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AITACHMENT B

MARKED-UP TSUP PAG~ DRESDEN NUCLEAR POWER STATION

LICENSE NOS. DPR-19 AND DPR-25

\.

I I

·- -

TABLE OF CONTENTS TOC

•• LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION

3/4.6 PRIMARY SYSTEM BOUNDARY

3/4.6.A Recirculation Loops ................... · ................ . 3/4.6-1

3/4.6.B Jet Pumps ...... · ...................... : ............. . 3/4.6-3

3/4.6.C Recirculation Pumps· ................................... . 3/4.6-5

3/4.6.D Idle Recirculation Loop Startup ........................... . 3/4.6-6

3/4.6.E Safety Valves ....................................... . 3/4.6-7

3/4.6.F Relief Valves ........................................ . 3/4.6-8

3/4.6.G Leakage Detection Systems ............................. . 3/4.6-10

3/4.6.H Operational Leakage . ~ ............................ , .... . 3/4.6-11

• 3/4.6.I Chemistry .......................................... .

Table 3.6.1-1, Reactor Coolant System Chemistry Limits

3/4.6-13

3/4.6.J Specific Activity ..................................... . 3/4.6-16

3/4.6.K Pressure/Temperature Limits . : ........................... . 3/4.6-19

Figure 3.6.K-1, Minimum Reactor Vessel Metal Temperature vs. Rx.Vessel Pressure

3/4.6.L

3/4.6.M

• 3/4.6.N

3/4.6.0

3/4.6.P

Reactor Steam Dome Pressure ........................... . 3/4.6-22

Main Steam Line Isolation Valves ......................... . 3/4.6-23

Structural Integrity .................................... . 3/4.6-24

Shutdown Cooling - HOT SHUTDOWN ...................... . 3/4.6-25

Shutdown Cooling - COLD SHUTDOWN ..................... . 3/4.6-27

DRESDEN - UNITS 2 & 3 VIII Amendment Nos.

•

•


LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION PAGE

3/4.10 REFUELING OPERATIONS

3/4. 1 O,A Reactor Mode Switch ................................ : ..

3/4.1 O.B Instrumentation •......................................

3/4. 1 O.C · Control Rod Position ..... . ............................ .

3/4.10.D ~· .. . ~~:.~:~~ ............ , .. .. · ..... _ ..

3/4.1 O.E Communications ........•............. · ................ .

3./4.1 O.F . DELETED ................ : ......................... .

3/4. 1 O.G· Water Level - Reactor Vessel .......... ; .................. .

3/4.10.H Water Level - Spent Fuel Storage Pool ...................... .

3/4.10.1 Single·Control Rod ~emoval .................. '. ........ ~ ..

3/4.-10.J Multiple Control Rod Removal ....................... .- .... .

3/4. 1 O.K Shutdown Cooling and Coolant Circulation -High Water Level ...... .

3/4.19.L Shutdown Cooling and Coolant Circulation - Low .water Level ...... .

3/4.10-1

3/4.10-3

3/4~ 10-5

3i4.10-6

3/4.10-7

3/4.10-8

3/4.10-9

3/4.10-10

3/4.10-11

3/4'.10-13

3/4..10~1s

3/4.10-16

DRESDEN - UNITS 2 & 3 XII Amendment Nos.

TABLE OF CONTENTS B TOC

•• BASES

SECTION


3/4.10.A Reactor Mode Switch .................................. . B 3/4.10-1

3/4.10.B 1.nstrumentation ....................................... . B 3/4.10-1

3/4.10.C Control Rod Position ....................•...... ; . ; ...... . B 3/4.10-2

3/4.10.D . . ·~· - -ET~~ . (Decay Tome ~ ~ ..................... ·: ................•. B 3/4.10-2

3/4.1 O.E CommunicatiQns B 3/4.10-2

3/4.1 O.F DELETED ... ·. , . , .................................. , . B 3/4.10-2

3/4.10.G Water Level - Reactor Vess~I . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .B 3/4.10-2

3/4.10.H Water Level - Spent Fuel Storage Pool ..................... , . B 3/4.10-2

•• 3/4.10.I

3/4.1 O.J

Single Control Rod Removal ............. ~ ............... ;· B 3/4. 10-3

Multiple ControLRod Removal .............................. · B 3/4.10-3

3/4.1.0.K Shutdown Cooling and Coolant Circulation - High Water Level . . . . . . . B 3/4. 10-3

3/4.10.L Shutdown Cooling and Coolant Circulation - Low Water Level B 3/4.10-3

•

DRESDEN - UNITS 2 & 3 xxv Amendment Nos.

Definitions 1 .0

,.'. 1.0 DEFINITIONS

MINIMUM CRITICAL POWER RATIO {MCPR) . The MINIMUM CRITICAL POWER RA TIO (MCPR) shall be the smallest CPR which exists in the core.

OFFSITE DOSE CALCULATION MANUAL (ODCM) The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm!Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Specification 6.8 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and 7i e Radioactive Effluent Release Reports required by Specification 6.9.

,A11l'f"l"a. \

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its specified safety function(s) are also capable of performing their related support function(s).

·.~:· .PERATIONAL MODE

An OPERATIONAL MODE, i.e., MODE, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1-2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1 ) described in Chapter 14 of the UFSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission. ·

PRESSURE BOUNDARY LEAKAGE .

• ):• - f

PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system cqmponent body, pipe wall or vessel wall •

DRESDEN - UNrTS 2 & 3 1-4 Amendment Nos.

~-,, •• \. l .• • ..

0 TABLE 4.1.A-1 ::0 :0 m m )> en REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

() 0 -i m 0 z ::0

c Applicable CHANNEL "'tJ ::0

z OPERATIONAL CHANNEL FUNCTIONAL CHANNEL1'1 0 =i Functional Unit MODES

-i Ul CHECK TEST CALIBRATION m

()

N :j !20 0 VJ 1. Intermediate Range Monitor: z

E''ff Ul

Neutron Flux - High s1b1 -< a. 2 s1u1c1, w101 en

-i 3,4, 5 s w101 Elorl m

s:: b. Inoperative 2, 3,4, 5 NA w101 NA

2. Average Power Range Monitor111:

w Setdown Neutron Flux - High 2 s1b1 - a. ~

_. .31

51m1

s,s~ I

-..J b. Flow Biased Neutron Flux - High 1 w w1d.•>, SA

c. Fixed Neutron Flux - High 1 . s w w1d1, SA

d. lnop.erative 1, 2, 3, 5iml NA w NA

3. Reactor Vessel Steam Dome Pressure - High 1, 2111 NA M Q

~

)> 4. Reactor Vessel Water level - Low 1, 2 D M E'h'

3 (l)

::J 1, 2lPI NA M a. 5. Main Steam Line Isolation Valve - Closure E 3 (l)

::J ...+ 1, 2111 s M Elql ::0

z 6. Main Steam Line Radiation - High "'tJ en

0 en w

Drywell Pressure - High 1, 21n1 NA M Q -7. ~ _.

)>

.,. ":. .

•

•

REACTOR PROTECTION SYSTEM RPS 3/4.1.A

TABLE 4. 1 .A-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

TABLE NOTATION

(a) Neutron detectors may be excluded from the CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least ( % ) decades during each startup after entering OPERATIONAL MODE 2 and the IRM and APRM channels shall be determined to overlap for at least ( % ) decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Within 24 hours prior to startup, if not performed within the previous 7 days. The weekly CHANNEL FUNCTIONAL TEST may be used to fulfill this requirement.

(d) This calibration shall consist of the adjustment of the APRM CHANNEL to conform, within 2% of RATED THERMAL POWER, to the power values calculated by a heat balance during OPERATIONAL MODE 1 when THERMAL POWER is ~25% of RATED THERMAL POW.ER. This adjustment must be accomplished: al within 2 hours if the APRM CHANNEL is indicating lower power values than the heat balan.ce, or bl wi.thin 12 hours if the .APRM CHANNEL is indicating higher power values than the heat balance. Until any required APRM adjustment has been accomplished, notification shall be posted on the reactor control panel.

Any APRM CHANNEL gain adjustment made in compliance with Specification 3.11.B shall not be included in determining the above difference. This calibration is not required when THERMAL POWER is < 25% of RATED THERMAL POWER. The provisions of Specification . 4.0.D are not applicable.

(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal. DE"L£ T.f"O .

(h) Trip units are calibrated at least once per 31 days and transmitters are calibrated at the frequency identified in the table .

(i) This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.12.A.

(j) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.10.1 or 3.1 O.J.

(k) This function may be bypassed, provided a control rod block is actuated, for reactor protection system reset in Refuel and Shutdown positions of the reactor mode switch.

DRESDEN - UNITS 2 & 3 3/4.1-9 Amendment Nos.



• REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

• •••

(I) This function not required to be OPERABLE when THERMAL POWER is less than 45% of RA TED THERMAL POWER.

(m) Required to be OPERABLE only prior to and during required SHUTDOWN MARGIN demonstrations performed per Specification 3.12.B.

(n) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(o) The provisions of Specification 4.0.D are not applicable to the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for a period of 24 hours after entering OPERATIONAL MODE 2 or 3 when shutting down from OPERATIONAL MODE 1.

(p) This function is not required to be OPERABLE when reactor pressure is less than 600 psig.

(q) A current source rovides an instrument channel alignment every 3 months.

(r) The CHANNEL CA' {RATION surveillance equirements shall be p/rformed if not performed within the previo~t::ven days. / ·

• lRESDEN - UNITS 2 & 3 3/4.1-10 Amendment Nos.

II ti

• l •

TABL,,2.8-1 ./

0 z ::D m ECCS ACTUATION INSTRUMENTATION (/)

U> SURVEILLANCE REQUIREMENTS -i

0 ::D

m c z ~

m

c CHANNEL Applicable z -i z CHANNEL FUNCTIONAL CHANNEL OPERATIONAL )>

=i Functional Unit CHECK TEST CALIBRATION MODEis) -i U> 0 N

.L. CORESPRAY(CS)SYSTEM z

fl<> w JD a. Reactor Vessel Water Level ,.. Low Low s M 1 , 2' 3' 4 (bl' 5 (b)

b. Drywall Pressure - Highldl NA M 0 1' 2, 3

c. Reactor Vessel Pressure - Low (Permissive) NA M 0 1' 2, 3, 41bl' 5lbl

d. CS Pump Discharge Flow - Low (Bypass) NA :-® Ql•l 1' 2, 3, 4lbl' 5lbl

L LOW PRESSURE COOLANT INJECTION {LPCI} SUBSYSTEM

a@ w ..._ Reactor Vessel Water Level - Low Low 1, 2, 3, 4lbl, 5lbl .f>. a. s M

N ' b. Drywell Pressure - Highldl NA M a 1' 2, 3 .....

00

c. Reactor Vessel Pressure - Low (Permissive) NA M 0 1, 2, 3, 4lbl' 5lbl

d. LPCI Pump Discharge Flow - Low (Bypass) NA Jr@ Ql•I 1, 2, 3, 4lbl, 5lbl

3. HIGH PRESSURE COOLANT INJECTION !HPCI} SYSTEM181

a@ a. Reactor Vessel Water Level - Low Low s M 1, 2, 3

b. Drywall Pressure - High1d1 NA M 0 1, 2, 3 m )> 0 3 c. Condensate Storage Tank Level - Low NA M NA 1, 2, 3 0 ct> (/) :::::J a. d. Suppression Chamber Water Level - High NA M

NA® 1, 2, 3 )>

3 1. Q~) ("')

ct> ..... c

:::::J e. Reactor Vessel Water Level - High (Trip) NA M 1, 2, 3 Q) .....

p/(Q· ':!.

z f. HPCI Pump Discharge Flow - Low (Bypass) NA 0 1, 2, 3

0 0 :::I (/) -~ w

g. Manual Initiation NA E NA 1, 2, 3 ..._ .f>. N ClJ

• !·· TABLE 4.- (Continued)

.;. 0 z :a ECCS ACTUATION INSTRUMENTATION en m -i en SURVEILLANCE REQUIREMENTS 0 :a m c z 5:

m c z

-i z CHANNEL Applicable )> =i CHANNEL FUNCTIONAL CHANNEL OPERATIONAL -i en 0 N Functional Unit CHECK TEST CALIBRATION MODE(s) z '20

4. AUTO MA TIC DEPRESSURIZATION SYSTEM181 w

a@ a. Reactor Vessel Water Level - Low Low s M 1, 2, 3

b. Drywell Pressure - High1dl NA M Q 1, 2, 3

c. Initiation Timer NA E E 1, 2, 3

d. Low Low Level Timer NA E E 1, 2, 3

e. CS Pump Discharge Pressure - High NA M Q 1, 2, 3 w (Permissive) -.... ~

f. LPCI Pump Discharge Pressure - High N NA M Q 1, 2, 3 I (Permissive)

(0

~ LOSS OF POWER

a. 4. 1 6 kv Emergency Bus Undervoltage NA E E 1, 2, 3, 4lcl' 5lcl

(Loss of Voltage)

b. 4. 1 6 kv Emergency Bus Undervoltage NA E E 1, 2, 3, 4lcJ' 5lcl

(Degraded Voltage)

)> m (')

3 (') ct> Cf) ::> a. )> 3 (')

.-+ ct> c ::> Q) .-+ !:!'. z 0 0 ::> C/'J

w -.... ~

N

OJ

•

INSTRUMENTATION ECCS Actuation 3/4.2.B

TABLE 4.2.B-1 (Continued)

ECCS ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS

TABLE NOTATION

(a) Not required to be OPERABLE when reactor steam dome pressure is s 150 psig.

(b) When the system is required to be OPERABLE per Specification 3.5.B.

(c) Required when the associated diesel generator is required to be OPERABLE per Specification 3.9.B.

(d) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required .


0 ::0 m (/)

0 m z

c z =i (/)

N

!20 w

~ 3 ct> :::J a. 3 ct> :::J ,.... z 0 (/)

• •/ ,.

