh drywellNPessure could occur as the result of nonUsafety-related causes such as not purging the drywell air during start @p, Total system isolation is not desirable for these conditions, and only the6alves h6roup 2)are required to close. The lodow water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes a trip of Group 1 primary system isolation valves.

Inseri- :

Mtched Venturi tubes are provided in the main steam [ines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a stean(ine break accident. In addition to monitoring steam flow, instrumentation is provided which causes a trip of Group 1 isolation valves. The primary function of the instrumentation is to detect a break in For the worst-case themainsteam[ine,(thusonlyGroup1valvesareclosed.

accident, main stea ine break outside the drywell, this trip Loss 9ofC.oolant accident analysis for Dresden Units 2 & 3 and Quad Cities Units 1 & 2, NED0-24146A, April, 1979.

3.2/4.2-7 Amendment No. 114

9 INSERT FOR TECHNICAL BPECIFICATION PAGE 3.2/4.2-7 111gh radiation in the drywell indicates an abnormal situation due to a line break or other abnormal occurrence. To preclude release of potentially highly contaminated material from the drywell, this high radiation isolation automatically closes the Group 2 isolation valves.

4

~

t QUAD-CITIES DPR-29 setting of 140% of ' rated steam flow, in conjunction with the flow limiters .

andmain-stearQinevalveclosure,limitsthemassinventorylosssuchthat fuel is not uncovered, fuel temperatures remain less than 1500 F, and release of radioactivity to the environs-is well below 10 CFR 100 guidelines

(reference SAR - Sections 14. 2. 3. 9 and ~ 14. 2. 3.10).

Temperature 3nonitoring instrumentation is provided in the main steam Iine tunnel to detect leaks in this area. Trips are provided on this

instrumentation and when exceeded cause closure of Group i iso tion valves. Its_ setting of 200'F is low enough to detect leaks W e order of 5 to 10 gpm; tnus it 1s capable of covering the entire spectrum of breaks.

For large breaks, it is a backup to high-steam flow instrumentation discussed above, and for small breaks with the resulting small release-of '

radioactivity, Sve*s isolation ** before the guidelines of 10 CFR 100 are

-exceeded. F*

High radiation monitors in the main stear (ine tunnel have been provided to detect gross fuel failure.- This instrumentation causes closure of Group 1 valves, the only valves required to close for this accident. With'the estabitsbed setting of 15 times normal background (without hydrogen addi-tion)_and main steamline isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident (reference SAR-Section 14.2.1.7).

Pressureinstrumentationisprovidedwhichtripswhenmainsteamline pressure drops below 825 psig. A trip of this instrumeatation results in closure of Group 1 isolation valves. IntheR///// ands ///U//H//S/#d//

esAATlWA(.$///f this trip function is bypassed. This function is provided primarily to provide protection against a pressure regulator malfunction which would causethecontroland/orbypassvalv@toopen. With the trip set at 825 psig, inventory loss =is limited so that fuel is-not uncovered and peak cladding temperatures are much less than 1500 F; thus, there are no fission

-products _available for release other than those_in the reactor water

-_(reference SAR Section 11,2.3),-

The RCIC and the HPCI high~ flow and temperature instrumentation are provided to detect a: break in their respective piping. Tripping of this

instrumentation results in actuation of the RCIC or of HPCI isolation

. . Tripping logic for this function'is the.same as that for the main

-.steang valve}ine isolation. valves, thus all sensors are required to be: operable or in a tripped condition to meet single-failure criteria. The trip settings p of 170'F and_300% of design flow and valve closure time are such that core l l_

uncovery_is prevented and fission product release is within limits.

l l

L l_ 3.2/4.2-8 Amendment No. 121 l

QUAD-CITIES OPR-29 The instrumentation which initiates ECCS action is arranged in a one-out-of-two taken twice logic circuit. -Unl4ke-the-teseter sc-mm efteu4%-howeve@here !c cre-tr4p-systsa-assoc 4ated-with-sach functdon pathe+-the the two trip nstems-4Hhe-eeac40%eetecrt4en--sy+ tem. The single-failure criteria are met by virtue of the fact that redundant core cooling functions are provided, e.g., sprays and automatic blowdown and highe$ressurecoolantinjection. T he-spec 4 f4 ca t4en-cequire&-that-i-f-a-teip tysteHeeomes-4Aoper-able, t ha sy$ tem which it activates is declared On<se d inopeeaMt. F0r-c*amph , i f - thc trip-system-for - core sprey-A-becomes AHached) 4eopePaMe r -ceee-6Pr+y-A-45-4eclared inoperable and tha e-ni service _

spee4ffeet4 ens-of-6pe+444 cation - 3. 5 gove rn. This specification preserves the effectiveness of the system with respect to the single-failure criteria even during periods when maintenance or testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not go below the MCPR /uel fladding

/ntegrity SM/f/ LS/l, The trip logiq e.g. , any trip on one of the six APRK_,,for this function is ons90utett9n;eight _# IRHDi~f fo in a rod block. The minimum instrument channel requirements assure sufficient instrumentation to assure that the single-failure criteria are me'. . The minimum instrument channel requirements for the RBM may be reduced by one for a short period of time to allow for maintenance, testing, or calibration. This time period is only

3% of the operating time in a month and does not significantly increase the risk of peevent4ng an inadvertent control rod withdrawal, fo/ung fogreven/

The APRM rod block function is flowEbiased and prevents a significant reduction in MCPR, especially during operation at reduced flow. The APRM provides gross core protection, i.e., limits the gross withdrawal of control rods in the normal withdrawal sequence.

crenartMML In the M/$ff and $Mf4f//$f f4f#f(/////, the APRM rod block function is set at 12% of rated power. This control rod block provides the same type of protection in the R$/Hf and SMfG//H$f M/fMy ($dif as the APRH flow-biased before rod block does in , prevents control rod withdrawal a scram is the reached.pdf, R// -

i.e. wg7mg

-a, rovides local protection of the core, i.e., the The RBM rod block funct*on p///M in a local region of the core for a single preventionofff/ffffff/f4 rod withdrawal error from a JM///M //i'Z/$f /A5 ///////. The trip point is flowsbiased. The worst-case single control rod withdrawal error is analyzed for each reload to assure that, with the specific trip settings, rod withdrawal is blocked before the MCPR reaches the fuel cladding integrity

//ftflIH11 3.2/4.2-9 Amendment No. 114

1 I

INSERT FOR TECHNICAL SPECIFICATION PAGE 3.2/4.2-9 The specification requires with the number of OPERABLE CHANNELS less than required by the minimum OPERABLE CHANNELS per TRIP SYSTEM requirement, either the CHANNEL (s) is placed on the trip condition or the associated system which it activates is declared inoperable and the ACTION statements of Specification 3.5 govern.

l

QUAD-CITIES DPR-29 Below 30% power, the worst-case withdrawal of a single control rod without g* rod block action will not violate the fuel cladding integrity ///////Pff//.

hus_the RBM rod block function is not required below this power level.

The IRM_ Prod _ block function provides local as well as gross core protection. The scaling arrangement is such that the trip setting is less than a factor of 10 above-the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the MCPR fuel cladding integrity

/MHf Mtt.

at A downscale indication on an APRM is an indication +5the instrument has failed or is not sensitive enough. In either cas ythe instrument will not respond to changes in control rod motion, and the control rod motion is thus prevente'd. The downscale trips are set at 3/125 of full scale.

x. seu +han ce squal is 1 100 CPS and the detector 4not ful pinserted assures Q a The SRM rod Rock that the SRrs cre not withdrawn from the core prior to commencing rod j withdrawal for startup. The scram discharge volumg)higk wateC% vel' block Thalarmsetpointhasbeen provid$

selected annunciation to provide adequatefor operator time toaction.Meurmination allo of the cause of for-the level increase and corrective action prior to automatic scram initiation.

'For effective emergency core cooling M mall pipe breaks,the HPCI system must function since reactor pressure-does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The automatic pressure rel_ief function is-provided as a backup to the HPCIg n i the event the HPCI does not operate.. The arrangement of the tripping contacts is.such as to provide this function when necessary and minimize spurious operation. The

. trip; settings given in-the specification are-adequate to assure the above .

criteria are met (reference SAR Section 6.2.6.3). The specification preserves-the effectiven'ess of the system during periods of maintenance, testing or calibration and also minimizes the risk of inadvertent operation, i.e. . only-one inttrument channel oucofeservice.

Two radiation monitors are provided on the refueling floor which initiate isolation of the reactor building and operation of the standb as treatment l systems. The trip logic is onesouteofetwo. Trip settings of 100 mR/hr for L - the monitors on the refueling floor are based upon initiating ormal ventilation isolation and standby gas treatment system operation so that none of the-activity released during the refueling accident le ives the reactor building via the normal ventilation stack but that all the activity is processed by the standby gas treatment system.

l less than or equa.I +o 3.2/4.2-10 Amendment No. 114 p

INSERT FOR TECHNICAL BPECIFICATION PAGE 3.2/4.2-10 If a peripheral control rod is selected, the neutron leakage is sufficiently high such that withdrawal of this rod will not violate the fuel cladding integrity SAFETY LIMIT. Thus, the RBM function is not required for withdrawal of peripheral control rods.

\

=

QUAD-CITIES t DPR-29 The instrumentation which is provided to monitor the post a'ccident condition is listed in Table 3.2-4. The instrumentation listed an( the // W T#

/M////8# /M //MM/M on these systems ensure adequate monitoring of the containment following a loss-of-coolant accident. Information from this instrumentation will provide the operator with a detailed knowledge of the can conditions make logicalresulting decisionsfrom regarding the accident; post @ccident based on this informatiorg+he efemlor recovery.

The specifications allow for post 5ccident instrumentation to be out-of-service for a period of 7 days. This period is based on the fact that several diverse instruments are available for guiding the operator should an accident occur, on the low probability of an instrument being ouieofO ervice and an accident occurring in the 7-day period, and on engineering judgment.

Thenormalsupplyofairforthecontrolroomventilationsystem/ rains"A" and "B" is outside the service building g g he event of an accident, this source of air may be required to be shu. _ , to prevent high doses of radiation in the control room. Rather than provide this isolation function i

on a radiation indiepte monitor an acciden1, i.e.installed in the %ntake air duct, signals whichfhssure, low wa

, Mgh drywell steaniline high flo G or high radiation in the reactor building ventilation

[ duct,'will cause isolation of the intake air to the control room. The above trip signals result in immediate isolation of the control room ventilation system and thus minimize any radiation dose. Manual isolation capability is also provided. Isolation from high toxic chemical concentration has been added as a result of the " Control Room Habitability _5tudy" submitted to the NRC in December 1981 in response to NUREG-0737 Item III D.3.4. As explained in Section 3 of this study, ammonia, chlorine, and sulphur dioxide detection f capability has been provided. The setpoints chosen for the control room ventilation isolation are based on early detection in the outside air supply at the odor threshold, so that the toxic chemical will not achieve toxicity limit concentrations in the Control Room.

The i editractiye liqui +end- gasecu'. ef thent-4nstraentation 4s-provided u mete 4hc rehnt of radioeet4ve--meteeiels- in liquid-end geseevs ef fluents

/ dtteing relents. The alarm setpoints for the instruments are provided to f ensure that the alarms will occur prior to exceeding the limits of 10 CFR 20.

ihe mstrumentadion f.ha+ monifca 4.he release. cf gesses from 6:he }

main e.ondenser of ga.s system provides trasenable anurance.Oc ihat the efluenh to unresfncted areas will no/ exceed guda n c. ef a cruset .Lo. )

drywell hlyh radiaHan, high radiaHch at the refu. cling loor monHans, y 3.2/4.2-11 Amendment No. 114 l!

QUAD-CITIES

-DPR-29 4.2 SURVEILLANCE REQUIREMENTS BASES T e instrumentation listed in Table 4.2-1 will be functionally tested anf ca ibrated at regularly scheduled intervals. Although this instrument 4 tion is et generally considered to be as important to plant safety as the/

reac r protection system, the same design reliability goal of 0.99 9 is genera ly applied for all applications of one-out-of-two taken twi logic.

Therefe e, on-off sensors are tested once every 3 months and bist le trips associat with analog sensors and amplifiers are tested once eek.

Those instr ents which, when tripped, result in a rod block ave their contacts arr ged in a one-out-of-n logic, and all are capa e of being bypassed. For such a tripping arrangement with bypass ca 111ty provided, there is an opt um test interval that should be maintai d in order to maximize the rel bility of a given channel (Reference . This takes account of the fac that testing degrades reliability The optimum interval between test is app ximately given by:

i= -r where:

1= the optimum interval bet en tests, t= thetimethetripcontactsaedigabledfromperformingtheirfunction while the test is in progress 96d r= the expected failure rate of h relays.

To test the trip relays requir that t channel be bypassed, the test made, and the system returned o its ini al state. It is assumed this task requires an estimated 30 mi tes to comple e in a thorough and workmanlike manner and that the relays /iave a failure r te of 10 8 failures per hour.

Using this data and the aFove operation, the ptimum test interval is:

i = 42(0.M =1x1 .iours 10 8

= s 40 days For addition margin a test interval of once per month 11 be used initially.

The sens/s and electronic apparatus have not been included re, as these are ana)og devices with readouts in the control room, and the ensors and elect nic apparatus can be checked by comparison with other li ins uments. The checks which are made on a daily basis are ade ate to as re operability of the sensors and electronic apparatus, and t test ipf.erval given above provides for optimum testing of the relay circ ts.

3.2/4.2-12 Amendment No. 114

QUAD-CITIES OPR-29 e above calculated test interval optimizes each individual channel, c sidering it to be independent of all others. As an example, assume that the e are two channels with an individual technician assigned to each, ach tech dcian tests his channel at the optimum frequency, but the two techna ians are not allowed to communicate so that one can advise th other that hi channel is under test. Under these conditions, it is pos ble for both cha els to be under test simultaneously. Now assume that t technician are required to communicate and that two channels at never tested a t same time.

Forbidding sim taneous testing improves the availability of the smtem over that which woul be achieved by testing each channel indep dently. These one-out-of-n trip ystems wili be tested one at a time in order to take advantage of this herent improvement in availability.

Optimizing each chann independently may not truly timize the system considering the overal rules of system operation, powever, true system optimization is a comple problem. The optimums te broad, not sharp, and optimizing the individual hannels is generally equate for the system.

The formula given above mini zes the unavail ility of a single channel which must be bypassed during esting. The inimization of_the unavailability is illustrated b curve 1 Figure 4.2-1, which assumes that a channel has a failure rate of 1 x 10 hour and 0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is required to 3 test it. The unavailability is a int at a test interval i, of 3.6 x 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

If two similar channels are used i a e-out-of-two configuration, the test interval for minimum availability hang as a function of the rules for testing. The simplest case is test ea h one independent of the other.

In this case, there is assumed o be a fin} e probability that both may be bypassed at one time. This c se is shown b curve 2. Note that the unavailability is lower, as xpected for a re ndant system, and the minimum occurs at the same test i erval. Thus, if th two channels are tested independently, the equat n above yields the tes interval for minimum unavailability.

A more usual case is hat the testing is not done in ependently. I" both

! channels are bypas d and tested at the same time, th result is shown in l- curve 3. Note th the minimun. occurs at about 40,000 ours, much longer than for Cases and 2. Also, the minimum is not nearly s low as Case 2, l which indicate that this method of testing does not take ull advantage of the-redundan channel, Bypassing both chanrels for simulta eous testing should be a oided.

l

-The most ikely case would be to stipulatt that one channel be passed, tested and restored, and then immediately following the second annel be bypas ed, tested, and restored. This is shown by curve 4. Note t at there is t true minimum. The curve does have a definite knee, and very little l

l re uction in system unavailability is achieved by testing at a short ihterval than computed by the equation for a single channel.

3.2/4.2-13 Amendment No. 114 L

This mon /hly calibnt&n/fanclurul fr.S/ jfrquerxy hq ben

- ettended ib paarterlyby Hre analy.scs con / aided m NE/X - stM QUAD-CITIES for trip anif.s in and/9 DPR-25 q fnf ysterns.

bU UTbb bEDb pFUbCUUIC U G kIIV h E eN ne h h p2 h y bNh teGie. Thi; is, if the test-4fitefd--(; 4 200ths7-tefrt enE Gid he-ether ch:nnel: every 2 months. Th!=s Js-shov+ in c'rve 5. The differeare between G:::: 4 :nd 5 i negM 4Mer---Thew-r,y4e-ether-argument:,

9 beweveythat more strengly-support the perfenly- neggered-test +r-4nehding-feduct4 ens-in IlWHISII CI I W3 e 4tre-tenclusion; te4e-4rawn-are-thes+:

a. A :n:-cut of-n :ystem ::y be treated the sem as e single channal '

in-terms-of chec;4c; : test interval,

t. " ore then :n; hafincl :h;uld-not be byp:ssed-for testing-st -any--

. sed , . . .- s. w. _ ,

It IU) nstruments X-263-73A & B HPCI high steam flow Reactorwater/ vel f(1).2-2389A-D instruments have been 2-2352&J-2353,andHPCIsteamlinelowpressureins modified to be analog trip systems. The analog trip system consists of an analog sensor (transmitter) and a master / slave trip l unit setup which ultimately drives a trip relay. The frequency of calibra- .i tion and functional testing for instrument loops of the analog trip system I

'has been established in Licensing Topical Report NED0-21617-A (December 1978).

With the one-out-of-two Caken5Gice logic, NED0-21617-A states that each trip. unit be subjected to a calibration / functional test frequen;y of one

<~ month.+ An adequate calibration / surveillance test interval for the transmitter is once per operating cycle.

nanchor wntrol room venHlanon syskrn iso /aHen, The radiation nonitors in the ventilation duct and on the refueling floor which initiate building isolation,and standby gas treatment operation are arranged in two one-out- oftawo logic systems. The bases given above for the rod blocks apply here a' so and were used to arrive at the functional testing frequency. y gg, Based on experience at Dresden Unit I with instruments of similar 'esign, a testing interval of once every 3 months has been found to be adequate.