TABLE 4.2.C-1

ATWS - RPT INSTRUMENTATION SURVEILLANCE REQUIREMENTS

Functional Unit

· 1 . Reactor Water Level - Low Low

2. Reactor Vessel Pressure - High

~

CHANNEL CHECK

s s

CHANNEL FUNCTIONAL

TEST

JK' {§j ~®

a Trip units are calibrated at least once pe~s and transmitters are calibrated at the frequency identified in the table.

CHANNEL CALIBRATION

z (/)

-t ::0 c s: m z -t ~ -t 0 z

~ -t ~ (/)

::0 '"C -t w .._ ~

N

n

0 :0 m U> 0 m z c z =i U> N

S20 w

w ..._ ~

• •,' , .

Functional Unit

~ SOURCE RANGE MONITORS

a. Detector not full intbl

b. Upscale tel

c. lnoperativetcl

TABLE 3.2.E-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION

Trip Set point

NA

s 1 x 105 cps

NA

Minimum CHANNEL(s) per Trip Functionli1

3 2

3 2

3 2

":" 4. INTERMEDIATE RANGE MONITORS w 0

a. Detector not full in@

)>

3 Cl> :::J 0.. 3 Cl> :::J .-+

z 0 (/)

b. Upscale

c. Inoperative

d. Downscale1d1

NA 6

:5108/125 6 of full scale

NA 6

2: 5/125 6 of full scale

Applicable OPERATIONAL

MODE Cs>

2 5

2 5

2 5

2, 5

2, 5

2, 5

2, 5

ACTION

51 51

51 51

51 51

51

51

51

51

z U> -i :0 c ~ m z -i )> -i 0 z

() 0 :::J .-+ .... 2. :0 0 0..

~ 0 n 7'" (/)

w ..._ ~

N m

•• M; •• • ,. I /'; _/ ~ ..

TAB 4.2.E-1 0 z :0 CONTROL ROD BLOCK INSTRUMENTATION CJ) m CJ) SURVEILLANCE REQUIREMENTS

-; 0 :0 m c z s:

m c CHANNEL Applicable z

-; z CHANNEL FUNCTIONAL CHANNEL OPERATIONAL )> =i Functional Unit CHECK TEST CALIBRATION 181 MODE(sl

-; CJ) 0 N z 12'> w .L ROD BLOCK MONITORS

a. Upscale NA S/U(b,c), Mic) Q 1 (d)

b. Inoperative NA S/Ulb,cJ, Mic) NA 1 (d)

c. Downscale NA S/Ulb,c), Mic) Q 1 (dJ

b AVERAGE POWER RANGE MONITORS

w a. Flow Biased Neutron Flux - High -... ~

S/Ulbl, M N 1. Dual Recirculation Loop Operation NA SA 1 I

w ~ 2. Single Recirculation Loop Operation NA S/Ulbl, M SA 1

b. Inoperative NA S/Ulbl, M NA 1, 2, 5lil

c. Downscale NA S/Ulbl, M Q 1

d. Startup Neutron Flux - High NA S/Ulbl, M SA{$; 2, 5lil

~ SOURCE RANGE MONITORS n 0

)> a. Detector not full in10 NA S/U'b', W E 2,lil 5 :::::l .... 3

.., Q_

CD b. Upscale1g1 NA S/Ulbl, W E 2,lil 5 :::::l :0 a. 0 3 c. lnoperative1g1 NA S/Ulbl, W NA 2,lil 5 a. CD OJ :::::l 0 .... z C1

'1' 0 (/) (/)

w -... ~

N m

• • ·:-.. • "·· --· TABLE 4.2.E'-1 (Continued)

0 z :0 CONTROL ROD BLOCK INSTRUMENTATION U> m _, U> SURVEILLANCE REQUIREMENTS 0 :0 m c z ~

m c z _, z. CHANNEL Applicable :l> =i CHANNEL FUNCTIONAL CHANNEL OPERATIONAL

_, U> 0 1--l Functional Unit CHECK TEST CALIBRATION 181 MODE(s) z f20 w

4. INTERMEDIATE RANGE MONITORS

a. Detector not full i~ NA S/Ulbl, W E 21il, 5

b. Upscale NA S/Ulbl, W v 21;1, 5

c. Inoperative NA S/Ulbl, W NA 21i1, 5 ~

d. Downscale1h1 NA Stu<bl, W ~ 21il, 5

w -- ~ SCRAM DISCHARGE VOLUME {SDV} .J:>. N

I Water Level - High NA Q NA 1, 2, 51"1 w a. (J1

b. SDV Switch in Bypass NA E NA 5<•1

() 0

:l> ::::J .....

3 .... 0

CD ::::J :0 a. 0 3 a. CD ~ ::::J ..... 0

z (")

7'" 0 CJ) CJ)

w --.J:>. 1--l m

••

. "-.-J ,.'!

•

INSTRUMENTATION Control Rod Blocks 3/4.2.E


CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS

TABLE NOTATION

(a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) Within 7 days prior to startup.

(c) Includes reactor manual control "relay select matrix" system input.

(d) With THERMAL POWER ;::30% of RATED THERMAL POWER.

(e) With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3. 10.1 or 3.10.J ..

(f)

(g)

(h)

(i)

This function shall be automatically bypassed if detector count rate is > 100 cps or the IRM channels are on range 3 or higher.

This function shall be automatically bypassed when the assoCiated IRM channels are on range 8 or higher. · ·

This function shall be automatically bypass~d when the IRM channels are on range 1.

The provisions of Specification 4.0.D are not applicable to the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for entry into the applicable OPERATIONAL MODE(s) from OPERATIONAL MODE 1 provided the surveillances are performed within 12 hours after such entry

(j) Required to be OPERABLE only during SHUTDOWN MARGIN demonstrations performed per Specification 3.12.B.

rveillance requireme s shall be performed ithin 1 2 hours upon eac entry into any OPERA ONAL MODE(s) from PERATIONAL MOD performed · hin the previous seven ays .


•

•

INSTRUMENTATION Accident Monitors 3/4.2.F

ACTION 60 -

ACTION 61-

ACTION 62-

TABLE 3.2.F-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATION

ACTION

a. With the numbe~ of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Required CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours.

b. With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Minimum CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 1 2 hours.

With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Minimum CHANNEL(s) shown in Table 3.2.F-1, initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours, and:

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

b. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.B 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 stat1,.1s:

a. With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) one less than the Required CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours.

b .


., ··4.2.F'1 0 :0 m Ul 0 m z

ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

c z ~ . Ul

I\)

j20

w

w ~ I\)

~

INSTRUMENTATION

1 . Reactor Vessel Pressure

2. Reactor Vessel Water Level

3 Torus Water Level

4. Torus Water Temperature

6. Drywall Pressure - Wide Range

6. Drywall Pressure - Narrow Range

7. Drywall Air Temperature

8. Drywall ~n Conc~ntratlon - Analyzer and Monitor

~ . p-safety/Rellef Valve Position Indicators 6) - Acoustic & Temperature

~Source Range) Neutron Monitors

~rywell Radiation Monitors .

~rus Pressure

z 0 (II

CHANNEL CHECK

M

M

M

M

M

M

M

M

M

M

M

IV\

CHANNEL CALIBRATION

SA~ SA .CJ)

A

A

E

a E

Ci

E

Qlbl

El•I

a

Q

. ... Applicable

OPERATIONAL MODE Isl

1, 2

1, 2

1, 2

1, 2

1, 2

1, 2

1, 2 .

1, 2

1, 2

1, 2, 3

1, 2

1, 2

z (/') -i ::0 c ~ m z -f }>

~ 0 z

}> n n a: Cb :::> r+

~ 0 :::>

.., (II

w -~ N "TI·

•-1:

'.- ..

•




TABLE NOTATION

(a) CHANNEL 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.

(b) Neutron detectors may be excluded from the CHANNEL CALIBRATION.

(c) CHANNEL CHECK of the Acoustic Monitors shall consist of verifying the instrument threshold levels.

P.11~..,._ 1~d,c11"t-<.PV' f:.siV' "t1'i<... td'1e1/1?j -\Y'4~Sl'11~1-ter 1.s ~~f,'Jy.._t.,,.c/ t<-tlhe... ~rc•t"'-tYl'-J

;4 t Ill.-\ ~ ~I e J I ~ ..,. "'e. T ~ b \e .


• INSTRUMENTATION B 3/4.2

BASES

3/4.2.E Control Rod Block Actuation Instrumentation

The control rod block functions are provided to prevent excessive control rod withdrawal so that the MINIMUM CRITICAL POWER RATIO (MCPRJ does not go below the MCPR fuel cladding integrity Safety limit. During shutdown con(jitions, control rod block instrumentation initiates withdrawal blocks to ensure that all control rods remain inserted to prevent inadvertent criticality.

The trip logic for this function is one-out-of-n; e.g., any trip of one of the six average power range monitors (APRMs), eight intermediate range monitors (IRMs), or four source range monitors (SRMs), will result in a rod block. The minimum instrument CHANNEL requirements assure sufficient instrumentation to assure that the single failure criterion is met. The minimum instrument CHANNEL requirements for the rod block monitor may be reduced by one for a short period of time to allow for maintenan~e, testing, or calibration.

The APRM rod block function is flow-biased 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.

In the REFUEL and STARTUP/HOT STANDBY OPERATIONAL MODE(s), the APRM rod block function setpoint is significantly reduced to provide the same type of protection in the REFUEL and STARTUP/HOT STANDBY OPERATIONAL MODE(s) 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 rod block monitor (RSM) 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. The trip setting is flow-biased. At low 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 the specified power level. 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.

The !RM 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 ten above the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the MCPR

· fuel cladding integrity Safety limit.

, A downscale indication on an APRM is an indication that the instrument has failed or is not A suff~ciently sensitive. In either ~ase: the instrument will not respond to changes in control rod • motion, and the control rod motion 1s thus prevented.

The SRM rod blocks of low count rate and the detector not fully inserted assure that the SRMs are not withdrawn from the core prior to commencing rod withdrawal for startup. The scram discharge volume, ~h water level rod block providt@mnunciation 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 .

• DRESDEN - UNITS 2 & 3 B 3/4.2-3 Amendment Nos.

•

-· a

REACTIVITY CONTROL

3.3 - LIMITING CONDITIONS FOR OPERATION

D. Maximum Scram Insertion Times

The maximum scram insertion time of each control rod from the fully withdrawn position to 90% insertion, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed 7 seconds.

APPLICABILITY:

OPERATIONAL MODE{s) 1 and 2.

ACTION:

With the maximum scram insertion time of one or more control rods exceeding 7 seconds:

1. Declare the control rod{s) exceeding the above maximum scram insertion time inoperable, and

2. When operation is continued with three or more control rods with maximum scram insertion times in excess of 7 seconds, perform Surveillance Requirement 4.3.D.3 at least once per 60 days of(Power operati-OV- ALL <.Af'~

With the provisions of the ACTION(s) above not met, be in at least HOT SHUTDOWN within 1 2 hours.

Maximum Scram Times 3/4.3.D

4.3 - SURVEILLANCE REQUIREMENTS


The maximum scram insertion time of the control rods shall be demonstrated through measurement with reactor coolant pressure greater than 800 psig and, during single control rod scram time tests, with the control rod drive pumps isolated from the accumulators:

1. For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER:

a. following CORE ALTERATION(s), or

b. after a reactor shutdown that is greater than 1 20 days,

2. For specifically affected individual control rods 181 following maintenance on or modification to the control rod or control rod drive system which could . affect the scram insertion time of those specific control rods, and

3. For at least 10% of the control rods, on a rotating basis, at least once per 120 days ot'i)"ower operation~

jALL ~JIPS

The provisions of Specification 4.0.D are not applicable provided this surveillance is conducted prior to exceeding 40% of RATED THERMAL POWER.


REACTIVITY CONTROL EGC 3/4.3.N

,. 3.3 - LIMITING CONDITIONS FOR OPERATION 4.3 - SURVEILLANCE REQUIREMENTS

N. Economic Generation Control (EGC) System N. Economic Generation Control (EGC) System

The economic generation control (EGC) The economic generation control system system may be in operation with automatic shall be demonstrated OPERABLE by flow control provided: verifying that core. flow is within 65% to

/,") ~ 100% of rated core flow and THERMAL L~ Core flow is within 65% to 100% of POWER is 2:::20% of RATED THERMAL

rated core flow, and (i~ PO:ER:

/f:1,, ~HERMAL POWER is 2::: 20% of RATED ~ 0 Prior to entry into EGC operation, and ~ THERMAL POWER.

~At least once per 1 2 hours while ly l7 ~perating in EGC.

••

•

APPLICABILITY

OPERATIONAL MODE 1.

ACTION:

With core flow less than 65% or gre·ater than 100% of rated core flow, or THERMAL POWER less than 20% of RATED THERMAL POWER, restore operation to within the limits within one hour. Otherwise, immediately remove the plant from EGC operation .

DRESDEN - UNITS' 2 & 3 3/4.3-20 Amendment Nos.

Reactivity Control B 3/4.3

•. BASES

3/4.3.A SHUTDOWN MARGIN

A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

The SHUTDOWN MARGIN limitation is a restriction to be applied principally to a new refueling pattern. Satisfaction of the limitation must be determined at the time of loading and must be such that it will apply to the entire subsequent fuel cycle. This determination is provided by core design calculations and administrative control of fuel loading patterns. These procedures include restrictions to allow only those intermediate fuel assembly configurations that have been shown to provide the required SHUTDOWN MARGIN.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed during xenon free conditions and adjusted to 68°F to accommodate the current moderator temperature. The generalized form is that the reactivity of the core loading will be limited so the core can be made subcritical by at least R + 0.38% f1k/k or R + 0.28% f1k/k, as appropriate, with the strongest control rod fully withdrawn and all others fully inserted. Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of determination of the highest control rod worth, either analytically or by test. This is due to the reduced uncertainty in the SHUTDOWN MARGIN test when the highest worth control rod is determined by demonstration. When SHUTDOWN MARGIN is determined by calculations not associated with a test, additional margin must be added to the specified SHUTDOWN MARGIN limit to account for uncertainties in the calculation. · ·

The value of R in units of % f1k/k is the difference between the calculated beginning-of-life core reactivity, at the beginning of the operating cycle, and the calculated value of maximum core reactivity at any time later in the operating cycle, where it would be greater than at the beginning. The value of R shall include the potential SHUTDOWN MARGIN loss assuming full 84C settling in all inverted poison tubes present in the core. R must be a positive quantity or zero and a new value of

f:2 R must be determined for each new fuel cycle.