The automatic pressure relief instrumentation can be considered to be a o -

one out-of-two logic system, and the discussion above applies to it also.

The instrumentation which is required for the posthecident condition will be tested and calibrated at regularly scheduled intervals. The basis for the calibration and testing of:this instrumentation is the same as was discussed above for the reactor protection sy!, tem and the emergency core cooling systems.

Eef;T;n;;$

1. B. Ep>icio and A. Strtfi, "'mprUving Avaiiebility and Readiness of Field Equigs t Tin uus Per ivdic Iu=peuuun", UCRL-L M51, Lam ence Radiad un l hebetetcry, p 10 Equ: ten (20, Aly 15,1968 i

L 3.2/4.2-14 Amendment No. 114 L _ -

~

. . _ _ _ _ ._ ____..._ _ _.__ _ m_ . . _ _ . _ _ . _ - _ _ _ . _ , _ _ _ _ . _ _ _ . . . _ _ _

INSERT FOR TECHNICAL SPECIFICATIONS PAGE 5.2/4.2=14 Surveillance requirements for the instrumentation in Technical Specification Section 3.2/4.2 are nelected in order to demonstrato proper function and operability. '4 ho surveillance intervals can be determined by different moano, such as 1) operating experience, 2) good engineering judgement, 3) reliabli. sty analysen, or 4) other analyses that are found acceptable to the NRC.

l l

1 .

1

QUAD-CITIES OPR-29 TABLE 3.2-1 INSTRUMENTATION THAT INITIA.YES PRIMARY CONTAINMENf ISOLATION FUNCTIONS Minimum Number of Operable or Tripped Instrument channels [1] Instruments Trio Level Settino Actioni?1 4 Reactor low water [5] >144 inches above top of A active fuel

4 Reactor low low water >B4 inthes abov. top of A irtive fuel

4 High drywell pressure [5] 12.5 prig [3] A 16 High flow main steamline[5] s140% of rated steam flow B 16 High temperature main 1200 F B steamline tunnel 4 Hiqh radiation main <15 x r:ormal rated power B steamline tunnel [6] Sackground (without hydrogen addition) 4 Low main steam pressure [4] >825 psig B 2 High flow RCIC steamline <300% of rated steam C Tiow[7]

4 RCIC turbine tuaa high 1170 F C temperature 2 High fiow HPCI steamline <300% of rated steam 0 llow[7]

4 HPCI area high temperature 1170 F 0 6 Nctes

[1] Whenever primary containment integrity is required, there shall be two oprable or tripped systems for each function, except for low pressure main steamline which only need be available in the Run position.

3.2/4.2-15 Amendment No. 121

QUAD C111ES DPR-29 (2) Action, if the first column cannot be met for one of the trip systems, that trip system shall be tripped.

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

A. Initiate an orderly shutdown and have the reactor in Cold Shutdown condition in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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

C. Close, isolation valves in RCIC system.

D. Close isolation valves in HPCI subsystem.

[3] Need not be nperable when primary containment integrity is not required.

[4] The isolation trip signal is bypassed when the mode switch is in Refuel or startup/ Hot Standby.

[5] Tne instrumentation also isolates the control room ventilation system.

[6] This signal also automatically closes the mechanical vacuum pump discharge line isolation valves.

[7] Includes a time delay of 3 1 t 1 9 seconds.

Top of active fuel is defined as 360" above vessel zero f or all water levels used in the LOCA analysis (see Bases 3.2).

3.2/4.2-16 Amendment No. 114

-_.m_.___..___._. -

.- m_

QUAD-CITIES DPR-29

-TABLE 3.2-2 INSTRUMENTATION lt!AT INITIATES OR CONTROLS THE CORE AND CONTAINMENT C00L]NG SYSTEMS $

Minimum Number of Operable or Tripped '

Instrument Trip Function Trip Level Setting Channels [1] Remarks 4 Reactor low low >84 inches above 1. Inconjunctionwithlow-water level Top of active reactor pressure initiates ,

fuel

core spray and LPCI.

2. In conjunction with high- c drywell pressure 120 second time delay and low-pressure core cooling interlock initi-ates auto blowdown.

3. Initiates llPCI and RCIC.

4. Initiates starting of diesel

-generators.

4 I43 High-drywell 12.5 psig 1. Initiates core spray, LPCI, pressure [2),[3] HPCI, and SBGTS.

2. In conjunction with low low water level, 120-second time delay, and low-pressure core cooling interlock i.nitiates auto bicwdown.

3. Initiates starting of diesel generators.

4. Initiates isolation of controi roum ventilation.

s 2 Reactor low - 300 psigsp1350 psig 1. Permissive for opening core pressure spray and LPCI admission valves.

2. In conjunction with low low

'* reactor water level initiates ,

core spray and LPCI -

Containment spray Prevents inadvertent operation interlock of containment spray during accident conditions.

2 2/3 core height ~ >2/3 core hei ht 4 containment - U.Spsigip11.hpsig-high pressure 2 -Timer auto <120 seconds Inconjunctionwithlowlow reactor water level, high-drywell blowdown pressure, and low pressure core

Top-of active fuel is defined at 360" above cooling' interlock initiates auto

-vessel zero for all water levels used in the blow-down.

LOCA analysis 3.2/4.2-17 Amendment No. 114 i

(

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

QUAD-CITIES DPR-29 TABLE 3.2-2 (Cont'd)

Minimum Number of Operable or Tripped Instrument Channels [1] Trip Function Trip Level Setting Remarks 4 Low pressure 100 psigsp5150 psig Defers APR actuation pending core cooling confirmation of low pressure pump discharge core 1:ooling system pressure operation.

2/BVS[5] Undervoltage on 3045 in volts 1. Initiates starting of emergency buses diesel generators.

2. Perm'ssive for starting ECCS pumps.

3. Removes nonessential loads from buses.

4. Bypasses degraded voltage timer.

2/ BUS (5) Degraded 3840 volts 12% Initiates alarm and picks up Voltage with 5 15% minute time delay relay. Diesel on 4 KV time delay and 7 i Generator picks up load if Emergency 20% second inherent degraded voltage not Buses time delay corrected after time delay.

NOTES

[1] For all positions of the reactor mode selector switch (except for the containment interlock) whenever any ECCS subsystem is required to be operable, there shall be two operable trip systems. If the first column cannot be met for one or both of the trip systems, the systems actuated shall be declared iimperable and Specificatior.s 3.5 or 3.9 shall govern.

[2] Need not be operable when primary containment integrity is not required.

[3] If an instrument is inoperable, it shall be placed (or simulated) in the tripped condition so that it will not prevent containment spray.

.[4] There are a total of eight high drywell pressure sensors. Four are used for core spray and LPCI initiation, and four are used for HPCI and auto blowdown initiation. This specification applies to each set of four sensors.

[5] With the number of operable channels one less than the total number of channels, operation may proceed until performance of the next required functional test, provided the inoperable channel is placed in the tripped cundition within one hour.

3.2/4.2-18 Amendment No. 114

QUAD CITIES DPR 29 TABLE 3.2-3 INSTRUMENTATION THAT INITIATES ROD BLOCK Minimum Number of Operable or Tripped Instrument Channels per Teip System [1] Instrument Trip Level Setting 2 APRM upscale (flow bias)(7) < FRP

~(0.58WD + 50)FUTFD (2) 2 APRM upscale (Refuel and 112/125 full scale Startup/ Hot Standby mode)

-2 APRMdownscale(7) 33/125 full scale 1 Rod block monitor upscale (flow (10] l bias)[7]

1 Rodblockmonitordownscale[7] 33/125 full scale 3 IRMdownscale[3][8] 33/125 full scale -

3 IRM upscale (8) 1108/125 full scale 2(5) SRM detector not in Startup 32 feet below core centerline position (4) 3 1RM detector not in Startup 12 feet below core centerline position (8) 5 2[5](6) SRM upscale 110 counts /sec 2

2(5) SRM downscale (9) 110 counts /sec 1 (per bank) High water level in scram 1 25 gallons (per bank) discharge volume (50V) ,

1 SDV high water level scram NA trip bypassed l

l u

i-3.2/4,2-19 Amendment No. 120 l

L

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

i QUAD ClTIES '

DPR-29 !

TABLE 3.2-3 (Con't)

Notes (1) For the Startup/ Hot Standby and Run positions of the reactor mode selector switch, there shall be two operable or tripped trip systems for each function except the SRM rod blocks.

IRM upscale and IRM downscale need not be operable in the Run position, APRM downscale, APRM upscale (flow biased), and RBM downstale need not be operable in the Startup/ Hot Standby mode. The RBM upscale need not be operable at less than 30% rated thermal power. One channel may be bypassed rod pattern above does not exist. 30% rated thermal power provided that a limiting control For systems with more than one channel per trip system, if the first column cannot be met for one of the two trip systems, this condition may exist for up to 7 days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than 7 days the system shall be tripped, if the first column cannot be met for oath trip systems, the systems shall be tripped.

(2) Wn is the percent of drive flow required to produce a rated core flow of 98 mTilion lb/hr. Trip level setting is in percent of rated power (2511 MWt).

(3) 1RM downscale may be bypassed when it is on its lowest range.

(4) This function is bypassed when the count rate is > 100 CPS. ,

(S) One of the four SRM inputs may be bypassed.

[6] This SRM function may be bypassed in the higher IRM ranges (ranges 8, 9, and

10) when the IRM upscale rod block is operable.

(7) Not required to be operable while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed SWt, (8) This Startup/ IRM function Hot Standby occurs position. when the reactor mode switch is in the Refuel or

[9] This trip is bypassed when the SRM is fully inserted.

(10) The Rod Block Monitor upscale setpoint shall be established as specified $n the CORE OPERATING LIMITS REPORT. j 3.2/4.2-20 Amendment No. 120

1 l

l QUAD CITIES DPR-29 TABLE 3.2-4 I23 POSTACCIDENT MONITORING INSTRUMENTATION REQUIREMENT $

Minimum Instrunient Number of Readout Operable Chan- Location Number nels(1) (3) Parameter Unit 1 Provided Range 1 Reactor pressure 901-5 1 0-1500 psig 2 0-1200 psig 1 Reactor water level 901-3 2 *243 inches +57 inches 1 Torus water temperature 901-21 2 0-200'F 1 Torus air temperature 901 21 2 0 600'F ,

Torus water level 901 3 1 -5 inches +5

. indicator inches (narrow range) 2(6) Torus water level 901-3 2 0-30 feet indicator (wide range)

Torus water level 1 18 inch range sight glass (narrow range) 1 Torus pressure 901-3 1 -5 inches Hg to 5 psig 2 Drywell pressure 901-3 1 -5 inches Hg to 5 psig

-10 inches Hg to 70 psig 2 0 to 250 psig 2- Drywell temperature 901-21 5 0-600'F 8

2 Neutron monitoring 901-5 4. 0.1-10 CPS ,

2 I43 Torus to drywell 2 0-3 psid differential pressure 1[8] Drywell Hydrogen 901-55, 56 2 0-4%-

. concentration 2

I73 Drywell radiation 901-55, 50 2 1 to 100 R/hr monitor 3.2/4.2-21 Amendment No. 114 i , . _ _ , . . . . - . ~ . , . . _ . _ . , , _ . . . . _ . . . , . , _ , . . _ , _ . . . , _ . , , . . _ . . . . _ . _ . _ _ . _.

QUAD-CITIES DPR-29 1ABLE 3.2-4 (Cont'd)

Minimum Instrument Number of Readout Operable Chan- Location Number nels[1] [3] Parameter Unit 1 Provided Range Main Steam RV posi- 901-21 1 per NA tion, acoustic monitor valve 2/ valve ($3 Main Eteam RV post- 901-21 1 per 0-600*F

, tion, temperature monitor valve Main Steam SV posi- 901-21 1 per NA tion, acoustic monitor valve 5)

Main Steam SV posi- 901 21 1 per 0-600'F tion, temperature monitor valve Notes (1) Instrument channels required during power operation to monitor postaccident conditions.

(2) Provisions are made for local sampling and monitoring of drywell atmosphere.

[3] In the event any of the instrumentation becomes inoperable for more than 7 .

days during reactor operation, initiate an orderly shutdown and be in the cold shutdown condition witnin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . See notes 4, 5, 6, 7, and 8 for exceptions to this requirement.

[4] from and after the date that one of these parameters is reduced to one indication, continued operation is not permissible beyond thirty days unless such instrumentation is sooner made operable. In the event that all indication of these parameters is disabled and such indication cannot be restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in twenty-four (24) hours.

(5) -If the number of position indicators is reduced to one indication on one or more valves, continued operation is permissible; however, if the reactor is in a cold shutdown condition for longer than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , it may not be started up until all position indication is restored. In the event that all position indication is lost on one or more valves and such indication cannot be restored in 30 days, an orderly shutdown shall be initiated, and the reactor shall be depressurized to less than 90 psig in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.2/4.2-22 Amendment No. 114

i i

QUAD CITIES :

DPR 29 TABLE 3.2 4 (Cont'd)

6. From and after the date that this parameter is reduced to either one narrow-range-indication or one wide-range indication continued reactor -

operation is not permissible beyond 30 days unless such instrument is sooner made operable. In the event that either all narrow-range indication or all wide range indication is disabled, continued reactor operation is not permissible beyond 7 days unless such instruments are sooner made operable.

in the-event that all indication for this parameter is disabled, and such ;

indication cannot be restored in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

- 7. With less than the minimum number of operable channels, initiate the pre planned alternate method of monitoring this parameter within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

c. Either restore the inoperable channel (s) to operable status within 7 days of the event, or r

b. Prepare and submit a special report to the NRC within 30 days following the event, outlining the action taken the cause of the inoperability, -

and the plans and schedule for restoring the system to operable st3tus.

-8. From and after the date that one of the drywell hydrogen monitors becomes .

inoperable, continued reactor operation is permissible.

as If both drywell hydrogen monitors are inoperable, continued reactor operation is permissible for up to 30 days provided that during this time the HRSS_ hydrogen monitoring capability for the drywell is operable,

b. If-all drywell hydrogen monitoring capability is lost,-continued reactor operation is permiss<ble for up to 7 days.

V 4

i l

l i

I 3.2/4.2-23 Amendment No. 114 b.egm...,-, , , - . , . A yr , . . _ , . . , , , ,,w_.,y y..,p 4,..,.w., g, .mm,,v y y __ , , , w..,,,,_n%..g,g %y%._,,,, . , ,,,,_5.,,,,,p_, ,;%-,

QUAD CITIES !

DPR 29 i TABLE 3.2 5 j RADI0 ACTIVE LIQUID EfTLUENT MONITORING INSTRUMENTATION !

Minimum No. ;

cf Operable Total No.

Channels of Channels Parameter Action 1 1 Service Water A Effluent Gross-Activity Monitor l 1- 1 Liquid Radwaste C Effluent Flow ,

Rate' Monitor i

1 1 Liquid Radwaste B Effluent Gross ;

Activity Monitor ,

,,No te s Action A: With less than the minimum number of operable channels, releases via this pathway may continue, provided that at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months grabsamplesarecollectedandanajyzedforbetaorgammaactivityat an LLD of less than or equal to 10 uCi/ml.

Action B: With less than' the minimum ~ number of operable channels, ef fluent releases via this pathway may continue, provided that prior to 1 initiating a release, at least 2 independent samples-.are analyzed in .

accordance with Specification 4.8.B.1, and at least 2 members of the~ ?

facility staff independently verify thJ release calculation and ,

,~

+ discharge valving. -Otherwise, suspend release of radioactive- ;

effluents via this pathway.

Action-C: _ With less than the minimum number of operable channels,' releases via this pathway may' continue, provided the flow rate is estimated at .

least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves may be utilized to estimate flow.

t l-3.2/4.2-24 Amendment No. 114 o

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

QUAD CITIES DPR-79 TABLE 3.2-6 RADIOACTIVE GASEOUS EFFLVENT MONITORING INSTRUMENTATION Minimum No.

ofOperab)g3 Total No.

Channels L of Channels Pa ameter Action 1 2 SJAE Radiation D Monitors 1 2 Main Chimney Noble A Gas Activity Monitor 1 1 Main Chimney lodine C Sampler 1 1 Main Chimney C Particulate Sampler *

,1 1 Reactor Bldg. Vent B Sampler flow Rate Monitor 1 1 Reactor Bldg. Vent C lodine Sampler 1 1 Reactor Bldg. Vent C Particulate Sampler 4

1 1 Main Chimney Sampler B Flow Rate Monitor 1 1 Main Chimney Flow B Rate Monitor 1 2 Reactor Bldg. Vent E Noble Gas Monitor 1 1 Main Chimney F High Range Noble. P Gas Monitor Notes

[1] For SJAE monitors, applicable during SJAE operation. For other instrumentation, applicable at all times.

[2] Action A: With the number of operable channels less than the minimum requirement, effluent releases via this pathway may continue, provided grab samples are taken at -least once per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months shif t

( and these samples are analyzed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l l 3.2/4.2-25 Amendment No. 114 l

QUAD-CITIES l DPR 29 ;

TABLE 3.2-6 (Cont'd) ;

Action B: With the number of operable channels less than the minimum required, efflunt releases via this pathway may continue provided that the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Action C: With less than the minimum channels operable, effluent releases l via this pathway may continue provided samples are continuously collected with auxiliary sampling equipment, as required in Table ;

4.8-1.

l Action D: With less than the minimum channels operable, gases from the main !

condenser off gas system may be released to the environment for i up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least one chimney monitor is operable; )

otherwise, be in hot stand-by in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Action E: With less than the minimum channels operable, immediately suspend release of radioactive effluents via this pathway. l Action F: With less than the minimum channels operable, initiate the l preplanned alternate method of monitoring the appropriate l parameter (s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

(1) Either restore the inoperable channel (s) to operable status "

i witnin 7 days of the event, or (2) Prepare and submit a Special Report to the Commission within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to operable status.

1 1

3.2/4.2-26 Amendment No. 114

QUAD-CITIES DPR-29 1ABLE 4.2-1 MINIMUM TEST AND CAllBRATION FREQUENCY FOR CORE AND CONTAINMENT COOLING SYSlEMS INSTRUMEN1ATION, R0D BLOCKS, AND ISOLATIONS[7)

Instrument Instrument functional Instrument Channel Test [2] Calibration (2) Check [2]

ECCS Instrumentation

1. Reactor low-low water level [1] Once/3 months Once/ day

2. Drywell high pressure 1:1) 0,ce/3 months None

3. Reactor low pressure ;1) Once/3 months None

4. Containment spray interlock

a. 2/3 care height [1] [10] (10) None

b. Containment pressure [1] Once/3 months None

5. Low pressure core cooling L1)

Once/3 months None pump discharge

6. Undervoltage 4-KV essential Refueling outage Refueling outage None buses

7. Degraded voltage Refueling outage Once/ month 4-KV essential buses Refueling]

outage [8 Rod Blocks

1. APRM downscale [1] (3) Once/3 months None 2, APRM flow variable (1) L3) Refueling outage None

3. IRH upscale (5) 13) (5) [3] None

4. IRM downstale [5] ;3) (5) [3] None

5. RBM upscale [1] :3] Refueling outage None

6. RBM downscale L1] :3] Once/3 months None

7. SRM upscale :5) :3) 1:5) [3] None

8. SRM detector not in startup ;5) ;3) ;6) None position

9. IRM detector not in startup [5] [6] None

- position

10. SRM downscale (5) [3] [5] [3] None

11. High water level in scram Once/3 months Not applicable None discharge volume (50V)

12. 50V high level trip bypassed Refueling outage Not applicable None Main Steamline Isulation

1. Steam tunnel high temperature Refueling outage Refueling outage None

2. Steamline high flow [1] Once/3 months Once/ day

3. Steamline low pressure [1] Once/3 months None

4. Steamline high radiation [1] [4] Refueling outage Once/ day

5. Reactor low low water level [1] [10] [10] Once/ day 3.2/4.2-27 Amendment No. 114

. QUAD CITIES OPR-29 TABLE 4.2-1 (Cont'd)

Instrument Instrument functional Instrument Channel Test [2] Calibration I23 Check (2]

RCIC Isolation Steamline high flow Once/3 months [9] Once/3 months [9] None 1.

Turbine area high temperature Refueling outage Refueling outage None 2.

3. Low reactor pressure Once/3 months Once/3 months None HPCI Isolatio'n

1. Steamline high flow (1) (9) Once/3 months None

2. Steamline area high Refueling outage Refueling outage None temperature Low reactor pressure [1] Once/3 months None 3.

Reactor Building Ventilation System Isolation and Standby Gas Treatment System Initiation Refueling floor radiation [1] Once/3 months Once/ day 1.

monitors Steam Jet Air Ejector Off-Gas Isolation

1. Off gas radiation monitors [1] (4) Refueling outage Once/ day .

Control Room Ventilation System I ulation Reactor low water level L1] Once/3 months Once/ day 1.

2, Orywell high pressure 1) Once/3 months None Main steamline high flow 1) Once/3 months Once/ day 3.

Toxic gas analyzers Once/ month Once/18 months Once/ day 4.

(chlorine, ammonia, sulphur dioxide)

Notes

[1] Initiallyongepermonthuntilexposurehours(HasdefinedonFigure4.11) are 2.0 X 10 ; thereaf ter, according to Figure 4.1-1 with an interval not less than 1 month nor more than 3 months. The compilation of instrument failure rate data may include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of Quad Cities Units 1 and 2.

[2] Functional tests, calibrations, and instrument checks are not required when ,

these instruments are not required to be operable or tripped.

i

! 3.2/4.2-28 Amendment No. 114 i

l

QUAD CITIES DPR*29 TABLE 4.2-1 (Cont'd)

(3) This instrumentation is excepted from the functional test definition. The functiontestshallconsistofinjectingasimulatedelectricsignalinto the measurement channel. <

[4] This instrument channel is excepted from the functional test definitions and shall be calibrated using simulated electrical signals once every 3 months.

[5] Functional tests shall be performed before each startup with a required frequency not to exceed once per week. Calibrations shall be performed during each startup or during controlled shutdowns with a required frequency not to exceed once per week.

[6] The positioning mechanism shall be calibrated every refueling outage.

[7] Logic system functional tests are performed as specified in the applicable section for these systems.

[8] Functional tests shall include verification of operation of the degraded voltage 5 minute timer and 7 second inherent timer.

[9] Verification of the time delay setting of 3 < t < 9 seconds shall be ~ ~

performed during each refueling outage.

l 3.2/4.2-29 Amendment No. 114 i

QUAD CITIES l DPR-29 i

TABLE 4.2-2 POSTACCIDENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Minimum Instrument Number of Readout Operable- Location Instrument Channels

Parameter Unit 1 Calibration Check 1 Reactor pressure 901-5 Once every 3 months Once per day 1 Reactor water level 901-3 Once every 3 itenths Once per day 1 lorus water 901-21 Once every 3 months Once per day temperature i Torus air 901-21 Once every 3 months Once per day temperature Torus water level 901-3 Once every 3 months Once per day indicator (narrow range) 2 Torus water level 901-3 Once every 18 months Once per 31 indicator (wide range) days Torus water level N/A None sight glass 1 Torus pressure 901-3 Once every 3 months Once per day 2 Drywell pressure 901-3 Once every 3 months Once per day 2 Drywell temperature 901-21 Once every 3 months Once per day 2 Neutron monitoring 901-5 Once every 3 months Once per day 2 Torus to drywell Unce overy 6 months None differential pressure 1 Drywell Hydrogen 901-55, 56 Once every 3 months Once per 31 concentration days 2 Drywell radlu ion 901-55, 56 Once every *** Once per 31 monitor 18 months days Main Steam RV 901 21 ** Once per position, acoustic 31 days monitor 2/ valve Main Stram RV 901-21 Once every 18 months Once per position, 31 days temperature monitor 3.2/4.2-30 Amendment No. 114

QUAD CITIES OPR 29 TABLE 4.2-2 (Con'd)

Minimum Instrument Number of Readout Operable Location Instrument Channels" Parameter Unit 1 Calibration Check

- Once per Main Steam SV 901-21 Position, acoustic 31 days t

2/ valve Main Steam SV 901-21 Once every 18 months Once per Position, 31 days temperature monitor

Instrument channels required during power operation to monitor postaccident conditions.

- Functional tents will be conducted before startup at the end of each refueling outage or af ter maintenante is perforrned on a particular safety or relief valve.

- - Calibration shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr; and a one-point calibration check of the detector below 10 R/hr with an installed or portable gamma source.

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3.2/4.2-31 Amendment No. 114

QUAD-CITIES OPR-29 TABLE 4.2-3 RADIOACTIVE LIQUID EFFLUENT MONITORING IN5TRUMENTATION SURVCit'. ANCE REQUIREMENTS Instrument functional Source Instrument Check [1] , Calibration IAN3) Test [1][2] Check [1]

Liquid Radwaste Effluent 0 R Q(7) (6)

Gross Activity Monitor Service Water Effluent 0 R Q[7] R Gross Activity Monitor Liquid Radwaste Effluent (4) R NA NA Flow Rate Monitor Notes

[1] O = once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months M = once per 31 days Q = once per 92 days R = once per 18 months S = once per 6 months (2) The Instrument. Functional Test shall also demonstrate that control room 61 arm annunciation occurs, if any of the following conditions exist, where appilcable,

a. Instrument indicates levels above the alarm setpoint.

b. Circuit failure.

c. Instrument indicates a downscale failure, 2nstreaent controls not set in OPERATE mode.

d.

[3] Calibration shall include performance of a functional test.

[4] Instrument Check to verify flow during periods of release.

(5) Calibration shall include performance of a source check.

(6) Source check shall consist of observing instrument response during a discharge.

(7) Functional test may be performed by using trip check and test circuitry associated with the monitor chassis.

3.2/4.2-32 Amendment No. 114

QUAD CITIES DPR 29 1

TABLE 4.2-4 ;

RADIDACTIVE GASEOUS EFfLVENT MON 110 RING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Calibra- Functional Source Instrument Mode (2) Check [1L tion [1][4] Test [1][3] Checkg))

Main Chimney Noble Gas B D R Q M Activity Monitor i Main Chimney Sampler B D R Q(6) NA Flow Rate Monitor Reactor Bldg. Vent Sampler B D R Q[6] NA Flow Rate Monitor Hain Chimney Flow Rate B D R Q NA Monitor Reactor Bldg Vent '8 0 R Q Q Activity Monitor SJAE Activity Monitor A D R Q R Main Chimney lodine and B D(5) NA NA NA Particulate Sampler Reactor Bldg. Vent lodine B D[5] NA NA NA and Particulate Sampler Main Chimney High Range B D(5) R Q M Noble Gas Monitor N!Lty

-C (1) D = once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months M = once per 31 days Q = once per 92 days R = once per 18 months (2) A = during SJAE operation B = at all times (3) The Instrument functional Test shall also demonstrate that control room alarm annunciation occurs, if any of the following conditions exist, where applicable:

a. Instrument indicates levels above the alarm setpoint s b. Circuit failure

c. Instrument indicates a downscale failure

d. Instrument controls not set in OPERATE mode 3.2/4.2-33 Amendment No. 114

QUAD CITIES DPR-29 TABLE 4.2-4 (Cont'd)

[4] Calibration shall include perfortnance of a functional test.

[5] Instrument check to verify operability of the instrument; that the instrument is in place and functioning properly.

[6] Functional test shall be performed on local switches providing low flow alarm.

3.2/4.2-34 Amendment No. 114

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QJAD-clTIES DPR-29 10 I I I I I 10*1 -- *'

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CURVE 1

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TESTINTERVAL(1) hours FIGURE 4.2-1 TEST INTERVAL VS. SYSTEM UNAVAILAPILITY Amendment No. 114

SIGNIFICANT HAZARDS CONSIDERATIONS AND ENVIRONMENTAL ASSESSMENT EVALUATION PROPOSED TS 3.2/4.2

' PROTECTIVE INSTRUMENTATION"

I EYALUATION. FOR BIGNIFICANT MA2 APD1_99FJilDIEATIDE Proposed Specification 3.2/4.2 Protective Instrumentation The proposed changes provided in this amendmunt requent 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 llazards consideration. The basis for our determination is documented as follows:

BASIS FOR NO SIGNIFICANT ltAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards consideration. In accordance with the criteria of 10 CFR 50.92(c) a proposed amendment to an operating licensu involves no significant hazards consideration if operation of the facility, in accordance with the proposed amendment, would not:

1) Involve a significant increase in the probability or consequences of an accident previous' avaluated, because:

a. The proposed changes to Specifications 3.2.A/4.2.A through 3.2.H/4.2.H are made to provide the user with a format that will allow quicker access to needed information and to provide concise LCO, Applicability, Action and Surveillance requirenants. The bland of requirements from the present Quad Cities Technical Specifications and later operating BWRs utilizes proven material and testing techniques.

The proposed changes to Table 3.2-1 include the addition of RCIC Steam Supply Line Low Pressure and HPCI Steam Supply Line Low Pressure isolation functions setpoints that have been a part of the original design basis for the system; however omitted from the Technical Specifications. The proposed setpoints have been evaluated to ensure that isolation will not occur prior to a decreasing reactor pressure assumed in the accident analysis of 150 psig. The remaining changes to Tables 3.2-1 through 3.2-6, and 4.2-1 through 4.2-4 do not alter any established setpoints cr reduce the minimum operable channels por trip system requirements.

Since the proposed changes are applicable for the Quad cities plant and have been utilized on other operating plants, they do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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b. The drywell high radiation isolation was added to the plant by design change in order to comply with the requirements of NUREG 0737 recommendations. The inclusion of the_ operability requirements in the technical specifications helps to ensure operability of the drywell high radiation isolation function when required to initiate a Group 2 valve isolation. Since this instrumentation was installed in accordance with design toquirements and has been operating in the plant for several years, the inclusion of operability Tequirements in the technical specifications cannot involve a significant increase in the probability or consequences of an accident previously evaluated. The requirements chosen to help ensure operability are similar to existing technical specification requirements for other similar instruments or for the same radiation sensors that perform a dual role in containment isolation and post accident monitoring.

c. The proposed changes to incorporate the Surveillance Testing Intervals and Allowed out of Service Intervals in Topical Reports HEDC-33677P, NEDC-30936P-A, and NEDC-30851P Supplement 1 do not degrade the reliability of the Isolation, ECCS or Control Rod Block systems, as demonstrated in the Topical Reports and corresponding plant specific analyses. Implementation of the extended surveillance intervals for Channel Functional and Channel Calibration Tests will not be made without factoring in appropriate drift information into the setpoint calculations. The proposed changes to the Channel check frequency increases present testing requirements. Since the changes do not degrade the reliability of the affected systems over present conditions, there is no significant increase in the probability or consequences of an accident previously evaluated.

I

d. 1ue NRC issued Generic Letter 89/01 on January 31, 389, in order to allow the technical specification provisions for the Radiological Effluent Technical Specifications (RETS) to be relocated to the Administrative Controls section of the technical specjfications (programmatic requirements) and to the ODCM or PCP (procedural details). New programmatic controls for radioactive effluents and radiological l environmental monitoring are relocated to Section 6.0 l of the technical specifications to conform to the l

regulatory requirements of 10 CFR 20.106, 40 CFR Part i 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50.

l Tho proposed changes to Section 3.2/4.2 of the Quad

Cities technical specifications relocates the l

provisions of present Tables 3.2-5, 3.2-6. 4.2-3, and l

4.2-4 and the provisions of Specifications 3.2.G/4.2.G l and 3.2.H/4.2.H to the ODCM. The present requirements

q for the SJAE are retained in the technical -

specifications in accordance with GL 89-01 guidelines.

The relocation of the present technical specification requirements is not intended to reduce the level of radiological effluent control. The proposed chango will allow programmatic controls to remain in the (

technical specifications with procedural details controlled by the controls for changos to the ODCH.

Since the proposed changos to implement CL 89/01 provisions moet NRC requirements and retain adequate programmatic controls for RETS in the technical --

specifications, the proposed change does involve a r significant increase in the probability or consequences (

of an accident previously evaluated.

i 2) Create the possibility of a now or difforont kind of accident from any previously evaluated because

a. The changes to Specifications 3.2.A/4.2.A through 3.2.H/4.2.H blend STS requirements with existing Quad Cities requirements to provide a user friendly format and presentation of requirements. The changes proposed to the Tables follow later operating BWR guidelines that are presently being utilized at those plants and have been evaluated and found acceptable for use at Quad Cities. Thorofore, the changos do not create the possibility of a now or different kind of accident from any previously evaluated.

The proposed changes to Tables 3.2-1 through 3.2-6 and 4.2-1 through 4.2-4 implement STS and later operating BWR plant Table arrangements, Notes and Actions. These _

changes are in use at plants with systems similar to those at Quad Cities. The proposed changes do not reduce the number of channels presently required to be operable and maintain operability of systems when required to perform their design function.

b. The drywell high radiation isolation function provides a closure signal to only the Gr,oup 2 isolation valves.

The addition of the operability requirements for this isolation function to the technical specifications does not affect any other accidents or transients other than primary containment isolation which is analyzed.

Therefore, the proposed changes to the technical specifications does not create the possibility of a now or different kind of accident from any previously evaluated,

c. The proposed changes to incorporate the Surveillance Testing Intervals and Allowed Out of Service Times in the Topical Reports discussed above, do not create the possibility of a new or different kind of accident from any previously evaluated because Isolation System,

_ . . . . . - - - . - - - - . - - - - - - - - - - - - - - - - - - - - - " - - - - ' ' ' ^ ' ' ~ ^ ' ^ ' ^ ^ ^ ^ ^ ^ ~ ~ ^ ^ ^ ^ ~

ECCS, and Control Rod Block System function and reliability is not degraded by these changes. No new modes of plant operation are involved. The implementation of STS and later operating plant Channel Calibration Test Frequenciew will only be made to the extent that the instrumentation drift characteristics allow the interval extensions. The change to STS shiftly 2hannel Checks will increase the frequency of l present testing requirements. I

d. The proposed changes to the Technical Specifications resulting from the implementation of Generic Letter 89/01 follow NRC guidelines for relocating RL'TS l procedural requirements to the ODCM and programmatic !

requirements to Section 6.0 of the technical specifications. Since the present level of radiological effluent monitoring is maintained by the proposed change, the proposed change does not cruate

.e possibility of a new or different kind of accident from any previously evaluated.

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

a. The changes to Specifications 3.2.A/4.2.A through 3.2.H/4.2.H implement an STS type of format while retaining the present two column layout. This two column layout has been in use at Quad Cities since initial licensing and is preferred by the majority of the technical specification users at the plant. The proposed LCO, Applicability, Actions, and Surveillance Requirements are modeled after STS requirements which have been evaluated and found to be acceptable for use at Quad Cities.

The current setpoints for the low steam pressure isolation functions of HPCI and RCIC are 90 psig and 50 psig respectively. The proposed setpoints of 100 for HPCI and 60 for RCIC are consistent with current methodologies for setpoint determinations and ensure that the systems will remain operable as assumed in the accident analysis and isolate when no longer required.