~ The value of 0 o k/k in the abo e expression is provided as a finite, demonstrable, subcriticality

•"1 margin. This mar ·n is verified u ·ng an in-sequence control rod withdrawal at the beginning-of-life

· fuel cycle condition . This assure su~ticality with not only the strongest fully withdrawn but at least an R + 0.28%~ 0.38% ®margin beyond this condition. This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion following a scram signal. Any control rod that is immovable as a result of excessive friction or mechanical interference, or is known to be unscrammable, per Specification 3.3.C, is considered to be incapable of insertion following a scram signal. It is. important to note that a

• control rod can be electrically immovable, but scrammable, and no increase in SHUTDOWN

- MARGIN is required for these control rods.

DRESDEN - UNITS 2 & 3 B 3/4.3-1 Amendment Nos.

EMERGENCY CORE COOLING SYSTEMS ECCS - Operating 3/4.5.A

• 3.5 - LIMITING CONDITIONS FOR OPERATION

APPLICABILITY:


b. Three LPCI pumps together develop a flow of at least 14,500 gpm against a test line pressure corresponding to a reactor vessel pressure of ~ 20 psig.

•

•

OPERATIONAL MODE(s) 1, 21b1 and 3<bl.

ACTION:

1. For the core spray system:

a. With one CS subsystem inoperable, provided that the LPCI subsystem is OPERABLE, restore the inoperable CS subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following 24 hours.

b. With both CS subsyste.ms inoperable, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.

2. For the LPCI subsystem:

a. With one LPCI pump inoperable<dl, provided that both CS subsystems are OPERABLE, restore the inoperable LPCI pump to OPERABLE status within 30 days, or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following 24 hours .

c. The HPCI pump develops a flow of at least 5000 gpm against a system head corresponding to reactor vessel pressure, when steam is being supplied to the turbine between 920 and 1005 psiglcl.

3. At least once per 18 months:

a. For the CS system, the LPCI subsystem, and the HPCI system, verify each system/subsystem actuates on an actual or simulated automatic initiation signal. Actual injection of coolant into the reactor vessel may be excluded from this test.

b. For the HPCI system, verifying that:

1) The system develops a flow of ~ 5000 gpm against a system head corresponding to reactor vessel pressure, when steam is being supplied to the turbine between 1 50 and 350 psig1c1•

Qth ev-w• Sc ) e.V\ -t-C.Y'

3, 5. A.) Ac1'10~ ~Le.

b The HPCI system and ADS are not required to be OPERABLE when reactor steam dome pressure is :s 150 psig.

d The provisions of Specification 3.9.A, Actions 4.a or 6.b are applicable to the LPCI subsystem such that with an inoperable diesel generator, for the remaining OPERABLE diesel generator, both LPCI pumps (and their associated flow path) associated with that OPERABLE diesel generator, shall be OPERABLE .

c The provisions of Specification 4.0.D are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.



• 3.5 - LIMITING CONDITIONS FOR OPERATION 4.5 - SURVEILLANCE REQUIREMENTS

b. With the LPCI subsystem otherwise inoperable1d1, provided that both CS subsystems are OPERABLE, restore the LPCI subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following 24 hours.

c. With the LPCI subsystem and one or both CS subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.

3. With the HPCI system inoperable, provided both CS subsystems, the LPCI subsystem, the ADS and the Isolation Condenser (IC) system are OPERABLE, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 1 2 hours and reduce reactor steam dome pressure to s 1 50 psig within the following 24 hours.

4. For the ADS:

a. With one of the above required ADS valves inoperable, provided the HPCI system, both CS subsystems and three LPCI pumps are OPERABLE, restore the inoperable ADS valve to OPERABLE

2) The pump suction is automatically transferred from the condensate storage tank to the suppression chamber on a condensate storage tank water level - low signal and on a suppression chamber water level - high signal.

c. Performing a CHANNEL CALIBRATION of the CS and LPCI system discharge line "keep filled" alarm instrumentation.

d. Deleted.

4. At least once per 18 months for the ADS:

a. Verify the ADS actuates on an actual or simulated automatic initiation signal. Actual valve actuation may be excluded from this test.

b. Manually opening each ADS valve when the. reactor steam dome pressure is ~ 1 50 psig1c1 and observing that either:

1) The turbine control valve or turbine bypass valve position responds accordingly, or

2) There is a corresponding change in the measured steam

,..-rc_"'--_._10 w. Q.,. he. ~wise:. , (!.VI 1't:.r" S'peG1 fi&4 ~; """

:,, S. A, A", va-11 ~ • e.. •

d The provisions of Specification 3.9.A, Actions 4.a or 6.b are applicable to the LPCI subsystem such that with an inoperable diesel generator, for the remaining OPERABLE diesel generator, both LPCI pumps (and their associated flow path) associated with that OPERABLE diesel generator, shall be OPERABLE .

• c The provisions of Specification 4.0.D are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.


------------------- - ------ -- -- ------

PRIMARY SYSTEM BOUNDARY

• 3.6 - LIMITING CONDITIONS FOR OPERATION ·- ... f-;/

• •

J. Specific Activity

The specific activity of the reactor coolant shall be limited to s0.2 µCi/gram DOSE EQUIVALENT 1-131.

APPLICABILITY:

OPERATION~L MODE(s) 1, 2 and 3f) J~~-:ic_4_1t1~-

ACTION: no1 i"solo.""to!«>l.

1. In OPERATIONAL MODE(s) 1, 2 or 3 c41' -'*@th the specific activity of the reactor

coolant > 0.2 µCi/gram DOSE EQUIVALENT 1-131 but s4.0 µCi/gram DOSE EQUIVALENT 1-131 for more

Specific Activity y-" A/o <::t'f~

CDhe specific activity of the reactor coolant

shall be e onstrt~o b~i7,h·n the ') im1ts by p rforma ce f th sa pli.ng and analysis ogram f ble .6. -1.

Vc""'~'u:l 10 be :=.. D· ';;/ r c.../Jf'aW'I~~ 00~~ Et>i lJIVlli,.£1\/T 'I-l31 o,.,c.e re"'_)

7 J4y~.

, J dc.1'C Y'l'Ylt Y1e.

J-(31 QY\C~f~Y' <-tV!oCAV'-' <.e>'ld

V'e ~:.-t c9 VIC. D Oj E E (;) V I V }\ LP: Jll'f

:!-111 ..,.o W,"tl-\•Y\ l1m11S (A-'l

ho<Lr"S

fe>l"ji"CCi.teVI "t"~A1'\ Lt~ h.ou.Y'.S)

I t • ·.,

01" W\'tl-, -t~e. 9 p e.C.1 ~\c a..:.111/l"t'-{

·6' -the re a.c,"tor cool ... 1"11" "> lf. 0

I' c.,/j;"CHYI oos f" i:: Q VI VA L~A)/

J, 1311 d"1e.l"rntnC: POSE

eQVI Vh LE t\/T 1-131 .,,,,C. pe.-r- 4 ·ho1.1.v-~J at1d tf>OLq,tc::, atl math

stc~wi l1tie5 wl-t"'·~ 1.1 h.a&A.Y'S,

or- be. ~ q;'i le..:..f>T HOT ~HL>iQCWI\/

-W 1-ihd1 -the n. e d \ '2. ~ou V':;> 4 vid (I"\ Leno SHUTOOl.(}N t.(/1"th1Y1 ~'1e

Co~~

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

3.(9. /) Go\V'C. Ytel't 'lff' !1 Cc.. bl(. .

3/4.6-16 Amendment Nos.


• 3.6 - LIMITING CONDITIONS FOR OPERATION · . ..__..,,..

••

a. THER AL POWER changed by more han 20 % of A TED THE MAL POWE in 1 hour, or

b. Th offgas lev. I, prior tot e h ldup line, i creased by

25,000 JiJ i/second i one hour uring ste dy state o eration at

release r es < 100,0 0 µCi/sec nd, or

The o gas level, rior to the hold line, incr ased by > 1 one our durin steady stat op ration at r ease rates > 00,000 µ i/second,

Perfo m the sa piing and an ysis requirements f Item 3.b of able 4.6.J-1 until e specific a ivity of the reactor cool nt is restored o within its limit,

-n-\ \ .s

Specific Activity 3/4.6.J


I tJ TE. rJTtO k!A LL'( L~ FT Bl ANK


0 ::n m U> 0 m z

c z =i Cl> N j20

w

<::. -i TT\ <::: -1 w --~

~

m ~ I

~ <X> ,.... r -< r-TTI ..., ~

00 r--:z:> e::. )> 7' 3

CD ::J a. 3 CD ::J r+

• •••• ,,

TABLE 4.6.J-1

REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE A ANALYSIS PROGRAM

Type of Measurement and Analysis

1 . Gross Beta and Gamma A ivity Determination

2. Isotopic Analysis for DOSE EQUIVALENT 1-1 31 Concentration

At least once per

At least once er 31 days

3. Isotopic Analysis for Iodine At leas once per 4 hours, whenever the speci c activity exceeds a limit, as

4. Isotopic Analysis of an Off-gas Sample Including Quantitative Measurements for Xe-133, Xe-135 and Kr-88

r u· ed by ACTION 2.

b) lea t one sample, between 2 and 6 ours to owing the change in THERMAL

POWER or ff-gas level, as required by ACTION 3.

t least once per 31 ays

Until the specific actiyity of th reactor coolant system is restored to within its limits.

OPERATIONAL MODE(s) in Which Sample

and Analysis Required

1, 2, 3

1 (al, 2tal 1 3tal

1

"'ti ::n ~ )> ::n -< Cl> -< U> -; m ~ ClJ 0 c z 0 )> ::n -<

Cl> 'O CD (')



• 0. Shutdown Cooling - HOT SHUTDOWN

Two 1•1 shutdown cooling (SOC) loops shall be OPERABLE and, unless at least one · recirculation pump is in operation, at least one shutdown cooling loop shall be in . operation1bHcJ, with each loop consisting of at least:

1 . One OPERABLE SOC pump, and

2. One OPERABLE SOC heat exchanger.

APPLICABILITY:

OPERATIONAL MODE 3, with reactor vessel coolant temperature less than the SOC cut-in permissive setpoint.

ACTION:

1 . With less than the above required SOC loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible. Within 1 hour and at least once per 24 hours thereafter, demonstrate the operability of at least one alternate method .capable of decay heat removal for each inoperable SOC loop. Be in at least COLD SHUTDOWN within 24 hours1dl •

SOC- HOT SHUTDOWN 3/4.6.0


0. Shutdown Cooling - HOT SHUTDOWN

At least one SOC loop, one recirculation pump or alternate method shall be verified to be in operation and circulating reactor coolant at least once per 1 2 hours.

• One shutdown cooling loop may be inoperable for up to 2 hours for surveillance testing provided the other loop is OPERABLE and in operation.

b A shutdown cooling pump may be removed from operation for up to 2 hours per 8 hour period provided the other loop is OPERABLE.

c The shutdown cooling 100~;8:.;;;;,: ~~ration during hydrostatic testing. loo

d Whenever two or more SOC re in e, rt unable to attain COLD SHUTDOWN as required by this a ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

~RESDEN - UNITS 2 & 3 3/4.6-25 Amendment Nos.

•

•

CONTAINMENT SYSTEMS PCIVs 3/4.7.D

3.7 - LIMITING CONDITIONS FOR OPERATION 4. 7 - SURVEILLANCE REQUIREMENTS

2. With one or more reactor instrumentation line excess flow check valves inoperable, operation may continue and the provisions of Specification 3.0.C are not applicable, provided that within 4 hours either: ·

a. The inoperable valve is restored to OPERABLE status, or

b. The instrument line is isolated and the associated instrument is declared inoper~ble.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.

DRESDEN - UNITS 2. & 3 3/4.7-7

a.

b.

At least once per 31 days by verifying the continuity of the explosive charge.

At least once per 18 months by removing at least one explosive squib from ~explosive valve such that each explosive squib -ffieach exialesive •;alve will be tested at least once er~ months, and 1rnt1atmg the removed explosive squib(s). The replacement charge for the exploded squib(s) shall be from the same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch successfully fired. No squib shall remain in use beyond the expiration of its shelf-life or operating life, as applicable.

6. At the frequency specified by the Primary Containment Leakage Rate Testing Program/ verify leakage for any one main steam line isolation valve when tested at Pt (25 psig) is ::s 11.5 scfh. ·

Amendment Nos.

CONTAINMENT SYSTEMS


.\P. Standby Gas Treatment System

Two independent standby gas treatment subsystems shall be OPERABLE.

APPLICABILITY:

OPERATIONAL MODE(s) 1, 2, 3 and *.

ACTION:

1 . With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:

a. In OPERATIONAL MODE(s) 1,2 or 3, be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following 24 hours.

b. In OPERATIONAL MODE •, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATION(s), and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0 .. C are not applicable.

Q. 9elete!:J g

SBGT 3/4. 7 .P

4. 7 - SURVEILLANCE REQUIREMENTS

P. Standby Gas Treatment System

Each standby gas treatment subsystem shall be demonstrated OPERABLE:

1. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours with the heaters operating.

2. At least once per 1 8 months or ( 1 ) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem by:

a. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of < 1 % and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 4000 cfm ±10%.

b. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-0-3803-89, for a methyl iodide penetration of < 10%, when tested at 30°C and 70% relative humidity; and

• When handling irradiated fuel in the secondary containment, during CORE ALTERA TJON(s), and operations a with a potential for draining the reactor vessel.

~RESDEN - UNITS 2 & 3 3/4.7-23 Amendment Nos.