The remaining changes to the Tables in Section 3.2/4.2 follow proven STS guidelines or provisions that have been implemented at other operating BWR plants. These changes have been evaluated for use at Tuad Cities with a determination that implementation at the plant will not involve a significant reduction in the margin of safety,

b. The proposed change to add the drywell high radiation icolation function to the technical specifications completcs the listing for the Primary Containment Group 2 isolation signals. The additions to the technical

_ ._. m _ . _

specifications of this signal and the inclusion of the Reactor Building Ventilation Exhaust Duct Radiation ,

Monitor in Table 4.7-1 will help to ensure operability '

of tLese isolation functions when required and thus cannot involve a significant reduction la the margin of safety.

c. The proposed changes based on the three General Electric Topical Reports decreases testing for the affected instruments in the Isolation, ECCS, and control Rod Block Systems. Allowed out of service times are increased for the affected systems as a result of the Topical Reports analyses. However, the requested changes do not degrade the reliability of the affected systems and thus the margin of safety is preserved. The results of the topical reports have been.found acceptable for plant use by NRC SERs with the stipulation that setpoint drift over the increased testing interval be conridered it setpoint calculations. Quad cit os vill consider the additional drift in the setpoint calculations before implementing ~

, the exter.ded surveillance testing Intervals and while the extended intervals are in use for both the Channel Functional and Channel Calibration tests. The changes proposed to the Channel Check frequencies will increase '

present testing requirements. Therefore, the changes do not involve a significant reduction in the margin of ,

safety.

d. The proposed implementation of the guidelines of Generic Letter 89/01 retains required RETS programmatic controls in Section 6.0 of the technical specifications and specific procedural controls in the ODCM. As such, the present margins of safety are preserved and the changes do not involve a significant reduction in the margin of safety.

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LNVIRONMENTAL; ASSESSMENT EVALUATION PROPOSED SPECIFICATION SBCTION 3.2/4.2 PROTECTIVE INSTRUMENTATION Commonwealth Edison has evaluated the proposed amendment in accordance with the requirements of 10 CFR 51.21 and has de+ 3rmined that the aRendacat meets the requirements for categorical exclusion as specified by 10 CFR Ul.22 (c) (9) .

Commonwealth Edison has detarmined 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 amendment 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 accupational radiation exposure. The proposed amendment co Section 3.2/3.2, " Protective Instrumentation",

contains administrative changes such as including appropriate applicability statements within the specifications to define the applicability during operating modo and the required actions to be implemented in the event tnat specification cannot be net. The information is consistent with the Standard Technical Specifications or later operating plants. In aidition, some existing requirements have been updated and new .equirements added to reflect the Standard Technical Specifications er later operating plant requirements.

QC-1 / QC-2 DIFFERENCES TS 3.2/4.2

' PROTECTIVE INSTRUMENTATION' M

COMPARIGON OF UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS FOR THE IDENTIFICATION OF TECHNICAL DIFFIRENCES SECTION 3.2/4.2 PROTECTIVE INSTRUMENTATION 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 Specifications into one document. The intent of the review was not to identify any differences in presentation style (e.g. table formats, use of capital letters, etc.), punctuation, or spelling errors but rather to identify areas which the Technical Specifications are technically or administratively different.

The review of Section 3.2/4.2 " Protective Instrumenta-tion" revealed no technical differences between the Unit 1 and the Unit 2 Specifications.

Several administrative differences were identified as follows:

Eaas 3.2/4.2-3 Paragraph 1 Unit 1: ... high main ateamline flow ...

Unit 2: .. high main streamline flow ...

Page 3.2/4.2-6 aUnit n and 2.2/4.2-Sa (Unit n First Paragraph Unit 1: ... instrumentation that initiate or control...

Unit 2: ... instrumentation that initiates or controls...

Pace 3.2/4.2-7 .(Unit H and 3.2/4.2-5a M M First Paragraph Unit 1: ... used in the LOCA analyses *.

Fourth Paragraph Unit 2: ... used in the LOCA analyses (NEDO-24146A, April 1979).

First Paragraph: Unit 1: ... 504 inches above vessel zero and a 60-second...

Fourth Paragraph Unit 2: ... 504 inches above vessel zero (144 inches above top of active fuel) and a 60-second...

Asterisk Footnote Unit 1: ... NEDO-24146A...

Unit 2: No footnote since reference already incorporated in text.

Pace 3.2/4.2-11 (Unit H and Pace 3.2/4.2-8 (Unit H Third Paragraph Unit 1: ... main steamline high flow ...

Fourth Paragraph Unit 2: ... main streamline high flow ...

Ea92 3.2/4.2-13 (Unit 11 And Eage 3.2/4.2-10 (Unit 21 Fourth Paragraph Unit 1: 6 Second Paragraph Unit 2:

rate of 0.1 x 106 / hour ...

rate of 0.1 x 10 / hour ...

(Typographical error - Unit i value is the correct one)

Engg 3.2/4.2-14 (Unit 11 RDd Paco 3.2/4.2-12 (Unit 11 Second Paragraph Unit 1: ... instruments 1-263-73A&B ...

1-2352&1-2353.. 1-2389A-D ...

... calibration and functional testing ...

Eighth Paragraph Unit 2: ... instruments 2-263-73A&B ...

2-2352&2-2353.. 2-2389A-D ...

... calibration and function testing ...

Eagg 3.2/4.2-27 .{.llait 11 and Enan 3.2/4.2-16, (Unit 21 ECCS Instrumentation (Item #6):

Unit 1: Undervoltage 4-KV essential buses Unit 2: Undervoltage 4-KV essential Page 3.2/4.2-30 (Unit 11 and Paqo 3.2/4.2-18 (Unit 21 Table 4.2-2 Unit 1: No clarifying brackets used.

Unit 2:- Clarifying brackets used on torus water level

! instruments and on relief and safety valve position

instruments.

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QUAD CITIES NUCLEAR POWER STATION TECHNICAL SPECIFICATION UPGRADE PROGRAM l

PROPOSED AMENDMENT SECTION 3.9/4.9 " Auxiliary Electrical Systems"

[_ _ _ _ :

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EXECUTIVE

SUMMARY

Proposed Changes to TS- 3.9/4.9

' AUXILIARY ELECTRICAL SYSTEMS'

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EXECUTIVE

SUMMARY

QUAD CITIES TECHNICAL SPECIFICATION UPGRADE PROGRAM The Quad Cities Technical Specification Upgrade Program was conceptualized in response to lessons learned from the Dresden Diagnostic Evaluation Team inspection and the frequent need for Technical Specification interpretations. A comparison of the existing Quad Cities Technical Specification and, Standard Technical Specifications and later operating plants' Technical Specification provisions was conducted to identify potential improvements in clarifying requirements and to identify requirements which are no longer consistent with current industry practicos. The comparison review identified approximately one-hundred and fifty suggested improvements. The Technical Specification Upgrade Program was not intended to be a complete adoption of the Standard Technical specifications. Overall, the

' Quad Cities custom Technical specifications provide for a safe operation of the plant and, therefore, only an upgrade was deemed appropriate.

The comparison study revealed a mix of recommended upgrades which included the relaxation of certain existing Technical Specification requirements, the addition of surveillances, the removal-of allowances which would no longer be allowed under new plant licensing, and better definition of appropriate action requirements in the event a Limiting Condition for Operation cannot-be met. The Technical Specification Upgrade Program also implements NRC recommended line item improvements to the Technical Specifications which were issued under Generic 7,etters.

In response to-an NRC recommendation, the Unit 1 and Unit 2 Technical Specifications are combined into one document. To accomplish the combination of the Units' Technical Specifications, a comparison of the Unit 1 and 2 Technical Specifications was performed to identify any technical differences. The technical differences are identified in the proposed amendment package for each section.

The Technical Specification Upgrade Program was identified as a Station top priority during the development of Quad Cities Station's Performance Enhancement Program (PEP). The Technical Specification Upgrade Program's goal is to provide a better tool to Station personnel to implement their responsibilities and to ensure Quad Cities Station is operated in accordance with current industry practices. The upgraded specifications provide for more safe and reliable operation of the plant. The program improves the operator's ability to use the Technical Specifications by more clearly defining Limiting Conditions for operations and required actions. The most significant improvement to the specifications is the addition of equipment operability requirements during shutdown conditions.

EXECUTIVE

SUMMARY

(continued)

Proposed-Technical' Specification Section-3.9/4.9, " Auxiliary Electrical Systems" The-proposed change deletes the present Objective statement and

-provides Applicability statements within each specification similar to the STS._ The proposed Applicability statements include the Reactor Modes and other conditions for which the LCOs must be satisfied. An STS type of format-is proposed

_ while retaining the present-two column layout.

The present Normal and Emergency A-C Auxiliary Powcr specification is separated into two specifications: A.C.

Sources - Operating; and Onbite Power Distribution Systems -

Operating.- The. proposed changes invclve the incorporation of present provisions into a new format and the integration of STS guidelines where no present requirements exist or improvements are needed.

In accordance with STS guidelines, the Applicability of A.C.

. Sources - Operating is expanded to include Operational Mode 3.

Also added from STS is_an Action for_both diesel generators inoperable. STS reactor shutdown time frames are added~in the proposed Actions. Consistent'with similar. changes proposed for

.the ECC Systems in Section 3.5/4.5 and the Standby Gas Treatment System in.Section 3.7/4.7, " verification of operability" is. proposed'in lieu of " demonstration of operability"ffor the ECC_ Systems on the remaining operable diesel generator when one diesel generator becomes inoperable.

Based on-STS-guidelines,' operability of_D.C. distribution systems is added to the proposed specifications.for onsite Power Distribution Systems - Operating. Proposed Applicability

-includes present intent (Operational Modes _1 and 2) and adds OperationalLMode 3-from STS.- STS' remedial Actions associated with:re-energizing inoperable'A.C. and D.C.-systems are added inithe' proposed. Present Surveillance; Requirements are replaced with STS guidelines.

Based on STS guidelines, proposed Specification 3.9.F/4.9.F, Onsite Power Distribution Systems --Shutdown, -applicable to operational Modes 14 and 5-and when handling irradiated-fuel in

-the secondary containment,-is included as-a new requirement not presently in the. technical specifications.

'STS Operational Mode 4'and 5 provisions are added to the proposed A.C. Sources = Shutdown specifications. Included is the requirement for an offsite power line to be operable in Operational: Modes 4_and 5, and when irradiated fuel is being handled in the secondary containment. STS remedial-Action statements are added.

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Present LCO provisions are maintained in the proposed Specification 3.9.C/4.9.C, Station Batteries - Operating.

Present Applicability is expanded to include Operational Modes-1, 2, and 3. STS guidelines are used to add an Action statement to address the inoperability of.the other battery systems. STS reactor shutdown time frames are added. STS Surveillance Requirements are added to replace present daily checks in SR 4.9.C. New battery charger and battery performance discharge test surveillances are added based on STS guidelines.

Based on STS guidelines, proposed Specification 3.9.D/4.9.D,

" Station Batteries - Shutdown", applicable to Operational Modes 4 and 5 and when handling irradiated fuel in the secondary containment, is included as a new requirement not presently in the technical specifications.

The proposed Reactor Protection Bus Power Monitoring System

, specifications change the present Applicability to "at all times". Present Action and Surveillance Requirements are retained except for channel functional testing provisions where later operating BWR plant provisions are proposed for incorporation. Present setpoints-and LCOs are maintained by

the proposed changes.

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SUMMARY

- OF CHANGES Proposed TS 3,9/4.9

' AUXILIARY ELECTRICAL SYSTEMS' l

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SUMMARY

OF CHANGES Proposed Specification 3.9/4.9 Auxiliary Electrical Systems I i

This amendment package is one in a series of proposnis that will j provide improvements to the present Quad Cities Technical ;

Specifications. The summary of changes includes a general section l to describe generic changes that are applicable to more than one '

section of the technical specifications and a section which provides the changes that are page by page specific.

GENERIC CHANGES Item 1:

The present Applicability and Objective statements at the beginning of each technical specification section are being deleted. The Applicability statement is being included after the LCO statement in each individual specification.

Item 2:

Each specification is rearranged to follow an STS type of format while retaining the present two column layout. Each specification will contain an LCO, Applicability, Action and Surveillance Requirement section, as applicable.

SPECIFIC CHANGES Item 1:

Pages 3.9/4.9-1 and 3.9/4.9-2, Specification 3.9.A/4.9.A, DPR-29

a. Present Specification 3.9.A/4.9.A, Normal and Emergency A-C Auxiliary Power, is rewritten and separated into two Specifications; 3.9.A/4.9.1 A.C. Sources - Operating, and 3.9.E/4.9.E, Onsite Power Distribution Systems -Operating.

b. Proposed LCO 3.9.A for A.C. Sources - Operating is taken from provisions of present Specifications 3.9.A and 3.9.D.

c. Proposed Applicability for 3.9.A implements the intent of present Specification 3.9.A and follows STS guidelines by requiring operability in Operational Modes 1, 2, and 3. An exception to operability for diesel maintenance is taken from present Specification 3.9.E.2.

d. Proposed Actions 3.9.A.1 and 3.9.A.2 for inoperability of one or both offsite lines are taken from present Specifications

3.9.C.1 and 3.9 C.2. Proposed Actions 3.9.A.3 and 3.9.A.4 for inoperability of one or both of the Unit diesel generators are taken from present Specifications 3.9.E.1, part of 4.9.E.1 and STS provisions. STS reactor shutdown time frames are added to the Actions and require the reactor to be in at least Hot Shutdown in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in Cold Shutdown in another 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e. Proposed Surveillance Requirement (SR) 4.9.A.1 for the offsite power lines is taken from present SRs 4.9.A.2 and 4.9.A.3. Proposed SR 4.9.A.2 for the emergency diesel generators is taken from present SRs-4.9.A.1 and 4.9.D in combination with the STS guidelines regarding test duration for the' monthly emergency diesel generator runs. Proposed SR 4.9.A 3 for refueling outage interval testing of the emergency diesel generators is taken from present SR 4.9.E.2.

f. Present' Specifications 3.9.A.4.a and 3.9.A.4.b for the Unit engineered safety features 4160-volt and 480-volt buses are moved to proposed LCO 3.9.E/4.9.E, Onsite Power Distribution Systems - Operating. Added to proposed Specification 3.9.E/4.9.E are requirements for D.C. power distribtition systems based on STS guidelines.

g. Proposed Applicabi1.ity of Operational Modes 1, 2, and 3 far

-Specification 3.9.E implements present intent in 3.9.A of meeting the specification requirements before making the reactor critical.

l h. Proposed Actions 3.9.E.1 and 3.9.E.2 are taken from STS provisions.

i. Proposed SR 4.9.E is based on STS guidelines and replaces present SR 4.9.A.4.

I j. Proposed Specification 3.9.F/4.9.F,- Onsite Power Distribution l Systems - Shutdown, is added based on STS guidelines.

Item 2:

Page 3.9/4.9-3, Specification-3.9.B/4.9.B, DPR-29

a. Present Specification 3.9.B/4.9.B, Station Batteries, is rewritten in a STS. format as proposed Specification 3.9.C/4.9 C, Station Batteries - Operating. The proposed LCO is taken from present requirements in Specification 3.9.D.

b. Proposed Applicability of Operational Modes 1, 2, and 3 implements present intent of Specification 3.9.B of requiring battery operability before the reactor can be made critical.

c. Proposed Action 3.9.C.1 for inoperability of one of the two 125/250-volt battery systems is taken from present Specification 3.9.C.3 provisions. Proposed Action 3.9.C.2 it:

e taken from STS guidelines to addresses the condition where battery systems are inoperable other than one of the 125/250-volt batteries.

d. Proposed SR 4.9.C.1 is' based on STS provisions and replaces present daily checks in SR 4.9.C. Proposed SRs 4.9 C.2 and 4.9.C.3 implement the present provisions of SR 4.9.B.1 and i 4.9.B.2. Proposed SR 4.9.C.4, modified to reflect STS guidelines, implements the intent of present SR 4.9.B.3. STS provisions are incorporated in new SRs 4.9.C.5, 4.9.C.6 and 4.9.C.7 associated with battery charger testing and battery performance discharge tests.

e. Proposed Specification 3.9.D/4.9.D, Station Batteries -

Shutdown is added based on STS guidelines.

Item 3:

J Pages 3.9/4.9-3 and 3.9/4.9-4, Specification 3.9.C/4.9.C,

'DPR-29

a. The provisions of present Specification 3.9.C/4.9.C, Electric Power Availability, are rewritten anc included in other proposed specifications as detailed in this amendment request.

b. Present requirements of Specifications 3.9.C.1 and 3.9.C.2 are moved to proposed Specification 3.9.A on A.C. Sources -

Operating, as described in IteL 1 above,

c. Present requirements of Specification 3.9.C.3 are moved to proposed Specification 3.9.C on Station Batteries,

d. Present SR 4.9.C requiring daily checks of the availability status of electric power is included in proposed-Specification 4.9.A.

Item 4:

Page 3.9/4.9-4, Specification 3.9.D/4.9.D, DPR-29 The present provisions of Specification 3.9.D/4.9.D, Diesel Fuel, are incorporated into proposed Specification 3.9.A/4.9.A, A.C. Sources - Operating and Specification 3.9.B/4.9.B, A.C. Sources - Shutdown.

Item 5:

Pages 3.9/4.9-5 and 3.9/4.9-6, Specification 3.9.E/4.9.E, DPR-29

a. The present provisions of Specifications 3.9.E/4.9.E, Diesel-Generator Operability, are incorporated into proposed Specifications 3.9.A/4.9.A and 3.9.B/4.9 B as discussed

below.

b. Present Specifications 3.9.E.1, 3.D.E.2, and 4.9.E.2 are incorporated into proposed Specification 3.9.A/4.9.A, A.C.