CONTAINMENT SYSTEMS

3.7 - LIMITING CONDITIONS FOR OPERATION •• ~· Delete£) '<---

• •• •

SBGT 3/4. 7 .P


c. Verifying a subsystem flow rate of 4000 cfm ± 10% during system operation when tested in

)~ accordance with ANSI N510-1980. ~~,~

3. After every'@i 4 40 hours of charcoal adsorber operation ·by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-D-3803-89, for a methyl iodide penetration of < 10%, when tested at 30°C and 70% relative humidity.

4. At least once per 18 months by:

a. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is

b.

< 6 inches water gauge while operating the filter train at a flow rate of 4000 cfm ± 10%.

Verifying that the filter train. starts and isolation dampers open on each of the following test signals:

1) Manual initiation from the control room, and

2) Simulated automatic initiation signal.

c. Verifying that the heaters dissipate 30 ± 3 kw when tested in accordance with ANSI N510-1989. This reading shall include the appropriate correction for variations from 480 volts at the bus.


CONTAINMENT SYSTEMS


With both standby gas eatment subsystems o herw1s inoperable in OPERATIONAL MODE(s) 1,2 or 3,

· restore at least one subsystem to OPERABLE status within one hour, or be in at least HOT SHUTDOWN within the next 1 ~ hours and in COLD SHUTDOWN within the following 24 hours.

C[)' With both standby gas treatment r--/ subsystems inoperable in ~ OPERATIONAL MODE*, suspend

handling of irradiated fuel in the secondary containment, CORE

••

AL TERATION(s), and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.C are not applicable .

SBGT 3/4. 7 .P


5. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the in-place penetration and leakage testing acceptance criteria of < 1 % in accordance with ANSI N510-1980 while operating the system at a flow rate of 4000 cfm ± 10%.

6. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the in-place penetration and leakage testing acceptance criteria of < 1 % in accordance with ANSI N 51 0-1 980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 4000 cfm ± 10%.

When handling irradiated fuel in the secondary containment, during CORE ALTERATION(s), and operations with a potential for draining the reactor vessel.

DRESDEN - UNITS 2 & 3 3/4.7-25 Amendment No.

•

,.--~.~ ,.

~ ~/'

•

PLANT SYSTEMS


2. In OPERATIONAL MODE *, with the control room emergency filtration system or the RCU inoperable, immediately suspend CORE AL TERATION(s), handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.

3. The provisions of Specification 3.0.C are not applicable in OPERATIONAL MODE*.

CREVS 3/4.8.D


b. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, . March 1978, meets the laboratory testing criteria of ASTM-D-3803-89, for a methyl iodide penetration of < 0.50%, when tested at 30°C and 70% relative humidity; and

c. Verifying a system flow rate of 2000 scfm ± 10% during system operation when tested in

· accordance with ANSI N510-1980. I --;... D ---~-----~

4. After every~ hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-D-3803-89, for a methyl iodide penetration of <0.50%, when tested at 30°C and 70% relative humidity.


a. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches water gauge while operating the filter train at a flow rate of 2000 scfm ± 10%.

b. Verifying that the filter train starts and isolation dampers close on manual initiation from the control room .


PLANT SYSTEMS CREVS 3/4.8.D


• DRESDEN - UNITS 2 & 3 3/4.8-8

c. Verifying that during the pressurization mode of operation, control room positive pressure is maintained at ;:::; 1 /8 inch water gauge relative to adjacent areas during system operation at a flow rate :s 2000 scfm.

d. Verifying that the heaters dissipate 12 ± 1.2 kw when tested in accordance with ANSI N510-+980. This reading shall include the appropriate correction for variations from 480 volts at the bus.

6. After each complete or partial replacement of an HEPA filter bank by verifying that the HEPA filter bank satisfies the in-place penetration and leakage testing acceptance criteria of <0.05% in accordance with ANSI N510-1980 while operating the system at a flow rate of 2000 s'cfm ± 10%.

7. After each complete or partial replacement of an charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the in-place . penetration and leakage testing acceptance criteria of <0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at flow rate of 2000 scfm ± 10% .

Amendment Nos.

PLANT SYSTEMS B 3/4.8

BASES

3/4.8.G Sealed Source Contamination

The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources, including startup sources end fission detectors, ere classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled ere required to be tested more often than those which ere not. Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring or boron measuring d.evices, are considered to be stored and need not be tested unless they ere removed from the shielded mechanism.

3/4.8.H Explosive Gas Mixture

This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the offgas holdup system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen end oxygen below their flammability limits. provides assurance that the releases of radioactive materials will be controlled in conformance with

!). _____ the requirements of General ~;sign Criterion 60 of Appendix A to 1 OCFR Part 50 •

. $ •. \ . :r

_ - 3/4.8.1 Main Condenser Offgas Activtty

Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction. of the limits of 1 OCFR Part 1 00 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 1 OCFR Part 50. The. release rates are detemiined at a more expeditious frequency following the determination of an increase of greater than 50%, as indicated by the air ejector noble gas monitor, after factoring out increases due to changes in THERMAL POWER level and off-gas flow in the nominal steady-state fission gas release from the primary coolant •

• 3/4.8.J _ Liquid Holdup Tanks

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations

( ,. would be less than the limits of 1 OCFR Part 20, Appendix 8, Tabl olumn 2, ~er@) .e__ Lt('C!"otabte water supply and the AHrest St:lrlece weta1 '' rpp(i:>in e{rnrestricted area~ulation of

the tank contents for the purpose of reducing the radioactive content is not considered to be an addition of radioactive material to the tank. ,. DRESDEN - UNITS 2 & 3 B 314.8-4 Amendment Nos.

ELECTRICAL POWER SYSTEMS

. • 3.9 - LIMITING CONDITIONS FOR OPERATION

• •

E. Distribution - Operating

The following power distribution systems shall be energized:

1. A.C. power distribution, consisting of:

a. Both Unit engineered safety features 41 60 volt buses:

1) For Unit 2~ Nos. 23-1 and 24-1,

2) For Unit 3, Nos. 33-1 and 34-1.

b. Both Unit engineered safety features 480 volt buses:

1) For Unit 2, Nos. 28 and 29,

2) For Unit 3, Nos. 38 and 39.

c. The Unit 120 volt Essential Service Bus and Instrument Bus.

2. 250 volt D.C. power distribution, consisting of:

3. For Unit 2, 125 volt O.C. power distribution, consisting of:

a. TB Main Bus Nos. 2A-1 and 3A,

b. TB Res. Bus Nos. 2B and 2B-1,

c. Reserve Bus No. 2, and

d. RB Distribution Panel No. 2 .

Distribution - Operating 3/4.9.E



Each of the required power distribution system divisions shall be determined energized at least once per 7 days by verifying correct breaker alignment and voltage on the busses/MCCs/panels.

Ci. f!>"' (.,{...,.,, ;() I 6 M'l a a11J R'B MC.C... ?.

?. foo'" (.L>'\1 1 3 } 1'1 MC(..~ und

Rt3 MC-<- 3.

DRESDEN ~ UNITS 2 & 3 3/4.9-17 Amendment Nos.



G. RPS Power Monitoring

Two Reactor Protection System (RPS) electric power monitoring CHANNEL(s) for each inservice RPS Motor Generator (MG) set or alternate power supply shall be OPERABLE.

APPLICABILITY:

O~IONAL MODE(s) 1, 2, 3, 4181 and

5. ~ .

ACTION:

1 . With one RPS electric power monitoring CHANNEL for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable power monitoring CHANNEL to OPERABLE status within 72 hours or remove the associated RPS MG set or alternate power supply from service.

2. With both RPS electric power monitorin'g CHANNEL(s) for an inservice RPS MG set or alternate power supply inoperable, restore at least one ·electric power monitoring CHANNEL to OPERABLE status within 30 minutes or remove the associated RPS MG set or alternate power supply from service.

With any control rod withdrawn.

RPS Power Monitoring 3/4.9.G



The specified RPS electric power monitoring CHANNEL(s) shall be determined OPERABLE:

1. By performance of a CHANNEL J@ FUNCTIONAL TESTf8ci"ch time the plant is in COLD SHUTDOWN for a period of more than 24 hours, unless performed in the previous 6 months.

2. At least once per 18 months by demonstrating the OPERABILITY of overvoltage, undervoltage, and underfrequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic, and output circuit breakers, and verifying the following setpoints:

a. Overvoltage :s 129.6 volts AC

b. Undervoltage ;::::; 105.3 volts AC

c. Underfrequency ;::::; 55.4 Hz


ELECTRICAL POWER SYSTEMS 8 3/4.9

BASES •

' ... ;::, .. , I . ! .. · :

redundancy in components and features not available, the plant must be placed in a condition for which the Limiting Condition for Operation does not apply.

The term verify as used toward A. C. electrical power sources means to administratively check by examining logs or other information to determ.ine if certain components are out-of-service for preplanned preventative maintenance, testing, or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component.

With one otfsite circuit and one diesel generator inoperable, individual redundancy is lost in both the offsite and onsite electrical power system. Therefore, the allowable outage time is more limited. The time limit takes into account the capacity and capability of the remaining sources, reasonable time for repairs, and the low probability of a design basis event occurring during this period.

With both of the required offsite circuits inoperable, sufficient onsite A.C. sources are available to maintain the unit in a safe shutdown condition in the event of a design basis transient or accident. In fact, a simultaneous loss of offsite A.C. sources, a loss-of-coolant accident, and a worst-case single failure were postulated as a part of the design basis in the safety analysis. Thus, the allowable outage time provides a period of time to effect restoration of all or all but one of the offsite circuits commensurate with the importance of maintaining an A.C. electrical power system capable of meeting its design intent.

With two diesel generators inoperable there are no remaining standby A.C. sources. Thus, with an assumed loss of offsite electrical power, insufficient standby A.C. sources are available to power the minimum required ESF functions. Since the offsite electrical power system is the only source of A.C. power for this level of degradation, the risk associated with continued operation for a very short time could be 'less than that associated with an immediate controlled shutdown, which could result in grid instability and possibly a loss of total A.C. power. The allowable time to repair is severely restricted during this condition. The intent here is to avoid the risk associated with an immediate controlled shutdown and to minimize the risk associated with this level of degradation. . .

Surveillance Requirements are provided which assure proper circuit continuity for the offsite A.C. electrical power supply to the onsite distribution network and availability of offsite A.C. electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure distribution buses and loads are connected to their preferred power source. The frequency is adequate since breaker position is not likely to change without the operator being aware of it and because status is displayed in the control room. Should the action provisions of this specification require an increase in frequency, this Surveillance Requirement assures proper circuit continuity for the available offsite A.C. sources during periods of degradation and potential information on common cause failures that would otherwise go undiscovered.

:"'!' DRESDEN - UNITS 2 & 3 .B 314.9-2 · Amendment Nos. 138 & 132

ELECTRICAL POWER SYSTEMS B 3/4.9

• BASES

/

. -~ • •

Each battery of the D.C. electrical power systems is sized to start and carry the normal D.C. loads plus all D.C. loads required for safe shutdown on one unit and operations required to limit the consequences of a design basis event on the other unit for a period of 4 hours following loss of all A.C. sources. The battery chargers are sized to restore the battery to full charge under normal (non-emergency) load conditions. A normally disconnected alternate 1 25 volt battery is also provided as a backup for each normal battery. If both units are operating , the normal 125 volt battery must be returned to service within the specified time frame since the design configuration of the alternate battery circuit is susceptible to single failure and, hence, is not as reliable as the normal station circuit. During times when the other unit is in a Cold Shutdown or Refuel condition, an alternate 125 volt battery is available to replace a normal station 125 volt batter on a continuous basis to provide a second available power source. With the a ernate 12~_t_lt battery

--i ervice, en rea on the DC Res~ Bus is placed · the open p'ttion and osted, · .e., "tagge out." ·

With one of the required D.C. electrical power subsystems inoperable the remaining system has the_ capaeity to support a safe shutdown and to mitigate an accident condition. However, a subsequent worst-case single failure would result in complete loss of ESF functions. Therefore, an allowed outage time is provided based on a reasonable time to assess plant status as a function of the inoperable D.C. electrical power subsystem and, if the D.C. electrical power subsystem is not restored to OPERABLE status, prepare to effect an orderly and safe plant shutdown.

Inoperable chargers do not necessarily indicate that the D.C~ systems are not capable of performing their post-accident functions as long as the batteries are within their specified parameter limits. With both the required charger inoperable and the battery degraded, prompt action is required to assure an adequate D.C. power supply.

ACTION(s) are provided to delineate the measurements and time frames needed to continue to assure OPERABILITY of the Station batteries when battery parameters are outside their identified limits.

Battery surveillance requirements are based on the defined battery cell parameter values. Category A defines the normal parameter limit for each designated pilot cell in each battery. The pilot cells are the average cells in the battery based on previous test results. These cells are monitored closely as an indication of battery performance. Category B defines the normal parameter limits for each connected cell. The term "connected cell" excludes any battery cell that may be jumpered out because of a degraded condition or for any other reason. Category B also defines allowable values for each connected cell. These values, although reduced, provide assurance that sufficient capacity exists to perform the intended function and maintain a margin of safety. When any battery parameter is outside the Category B allowable value, the assurance of sufficient capacity as described above no longer exists and the battery must be declared inoperable.

Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations.


•

,.,. ........ 7 .

·=·

•

REFUELING OPERATIONS ~-~~ --------- ---- j) e. I <. -t :.C- cl 1 .. )-, ,-- • I

Decay Time 3/4.1 O.D

3.10 --GMITING CONDITIONS FOR OPERATION 4.10,..- SURVEILLANCE REQUIREMENTS ·'

Di' Decay Time

The reactor 24 hours.

DRESDEN - UNITS 2 & 3

5, during ovement he reactor pressure

than 2

3/4.10-6

e reactor s all be det

verificati subcriti ality prior o moveme irradi ed fuel in ves I.

Amendment Nos.

REFUELING OPERATIONS B 3/4.10

• BASES

•

3/4.1 o.c Control Rod Position

The requirement that all control rods be inserted during other CORE AL TERATION(s) ensures that fuel will not be loaded into a cell without an inserted control rod.