Sources - Operating.

c. The present part of SR 4.9.E.1 requiring demonstration of the operability of ECC systems with one DG inoperable, is deleted. The part of SR 4.9.E.1 requiring immediate and daily demonstration of operability of the remaining operable Unit or Unit 1/2 diesel generator is retained and incorporated in proposed Action 3.9.A.3.

d. Present Specification 3.9.E.3 is incorporated into proposed Specification 3.9.B/4.9.B, A.C. Sources -Shutdown.

1) Proposed LCO 3.9.B is based on present provisions in 3.9.E.3 which requires one diesel generator operable.

Added to the proposed LCO is a requirement to also have one offsite power line operable in accordance with STS guidelines.

2) Proposed Applicability is Operational Modes 4, 5 and when handling irradiated fuel in the secondary containment in accordance with GTS guidelines.

3) Proposed Action 3.9.B and SR 4.9.B are based on STS guidelines.

Item 6:

Pages 3.9/4. 6, 3.9/4.9-7 and 3.9/4.9-8, Specification 3.9.F/4.9.F, DPR-29

a. The present provisions of Specification 3.9.F/4.9.F, Reactor Protection Bts Power Monitoring System, are incorporated into proposed Specification 3.9.G/4.9.C.

b. Proposed LCO 3.9.G is taken from present requirements and requires two RPS electric power monitoring channels to be operable for each inservice RPS MG set or inservice alternate power source.

c. Present Applicability is all Operational Modes except Shutdown. The proposed Applicability is "At all times" in accordance with STS guidelines,

d. Proposed Actions 3.9.G.1 and 3.9.G.2 are taken from present Specification 3.9.F.2.

e. Proposed SRs 4.9.G are taken from present provisions of SR 4.9.F and provisions from later operating BWRs. Present SR 4.9.F.1.a which requires a channel functional test at least once per 6 months is modified in accordance with later BWR

provisions to require testing each time the unit is in Cold Shutdown for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous six months.

Item 7:

Pages 3.9/4.9-9 through 3.9/4.9-10, Bases for Specification 3.9/4.9, DPR-29 The proposed changes to the Bases for Specification 3.9/4.9 are made in order to impicment the proposed changes discussed in the items above.

DESCRIPTION OF CHANGES Proposed TS 3.9/4.9

' AUXILIARY ELECTRICAL SYSTEMS'

1 DESCRIPTION OF PROPOSED AMENDMENT REQUEET Proposed Specification 3.9/4.9 Auxiliary Electrical Systems The changes proposed in this amendment request are made to 1) improve the understanding and usability of the present technical specifications, 2) incorporate technical improvements, and 3) include some provisions from later operating BWR plants.

An item by item description of the proposed changes requested is provided below. The Summary of Changes section can be referred to in order to reference back to a given change and its affected page.

GENERIC CHANGES Item 1 The present Quad Cities technical specifications contain Applicability and Objective statements at the beginning of most sections. These statements are generic in nature and do not provide any useful information to the user of the technical specifications. The proposed change will delete the Objective statement and provide Applicability statements within each specification similar to the STS. The proposed Applicability statement to be included in each specification will include the Reactor Modes or other conditions for which the LCO must be satisfied.

Item 2 The changes proposed in this item will provide an STS type of format while retaining the present two column layout. The present format does not provide a separation of LCO, Applicability, and Action requirements that are easily understood and identified.

The two column layout has been utilized at Quad Cities since initial licensing of the plant and is preferred by the plant over the single column STS layout.

SPECIFIC CHANGES Item 1 The proposed changes described in Item 1 involve the rewrite of present Specification 3.9.A/4.9.A, Normal and Emergency A-C Auxiliary Power, into two Specifications; 3.9.A/4.9.A, A.C.

Sources - Operating, and 3.9.E/4.9.E, Onsite Power Distribution Systems - Operating using an STS format. The proposed LCO for Specification 3.9.A is taken from present provisions in Specifications 3.9.A and 3.9.D. Proposed LCO 3.9.A requires two I

offsite lines, associated transformers and buses to be operable along with the Unit and Unit 1/2 diesel generators with a minimum of 10,000 gallons of diesel fuel on site. The present Applicability for the Normal and Emergency A-C Auxiliary Power sources is before making the reactor critical which corresponds to Operational Modes 1 and 2. In order to address the Hot Shutdown condition (or Operational Mode 3) concerns for required operable A.C. Sources, Operational Mode 3 is added to proposed Specification 3.9 A in accordance with STS guidelines. Present Specification 3.9.E.2 provisions, that allow 1-1/2 hours for diesel generator maintenance without declaring the diesel inoperable, are added to the proposed Applicability for 3.9.A.

Proposed Action 3.9.A.1 is based on present provisions in 3.9.C.1 and addresses the condition where one of the required offsite lines is inoperable. Proposed Action 3.9.A.2 is based on present provisions in 3.9.C.2 and addresses the condition where both of the required offsite lines are inoperable. Since present Specifications 3.9.C.1 and 3.9.C.: do not contain specific reactor shutdown time frames if the remedial steps are not met, STS reactor shutdown time frames of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to Hot Shutdown and another 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to Cold shutdown are added. Proposed Action 3.9.A.3 is based on present provisions in 3.9.E.1 and part of 4.9.E.1 to address the condition where either the Unit or Unit 1/2 diesel generator is inoperable. The part added from present 4.9.E.1 requires demonstration of operability of the remaining operable diesel generator, immediately and daily thereafter.

Present reactor shutdown provisions to initiate an orderly shutdown and to be in the cold condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is replaced with the STS provisions to be in at least Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in Cold Shutdown within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The STS reactor shutdown time frames are being used throughout the revised Quad Cities Technical Specifications. Present Specifications do not address the condition where both the Unit and Unit 1/2 diesel generators are inoperable. Proposed Action 3.9.A.4 is added from the STS to address the condition where both diesel generators are inoperable and requires verification of operability of both of the required offsite lines and restoration of one diesel generator within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the reactor is required to be shutdown. Both inoperable diesel generator - are required to be restored to operable within 7 days from initic loss or the reactor is required to be shutdown. Seven days is used as the allowed out of service period in proposed action 3.9.A.4 instead of tne STS guideline of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months since present specifications use the 7 day allowance for one diesel generator being inoperable.

Proposed Surveillance Requirements for A.C. Sources - Operating are taken from present provisions. Proposed SR 4.9.A.1 for offsite power lines is taken from present SRs 4.9.A.2 and 4.9.A.3.

Proposed SR 4.9.A.2 for the emergency diesel generators is developed from present SRs 4.9.A.1 and 4.9.D and incorporates the STS, minimum 60 minute, under load, test duration for the monthly emergency diesel generator runs. Proposed SR 4.9.A.3 for the refueling outage interval testing of the emergency diesel I

i 4

generators is taken present SR 4.9.E.2.

Present Specifications 3.9.A.4.a and 3.9.A.4.b are used in the development of proposed Specification 3.9.E/4.9.E, Onsite Power Distribution Systems - Operating. The LCO scope of this specification is defined by present provisions for A.C. power distribution, as 4150-volt buses 13-1 and 14-1 for Unit 1, and 23-1 and 24-1 for Unit 2; and, 480-volt buses 18 and 19 for Unit 1, and 28 and 29 for Unit 2. Present technical specifications do not address D.C power distribution systems. In accordance with STS l guidelines, D.C. 125-volt buses and 250-volt motor control centers '

are added. These include Unit 1, Division 1, 125-volt Turbine Building Buses 1A, 1A1, lA2 and Reactor Building Distribution j Panel 1. For Unit 2, Division 2, 125-volt Turbine Building Reserve Buses 1B, 1B1, 1B2 and Turbine Building Main Bus 2A are i included. Motor Control Centers for 250-volt D.C. that are required operable include Reactor Building MCCs 1A and 1B for Unit 1; and either Turbine Building MCC 1 for Unit 1 or Turbine Building 250-volt MCC 2 for Unit 2. The energization of the specified D.C. distribution systems provides necessary power to systems that are r~equired operable in Operational Modes 1,'2, and

3. Proposed Applicability of Operational Modes 1, 2, and 3 implements present intent of 3.9.A such that buses are energized before making the reactor critical, and STS guidelines. Proposed Actions for the Onsite Power Distribution Systems -Operating are added from STS guidelines since no present provisions exist. The proposed action allows 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to re-energize a required A.C.

distribution system or initiate shutdown of the reactor. With one of the required D.C. distribution systems inoperable, 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is allowed to re-energize the system or a reactor shutdown is required. Present SR 4.9.A.4 requires a daily check of the Unit engineered safety features _4160-volt and 480-volt buses. This present SR is replaced with the STS SR which is a seven day test for verification of energization of the required buses by checking for correct breaker alignment and proper voltage. The STS seven day test frequency in conjunction with the added STS Action provisions help to provide assurance of required onsite power distribution system operability.

Present specifications do not contain provisions for Onsite Power Distribution Systems while the plant is shutdown or while conducting handling of irradiated fuel in the secondary containment. Proposed Specification 3.9.F/4.9.F addresses these conditions and ados requirements based on STS guidelines.

Proposed LCO 3.9.F includes requirements for one A.C. and one D.C.

distribution system to be operable during Applicability of Operational Modes 4, 5 and when handling irradiated fuel in the secondary containment. Proposed Action 3.9.F.1 specifies that with less than the required systems energized, Core Alterations are suspended and handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel are stopped. Proposed Action 3.9.F.2 implements STS guidelines by providing an exception to the reactor shutdown provisions of proposed Specification 3.0.C. This proposed

exception applies only to the condition of moving irradiated fuel in the secondary containment when all fuel has been removed from the reactor vessel and the plant is not considered to be in any Operational Mode. Proposed Surveillance Requirement 4.9.F repeats the test requirements in 4.9.E for Onsite Power Distribution Systems - Operating.

Item 2 This item describes the rewrite of present Specification 3.9.B/4.9.B, Station Batteries, into proposed Specification 3.9.C/4.9.C, Station Batteries - Operating. Proposed LCO 3.9.C is based on present provisions and requires operability of the Unit 24/48-volt batteries, the two station 125-volt batteries, the two station 250-volt batteries, and one battery charger for each required battery. The proposed Applicability of Operatjonal Modes 1, 2, and 3-implements present intent of before the reactor can be made critical and STS guidelines which also includes conditions where the reactor is in the Hot Shutdown Operational Mode.

Proposed Action 3.9.C.1 is based on present provisions in 3.9.C.3 and requires that with one of the two 125/250-volt battery systems inoperable, three days is allowed for restoration and then the reactor is required to be shutdown. Proposed Action 3.9.C.2 is based on STS guidelines to address the condition where battery systems are inoperable other than those hddressed in proposed Action 3.9.C.1. Proposed Action 3.9.C.2 allows 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore the inoparable systems and then the reactor is required to be shutdown..

Proposed SR 4.9.C.1 is taken from STS guidelines and replaces present daily status checks in 4.9.C with weekly checks by verifying breaker alignment and voltage. Weekly verification of system str,tus is adequate since breaker alignment should not change and major voltage changes will be noticed by routine operator rounds and chccks if they occur within the 7 day test period.- Proposed SRs 4.9.C.2 and 4.9.C.3 are taken from present provisions in SRs 4.9.B.1 and 4.9.B.2. Proposed SR 4.9.C.4, modified to reflect STS guidelines by requiring performance of battery service tests at refueling outage intervals, implements the intent of present SR 4.9.B.3 which prescribes performance of rated load discharge to:sts. Present Quad Cities Technical Specifications do not contain surveillance requirements for battery chargers. Therefore, new SR 4.9.C.5, based on STS guidelines, is proposed in order to ensure adequate charger performance during the operating cycle. Also, new SRs 4.9.C.6 and 4.9.C.7, associated with battery performance discharge tests, incorporate STS provisions and are intended to determine overal'1 batterv legradation due to age and usage.

Present Quad Cities Technical Specifications do not contain provisions for Station Batteries during shutdown and when handling irrhdiated fuel in the secondary containment. Proposed Specification 3.9.D/4.9.D is proposed from STS guidelines in order to add necessary requirements for these conditions. Proposed LCO 3.9.D requires operability of the Unit 24/48-volt batteries, one station 125-volt battery, and one battery charger for each required battery. The proposed LCO provides the minimum level of Station Battery operability during the proposed Applicability of Operational Modes 4, 5 and when handling irradiated fuel in the secondary containment. Proposed Action 3.9.D.1 requires that with less than the required D.C. electrical power sources operable, Core Alterations be suspended and handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel be stopped. Similar to proposed Specification 3.9.F, Onsite Power Distribution Systems - Shutdown, proposed Action 3.9.D.2 is added to provide a necessary exception to the reactor shutdown provisions of proposed Specification 3.0.C. This exception will only apply when handling irradiated fuel in the secondary containment when all fuel is removed from the reactor and the plant is not is any Operational Mode.

Proposed SR 4.9.D references SR 4.9.C, Station Batteries -

Operating, for required testing.

Item 3 Item 3 describes the inclusion of present provisions in Specification 3.9.C/4.9.C, Electric Power Availability, into cther specifications proposed for Section 3.9/4.9. The Summary of Changes provides the list of present provisions and their new locations.

Item 4 Item 4 describes the inclusion of present provisions of Specification 3.9.D/4.9.D, Diesel Fuel, into other proposed Specifications in Section 3.9/4.9. The Summary of Changes lists the proposed relocation of requirements.

Item 5 Item 5 describes the inclusion of present provisions of Specification 3.9.E/4.9.E, Diesel-Generator Operability, into other proposed Specifications in Section 3.9/4.9. The Summary of Changer lists the proposed relocation of requirements. Part of present SR 4.9.E.1 is deleted which contains Action provisions with one diesel generator inoperable such that all low-pressure core cooling systems and all loops of the containment cooling modes of the RHR system associated with the operable diesel generator shall be demonstrated to be operable immediately and daily thereafter. Proposed Action 3.9.A.3 includes a provision to require operability of the ECC systems associated with the operable diesel generator but does not contain the present provision in SR 4,9.E.1 to demonstrate such operability. As described in proposed changes to Section 3.5/4.5, verification of operability of systems is an adequate determination of operability without having to actually operate the affected systems. Also, present SR 4.9.E.1 requires the operable diesel generator to be

demonstrated operable immediately and daily thereafter and this part is included in proposed Action 3.9.A.3.

Present Specification 3.9.E.3 addresses diesel generator requirements in Operational Modes 4 and 5. Proposed Specification 3.9 B/4.9.B, A.C. Sources - Shutdown, addresses the present provisions in 3.9.E.3 and adds STS guidelines. In addition to the present requirement of one operable diesel generator, proposed LCO 3.9.B also includes the STS provision for operability of one offsite power line including associated switchgear and transformers, capabic of carrying power to the Unit. Present Applicability in 3.9.E.3 is in the Cold Shutdown or Refueling mode, whenever any work is being done which has the potential for draining the vessel, secondary containment is required, or a core or containment cooling system is required. Proposed Applicability is Operational Modes 4, 5, and when handling irradiated fuel in the secondary containment. The proposed Applicability includes the conditions of the present Applicability since the specification is applicable during all operations in Operational Modes 4 or 5. Proposed Action 3.9 B is added from STS guidelines since no present provisions exist. The proposed Action requires that with less than the required A.C. electrical power sources operable, core alterations and handling of irradiated fuel in the secondary containment are suspended, operations with a potential to drain the reactor vessel and crane operations over the spent fuel storage pool whoo fuel assemblies are stored therein are stopped. In addition, the proposed Action requires that when in Operational Mode 5 with the water level in the spent fuel storage pool less than 33 feet, corrective action is required immediately to restore the required power sources to operable status as soon as practical. Proposed SR 4.9.B references the SRs in 4.9.A as required testing for the operable A.C. Sources.

Item 6 Present Specification 3.9.F/4.9.F, Reactor Protection Bus Power Monitoring System, is rewritten as proposed Specification 3.9.G/4.9.G. Proposed LCO 3.9.G is taken from present provisions that require two RPS electric power monitoring channels fer each inservice RPS MG set or alternate power source. Present Applicability of all Operational Modes except Shutdown is replaced with the STS provision of at all times. The proposed change to the Applicability ensures that at all times when an RPS MG set or alternate power source is in operation, that proper controls over frequency and voltage are maintained. Present Action provisions are retained.

Present SR 4.9.F.1.a requires a functional test of the RPS Bus power monitoring system instrumentation at least once per 6 months. This present SR is proposed to be changed to allow performance of the test in Cold Shutdown in order to avoid the potential of tripping an RPS MG set at power and the potential for inadvertent reactor scrams. This testing method has been implemented at some later operating BWRs with RPS systems similar

to Quad cities and has been demonstrated to provide adequate assurance of system operability. The remaining SRs for proposed 4.9.G implement present provisions in 4.9.F.

Item 7 The proposed changes to the Bases for Specification 3.9/4.9 are limited to those necessary to implement the proposed changes discussed in Items 1 through 6 above.

i PROPOSED TECH SPEC TS 3.9/4,9

' AUXILIARY ELECTRICAL SYSTEMS' l

QUAD CITIES UNITS 1& 2 DPR-29 & DPR-30 3.9/4.0 AUXILIARY ELECTRICAL SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS A. A.C. Sources - Operating A. A.C. Sources - Operating As a minimum, the following A.C. 1. Each of the required of fsite electrical power sources shall be power lines shall be OPERABLE: demonstrated OPERABLE by:

1. One 345-KV line, associated a. At least daily check-switchgear, and the reserve ing the status of the auxiliary power transformer 345-KV lines, associ-capable of carrying power to ated switchgear and the unit. the reserve auxiliary transformer.

2. One other 345-KV line and unit reserve auxiliary b. At least daily check-transformer capable of ing the status of.the carrying auxiliary power to additional source of an essential electrical bus power via the 4160 of the unit through the 4160 volt bus tie.

volt bus tie.