3/4.10.D

um requirement ~ reactor subcriti ality prior to fuel mroment ens res that suffic" nt elapsed to allow t radioactive de ay of the short lived f sion prod cts. This dee

1stent with the assu ptions used in e accident analyses. is con

3/4.1 O.E Communications

The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status regarding core reactivity conditions during movement of fuel within the reactor pressure vessel.

3/4.10.F DELETED

3/4.10.G Water Level - Reactor Vessel

3/4.1 O.H Water Level - Spent Fuel Storage Pool

The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. This minimum water depth is consistent with the assumptions of the accident analysis .


REFUELING OPERATIONS

•. 10 - LIMITING CONDITIONS FOR OPERATION

H. Water Level - Spent Fuel Storage Pool

• •

The pool water level shall be maintained at a level of. 33 feet.

APPLICABILITY:

Whenever irradiated fuel assemblies are in the spent fuel storage pool.

ACTION:

With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the spent fuel storage pool area after placing the fuel assemblies and crane load in a safe condition. The provisioris of Specification ~;O.~ are not applicable •

Pool Water Level 3/4.1 O.H



The water level in the spent fuel storage pool shall be determined to be at least at its minimum required depth at least once per 7 days.


Procedures and Programs 6.8

ADMINISTRATIVE CONTROLS •=====================================

• ': • . • . ,

6.8.A

6.8.C

6.8.D

I

PROCEDURES AND PROGRAMS

Written procedures shall be established, implemented, and maintained covering the activities referenced below:

1. The applicable procedures recommended in Appendix A, of Regulatory Guide 1.33, Revision 2, February 1978,

· 2. The Emergency Operating Procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Section 7. 1 of Generic Letter No. 82-33,

3. Station Security Plan implementation,

4. Generating Station Emergency Response Plan implementation,

5. PROCESS CONTROL PRO<;lRAM (PCP) implementation,

6. OFFSITE DOSE CALCULATION MANUAL (ODCM) implementation, and

7. Fire Protection Program implementation.

Deleted.

Deleted

The following programs shall be established, implemented, and maintained:

1 • Reactor Coolant Sources Outside Primary Containment

This program provides controls to minimize leakage from those portions of systems outside primary containment that could contain highly radioactive fluids during a serious transient or accident to as ow as practical levels. The systems include CS, HPCI, LPCI, IC, process samplin , containment monitoring, and standby gas treatment systems. The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements, and

b. Leak test requirements for each system at a frequency of at least once per operating cycle.

DRESDEN - UNITS 2 & 3 6-9 Amendment Nos.

'I .>

. ·,,.:

•

••


ADMINISTRATIVE CONTROLS

4. Radioactive Effluent Controls Program

A program shall be provided conforming with 10 CFR 50. a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF C,..0 .'1-J>J__)

PUBL from radioactive ·e.ffluents as low as reasonably achievable. The program (1) shall be contained in the OOCM, (2) shall be implemented by station procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the operabilhy of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

,.,. cAl"5

b. Limitations on the instantaneou concentrations of radioactive material released in_ liquid effluents to lUNRESTRIC1 ED AREAS conforming to ten (10) times the concentration values in 10 CFR Part 20, Appendix 8, Table 2, Column 2 to 10 CFR Part 20.1001 - 20.2402,

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM,

d~ Limitations on the annual and quarterly doses to a NLC;oo.c.:~.H=--f1~~.i.c.Uj,,J radioactive materials in liquid effluents released from each Unit conforming to Appendix I to 10 CFR Part 50, ·

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days,

a A M F THE PUBLIC shall be an individual in a CONTROu.ED or UNRESTRICTED AREA is not a MEMB HE PUBLIC during any period in which the individual receives a.~---

b The CONTROllED AREA shall area, outside of a RESTRICTED ARi.::.--~u• access to which can be limited by the be

d to which is limited by the ee for the purpose of protecting individuals against undue risks exposure to radiation and radioactive maten ESTRICTED AREA!sl do not include areas used idential quarters, but separate rooms in a residential building a RESTRICTED

BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

DRESDEN - UNITS 2 & 3 Amendment Nos . 6-11

• Procedures and Programs 6.8


f. Limitations on the operability and use of the liquid and gaseous effluent treatment systems· to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31 -day period would exceed 2 percent of the guidelines for the annual dose conforming to Appendix I to 10 CFR Part 50_, .

g. Limitations on the dose rate resulting from radioactive materials released in

h.

gaseous effluents from the site to areas at or beyond the UNDA shall be limited to the following:

al For noble gases: less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, and

bl For lodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: less than or equal to a dose rate of 1 500 mrem/yr to any organ.

Limitatio~s on the annual and quarterly air doses resulting from noble gases ~ released m gaseous effluents from each Unit to areas beyond the ~1

@---! BOUNOAR'i conforming to Appendix I to 10 CFR Part 50, ~

·.• i. Limitations on the annual and quarterly doses to a(MCMaes OE THE PUBLI~ lodine-131, lodine-133, tritium, and all radionuclides in particulate form with halflives greater than 8 days in gaseous effluents released from each Unit conforming to Appendix I to 10 CFR Part 50,

j. Limitations on the annual dose or dose commitment to any (X!JE;.!M~B~u..i~H"tc~-~due to releases of radioactivity and to radiation from uranium fuel cycle ~ sources conforming to 40 CFR Part 190.

; ~RESDEN - UNITS 2·& 3 Amendment Nos. 6-12

Reporting Requirements 6.9


•

).

a..9. BEPORTING REQUIREMENTS

In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Regional Administrator of the appropriate Regional Office of the NRC unless otherwise noted.

6.9.A. Routine Reports

1. Deleted

2. Annual Report

Annual reports covering the activities of the Unit for the previous calendar year, as described in this section shall be submitted prior to of each year.

The reports required shall include:

a. Tabulation of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem/year and their associated · person rem exposure according to work and job functions, e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter or TLD. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions.

b. The results of specific activity analysis in which the reactor coolant exceeded the limits of Specification 3.6.J. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit. wes exceeded; (2) res~lts of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph ·of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the. steady-state level; and (5) The time duration when the specific activity of the·reactor coolant exceeded the radioiodine limit.

' DRESDEN - UNITS 2 & 3 Amendment Nos. 6-13

• Reporting Requirements 6;9


3. Annual Radiological Environmental Operating Report

The Annual Radiological Environmental Operating Report covering the operation of the Unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in ( 1) the ODCM and (2) Sections IV.B.2, IV.8.3, and IV.C of Appendix I to 10 CFR Part 50.

4. Radioactive Effluent Release Report

The Radioactive Effluent Release Report covering the operation of the facility during the previous calendar year shall be submitted prior to April 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility. The material provided shall be ( 1 ) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

5. Monthly Operating Report

Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety valves or safety/relief valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a. copy to the Regional Administrator of the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report.

6. CORE OPERA TING LIMITS REPORT

a. Core operating limits sh~ll be established and documented in the CORE . . OPERATING LIMITS REPORT before each reload cycle.or any remaining part of a reload cycle for the following:

(1) The Control Rod Withdrawal Block Instrumentation for Table 3.2.E-1 of Specification 3.2.E.

(2) The Average Planar Linear Heat Generation Rate (APLHGR) Limit for Specification 3. 11 .A. ,-r_rs)

(3) The~dy State Linear Heat Generation Rate ~Gm for Specification 3.11.D.

(4) The Minimum Critical Power Operating Limit (including 20% scram insertion time) for Specification 3.11 .C. This includes rated and off-rated flow conditions .

• DRESDEN - UNITS 2 & 3 6-14 Amendment Nos.

•• High Radiation Area 6.1 2


.fLI.2 HIGH RADIATION AREA

6. 12.A Pursuant to 10 CFR 20.1601 (c), in lieu of the requirements of paragraph 20.1601 of 10 CFR Part 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr at 30 cm (12 in.) shall be bi:!rricaded and conspicuously posted as a high radiation area and entranc thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP or equivalent document). Any individual or group of individuals permitted to enter such areas shall be rovided with or accompanied by one or more of the following:

ca) . 1 • A radiation monitoring evice which continuously indicates the radiation dose rate in

the area.

2. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or

3. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the RWP (or equivalent document).

~Health Physics penonnel or personnel escorted by health physics penonnel Shall be exempt from the RWP issuance requirements during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas . • DRESDEN - UNITS 2 & 3 Amendment Nos.

6-18

High Radiation Area 6.12

ADMINISTRATIVE CONTROLS •====-==-=-=---=-===-===-=-======-========

' ·•

6.12.B In addition to the requirements of 6.1 ~.A, areas accessible to personnel with radiation levels greater than 1000 mrem/hr at 30 cm (12 in.) from the radiation source or from any surface which the radiation penetrates shall require the following:

.· 1 . Doors shall be locked to prevent unauthorized. entry and shall not prevent individuals

from leaving the area. In place of locking the door, direct or electronic surveillanc.e that is capable of preventing unauthorized entry may be used. The keys shall be maintained under the administrative control of the Shift Manager on duty and/or health physics supervision.

2. Personnel access and exposure control requirements of activities being performed within these areas shall be specified by an approved RWP(or equivalent document).

3. Each. person entering the area shall be provided with an alarming radiation monitoring device that continuously integrates the radiation dose rate (such as an electronic dosimeter.) Surveillance and radiation monitoring by a Radiation Protection Technician may be substituted for an alarmi~g dosimeter.

4.

5. For individual HIGH RADIATION AREAS accessible to personnel with radiation levels of greater than 1000 mrem/h at 30 cm (12 in.) that are located within large areas where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual areas, then such individual areas shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device.


PCP 6.13


•

• \. / .

.6....1.J PROCESS CONTROL PROGRAM cpcp1

6. 13.A Changes to the PCP:

1 • Shall be documented and records of reviews performed. shall be retained. This documentation shell contain:

a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and,

b. A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable· regulations. L.-~ ·

2. Shall become effective after approvalV the ~ation Manager •

DRESDEN - UNITS 2 & 3 6-20

Amendment Nos.

ODCM 6.14

ADMINISTRATIVE CONTROLS ••

).

~ OEESJTE DOSE CALCULATION MANUAL tODCM)

6. 14.A Changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:


b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20. 1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix 1· to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoi t calculations.

~ 2. Shall become effective after: a:p:p:r:o:va=I~:::::=::..:=~~~----------. 3. Shall be submitted to the Commission in the form of a complete, legible copy of the

entire OOCM as a part of or concurrent with the Radioactive Effluent Report for the period of the report in which any change to the OOCM was made effective. Each change shall be identified by markings in the margin of the affected pages,· clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

- DRESDEN - UNITS 2 & 3 Amendment Nos. 6-21

•

• • c:llluplcleanestwpf

A TTACllMENT B (continued)

MARKED-UP TSUP PAGES QUAD CITIES NUCLEAR POWER STATION



• DEFINITIONS

SECTION PAGE

Section 1 DEFINITIONS

ACTION ............................................ . 1-1

AVERAGE PLANAR EXPOSURE (APE) ...................... . 1-1

AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) .... 1-1

CHANNEL .......................................... . 1-1

CHANNEL CALIBRATION ............................... . 1-1

CHANNEL CHECK .. · .................................. . 1-1

CHANNEL FUNCTIONAL TEST ........................... . 1-2

CORE ALTERATION ................................... . 1-2

•~· . CORE OPERATING LIMITS REPORT (COLR) .................. . 1-2

CRITICAL POWER RATIO (CPR) .......................... . 1-2

DOSE EQUIVALENT 1-131 ............................... . . 1-2

FRACTION OF LIMITING POWER DENSITY. (FLPD) .............. . 1-3

FRACTION OF RATED THERMAL POWER (FRTP) .............. . 1-3

~ FREOUEN~ NOTATION ............... : ............... .

~EL DESISM LIMITING RATIO (FOU"Xl ..................... .

~Eb OESIGN LIMITl~G RATIO fer CENTERLINE MELT (rOLAC)-.... . ,D 1-3

• IDENTIFIED LEAKAGE 1-3

LIMITING CONTROL ROD PATTERN (LCRP) .................. . 1-3

LINEAR HEAT GENERATION RATE (LH.GR) ................... . 1-3

LOGIC SYSTEM FUNCTIONAL TEST (LSFT) .................. . 1-3

MINIMUM CRITICAL POWER RATIO (MCPR) 1-3

Amendment Nos.


• LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

•

• •

SECTION

3/4.6

3/4.6.A

3/4.6.B

3/4.6.C

3/4.6.D

3/4.6.E

3/4.6.F

3/4.6.G

3/4.6.H

3/4.6.1

3/4.6.J

----e

3/4.6.K

3/4.6.L

3/4.6.M

3/4.6.N

3/4.6.0

3/4.6.P


Recirculation Loops .................................. .

Jet Pumps ........................................ .

Recirculation Pumps ................................. .

Idle Recirculation Loop Startup .......................... .

Safety Valves .......................... · ............ .

Relief Valves ....................................... .

Leakage Detection Systems ....... · ..................... .

Operational Leakage ................................. .

Chemistry ........................................ .


Specific Activity . . . . . . . . . . .. ·. . . . . . . . . . . . . . . . . . . . . . . . .

3/4.6-1

3/4.6-3

3/4.6-5

3/4.6-6

3/4.6-7

3/4.6-8

3/4.6-10

3/4.6-11

3/4.6-13

3/4.6-16

('.Table 4.6.J.·1, Reactor Cuolent Specifie Aeti'1it'f SaA'lple and A11al9sis Pre9ra~

Pressure/Temperature Limits ............................ . 3/4.6-19


Reactor Steam Dome Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . .

Main Steam Line Isolation Valves ........................ .

Structural lnte.grity ...........•.......................

Residual Heat Removal - HOT SHUTDOWN .................. .

Residual Heat Removal - COLD SHUTDOWN ................. .

3/4.6-22

3/4.6-23

3/4.6-24

3/4.6-25

3/4.6-27

QUAD CITIES - UNITS 1 & 2 VIII Amendment Nos.


• LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION PAGE

3/4. 10 REFUELING OPERATIONS

3/4.1 O.A Reactor Mode Switch ..................... ~ .......... . 3/4.10-1

3/4.10.B Instrumentation .................................... . 3/4.10-3

3/4.10.C Control Rod Position ......................... ~ ....... ; 3/4.10-5

3/4.10.D J)€L€(t:0:>

3/4.10-6

3/4.1 O.E Communications ................................... · .. 3/4.10-7

3/4.1 O.F DELETED ......................................... . 3/4.10-8

3/4.10.G Water Level - Reactor Vessel ........ ' ................... . 3/4.10-9

3/4.10.H Water Level - Spent Fuel Storage Pool ..................... . 3/4.10-10

/'- 3/4.10.1 !. . 3/4.1 O.J

Single Control Rod Removal ........................ ~· ... .

Multiple Control Rod Removal .................. '. ........ .

3/4.10-11

3/4.10-13

3/4.10.K Residual Heat Removal and Coolant Circulation - High Water Level ... 3/4.10-15

3/4.1 O.L Residual Heat Removal and Coolant Circulation - Low Water Level 3/4.10-16

• •

QUAD CITIES - UNITS 1 & 2 XII Amendment Nos.

TABLE OF CONTENTS

• BASES

SECTION

3/4.10

3/4.10.A

3/4.10.B

3/4.10.C

3/4.1 O.D

3/4.1 O.E

3/4.1 O.F

3/4.10.G

3/4.10.H

,....,.~ 3/4.10.1 • 3/4.10.J

3/4.10.K

3/4.1 O.L

•


Reactor Mode Switch ................................ .

Instrumentation .................................... .

Communications .................................... .

DELETED

Water Level - Reactor Vessel ............................ ·

Water Level - Spent Fuel Storage Pool ..................... .

Single Control Rod Removal ............................ .

Multiple Control Rod Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Shutdown Cooling and Coolant Circulation - High Water Level ..... .

Shutdown Cooling and Coolant Circulation - Low Water Level ..... .

B TOC

B 3/4.10-1

83/4.10-1

B 3/4.10-2

B 3/4.10-2

B 3/4.10-2

B 3/4.10-2

B 3/4.10-2

B 3/4.10-2

B 3/4.10-3

B 3/4.10-3

B 3/4.10-3

B 3/4.10-3

QUAD CITIES - UNITS 1 & 2 xxv Amendment Nos.

Definitions 1.0

• 1.0 DEFINITIONS

· rf·\ •

• •

OFFSITE DOSE CALCULATION MANUAL (ODCM) The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain ( 1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Specification 6.8 and (2) descriptions of the infer ation that should be included in the Annual Radiological Environmental Operating and nnual Radioactive Effluent Release Reports required by Specification 6.9.


OPERATIONAL MODE An OPERATIONAL MODE, i.e., MODE, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1-2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 . of the UFSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system component body, pipe wall or vessel wall.

QUAD CITIES - UNITS 1 & 2 1-4 Amendment Nos.

• • ., • 0 TABLE 4.1.A-1 :0 c m )> )>

0 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS () -;

() 0 -; :0 m Applicable CHANNEL ""CJ (/) :0

OPERATIONAL CHANNEL FUNCTIONAL CHANNEL181 0 -; c Functional Unit MODES CHECK TEST CALIBRATION m z ()

-; -; (/) 0 ..... 1 . Intermediate Range Monitor: z Ro

~:? (/)

N s1b1 -< a. Neutron Flux - High 2 S/Ulcl, w101 (/)

-; 3,4, 5 s w101 m

s:: b. Inoperative 2, 3, 4, 5 NA w101 NA

2. Average Power Range Monitor1'1:

w Setdown Neutron Flux - High 2 s1b1 s1u1c1, w101 - a. ~

..... 3, 51ml. s w I

s,~ "' w1d,e1, SA b. Flow Biased Neutron Flux - High 1 w

c. Fixed Neutron Flux - High 1 s w w1d1, SA

d. Inoperative 1, 2, 3, 5lml NA w NA

3. Reactor Vessel Steam Dome Pressure - High 1, 21il NA M Q

)> 4. Reactor Vessel Water Level - Low 1, 2 D M Elhl

3 CD ::J a. 3 5. Main Steam Line Isolation Valve - Closure 1 NA M E CD ::J ....+

z 6. Main Steam Line Radiation - High 1, 21il s M El Pl :0 0 ""CJ en (/)

w 7. Drywell Pressure - High 1, 21"1 NA M Q -~

..... )>

• REACTOR PROTECTION SYSTEM RPS 3/4.1.A

TABLE 4.1.A-1 (Continued)


TABLE NOTATION




(d) This calibration shall consist of the adjustment of the APRM CHANNEL to conform, within 2% of RATED THERMAL POWER, to the power values calculated by a heat balance during OPERATIONAL MODE 1 when THERMAL POWER is :?:25% of RATED THERMAL POWER. This adjustment must be accomplished: a) within 2 hours if the APRM CHANNEL is indicating lower power values than the heat balance, or b) within 12 hours if th.e APRM CHANNEL is indicating higher power values than the heat b.alance. Until any required APRM adjustment has been accomplished, notification shall be posted on the reactor control panel.

Any APRM CHANNEL gain adjustment made in compliance with Specification 3.11.B shall not be included in determining the above difference. This calibration is not required when THERMAL POWER is <25% of RATED THERMAL POWER. The provisions of Specification 4.0.D are not applicable.

(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.

(f) The LPRMs shall be calibrated at least once per 2000 effective full power hours (EFPH).

(g)

(h) Trip units are calibrated at least once per 31 days and transmitters are calibrate frequency identified in the table .

• (i) This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3. 12.A.


.(k) This function may be bypassed, provided a control rod block is actuated, for reactor protection system reset in Refuel and Shutdown positions of the reactor mode switch.

QUAD CITIES - UNITS 1 & 2 3/4.1-9 Amendment Nos.





(m) Requireq to be OPERABLE only prior to and during required SHUTDOWN MARGIN demonstrations performed per Specification 3.12.B.


(o) The provisions of Specification 4.0.D are not applicable to the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for a period of 24 hours after entering

. OPERATIONAL MODE 2 or 3 when shutting down from OPERATIONAL MODE 1.

(p) A current source provides an instrument channel' alignment every 3 months. cs The CllllNNEL _CALIBRATION s•rvaillaoce ••qui•o1Raoto s~all be perfor"1ed if Rat perfe~ -within the 13revm1:1s seven days _ .

•

• •

QUAD CITIES - UNITS 1 & 2 3/4.1-10 . Amendment Nos.

• .! .,, • 0 TABLE 3.2.A-1 z c CJ) )> -i Cl ISOLATION ACTUATION INSTRUMENTATION :::c () c -i ~

m m z CJ) -i

Minimum Applicable )>

c Trip CHANNEL(s) per OPERATIONAL -i

z 0 =i Functional Unit SetE2ointm TRIP SYSTEM'11 MODE(s) ACTION z CJ)

....... 1 . PRIMARY CONTAINMENT ISOLATION f20 N a. Reactor Vessel Water Level - Low ~144 inches 2 1, 2, 3 20

b. Drywell Pressure - High'd1 ~2.5 psig 2 1, 2, 3 20

c. Drywell Radiation - High ~100 R/hr 1 1, 2, 3 23

b SECONDARY CONTAINMENT ISOLATION

w a. Reactor Vessel Water Level - Low!c.kl ~144 inches 2 1, 2, 3 & * 24 --.J::> b. Drywell Pressure - High!c,d,kl ~2.5 psig 2 1, 2, 3 24 N

I

~ w c. Reactor Building Ventilation Exhaust 2 1, 2, 3 & ** 24

Radiation - High'c,kl

d. Refueling Floor Radiation - High'c,kl ~100 mR/hr 2 1, 2, 3 & * * 24

3. MAIN STEAM LINE {MSL) ISOLATION

a. Reactor Vessel Water Level ~84inches 2 1, 2, 3 21 - Low Low (/)

0 )> Cl)

b. MSL Tunnel Radiation - High'b' ~15'h' x normal 2 1, 2, 3 21 ..... 3 5· CD background :::J :::J a. )> 3 c. MSL Pressure - Low ~825 psig 2 1 22 0 ..... CD c: :::J Cl) ..... d. MSL Flow - High'kl ~140% of rated 2/line 1, 2, 3 21 ..... z 5· 0 :::J (/) e. MSL Tunnel Temperature - High ~200°F 2 of 4 in 1, 2, 3 21 w

each of 2 sets --.J::> N )>

• e:·j • • p TABLE 3.2.E-1 (Continued) z c en )> --l 0 CONTROL ROD BLOCK INSTRUMENTATION :0 n c -I ~

m m z en -I


c Trip CHANNEL(s) per OPERATIONAL -I z o.

' -I Functional Unit Setpoint Trip Function11l MODE!sl ACTION z en --& ~· SOURCE RANGE MONITORS RO

Detector not full in!bl N a. NA 3 2 51 2 5 51

b. Upscale!cl ~1x105 cps 3 2 51 2 5 51

c. lnoperative!cl NA 3 2 51 2 5 51

w .._ ~

N I

w 4. INTERMEDIATE RANGE MONITORS --&

a. Detector not full irS- NA 6 2, 5 51

b. Upscale ~108/125 6 2, 5 51 of full scale

c. Inoperative NA 6 2, 5 51

d. Downscale!dl ~3/125 6 2, 5 51 n of full scale 0

)> :J .... 3 .....

0 (I> :J :0 a. 0 3 a. (I> CJ :J .... 0 z 0

7'" 0 CJ) CJ)

w .._ ~

N m

• • • .,I I

TA' 4.2.E-1 p z c CONTROL ROD BLOCK INSTRUMENTATION C/) )> -I 0 SURVEILLANCE REQUIREMENTS :::0 () c -I ~

m m CHANNEL Applicable z C/) -I

CHANNEL FUNCTIONAL CHANNEL OPERATIONAL )>

c Functional Unit CHECK TEST CALIBRATION181 MODE(s) -I z 0 =i z C/) __. 1. ROD BLOCK MONITORS S20 I\.) a. Upscale NA S/Ulb,cl, Mlcl Q 1 (di

b. Inoperative NA $/Ulb,cl, Mlcl NA 1 (d)

c. Downscale NA S/U(b,cl, Mic) Q 1 (di

2. AVERAGE POWER RANGE MONITORS

w a. Flow Biased Neutron Flux - High -~ S/Ulbl, M I\.) 1. Dual Recirculation Loop Operation NA SA 1

I

w 2. Single Recirculation Loop Operation NA S/Utbl, M SA 1 (.Tl

b. Inoperative NA S/U'b', M NA 1, 2, 5m

c. Downscale NA S/U'b', M

s5V 1

d. Startup Neutron Flux - High NA S/Utbi, M 2, 5lil

1:. SOURCE RANGE MONITORS () 0

)> a. Detector not full in10 NA S/U'b', w E 210, 5 :J .-+

3 .., 0

Cl> b. Upscale191 NA S/U'b', W E 2u1, 5 :J :::0 a. 0 3 c. lnoperative191 NA S/U'b', w NA 2u1, 5 a. Cl> !B :J .-+ 0

z 0 '1'

0 CJ) CJ)

w -~· I\.)

m

~

• • .\ ..,, • TABLE 4.2.E-1 (Continued) p z c CONTROL ROD BLOCK INSTRUMENTATION (/) )> -I 0 SURVEILLANCE REQUIREMENTS :a () c =i ~

m m z (/)

CHANNEL -I

Applicable )> c CHANNEL FUNCTIONAL CHANNEL OPERATIONAL -I z

Functional Unit CHECK TEST CALI BRA TION181 MODE(s) 0 -I z (/)

...... !20 4. INTERMEDIATE RANGE MONITORS

"' Detector not full i~ a. NA S/U'b', W E 201, 5

b. Upscale NA s1u1b11 w E~ 2lil, 5

c. Inoperative NA S/Ulbl, W NA 21il, 5

d. Downscale1h1 NA S/U'b', W ~ 21il, 5

(,.) -- §..:. SCRAM DISCHARGE VOLUME !SDVl .i::.

"' I

Water Level - High NA Q NA 1, 2, 51e1 (,.) a. en

b. SDV Switch in Bypass NA E NA 51eJ

() 0

)> :::J r+

3 ..., 0

CD :::J :a c. 0 3 c. CD CD :::J r+ 0 z 0

'7' 0 (/) (/)

(,.) --~ "' m

•

.....

•

• •




TABLE NOTATION




(d) With THERMAL POWER ~30% of RATED THERMAL POWER.

· (e) With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.10.1 or 3.1 O.J.

(f)

(g)

(h)

(i)


This function shall be automatically bypassed when the associated IRM channels are on range 8 or higher.

This function shall be automatically bypassed when the IRM channels are on range 1.

The provisions of Specification 4.0.D are not applicable to the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for entry into the applicable OPERATIONAL MODE(s) from OPERATIONAL MODE 1 provided the surveillances are performed within 12 hours after such entry.

(j) Required to be OPERABLE only during SHUTDOWN MARGIN demonstrations performed per Specification 3.12.B.

k.)..-fhe Cl IA t-'1'.(EL CALIBRATION s11rvema 0 ce re~1:1ireR'leflts snail be 13erferrRed within 12 l=leursup0n easl=I efltry i11to a11y OPERATIONAL MODE(s) f10111 OPERATIONAL MODE I If not perfenT1ed witl"lifl tl"le ('3fe'lfieus seven days

QUAD CITIES.- UNITS 1 & 2 3/4.2-37 Amendment Nos.

••

• ••




ACTION

ACTION 60 - a. With the number of OPERABLE accident monitoring instrumentation

ACTION 61-

ACTION 62-

CHANNEL(s) less than the Required CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 1 2 hours.

b. With.. the number of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Minimum CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours.

With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) . less than the Minimum CHANNEL(s) shown in Table 3.2.F-1, initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours, and:

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

b. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.B 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.

a. With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) one less than the Required CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 1 2 hours.

b .

f?:'"\~ With the number of OPERABLE accident monitoring instrumentation ~ CHANNEL(s) less than the Minimum CHANNEL(s) shown in Table 3.2.F-1 ·, . .

restore at least one inoperable CHANNEL to status within 7 days or be in at least HOT SHUTDOWN within the next 1 2 hours.