2. The required emergency

3. ine Unit diesel generator diesel generators shall be and the Unit 1/2 diesel demonstrated OPERABLE at generator each with a mini- least cnce per month by:

mum of 10,000 gallons of diesel fuel supply on site. a. Manually starting and loading each diesel APPLICABILITY : generator. The test shall continue until OPERATIONAL MODES 1, 2 and 3. A full load output has diesel generator may be inoper- been maintained for a able for a period of time not to period of at least one exceed 1\ hours for the purpose hour.

of conducting preventative main-tenance provided both of the b. Checking the diesel above required offsite power starting air compres-lines are OPERABLE and the alter- sor for operation and nats diesel generator has been its ability to re-demonstrated to be OPERABLE. charge air receivers during the monthly BCTION: generator test.

1. With one of the above re- c. Operating the diesel quired offsite lines inop- fuel oil transfer erable and providing both pumps during the mon-the Unit and Unit 1/2 diesel thly generator test.

3.9/4.9-1

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 generators are OPERABLE, d. Logging the quantity

]- continued reactor operation of diese1 fue1 is permissible. Restore at available, least two off-site lines to OPERABLE status within 7 e. Checking a sample of days. Otherwise, be in at diesel fuel for least HOT SHUTDOWN within 12 quality.

hours and in COLD SHUTDOWN within the following 24 3. The required emergency hours. diesel generators shall be demonstrated OPERABLE at

2. With both of the above re- least once during each quired offsite lines inop- REFUELING OUTAGE by simu-erable, continued reactor lating a loss of offsite operation is permissible, power in conjunction with an provided: ECCS initiation signal test on the 4160 volt omorgency

a. Both the Unit and Unit bus by:

1/2 emergency diesel generators are opera- a. Verifying deenergi-ting, and zation of the emergen-cy buses, and load

b. All core and contain- shedding from the ment cooling systems emergency buses, are OPERABLE, and

b. Verifying the diesel

c. Reactor power level is starts from ambient reduced to 40% of RA- condition on the auto-TED THERMAL POWER or start signal, energi-less, and zes the emergency buses with permanently

d. The NRC is notified connected loads, _

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the energizes the auto-situation, the precau- connected emergency tions to be taken du- loads through the load ring this period and sequencer, and oper-the plans for prompt atos for greater than restoration of 5 minutes while its incoming power. generator is loaded with the emergency otherwise, be in at least loads.

HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUT-DOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3. With either the Unit or Unit 1/2 emergency diesel genera-tors inoperable, continued 3.9/4.9-2

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

i

-1 II QUAD CITIES 1 UNITS 1 & 2-DPR-29.& DPR-30 reactor operation is permis-sible provided:

a.- All ~ the . low-pressure-core cooling and all loops-of the ment cooling - contain- mode of the RHR - system asso- l ciated with the -;

OPERABLE diesel gene- l rator are -OPERABLE, -l and ,

b' '. - The two>above required

-offsite powerf lines i are OPERABLE, and c.- With. one .-uf the re-quired diesel ~ genera- l tors- inoperable, the

0PERABLE_ diesel gene-rator shall be demons-trated to be OPERABLE!

within Lone- hour' and' .

-daily thereafter.-

' Restore the inoperable die-

.sel generator ~ to OPERABLE-status. -within -7 days.

Otherwise,- be in . at.' least.

-HOT . SHUT-DOWN- -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- fol--

lowing 24_ hours. -

4;- With :-- - both z of the Unit and

~ Unit' -1/2 . emergency . diesel generators inoperable, fverify: the OPERABILITY - of the two. above ' required

offsite power lines by performing SR 4.9 A.1 within-one houriand'at least once

{per 8 .- hou rs thereafter; restore.at'least one of the -

above' required inoperable

! diesel generators- to

OPERABLE . status- within '2 . j.

hours or be in at least-HOT 3.9/4.9-3 i-l Wh:: _ ~ . . . - . . , _ _ . _ . . _ , . _ , _ . . . _ . _ . . . . _ . . , _ _ . . _ , . _ . - . . - _ , - . . . _ _ _ . . _ - . . . . _ , . _ . , , , , - . . . . _ . - . . _

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 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 . Restore both of the above required diesel gene-rators to OPERABLE status within 7 days from time of initial loss 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 andthe in COLD SHUTDOWN within following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. A.C. Sources - Shutdown B. A.C. Sources - Shutdown As a minimum, the following A.C. At least the required A.C. elec-electrical power sources shall be trical power sources shall be demonstrated OPERABLE by parfor-OPERABLE: ming the surveillance require-One offsite power line and ments of Specification 4.9.A at 1.

associated switchgear and the EPOCified frequencies.

transformer, capable of carrying power to the Unit, and

2. One diesel generator (either the Unit or the Unit 1/2) with a minimum of 10,000 gallons of diesel fuel supply on ite.

APPLICABILITY:

OPERATIONAL MODES 4, 5 and when handling irradiated fuel in the secondary containment.

ACTION:

With less than the above required A.C. electrical power sources OPERABLE, suspend CORE ALTERA-TION',, handling of irradiated fue.

in the secondary contain-ment, operations with a potential for draining the reactor vessel and crane operations over the spent fuel storage pool when fuel assemblies are stored therein.

3.9/4.9-4

}

i QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 In addition, when in OPERATIONAL MODE 5 with the water icycl less than 33 foot in the spent fuel storage pool, immediately initiato correctivo action to restore the required power sourceu to OPERABLE status as soon as practical.

C. Station Batteries - Operating C. Station Batterios - Operating As a ninimum the following D.C. Each of the required D.C.

slectrical power cources shall be electrical power sources shall be OPERABLE: demonstrated OPERABLE:

1. The Unit 1 .8-volt 1. At least enco per week by battories. verifying correct breaker alignment and voltage.

2. Two stat.1 125-volt batteries. 2. At least once per week by meenuring:

3. Two station 250-volt batterios. a) the specific gravity and voltage of each

4. One batterv charger for each pilot cell and the required t.Sttery. temperature of adja-cent cells, and AfELLCABILIU1 b) overall battery OPERATIONAL MODES 1, 2 and 3. Voltage.

ACTIOHi 3. At lesst once por quarter by measuring the voltage of

1. With e e of the two 125/250 each cell to the nearest volt battery syatems inop- 0.01 volt, the specific erablo, cantinued reactor gravity of each cell, and operation is permissible. the temperature of overy Restore the inoperable bat- fifth cell.

tory system to OPERABLE sta*us within 3 days or bo

. 4. At least overy refueling in at least HOT SHUTDOWN outage, during shutdown, by within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and verifying that the battery in COLD SHUTDOWN within the capacity is adequate to following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , supply and maintain in OPERABLE status all of the

2. With the battery systems actual or simulated emergen- .

otherwise inoperable, cy loads for the 240 minute restore the inoperable design cycle when the 3.9/4.9-5

l l

QUAD CITIES Ul4ITS 1 & 2 DPR-29 & DPR-30 system to OPERABLE status battery 10 subjected to a '

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at battery service test.

least il0T SilUTDOWii within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in 5. At least ovory refueling COLD SIIUTDOW11 within the outago by verifying that the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 125-volt and 250-volt bat-tory chargt.ra will supply a load equal to the manufac-

turer's rating for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

6. At least ovary refueling !

outage, during shutdown, i performanco dischargo testo of battery capacity shall be -

given to any battery that shows signs of degradation or has roached 85% of the service life expected for the application.

Dogradation is indicated when the battory capacity dropo more than 10% of rated capacity from its averago on previous performanco tosto, or -is below 90% of the manufacturer's rating.

7. At least onco por 60 months, during shutdown, by verify-ing that the battery capaci-ty is at least:

a) 00% of the manufac-turor's rating for the Unit 24/48-volt batterica, and b) 80% of the manufac-turor's rating for the Station 125-volt batteries, and c) groater than or - equal to the minimum acceptable battery capacity from the latent revision of-the D.C. Electrical Load l 3.9/4.9-6

i QUAD CITIES Vi1ITS 1&2 DPR-29 & DPR-30 Honitoring System (ELMS) for the Station 250-volt batteries, when subjected to a perfor-manco discharge test. At this onco per 60 month interval, this performanco discharge test may be performed in lieu of the.

battery service tost.

D. Station Batterior - Shutdown D. Station Batteries - Shutdown As a minimum, the following D.C. At least the rcquired batteries electrical power _ sources shall be and chargerr. shall bo demons-

. OPERABIII trated OPERABLE per Surveillance i Requirement 4.9.C. ;

1. The Unit 24/48-volt batteries.

2. One station 125-volt battery.

3. One battery charger for each required battery.

APPLICABILITYt-OPERATIONAL MODES 4, 5 and when-handling irradiated fuel in the secondary containment.-

ACTION:

1

1. With less than the required l__ D.C. electrichl power sources OPERABLE, suspend CORE ALTERATIONS, handling of. irradiated fuel in the secondary containment, and operations with a potential for draining the reactor vessel.

2. The provisions of Specifi-cation 3.0.C are not applicable.

3.9/4.9-7 M. -

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 E. Onsite PoWor Distribution Systems E. Onsite Power Distribution Systems

- Operating - Operating The following power distribution Each of the required power systems shall be energized: distribution systems shall be ;

datormined energized at least '

1. A.C. power distributions onco per 7 days by verifying correct breaker alignment and

a. The Unit engincorod proper voltago.

safety features 4160 volt buses (13-1 and-14-1, Unit 11 23-1 and 24-1 Unit 2).

b. The Unit engineered safety- features 480 volt-buses (18 and 19, Unit il 28 and 29, Unit 2).

2. 125-volt D.C. power i distributions

a. Division 1, consisting !

of TB Main Bus 1A, 1A1, 1A2 ar.d Reactor :

Building Dist Pn1 1, Unit 1

b. Division 2, consisting of TB Res Bus 1B, IB1, 1B2 and TB Main Bus 2A, Unit 2.

3. 250-volt D.C. power distribut' ton for Division 2 consisting of Reactor Bldg MCCs IA and IB, Unit 1; and either TB MCC 1, Unit 1 or TB MCC 2, Unit 2.

-APPLICABILITY:

OPERATIONAL MODES'1, 2 and 3.

ACTION:

1. With one of the above required A.C distribution 3.9/4.9-P

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

i QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 systems not energized re-energize the system within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least il0T S!!UTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in C(tLD GHUTDOWN within the following 24 I hoars.

2. With one of the above i required D.C. distribution '

systems not energized, reenergize the system within

- 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least

!!OT SilUTDOWN within the next 12 hours and ' -

Cold SilUTDOWN withi, the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

F. onsite Power Distribution Syst, - F. Onsite Power Distribution Systems

- Shutdown - Shutdown As a minimum, the following power Each- of the required power ,

distribution systems shall be distribution syntoms shall be OPERABLEt determined energized at least once per 7 days by verifying

1. For A.C. power distribution, correct breakor alignment and '

at least one of the fol- proper voltage.

lowing 4160-volt buses and associated 480-volt busent

a. Unit engineered

~

The safety features 4160 volt buses (13-1 and 14-1, Unit 11 23-1 and 24-1 Unit 2).

b. The Unit engineered l cafety features 480

-Volt buses (18 and 19, Unit 1; 28 and 29, Unit'2).

2. For D.C. power distribution systems, at least one of the following Divisions of 125 volt buses

a. Division 1, consisting of 125-volt buses (TD 3.9/4.9-9

, -- , , - . _ , .,_,.,.n-a-. ,---.m.+,. .,,,,4+-.-,y.-e..,ww.. 3,wv y,w -4 ,,.,,.,v.

-,--,w-,.w,y, ,9 w7,+,.,,,,-g, e-y. --+v- s --

3.,, ,,#--w- - -w.

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 Main Bus lA, lA1, 1A2 and Reactor Bldg Dist Pnl 1, Unit 1).

b. Division 2, consisting of 125-volt buses (TB Ros Bus 1B, 181, 182 and TB Main Bus 2A, Unit 2).

APPLICABILITY OPERATIONAL MODES 4, 5, and when handling irradiated fuel in the secondary containment.

ACTION 1

1. With less than the required A.C. or D. O. power distri-bution systems enorgized, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary 'ontain-mont, and .operatir a with a potential for drt.ning the reactor vessel.

2. The provisions of Specifica-tion 3.0.C are not applicable.

G. Reactor Protection Bus Power G. Reactor Protection Bus Power Monitoring System Monitoring System Two RPS olectric power monitoring The RPS Bus power monitoring channels for each inservico RPS system instrumentation shall be MG set or inservice alternato determined OPERABLE:

power source chall bo OPERABLE.

1. By performance of a CHANNEL APPLICABILITY 1 FUNCTIONAL TEST oach time the unit is in COLD SHUTDOWN .

At all timos. for a period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unless performed ACTION: within the previous six months, and

1. With one RPS olectric power monitoring channel for an 2. At least once per OPERATING inservice RPS MG set or CYCLE by demonstrating the ]

3.9/4.9-10 1

QUAD CITIES UNITS 1&2 DPR-29 & DPR-30 inservice alternate power operability of evervoltago, source inoperablo, restore undervoltage, and undorfro-the inoperable channel to quency protective instru-OPERABLE status within 72 montation by performance of hours or remove the asso- a CllANNEL CALIBRATION inclu-ciated RPS MG set or alter- ding simulated automatic nato power source from activation of the protectivo service. relays, tripping logic, and output circutt breakers, and

2. With both RPS olectric power verifying the following monitoring channels for an sotpoints:

inservico RPS MG set or inservice alternato power a. Overvoltago source inoperable, rostore 126.5 V i 2.5%

at least one channel to Min. 123.3 V OPERABLE status within 30 Max. 129.6 V minutos, or remove the asso-ciated RPS MG set or alter- b. Undervoltage nato power source from 108 V i 2.5%

service. Min. 105.3 V Max. 110.7 V

c. Undorfrequency 56.0 Hz i 1% of 60 llz Min. 55.4 Itz

=

Max. 56.6 Hz 3.9/4.9-11

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 3.9 LIMITING CONDITIONS roR OPERATION BASES A. The general objective of this specification is to assure an adequate sourco of electrical power to operato the t.oxiliarios during plant operation, to -

operato facilitics to cool and lubricate the plant during shutdown, and to oporato the engineered safety features following an accident. There are two sourcos of electrical energy available, i.e.,

the 345-kV transmission system and the diesel generators.

The ACTION requirements specified for the levels of degradation of the power sources provido restriction upon continued facility operation commensurate with the lovel of degradation. ACTION statements have been included in the specificaticn to cover all situations where either ono AC sourco or a combination of two AC uources are inoperable.

ACTION statement 4 is intended to be followed to completion once entered and should not be exited until both AC sources are restored. The OPERABILITY of the power sources is consistent with the initial condition assumptions of the accident analyses and is based on maintaining at least one train of the onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident condit # "ns coincident with an assumed loss of offsi ) wor and single failure of one of the two onsito C. sources.

Auxiliary power for the Unit is supplied from two sources, either the Unit auxiliary transformer or the Unit reserve auxiliary transformer. Both of these transformers are sited to carry 100% of the auxiliary load. If the reserve auxiliary transformer is lost, the unit can continue to run for 7 days, since the Unit auxiliary transformer is available and both diesel generators are operational. A 7-day period is provided if one source of of fsite power is lost. This period is based on having two diesels operable which are adequate to handle an accident assuming a single failure. In addition, auxiliary power from the other unit can be obtained through the 4160-volt bus tie. If both offsite lines are lost, power is reduced to 40% of rated so that the turbine bypass system could accept the steam flow without reactor trip, should the generator be separated from the B 3.9/4.9-1

. .. __ _ - - _ . ~ _

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 system or a turbino trip occur. In this condition, the turbino-generator is capable of supplying house load and ECCS loud if necessary through the unit auxiliary transformer. If the unit were shutdown on loss of both lines, fewer sources of power would be available than for sustained operation at 40% power.

Attention will to given to restoring normal offsito power to minimize the length of time operation is allowed in a condition where both sources are unavailable.

In the normal mode of operation, the 345-kV system is OPERABLE and two diesel generators are OPERABLE.

One diesel generator may be allowed out of service for a short period of timo to conduct proventativo --

maintenanco provided that power is available from the 345-kV system through a 4160-volt bus tio to supply the emergency buses, and the alternato diesel generator is proven OPERABLE. Offsite power is quito reliable, and in thu last 25 years there has been only one instance in which all offsite power was lost at a Commonwealth Edison Generating Station. When the unit or shared dicsol generator is made or found inoperabib for reasons other than preventative maintenance, the remaining diesel generator and its associated low-pressure core cooling and containment cooling systems, which provide sufficient engineered safety features coulpment to cover all breaks are required to be OPERABLE.

The diesel fuel supply of 10,000 gallons will supply each diesel generator with a minimum of 2 days of full load operation or about 4 days at 1/2 load.

i Additional diesel fuel can be obtained and delivered to the site within an 8-hour period; thus a 2-day supply provides for adequate margin.

B. lho OPERABILITY of the minimum specified A.C. power sources during shutdown, refueling, and when handling irradiated fuel in the secondary containment, ensures that the facility can be maintained in these conditions for extended time periods and that sufficient instrumentation and control capability is available for monitoring and maintaining the unit status. Requiring OPERABILITY of the minimum specified power sources when handling irradiated fuel in the secondary containment helps to ensure that systems needed to mitigate a fuel B 3.9/4.9-2

--__ I

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 handling accAdont are available.

C. The D.C. supply is required for control and motivo power for switchgear and engincored safety features.

The electrical power required provides for the maximum availability of power, i.e., one activo offsite source and one backup source of offsito power and the maximum numbers of onsito sources.

D. Tho OPERABILITY of the minimum specified D.C. power sources, during OPERATIONAL MODES 4, 5 and when handling irradiated fuol in the secondary containment, ensures that the facility can be maintained in those conditions for extended timo periods and sufficient instrumentation and control capability is available for monitoring and maintaining the unit status. Requiring OPERABILITY of the minimum specified D.C. power sources when handling irradiated fuel in the secondary containment helps to ensure thet systems nooded to mitigate a fuel handling accident are available. ,

E. Tho OPERABILITY of the A.C. and D.C. onsito pcwor distribution systems during OPERATIONAL MODFG 1, 2, and 3 onsures that sufficient power will be available to supply the safoty related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility..