• TA •. 2.F-1 0 c )> 0 ()

-t m en


c z =i en

w ~ N

I

~

INSTRUMENTATION

1. Reactor Vessel Pressure

2. Reactor Vessel Water Level

3 Torus Water Level

4. Torus Water Temperature

5. Drywell Pressure - Wide Range.

6. Drywell Pressure - Narrow Range

7. Drywell Air Temperature

8. Drywell~ Concentration - Analyzer and Monitor

~ ~Safety & Relief Valve Position Indicators ~- - Acoustic & Temperature

~ource Range) Neutron Monitors

©-eforvwell Radiation Monitors

~orus Air Temperature

®--efrorus Pressure )> 3 <l> ::::J c. 3 <l> ::::J r+

z 0 en

CHANNEL CHECK

M

M

M

M

M

M

M

M

M

M

M

M

M

• z en -t :a c s: m

CHANNEL Applicable z

-t OPERATIONAL )>

CALIBRATION MODE(s) ::::! 0

E 1, 2 z

E 1, 2

E 1, 2

E 1, 2

E 1, 2

E 1, 2

E 1, 2

Q 1, 2

E 1, 2

E'bl 1, 2 E1a1 1, 2, 3

E 1, 2

E 1, 2 )> 0 0 a: <l> ::::J r+

s:: 0 ::::J ;:::+" 0 ., (/)

w -~ N "Tl

INSTRUMENTATION B 3/4.2

• BASES

•

• •


The control rod block functions are provided to prevent excessive control rod withdrawal so that the MINIMUM CRITICAL POWER RATIO (MCPR) does not go below the MCPR fuel cladding integrity Safety Limit. During shutdown conditions, control rod block instrumentation initiates withdrawal blocks to ensure that all control rods remain inserted to prevent inadvertent criticality.

The trip logic for this function is one-out-of-n; e.g., any trip on one of the six average power range monitors (APRMs), eight intermediate range monitors (IRMs), or four source range monitors (SRMs), will result in a rod block. The minimum instrument CHANNEL requirements assure sufficient instrumentation to assure that the single failure criterion is met. The minimum· instrument CHANNEL requirements for the rod block monitor may be reduced by one for a short period of time to allow for maintenance, testing, or calibration.


In the REFUEL MODE during SHUTDOWN MARGIN demonstrations and the STARTUP/HOT STANDBY OPERATIONAL MODE, the APRM rod block function setpoint is significantly reduced to provide the same type of protection in the REFUEL and STARTUP/HOT STANDBY OPERATIONAL MODE(s) 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 rod block monitor (RBM) 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. The trip setting is flow-biased. At low 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 the specified power level. 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.

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 ten above the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the MCPR fuel cladding integrity Safety Limit .

A downs~~;~n-on an APRM is an indication that the instrument has aile or is not sensitive · n either case, the instrument will not respond to changes in control rod motion, and the control rod motion is thus prevented.

The SRM rod blocks of low count rate and the detector not fully inserted assure that the SRMs are not withdrawn from the core prior to commencing rod withdrawal for startup. The scram

QUAD CITIES - UNITS 1 & 2 B 3/4.2-3 Amendment Nos.

REACTIVITY CONTROL Maximum Scram Times 3/4.3.D





•

•

•

The maximum scram insertion time of each control rod from the fully withdrawn position to 90% insertion, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed 7 seconds ..

APPLICABILITY:

OPERATIONAL MODE(s) 1 and 2.

ACTION:


1. Declare the control rod(s) exceeding the above maximum scram insertion time inoperable, and

2. When operation is continued with three


1 . For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER:


b. after a reactor shutdown that is greater than 120 days,

2. For specifically affected individual control rods<al following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods, and

or more control rods with maximum 3. For at least 10% of the control rods, on a rotating basis, at least once per 120 days of@:ower operati~

scram insertion times in excess of 7 seconds, perform Surveillance Requirement 4.3.D.3 at least once per 60 days of@wer operat1aj))------ Au.. CAfJs

· /+-LL cA-/>;>

With the provisions of the ACTION above not met, be in at least HOT SHUTDOWN within 1 2 hours .

a The provisions of Specification 4.0.D are not applicable provided this surveillance is conducted prior to exceeding 40% of RATED THERMAL POWER.


• REACTIVITY CONTROL


N. Economic Generation Control (EGC) System

The economic generation control (EGC) system may be in operation with automatic flow control provided:

{ij~core flow is within 65% to 100% of rated core flow, and

EGC 3/4.3.N



The economic generation control system shall be demonstrated OPERABLE by verifying that core flow is within 65 % to 100% of rated core flow and THERMAL POWER is ~20% of RA TED THERMAL POWER:

~THERMAL POWER is ~20% . THERMAL POWER.

APPLICABILITY

of-RATED ~rior to entry into EGC operation, and

©/ _ _/At least once per 12 hours while @ - operating in EGC.

•

•

OPERATIONAL MODE 1.

ACTION:

With core flow less than 65% or greater than 100% of rated core flow, or THERMAL POWER less than 20% of RA TED THERMAL POWER, restore operation to within the limits within one hour. Otherwise, immediately remove. the plant from EGC operation .

QUAD CITIES - UNITS 1 & 2 3/4.3-20 --~9-

Amendment Nos. ~~


• BASES

••

• •


·A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

The SHUTDOWN MARGIN limitation is a restriction to be applied principally to a new refueling pattern. Satisfaction of the limitation must be determined at the time of loading and must be such that it will apply to the entire subsequent fuel cycle. This determination is provided by core design calculations and administrative control of fuel loading patterns. These procedures include restrictions to allow only those intermediate fuel assembly configurations that have been shown to provide the required SHUTDOWN MARGIN.

Since core reactivity values will vary through core life as a fl.J"nction of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed during xenon free conditions and adjusted to 68°F to accommodate the current moderator temperature. The generalized form is that the reactivity of the core loading will be limited so the core can be made subcritical by at least R + 0.38% &k/k or R + 0.28% &k/k, as appropriate, with the strongest control rod fully withdrawn and all others fully inserted. Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of determination of the highest control rod worth, either analytically or by test. This is due to the reduced uncertainty in the SHUTDOWN MARGIN test when the highest worth control rod is determined by demonstration. When SHUTDOWN MARGIN is determined by calculations not associated with a test, additional margin must be added to the specified SHUTDOWN MARGIN limit to account for uncertainties in the calculation.

The value of R in units of % &k/k is the difference between the calculated beginning-of-life core reactivity, at the beginning of the operating cycle, and the calculated value of maximum core reactivity at any time later in the operating cycle, where it would be greater than at the beginning. The value of R shall include the potential SHUTDOWN MARGIN loss assuming full B4 C settling in all inverted poison tubes present in the core. R must be a positive quantity or zero and a new value of R must be determined for each new fuel cycle.

The value of % &k/k ·n the above expression is provided as a finite, demonstrable, subcriticality margin. This margin is verified using an in-sequence control rod withdrawal at the beginning-of-life fuel cycle conditions: This assures subcriticality with not only the strongest fully withdrawn but at least an R + 0.28% (or 0.38% margin beyond this condition. This reactivity characteristic has been a basic ssumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion following a scram signal. Any control rod that is immovable as a result of excessive friction or mechanical interference, or is known to be unscrammable, per Specification 3.3.C, is considered to be incapable of insertion following a scram signal. It is important to note that a control rod can be electrically immovable, but scrammable, and no increase in SHUTDOWN MARGIN is required for these control rods.



• 3. 5 - LIMITING CONDITIONS FOR OPERATION 4.5 - SURVEILLANCE REQUIREMENTS

•

• b

f

4. The automatic depressurization system (ADS) with at least 5 OPERABLE ADS valves.

APPLICABILITY:

OPERATIONAL MODE(s) 1 , 2'b1 and 3!bl.

ACTION:


a. With one CS subsystem inoperable, provided that the LPCI subsystem is OPERABLE, restore the inoperable CS subsystem to OPERABLE status within 7 days, or be in at least HOT SHUT.DOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours .

b. With both CS subsystems inoperable, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.

2: For the LPCI subsystem:

a. With one LPCI pump inoperable'n, provided that both CS subsystems are OPERABLE, restore the inoperable LPCI pump to OPERABLE status within 30 days,

2. Verifying that, when tested pursuant to Specification 4.0.E:

a. The CS pump in each subsystem develop a flow of at least 4500 · gpm against a test line pressure corresponding to a reactor vessel pressure of ;:::go psig.

b. Two LPCI pumps together develop a flow of at least 9,000 gpm against a test line pressure corresponding to a reactor vessel pressure of ;:::20 psig.

c. The HPCI pump develops a flow of at least 5000 gpm against a system head corresponding to reactor vessel pressure, when steam is being supplied to the turbine between 920 and 1005 psig!cl.


a. For the CS system, the LPCI subsystem, and the HPCI system, verify each system/subsystem actuates on an actual or simulated automatic initiation signal. Actual injection of coolant into the reactor vessel may be excluded from this test.

The HPCI system and ADS are not required to be OPERABLE when reactor steam dome pressure is s150 psig.

The provisions of Specification 3.9.A, Actions 4.a or 6.b are applicable to the LPCI subsystem such. that with an inoperable diesel generator, 'for the remaining OPERABLE diesel generator, both LPCI pumps (and their associated flow path) associated with that OPERABLE diesel generator, shall be OPERABLE . ..---......__ ___ _

The provisions of Specification 4.0.D are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test .

. QUAD CITIES - UNITS 1 & 2 3/4.5-2 Amendment Nos.

EMERGENCY CORE COOLING SYSTEMS


or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following 24 hours.

••

b. With the LPCI subsystem otherwise inoperable''', provided that both CS subsystems are OPERABLE, restore the LPCI subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following

24houre@

c. With the LPCI subsystem and one or both CS subsystems inoperable, be iii at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours!dJ.

3. With the HPCI system inoperable, provided both CS subsystems, the LPCI subsystem, the ADS and the Reactor Core Isolation Cooling (RCIC) system are OPERABLE, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to ~1 50 psig within the following 24 hours .

ECCS - Operating 3/4.5.A



1) The system develops a flow of ~5000 gpm against a system head corresponding to reactor vessel pressure, when steam is being supplied to the turbine between 150 and 180 psig'c'.


c. Performing a CHANNEL CALIBRATION of the ECCS discharge line "keep filled" alarm instrumentation.

d. Deleted.

of~eri.v \~e 1 e'\-\-e.r

Ad;~ f.c.

• f The provisions of Specification 3.9.A, Actions 4.a or 6.b are applicable to the LPCI subsystem such that with an inoperable diesel generator, for the remaining OPERABLE diesel generator, both LPCI pumps (and their associated flow path) associated with that OPERABLE diesel generator, shall be OPERABLE. 4--------

~ Whenever the two required RHR SDC mode subsystems are inoperable, if unable to attain COLD SHUTDOWN V --~ as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat

removal methods.

• c The provisions of Specification 4.0.D are not applicable provided the surveillance is performed within 12 hours

after reactor steam pressure is adequate to perform the test .


•

- ...

•




The specific activity of the reactor coolant shall be limited to ~0.2 µCi/gram DOSE EQUIVALENT 1-131 .

APPLICABILITY:

ONAL MODE(s) 1, 2 and 3. J 1,,v,'(,/, .J-"a.11'f !"f-a.1i1 sfe ~m lt11 e

1101 iSo/~fec:I.

ACTION:

1 . 4ii OPERATIONAi MOOE(s;) 1 2 or g;,A-_fith the specific activity of the reactor coolant > 0.2 µCi/gram DOSE EQUIVALENT 1-131 but ~4.0 µCi/gram DOSE EQUIVALENT 1-131 ffM:mftMI

ith the specific activity of the reactor coolant > 0.2 µCi/gram DOSE EQUIVALENT 1-1 31;

~t'Sal~~~aRalfYst!M'En:lrmremertts-c




.; defermr'11<!.- [}?SC EQutvt4LGNI I-131 011c:.e ;e.r 4 ~wrs aMJ ~esiore DbSc ~VIVA-LEN/ ']:"-131 -/e wt'/-/i,',, l1'M1'fs IN llftti:... 4~ hotAfS (a)

P fer ,/LM 48 Ao1,4IS) or -tel"' 3r~~ . 't1" . ,

wlfM fl..e. .speci~ic ac..flJ 1'ty of' #i.e. red.c.hr c"o\~"'t --, 4.o./-"c•/3r~Wl l)oS~ E(iu1vALENI r-13/} deter~~r1e.. bc.5€ E&utVALEAll £'-13/ once {'€(' 4 VlOlA.ITS) 'twl~ tSol.J-e -d. II VV\ ~~"' ste ~ l l f\eS w itlA \ "-1Z ho 1..<f> > or be '"' ~ l-ea.s+-1-loT 5t.llATDowN wd-~lf' tit~ f\e~t lZ. ho~rs eM.tfi I" CoL.l> SI-' IAT C)()NiJ t.>J\""" in ~e. .klfow~tl!j :z4 ~ I( \I'S •

QUAD CITIES - UNITS 1 & 2 3/4.6-16 Amendment Nos.·~



• ••

a.

he

mpling an analysis requiremen of Item .b of Table

or

in

4.6.J-1 u9til the sp ific activity reactor coolant is estored to . hin its limit .

·Specific Activity 3/4.6.J

4,6 - SURVEILLANCE REQUIREMENTS

QUAD CITIES - UNITS 1 & 2 3/4.6-17 Amendment Nos. <fR<"rfil

m en

~ c z

G\ =i \t\ en

..... ~

.......... "->

~ ::-\ Cf\ ~ ~

~ <..> ~

...... ~

\" O>

"" I .....