F. The OPERABILITY of the minimum specified A.C. and D.C. onsito power distribution systems, during OPERATIONAL MODES 4, 5 and when handling irradiated fuel in the secondary containmont, ensures that the facility can be maintained in these conditions for extended time periods and sufficient instrumentation and control capability is available for monitoring.

and maintaining the unit status. Requiring OPERABILITY of the minimum specified onsito power distribution systems when handling irradiated fuel in the secondary containment helps to ensure that systems needed to mitigate a fuol handling accident are available.

G. Specifications are provided to ensure the OPERABILITY of the RPS Bus electrical protection assemblies (EPAs). Each RPS MG set and the alternate power sourco has 2 EPA channels wired in series. A trip of either channel from either B 3.9/4.9-3

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

l i

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 overvoltage, undervoltago, or undorfrequency will trip the associated MG set or alternato power DOurCO.

l B 3.9/4.9-4

QUAD CITIES UllITS 1 & 2 DPR-29 & DPR-30 >

4.9 SURVEILLA14CE REQUIREMEllTS LASES A&D The monthly test of the diesel generator is conducted to check for equipment failures and deterioration. Testing is conducted up to equilibrium operating conditions to demonstrato proper operation at tnese conditions. Tho dicsol will be manually started, synchronized to the bus, and load picked up. The diosol shall bo loaded to at least half load to prevent fouling of the engine.

It is expected that the diesel generator will be run for 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Diesel onerator experience at other commonwealth Edison generating stations indicates that the testing frequency is adequato and providos a high reliability of operation should the system be required. In addition, during the test, the generator is synchronized to the offsite power sources and thus not completely indopondent of this source. To maintain the maximum amount of independone:e, a 30-day testing interval is also desirable.

Each dioscl generator. has two air compressors and four air tanks. Two air tanks are piped together to form an air receiver. Each air compressor supplies an air receiver. This arrangement provides redundancy in starting capability. It is expected that the air compressors will run only infrequently.

During the monthly check of the diesel, the receivers will be drawn down below the point at which the compressor automatically starts to check operation and the ability of the compressors to rocharge the receivers. Pressure indicators are provided on each of the receivers.

Following the monthly test of the diesels, the fuel oil. day tank will be approximately half full based -

on the 2-hour test at full load and 205 gph at full load. At the end of the monthly load test of the diesel generators,-the fuel oil transfer pumps will be operated to refill the day tank and to check the operation of these pumps from the emergency source.

The test of the emergency diesel generator during the refueling outage will be more comprehensivo in that it will functionally test the system, i.e., it will check diesel starting, closure of diesel breaker, and sequencing of loads on the diesel. The B 3.9/4.9-5

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 diesel will be started by simulation of a loss-of-coolant accident. In addition, an undervoltage condition will be impos6d to simulate a loss of offsite power. The only load on the diesel is that due to friction and Windage and a small amount of bypass flow on each pump.

Periodic tests between refueling outages verify the ability of the diesel to run at full load and the core and containment cooling pumps to deliver full flow. Periodic testing of the various components plus a functional test at the refueling interval are sufficient to maintain adequate reliability.

The diesel fuel oil quality must be checked to

-ensure proper operation of the diesel generators.

Water content should be minimized, because water in the fuel would contribute to excessive corrosion of the system, causing decreased rollability. The growth of micro-organisms results in slime formations, which are one of the chief causes of jellying in hydrocarbon fuels. Minimizing such climes is also essential to assuring high reliability, c&D Although station batteries will deteriorate with time, utility experience indicates there is almost no possibility of precipitous failure. The type of surveillance described in this specification is that -

which has been demonetrated -over the years to provide an indication of a cell becoming irregular or unserviceable long before it becomes a failure.

In addition, the checks described also provide adequate indication that the batteries have the specified ampere-hour capability. Verifying specific gravity, voltage and temperature of cells, the overall battery terminal voltage ari performing

, battery service and discharge tests ensures the l effectiveness of the charging system, the ability to l handle high discharge rates and compares the battery l

capacity at that time with the rated capacity.

E&F Decause~

the availability of_ electricity to the l system is a normal operating function, a check of the status of these systems provides adequate surveillance.

l B 3.9/4.9-6

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 -I G. Surveillanco requirements are provided for the RPS l EPA's_ to demonstrato their operability. The sotpoints for overvoltage, undervoltage, and i undorfrequency have been chosen based on analysis (ref. February 4, 1983 lotter to H. Donton from T.

Rausch).

i 5

r 4

l l

l B 3.9/4.9-7 t.:

EXISTING TECH SPEC :

+

L e

TS 3.9/4.9

' AUXILIARY ELECTRICAL SYSTEMS' ;

h i

+

f b

l 4

6 9

f a

h

_t b

T b

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

1 i

QUAD 011115 OPR 29 3.9/4.9 AUX 1LIARY [L(C1RICAL SYS1[M$

LIMITING CONDITIONS FOR OP[RA110N $URv[lLLANC[ RlQUIR[MENI5 Applicability: Appilcability:

Applies to the auxillery electrical power Applies to the periodic testing system, requirernent of the auxiliary electrical system.

Objective: Objective:

To assure an adequate supply of electri- 10 verify the operability of the auxil-cal power during plant operation. lary electrical system SPECIFICA110NS A. Normal and Emergency A-C Auxiliary A. Normal and Emergency A C Auxiliary Power Power The reactor shall not be made criti-cal unless all the following require-ments are satisfied.

1. The Unit diesel generator and 1. a. Each diesel generator shall the Unit 1/2 diesel generator be manually started and ,

shall be operable. loaded once each u nth to demonstrate operational readiness. The test shall continue until both the diesel engine and the generator are at equilibrium ,

conditions of temperature while full load output is maintained.

b. During the monthly generator test, the diesel-starting air compressor shall be checked for operation and its ability to recharge-air receivers, u

3.9/4.9-1 Amendment No. 114

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

QUAD Clil[$

DPR-29

c. During the monthly generator test the diesel fuel oil transfer pumps shall be operated.

2. One 345-kV line, associated Switchgear, and the reserve

2. The status of the 345-kV lines, associated st<lichgear, and the auxiliary power transformer reserve auxiliary power capable of carrying power to the transformer shall be checked unit shall be available, daily.

3. One other 345 LV line and unit j

3. The status of the additional reserve aux transformer capable source of power via the of carrying auxiliary power to 4160 volt bus tie shall be

' an essential electrical bus of chected daily.

i the unit through the 4160-volt bus tie shall be available.

4. a. The Unit engineered safety 4. The Unit engineered safety

' features 4160-volt buses features 4160-volt and 480 volt (13-1 and 14 1, Unit 1; 23-1 buses shall be checked daily.

and 24-1, Unit 2) are energized,

b. The Unit engineered safety features 480-volt buses (18 and 19, Unit 1; 28 and 29, Unit 2) are energized.

i l

3.9/4.9-2 Amendment No. 114

QUAD-Clill'5 DPR 29 B. Station Batteries B. Station Batteries unit 24/48+ volt batteries, two 1. Every weet the specific gravity clon 125 volt batteries, the two and voltage of the pilot cell, ation 250-volt batteries, and a the temperature of adjacent oattery charger for each required cell, and overall battery battery shall be operable before the voltage shall bt measured.

reactor can be made critical.

2. (very 3 months the measurement shall be made of the voltage of each cell to the nearest 0.01 Solt, the specific gravity of each cell, and the temperature of every fifth cell.

3. Lvery ref ueling outage, the station batteries shall be

- subjectedtoaratedload discharge test. Specific gravity and voltage of each cell shall be determined af ter the discharge.

C. Electric Power Availability C. Electric Power Availability Whenever the reactor is in the Run The availability status of electric mode or for startup from a hot shut- power shall be checked daily.

down condition, the availability of electric power shall be as specified i in Specifications 3.9.A and 3.9.B

except as stated in Specifications 3.9.C.1, 3.9.C.2, 3.9.C.3, and 3.9.E.

l i

3.9/4.9-3 Amendment No, 114

=+, -

, - - . ---nm.- -e.. , , ,. ,- , , _ w a , , -, --.a--,,- -, ,w--

QUAD Clil[5 DPR-29

1. From and after the date that incoming power is available from only one of the lines specified in 3.9.A. continued reactor operation is permissible only during the succeeding 7 days, unless the second line is sooner made available, providing both the Unit and Unit 1/2 emergency diesel generators are operable.

2. From and after the date the incoming power is not available from any line, continued reactor operation is permissible pro-viding both the Unit and Unit 1/2 emergency diesel generators are operating, all core and containment cooling systems are operable, reactor power level is reduced to 40% of rated, and the NRC is notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the situation, the precau-tions to be taken during this period,and the plans for prompt restoration of incoming power.

3. From and after the date that one of the two 125/250-volt battery systems is made or found to be inoperable for any reason, continued reactor operation is permissible only during the succeeding 3 days unless such battery system is sooner made operable.

D. Diesel fuel D. Diesel fuel There shall be a minimum of 10,000 Once a month the quantity of diesel gallons of diesel fuel supply on site fuel available shall be logged.

for each diesel generator.

Once a month a sample of diesel fuel shall be checked for quality.

l l

l 3.9/4.9-4 Amendment No. 114

QUAD ClllES OPR-29 E. Diesel Generator Operability E. Diesei-Generator Operability

1. Whenever the reactor is in the 1. When it is determined that Startup/ Hot Standby or Run mode either the unit or shared diesel and the unit or shared diesel generator is inocerable, all generators and/or their respec- low pressure cors cooling tive associated buses are made systems and all loops of the or found to be inoperable for containment cooling modes of the any reason, except as specified RHR system associated with the in Specification 3.9.E.2 below, operable diesel generator shall continued reactor operation is be demonstrated to be operable permissible only during the suc- immedia'.ely and daily ceeding 7 days provided that all thereafter. The operable diesel of the low pressure core cooling generator shall be demonstrated and all loops of the containment to be operable immediately and coolbg mode of the RHR svste.i. daily thereafttr.

associated with the operable diesel generator shall be oper-able, and two offsite lines as specified in 3.9.A are available, if this requirement cannot be met, an orderly shut-down shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. Specification 3.9.E.1 shall not 2. During each refueling outage, a apply when a diesel generator simulated loss of off-site power has been made inoperable for a in conjunction with an ECCS period not to exceed 1-1/2 hours initiation signal test shall be for the purpose of conducting performed on the 4160 volt emer-preventative maintenance. Addi- gency bus by:

tionally, preventative mainten-ance shall-not be undertaken

. unless two off site lines as specified in 3.9.A are available and the alternate diesel generator has been demon-strated to be operable.

l 3.9/4.9-5 Amendment No. 114

QUAD Clllf5 DPR 29

a. Verifying de-energization of the emergency buses, and load shedding from the emergency buses,

b. Verifying the diesel starts f rom ambient condition on the auto-start signal, energizes the emergency buses with permanently connected loads, energizes the auto connected emergency loads through the load sequencer, and operates for greater than 5 minutes while its generator is loaded with the emergency loads.

3. When the reactor is in the Cold Shutdown or Refueling mode, a ,

minimum of one diesel generator !

(either the Unit diesel genera- I tor or the Unit 1/2 diesel generator) shall be operable whenever any work is being done which has the potential for

. draining the vessel, secondary containment is required, or a core or containment cooling system is required.

F. REACTOR PROTECTION BUS POWER F. REACTOR PROTECTION BUS POWER MONITORING SYSTEM MONITORING SYSTEM

1. Two RPS electric power monitor- 1. The RPS But, power monitoring ing channels for each inservice system instrumentation shall be RPS HG set or inservice alter- determitied OPERABLE:

nate power source shall be OPERABLE except when the reactor is in the SHUTDOWN mode.

3.9/4.9-6 Amendment No. 114

QUAD-CITl[5 DPR-29

a. At least once per 6 months by performing a channel functional test, and

b. At least once per operating cycle by demonstrating the operability of overvoltage, undervoltage, and underfre-quency protective instru-mentation by performance of b channel calibration including simulated auto-matic activation of the protective relays, tripping logic, and output circuit breakers, and verifying the following setpoints:

(1) overvoltage 126.5 V 2.5%

Min. 123.3 V Max. 129.6 V (2) undervoltage 108 V = 2.5%

Min. 105.3 V P .x. 110.7 V (3) underfrequency $6,0 llz 11%

of 60 Hz Min. 55.4 ilz Max. 56.6 Itz 3.9/4.9-7 Am:ndment No. 114

QUAD-CITIES OPR-29

2. a. With one RPS electric power monitoring channel for an inservice RPS HG set or inservice alternate po'av source inoperable, restore the inoperable channel to OPPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or remove the associ-ated RPS MG set or alter-nate power source from service,

b. With both RPS electric power monitoring channels for an inservice RPS HG set or inservice alternate power source inoperable, restorc-at least one channel to OPERABLE status within 30 minutes, or remove the associated RPS MG set or alternate power source from service.

i l

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l 3.9/4,9 8 Amendment No. 114

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QUAD Cll!!S OPR-29 3.9 LIMITING CONDITIONS FOR OPERATION BAS [$

A. The general objective of this specification is to assure an adequate source of electrical power to operate the auxiliaries during plant operation, to operate facilities to cool and lubricate the plant during !

Insert A shutdown, and to operate the engineered safety features following an accident. There are two sources of electrical energy available,i.e ,

y p ly, the 345 kV transmission system and the diesel generators.

8 M3. 4 Mku.hed pwwph B -

C. 5 The d< supply is required for control and motive power for switchgear and engineered safety features. The electrical power required provides for the maximum availability of power, i.e. , one active of fsite source and one backup source of of f site power and the maximum numbers of-onsite sources.

C. ~k'u'xiliary power for the Unit is supplied f rom two sources, either the ;

Unit auxiliary transformer or the Unit reserve auxiliary transformer.

Both of these transformers are sized to carry 100% of the auxiliary-load. If the reserve auxiliary transfouer is lost, the unit can continue to run for 7 days, since the Unit auxiliary transformer is-available and both diesel generators are operational. A 7 day period

is provided if one source of offsite power is-lost. This period is based on having two diesels operable which are adequate to handle an accident assuming a single failure. -In addition, auxiliary power f rom

-the.other unit can be obtained through the 4160-volt bus tie. If both of fsite lines are lost, power is reduced to 40% of rated so that the turbine bypass syste'n could accept the steam flow without reactor trirg should the generator be separated from the system or a turbine trip ;

occur. In this condition, the turbine-generator is capable of supplying house load and ECCS load if necessary through the unit '

auxiliary transformer. If the unit were shutJown on loss of both lines, fewer sources of- power would be available than for sustained operation at 40% power. Attention will be given to restoring normal offsite power to minimize the length of time operation is allowed in a condition where both sources are3available, un

- In the normal mode of operation the 345-kV system-is /M///// and two

.. diesel generators are M // M M., One diesel generator may be allowed

(

out of service for a'short period of time to conduct preventative maintenance provided that power is available from_the 345-kV-system through a 4160-volt bus tie to supply the emer alternate diesel generator is proven #ff#1&.gency buses, and theOffsite powe reliable, and in the last 25 years there has been only one instance in which all offsite power was lest at a Commonwealth Edison Generating Station. When the unit or shared diesel generator is made or found inoperable for reasons other than preventative maintenance, the

, -remaining diesel generator and its associated low-pressure core cooling 3.9/4.9-9 Amendment No.-114

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

INSERT A FOR TECHNICAL SPECIFICATIONS PAGE 3.9/4.9-9 The ACTION rc quirements specified for the levels of degradation of the power sources provide restriction upon continued f acility operation commensurate with the level of degradation.

ACTION statements have been included in tne specification to cover all situations who. - either one AC source or a combination of two AC source. are inoperable. ACTION statement 4 is intended to be followed to completion once entered and should not be exited until both AC sources are restored. The OPERADILITl of the power sources is consistent with the initial condition assumptions of the accident analyses and is based on maintaining at least one train of the onsite A.C. and D.C.

power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of one of the two onsite A.C.

sources.

PARAGRAPH B INSERT FOR TECHNICAL SPECIFICATIONS PAGE 3.9/4.9-9 B. The OPERABILITY of the minimum specified A.C. power sources during shutdown, refueling, and when handling irradiated fuel in the secondary containment, ensures that the facility can be maintained in these conditions for extended time periods and that sufficient instrumentation and control capability is available for monitoring and maintaining the unit status. Requiring OPERABILITY of tho minimum specified power sources when handling irradiated fuel in the. secondary containment helps to ensure that systems needed to mitigate a fuel handling accident are available.

t 5

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1 QUAD-C111[$

DPR 29 are rvquired to be]

gg,e g and containeent cooling systems, which provide ufficient engineered safety features equipment to cover all breaks, 41' be pro m ////////.

g. The diesel fuel supply of 10,000 gallons will supply each diesel generator with a minimum of 2 days of full load operation or about 4 days at 1/2 load. Additional diesel fuel can be obtained and delivered to the site within an 8-hour period; thus a 2-day supply provides for adequate margin.

Miesel-generatoe-operability-4s-d4secessed-in-Paragraph-3-9-C above.

C; f. Specifications are provided to ensure,_1he /////////t*/ of the RPS Bus electricalprotectionassemblies(EP/Fs). Each RPS HG set and the alternate power source has 2 EPA channels wired in series. A trip of either channel from either overvoltage, undervoltage, or underf requency will trip the associated MG set or alternate power source.

% 0, tt. e f Insert _per altat. hec) 3.9/4.9-10 Amendment No. 114

I!1 SERT FOR TECilllICAL SPECIPICATIOllS PAGE 3.9/4.)-10 D. The OPERABILITY of the minimum specified D.C. power soutCos, during OPERATIC 11AL MODES 4, 5 and when handling irradiated fuel in the secondary containment, ensures that the f acility can bo rnaintained in those conditions for extended timo periods and sufficient instrumentation and control capability is availablo for monitoring and maintaining tho unit status. Requiring OPERABILITY of the minimum specified D.C. powor sourcos when handling irradiated (uol in the secondary contai mont healps to ensure that systems needed to mitigato a fuel handling accident are available.

E. The OPERABILITY of the A.C. and D.C. onsito power distribution systems during OPERATIOliAL MODES 1, 2, and 3 onsures that suf ficient power will be available to supply the safety related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of acciuent conditions within the facility.

F. The OPERABILITY rf the minimutn specifiod A.C. and D.C.

onsite power distribution systems, during OPERATIO!1AL MODES 4, 5 and when handling irradiated fuel in the secondary containment, ensures that the facility can be maintained in those conditions for exterided timo periods and sufficient instrumentation and control capability is available for monitoring and maintaining the unit status.

Roqujring OPERABILITY of the minimum speci.fied onsite power distribution systems when handling irradiated fuel in the secondary containment helps to ensure that systems needed to mitigato a fuel handling accident are available.

- - _ _ _ _ _ _ . _ _ _ _________._m __ _-_.__m_- . _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ - _ _ _ . - _ - _ - - _ _ a._,

QUAD-CITIES OPR-29 4.9 SURVEILLANCE REQUIREMENTS BASES K The monthly test of the diesel generator is conducted to eneck for A ' & S. equipment failures and deterioration. Testing is conducted up to equilibrium operating conditions to demonstrate proper operation a'.

these conditions. The diesel will be manually started, synchronized to the bus, and loed picked up. The diesel shall be loaded to at least half load to prevent fouling of the engine, it is expected that the diesel generator will be run for 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Diesel generator experience at other Commonwealth Edison generating stations indicates that the testing frequency is adequate and provides a high reliability of operation should the system be required. In addition, during the test, thr: generator is synchronized to the of fsite power sources and thus not completely independent of this source. To maintain the maximum arount of independence, a 30-day testing interval is also desirable.

Each diesel generator has two air compressors and four air tanks. Two air tenks are piped together to form an air receivet. Each air compressor supplies an air receiver. This arrangement provides redundancy in starting capability, it is expected that the air compressors will run only infrequently.

During the monthly check of the diesel, the receivers will be drawn down belw the point at which the compressor automatically starts to chcck operation and the ability of the compressors to recharge the

, receivers. Pressure indicators are provided on each of the receivers.

Following the monthly test of the diesels, the fuel oil day tank will be approximately brl' full based on the 2-hour test at full load and 205 gph at full load. ? the end of the monthly load test of the diesel generators, Ehe ael oil transfer pumps will be operated to refill the day tank and to check the opt ration of these pumps from the emergency source.

The test of the emergency diesel generator during the refueling outage vill be h; ore comprehensive in that it will functionally test the system, i.e., it will check diesel starting, closure of diesel breaker, and sequencing of loads on the diesel. The diesel will be started by rimulction of a loss-of-coolant accident. In addition, an undervoltage

. x ndition will be imposed to simulate a hesstof th tim r%e red. The only load on the diesel is that due to friction and windage and a small amount of bypass flow on each pump. offSH e Po N -

Periodic tests between refueling outages verify the ability of the diesel to run at full load and the core and containment cooling pumps to deliver full flow. Periodic testing of the various components plus a functional test at the refueling interval art sufficient to maintain adequate reliability.

l % Although station batteries will deteriorate with time, utility experience l C. 4 D. indicates there is almost no possibility of precipitous failure. The i tyoe of surveillance described in this specification is that which has je been duonstrated over the years to providt an indication of a cell becoming irregular or unserviceabic long before it becomes a failure.

3.9/4.9-11 Amendment No, 114 i

i

QUAD-CITIES DPR-29 In addition, the checks described also provide adequate indication that -

hrt 8 % the batteries have the specified ampere-hour capability,

c. ( F. f. Because the availability of electricity to the system is a normal-operating function, a check of-the si 'us of these systems provides adequate surveillance, y3eyg )L The diesel' fuel oil T'al4+y must c .ecked to ensure proper operation of the diesel generator .. Water content should be minimized, because 70 water in the fuel would contribute to excessive corrosion of the

3. 9. A system, causing decreased reliability. The growth of micro-organisms results in slime formations, which are one of the chief causes of jellying in hydrocarbon fuels. Minimizing-of such slimes is also essential to assuring hign reliability.

1 jteselynereter Operetd-Hty sgrveillence is discussed-in-Paragraph

.a.n ...,..

Q f.

f Surveillance requirements are provided for the RPS EPAO f to demonstrate their oparability. The setpoints for overvoltage, undervoltage, and-underfrequency have been chosen based on-analysis (rei. February 4, 1983-letter to H. Denton from T. Rausch).

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3.9/4.9-12 Amendment No. 114

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

INSERT B FOR TECHNICAL SPECIFICATIONS PAGE 3.9/4.9-12 ,

)

1 Verifying specific gravity, voltage and temperature of cells, i the overall battery terminal voltage and performing battery service and discharge tests ensures the offectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated !

capacity.

4 -.4.-.-,e,,,,_ =mm. ,. me D. 4 - A -n--

l '.

. SIGNIFICANT HAZARDS CONSIDERATIONS

-AND ENVIRONMENTAL ASSESSMENT EVALUATION

' PROPOSED TS 3.9/4.9

' AUXILIARY ELECTRICAL SYSTEMS' 4

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d 4

4 r- -

-, -- * = =-- -

RVALUATION FOR SIGNIFICANT HAZARDS CONSIDERATION Proposed Specification 3.9/4.9 Auxiliary Electrical 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 ave been reviewed by Commonwealth Edison and we believe that the do not present a Significant Hazard. Consideration. The basis for our determination is documented as follows:

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION Common salth Edison has evaluated this proposed amendment and determined that it involves no significant hazards consideration.

In accordance with the criteria of 10 CFR 50.92(c) a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility, in accordance with the proposed amendment, would not:

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

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

b. The proposed changes to the present Normal and Emergency A-C Auxiliary Power-specifications include the separation into two specificatio:.s, ; C. Sources -

Operating and Onsite Power Distributjsn Systems -

Operating. Tha proposed changes involve the incorporation of present provisions ir.co a new format and the integration of S"S guidelines where no present requirements exist or improvements need to be made.

Proposed A.C. Sources - Operating 5pecifications use present provisions for LCO requirements. The proposed LCO requires operability of two independent offsite 345-kV lines with transmission ca= ability to the Unit and operability of both the Unit and Unit 1/2 diesel generators. Present intent is for Applicability in Operational Modes 1 and 2, which is retained, and expanded to include Operational Mode 3 in accordance with STS guidelines. Proposed Actions based on present provisions include those for one or both of the required offsite lines inoperable and those for one diesel generator inoperable. Added from STS guidelines is an

r- 1 i

Action for both diesel generators inoperable. STS reactor shutdown time frames of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to Hot Shutdown and another 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to Cold Shutdown are used in the proposed Actions. These proposed reactor shutdown time frames are being implemented throughout the Quad Cities Technical Specifications and have been proven in use at operating BWRs with systems similar to those at Quad Cities. Since present Actions for one or both offsite lines being inoperable do not contain reactor shutdown time frames, reliance on the general provisions of present 3.0.A would result in the same shutdown time frame requirements as proposed from the STS. Proposed Action for one diesel generator inoperable does not contain the present provision to demonstrate the operability of the ECC systems on the remaining operable diesel generator. Verification of operability of the ECC systems is retained and is adequate to assure ECC systems are available without actually running the systems. This change in testing requirements follows similar changes proposed for tne ECC Systems in Section 3.5/4.5 and the Standby Gas Treatment System in Section 3.7/4.7. Proposed SRs for A.C. Sources - Operating implement present provisions. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Proposed Specifications for Onsite power Distribution Systems - Operating are based on present provisions and STS guidelines. Present provisions requiring operability of Unit engineered safety features 4160-volt and 480-volt buses are used in the proposed LCO. Added to the present provisions are requirements, based on STS guidelines, for operability of D.C. distribution systems. The proposed Applicability is operational Modes 1, 2, and 3.which includes present intent and adds operational Mode 3 by STS guidelines. Since no remedial Actions exist in the present specifications, STS guidelines are used which allow 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to-re-energize the inoperable A.C. system and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to re-energize the inoperable D.C. system or the reactor is required to be shutdown. -Present Surveillance Requirements involve a daily check of Unit engineered safety features buses.

The present SR is replaced with STS guidelines for a weekly test for verification of the energization of required buses by checking for correct breaker alignment and proper voltage. The proposad changes maintain necessary operability of the required electrical buses and implement STS guidelines where none presently exist or where improvements in present requirements can be realized. Since the present level of Unit engineered safety features buses is maintained by the proposed changes, the changes do not involve a significant increase in the probability or conaequences of an

e accident previously evaluated.

The addition of Specification 3.9.F/4.9.F, Onsite Power Distribution Systems - Shutdown, represents new

~}

requirements that are not presently in the technical specitications. Proposed Specification 3.9.F/4.9.F is based on STS guidelines and requires operability of necessary A.C. and D.C. power distribution systems while in Operational Mode s 4, 5 and when handling irradiated fuel in the secondary containment. Proposed Actions require suspension of Core Alterations, handling of irradiated fuel in the secondary containment, and operations with a potential for draining the reactor vessel when required A.C. or D.C systems are not energized. The proposed exception to the provisions of specification 3.0.C only applies to the handling of irradiated fuel in the secondary containment when all the fuel has been removed from the reactor thereby resulting in the plant not being in any Operational Mode. Proposed Surveillance requirements repeat those for Onsite Power Dfstribution Systems - Operating. The proposed addition of Specification 3.9.F/4.9.F represents rerarictions that are not prese;.tly in the technical specifications and helps to orovide assurance of the availability of necessary onsite power distribution systems when requir:pd. Therefore, the proposed addition of these specifications cannot involve a significant increase in the probability or consequences nf an accident previously evaluated.

c. Proposed Changes to A.C. Sources - Shutdown Specifications Present provisions for A.C. Sources while the plant is in Operational Modes 4 or 5 are not complete and need STS provisions to be added. The STS provisions added include requiring, in addition to the present provision of one diesel generator, an offsite power line to be operable while the plant is in Operational Modes 4, 5, and when irradiated fuel is being handled in the secondary containment. STS actions are needed since there are no present remedial Action statements. The added Action provisions will ensure that without the required A.C. Sources, core alterations, handling of irradiated fuel in the secondary containment, operations with a potential for draiaing the reactor vessel, and crans operations over the spent fuel storage pool when fuel assemblies are stored therein, are suspended. The proposed Actions also require restoration of the inoperable A.C. Sources as soon as practical when in Operatienal Mode 5 and the spent fuel storage pool water level is less than 33 feet. The Surveillance Requirements for A.C. Sources - Shutdown are the same as for A.C. Sources - Operating. The proposed changes

retain the intent of the present provisions and add restrictions on operations that are not currently in the Specifications. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated,

d. Proposed Changes to the Station Batteries Specifications Present LCO provisions are maintained for proposed Station Batteries - Operating Specification, 3.9.C/4.9.C, and require operability of the Unit 24/48-volt batteries, two station 125-volt batteries, two station 250-volt batteries, and one battery charger for each required battery. Present Applicability of l battery operability before the reactor can be made i critical, is retained and expanded to include )

Operational Modes 1, 2, and 3. Present Actions for the l 125/250-volt batteries are retained and STS guidelines are used to add an Action provision to address the inoperability of the other battery systems. STS reactor shutdown time frames of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to Hot Shutdown and another 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to Cold Shutdown are added to the Actions in order to implement consistent, proven shutdown provisions throughout the Quad Cit ies Technical Specifications. STS surveillance requirements are added !

to replace present daily checks of availability of electric power in SR 4.9.C with weekly verifications of breaker alignment and voltage. New battery charger surveillance requirements, based on STS guidelines, are proposed in order to ensurc adequate charger performance during the operating cycle. Also, new surveillance requirements incorporating STS provisions and associated with battery performance discharge tests, are intended to determine overall battery degradation due to age and usage. The proposed changes enhance present provisions-and help to ensure operability of the Station Batteries when they are required to be operable. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed addition of Specification 3.9.D/4.9.D, Station Batteries - Shutdown, represents requirements that do not presently exist in Quad Cities Technical Specifications. These new provisions are based on STS guidelines and require operability of minimum D.C electrical power sources in Operational Modes 4, 5 and when handling irradiated fuel in the secondary containment. Proposed Actions require suspension of Core Alterations, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel, when less than the required D.C. systems are operable. Also added is an e:meption to the provisions of Specification 3.0.C.

This exception will apply only to handling of irradinted fuel in the secondary containment during the condition where all fuel is removed from the reactor vessel and

'-] the reactor is in no Operational Mode. Propored Surveillance Requirements for 4.9.D reference the SRs for 4.9.C, Station Batteries - Operating. The proposed addition of Specification 3.9.D/4.9.D adds restrictions not presently in the Quad Cities Technical Specifications and helps to ensure operability of necessary D.C. power sources in the shutdown condition.

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

e. Proposed Changes to tae Reactor Protection Bus Power Monitoring System Specifications Present provisions for RPS Bus Power Monitoring are improved with STS and later operating plants' requirements. The present LCO requirements are retained and require two instrumentation channels for each inservice RPS MG set or inservice alternate power source. The present Applicability of except when the reactor is in the Shutdown Mode is changed to "At all times." The proposed change in the Applicability requires proper controls to be maintained over frequency and voltage any time a RPS MG set or alternate power source is in operation. Present Action and Surveillance Requirements are retained except for channel functional testing pr0 Visions. Present SRs require a channel functional test at least once per six months.

Performance of this channel functional test when the reactor is operating increases the potential for inadvertent reactor trips. Some later operating BWR plants have modified this testing provisicn to allow the '

channel functional test to be performed each time the reactor is in Cold Shutdown for a period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unless the test has been performed in the last six months. Quad cities has evaluated this proposed change to the channel functional testing frequency and found it to be accentable for use based on system design similarities with later operating plants and demonstration by use at the later operating plants that necessary system operability is maintained. Present setpoints and LCOs are maintained by the proposed changes. The changes add more restrictive Applicability requirements and modify SR requirements to help avoid inadvertent reactor trips while maintaining necessary system operability. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2) create'the possibility of a'new or different kind of accident from any previously evaluated because:

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. The proposed changes for Quad Cities Technical Specification Section 3.9/4.9 are based on present.

provisions and STS guidelines or later operating BWR plants' changes which have received NRC approval. These proposed changes have been reviewed-for acceptability at

-the_ Quad cities Nuclear Station considering similarity of system or component design versus the STS or later operating BWRs. No new modes of operation are introduced-by the proposed changes, considering the acceptable Operational Modes in present specifications, the STS, or later operating BWRs. The proposed changes do not modify _ existing setpoints or design-ansumptions for system or component operation. Surveill ance requirements are changed to-reflect-improvements in technique,_ frequency of performance or operating experience at later plants. Proposed changes to Action statements in many places add requirements that are not in the present technical specifications or adopt requirements that have been used successfully at other operating BWRs with designs similar to Quad Cities. The

-proposed changes either maintain at-least the present.

level-of operability.or adopt relaxations to present requirements.which still-provide a proven. acceptable level of_ operability. Therefore, the proposed changes do not create the possibility of a new or different kind of accident.from any previously evaluated.

3) Involve a significant reduction in-th margin of safety

, because:

a. Due to the administrative nature of the Generic Changes, they do not-involve a significant reduction in the margin of safety,

b. The proposed changes to Technical-Specification Section 3.9/4.9 implement present requirements, the intent of present requirements, or provisions that have been found

~

acceptable for use on other operating BWRs with system

-designs similar to that at Quad' Cities. The proposed changes are intended to improve readability, usability,

-and the understanding of: technical specification requirements while maintaining acceptable levels'of safe operation. The proposed' changes-have been evaluated and found to be acceptable for use at Quad Cities based on system design, safety analysis requirements and operational performance. Since the proposed changes are

' based on NRC accepted provisions at other operating plants that are applicable at Quad cities and maintain necessary levele of system, component or parameter operability, the proposed changes do not involve a significant reduction in the margin of safety.

ENVIBONMENTAL ASSESSMENT EVALUATION PROPOSED SPECIFICATION SECTION 3.9/4.9 AUXILIARY ELECTRICAL 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 amendment 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 preposed amendment to Section 3.9/4.9, " Auxiliary Electrical 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 be met. The information is consistent with the Standard Technical Specifications or later operating plants. In addition, some existing requirements have been updated and new requirements added to reflect the Standard Technical Specifications or later operating plant requirements.

l 1

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QC-1/ QC-2 DIFFERENCES TS 3.9/4.9

' AUXILIARY ' ELECTRICAL SYSTEMS' i

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1 CONPARISON OF UNIT 1 AND UNIT 2 TBCHNICAL-SPECIFICATIONS

'FOR THE IDENTIFICATION OF TECHNICAL DIFFERENCES SECTION 3.9/4.9 AUXILIARY ELECTRICAL SYSTEMS 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 Specifications into one document. The intent of the review was not to : identify any differences in presentation style-(e.g. table formats, use of capital letters, etc.), punctuation,uor spelling errors but rather to identify areas which the Technical Specifications are

. technically or administratively different.

The review of Section 3.9 Systems" revealed the followin/4.9 " Auxiliary Electricalg administrative differenca:

Pace 3.9/4.9-2 4.9.B.2 Unit 2: each cell to the nearest Unit.2: each cells to the nearest 4

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ML20079G104

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