""" 00

~ ~ ~ )>

~ 3 CD :I a. 3 CD :I r+

z 0 en

Type. of Measurement ~ and al sis ~ 1. eta .and Gamma Activit0

Determi tion

2. Isotopic Ana sis for DOSE EQUIVALENT I- 31 Concentration

3. Isotopic Analysis fo~dine.

4. Isotopic Analy ·s of an Off-gas Sample lncludin Quantitative Measurements for e-133, Xe-135 and Kr-88

TABLE 4.6.J-1

At least once per 72 ti urs

st once per 31 days

a) At lea once per 4 hours, whe specific a tivity exceeds a limit, a required by CTION 2.

b) At least one sam e, between 2 and 6 hours following the hange in THERMAL POWER or off-gas lev as required by . ACTION 3 ..

Until the specific activity of the reactor coo nt system is restored to wit ·n its limits.

\,

• ""C :::0

s:: );> :::0 -< CJ)

-< CJ) -i m s:: CJ 0 c z 0 )> :::0 -< 1

1 la) 1

21•11

,3lal

1 I•>, 21•>

1

•

• •

CONTAINMENT SYSTEMS PCIVs 3/4. 7 .D

3. 7 - LIMITING CONDITIONS FOR OPERATION 4. 7 - SURVEILLANCE REQUIREMENTS

2. With one or more reactor instrumentation line excess flow check valves inoperable, operation may continue and the provisions of Specification 3.0.C are not applicable, provided that within 4 hours either:


b. The instrument line is isolated and the associated instrument is declared inoperable.


QUAD CITIES - UNITS 1 & 2 3/4.7-7

a. At least once per 31 days by verifying the continuity of the explosive charge.

At least once per 18 months by removin at least o explosive squib from xplosive valv~- .JZsuch that each expl~ squib {Dr

@ah cvf'lesj·1e V.aJ¥g w· I be tested at least once e months, and initiating the removed explosive squib(s). The replacement charge for the exploded squib(s) shall be from the same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch suc~.Jssfully fired. No squib shall remain in use beyond the expiration of its shelf-life or operating life, as applicable.

6. At the frequency specified by the Primary Containment Leakage Rate Testing Program, verify leakage for any one main steam line isolation valve when tested at Pt (25 psig) is ~11.5 scfh .

Amendment Nos. 169 & 165

CONTAINMENT SYSTEMS

• 3. 7 - LIMITING CONDITIONS FOR OPERATION

•



APPLICABILITY:


ACTION: .

1 . With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:


b. In OPERATIONAL MODE *, · suspend handling of irradiated fuel

in the secondary containment, CORE ALTERATION(s), and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.C are not applicable.

SBGT 3/4. 7. P




1 . At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours with the heaters operating.

2. At least once per 18 months or ( 1 ) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem by:

a. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of < 1 % and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and· the system flow rate is 4000 cfm ±10%.

b. Verifying within 31 days after removal that a laboratory analysis of a representative carbon· sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1 ~52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-D-3803-89, for a methyl iodide penetration of < 10%, when tested at 30°C and 70% relative humidity; and

•* When handling irradiated fuel in the secondary containment, during CORE ALTERATION(s), and operations

,- . with a potential for draining the reactor vessel. ·


•

• •

CONTAINMENT SYSTEMS SBGT 3/4. 7 .P

3. 7 - LIMITING CONDITIONS FOR OPERATION

ca=: Dole~ 0



c. Verifying a subsystem flow rate of 4000 cfm ± 10% during system operation when tested in accordance with ANSI N510-1980.

3". After every 1440 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-D-3803-89, for a methyl iodide penetration of < 10%, when tested at 30°C and 70% relative humidity.


a. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches water gauge while operating the filter train at a flow rate of 4000 cfm ± 10%.

b. Verifying that the filter train starts and isolation dampers open on each of the following test signals:



c. Verifying that the heaters dissipate 30 ± 3 kw when tested in accordance with ANSI N510-1989. This reading shall include the appropriate correction for variations from 480 volts at the bus .

Amendment Nos.

CONTAINMENT SYSTEMS

• _ 3. 7 - LIMITING CONDITIONS FOR OPERATION

With both standby gas treatment subsystems otherwise inoperable in OPERATIONAL MODE(s) 1,2 or 3, restore at least one subsystem to OPERABLE status within one hour, or tie in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.

~With both standby gas treatment ~ subsystems inoperable in

OPERATIONAL MODE •, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATION(s), and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.C are not applicable.

SBGT 3/4.7.P



6. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the in-place penetration and leakage testing acceptance criteria of < 1 % in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 4000 cfm ± 10%.

When handling irradiated fuel in the secondary containment, during CORE ALTERATION(s), and operations with a potential for draining the reactor vessel. --I

QUAD CITIES- UNITS 1 & 2 3/4.7-26 Amendment Nos.~

•

•

• •


3.8 - LIMITING CONDITIONS FOR OPERATION 4.8 - SURVEILLANCE REQUIREMENTS

2 .. In OPERATIONAL MODE*, with the control room emergency filtration system or the RCU inoperable, immediately suspend CORE ALTERATION(s), handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.

3. The provisions of Specification 3.0.C are not applicable in OPERATIONAL . MODE*.


b. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratorytesting criteria of ASTMD-3803-89, for a methyl iodide penetration of <0.50%, when tested at 30°C and 70% relative humidity; and

c. Verifying a system flow rate of 2000 scfm ± 10% during system operation when tested in ace~ with ANSI N510-1980.

4. After every~rs of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-D-3803-89, for a methyl iodide penetration of <0.50%, when tested at 30°C and 70% relative humidity.


a. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches water gauge while operating the filter train at a flow rate of 2000 scfm ± 10% .

Amendment Nos.


. • 3.8 - LIMITING CONDITIONS FOR OPERATION 4.8 - SURVEILLANCE REQUIREMENTS

•

QUAD CITIES - UNITS 1 & 2 3/4.a~a

b. Verifying that the isolation dampers close on each of the following signals:


2) Simulated automatic isolation signal.

c. Verifying that during the pressurization mode of operation,

. control room positive pressure is maintained at :=::1 /8 inch water gauge relative to adjacent areas during system operation at a flow rate :s;2000 scfm.

d. Verifying that the heaters dissipate 12 ± 1.2 kw when tested in accordance with ANSI N510 ~~ This reading shall include the appropriate correction for variations from 480 volts at the bus.

6. After each complete or partia_I replacement of an HEPA filter bank by verifying that the HEPA filter bank satisfies the in-place penetration and leakage testing acceptance criteria of <0.05% in accordance with ANSI N510-1980 while operating the system at a flow rate of 2000 scfm ± 10%.

7. After each complete or partial replacement of an charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the in-place penetration and leakage testing acceptance criteria of <0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbo.n refrigerant test gas while operating the system at flow rate of 2000 scfm ± 10%.

Amendment Nos.

•

.• ! " !

•






a. Both Unit engineered safety features 4160 volt buses:

1) For Unit 1 , Nos. 13-1 and 14-1 ,

2) For Unit 2, Nos. 23-1 and 24-1 .



2) For Unit 2, Nos. 28 and 29, and

c. The Unit 1 20 volt Essential Service Bus and Instrument Bus.

2. 250 volt D.C. power distribution, . consisting of: ·

a.

) ;-_/

b. 1) For Unit 1 , RB MCC Nos. 1 A and 1 B,

2) For Unit 2, RB MCC Nos. 2A and 2B.·

3. For Unit 1, 125 volt D.C. power distribution, consisting of:

a. TB Main Bus Nos. 1 A, 1 A-1 and 2A,

b. TB Reserve Bus Nos. 1 B and 1 B-1 , and

c. RB Distribution Panel No. 1.





For u~;~ i I

No. I,

1-o.- ~2. /.J" < 2..

TB /Y\cc


•

•








The specified RPS electric power monitoring CHANNEL(s) shall be determined OPERABLE·

{t,)

. @) APPLICABILITY: ';J

1. By performance of a CHANNEL FUNCTIONAL TEST each time the plant is in COLD SHUTDOWN for a period of more than 24 hours, unless performed in the previous 6 months.

OPERATIONAL MODE(s) 1, 2, 3, 41• 1 and 5.

ACTION:

· 1 . With one RPS electric power monitoring· CHANNEL for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable power monitoring CHANNEL to OPERABLE status within 72 hours or remove the associated RPS MG set or alternate power supply from service.

2. With both RPS electric power monitoring CHANNEL(s) for an . inservice RPS MG set or alternate power supply inoperable, restore at least one electric power monitoring CHANNEL to OPERABLE status within 30 minutes or remove the associated RPS MG set or alternate power supply from service .



2. At least once per 18 months by demonstrating the OPERABILITY of overvoltage, undervoltage, and underfrequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic, and output circuit breakers, and verifying the following setpoints:

a. Overvoltage ~129.6 volts AC

b. Undervoltage ~105.3 volts AC

c. Underfrequency ~55.4 Hz

Amendment Nos. 160 & 156


• ==BA=S=E=S===============================================================

-·

With suitable redundancy in components and features not available, the plant must be placed in a condition for which the Limiting Condition for Operation does not apply.

The term verify as used toward A.C. electrical power sources means to administratively check by examining logs or other information to determine if certain components are out-of-service for preplanned preventative maintenance, testing, or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component.

With one offsite circuit and one diesel genera.tor inoperable, individual redundancy is lost in both the offsite and onsite electrical power system. Therefore, the allowable outage time is more limited. The time limit.takes into account the capacity and capability of the remaining sources, reasonable time for repairs, and the low probability of a design basis event occurring during this period.

With both of the required offsite circuits inoperable, sufficient onsite A.C. sources are available to maintain the unit in a safe shutdown condition in the event of a design basis transient or accident. In fact, a simultaneous loss of offsite A.C. sources, a loss-of-coolant accident, and a worst-case single failure were postulated as a part of the design basis in the safety analysis. Thus, the allowable outage time provides a period of time to effect restoration of all or all but one of the offsite circuits commensurate with the importance of maintaining an A.C. electrical power system capable of meeting its design intent.

With two diesel generators inoperable there are no remaining standby A.C. sources. Thus, with an assumed loss of offsite electrical power, insufficient standby A.C. sources are available to power the minimum required ESF functions. Since the offsite electrical power system is the only source of A.C. power for this level of degradation, the risk associated with continued operation for a very short time could be less than that associated with an immediate controlled shutdown, which could result in grid instability and possibly a loss of total A.C. power. The allowable time to repair is severely restricted during this condition. The intent here is to avoid the risk associated with an immediate controlled shutdown and to minimize the risk associated with this level of degradation.

Surveillance Requirements are provided which assure proper circuit continuity for the offsite A.C. electrical power supply to the onsite distributfon network and availability of offsite A.C. electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure distribution buses and loads are connected to their preferred power source. The frequency is adequate since breaker position is not likely to change without the operator being aware of it and because status is displayed in the control room. Should the action provisions of this specification require an increase in frequency, this Surveillance Requirement assures proper circuit continuity for the available offsite A.C. sources during periods of degradation and potential information on common cause failures that would otherwise go undiscovered.

QUAD CITIES - UNITS 1 & 2 B 3/4.9-2 Amendment Nos.~


• BASES

••

•

Each battery of the D.C. electrical power systems is sized to start and carry the normal D.C. loads plus all D.C. loads required for safe shutdown on one unit and operations required to limit the

·consequences of a design basis event on the other unit for a period of 4 hours following loss of all A.C. sources. The battery chargers are sized to restore the battery to full charge under normal (non-emergency) load conditions. A normally disconnected alternate 125 volt battery is also provided as a backup for each normal battery. If both units are operating , the normal 125 volt battery must be returned to service within the specified time frame since the design configuration of the alternate battery circuit is susceptible to single failure and, hence, is not as reliable as the normal station circuit. During times when the other unit is in a Cold Shutdown or Refuel condition, an alternate 1 25 volt battery is available to replace a normal station 1 25 volt battery on a continuous basis to rovide a se wer source. n service, the nermally open breaker gn the DC Reserve 8bls is placed in the open position ar:id imsted i e., "tagged out."

With one of the required D.C. electrical power subsystems inoperable the remaining system has the capacity to suppo~ a safe shutdown and to mitigate an accident condition. However, a subsequent worst-case single failure would result in complete loss of ESF functions. Therefore, an allowed outage time is provided based on a reasonable time to assess plant statu.s as a function of the inoperable D.C. electrical power subsystem and, if the D.C. electrical power subsystem is not restored to OPERABLE status, prepare to effect an orderly and safe plant shutdown .

Inoperable chargers do not necessarily indicate that the D.C. systems are not capable of performing their post-accident functions as long as the batteries are within their specified parameter limits. With both the required charger inoperable and the battery degraded, prompt action is required to assure an adequate D.C. power supply.


Battery surveillance requirements are based on the defined battery cell parameter values. Category A defines the normal parameter limit for each designated pilot cell in each battery. The pilot cells are the average cells in the battery based on previous test results. These cells are monitored closely as an indication of battery performance. Category B defines the normal parameter limits for each connected cell. The term "connected cell" excludes any battery cell that may be jumpered out because of a degraded condition or for any other reason. Category B also defines allowable values for each connected cell. These values, although reduced, provide assurance that sufficient capacity exists to perform the intended function and maintain a margin of safety. When any battery parameter is outside the Category B allowable value, the assurance of sufficient capacity as described above no longer exists and the battery must be declared inoperable.

Verifying battery terminal voltage while on float charge for the batteries helps to ensure the· effectiveness of the charging system and the ability of the batteries to perform their intended function. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations.



• 3. 10 ~.LIMITING CONDITIONS FOR OPERATION

·D. --0~~~~


•

•

vessel.

ACTION:

hou , suspend all operations involving vement of irr(ldiated fuel in the reacto

frll s /;rec !NTE/1.i/ftJN/fLL'/ L&PT- 8UIJJZ


REFUELING OPERATIONS Pool Water Level 3/4.10.H

• 3. 10 - LIMITING CONDITIONS FOR OPERATION 4. 10 - SURVEILLANCE REQUIREMENTS

•


The pool water level shall be maintained at a level o~f 33feet.

, > -APPLICABILITY:


ACTION:

With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the spent fuel storage pool area after placing the fuel assemblies and crane load in a safe condition. The provisions- of Specification 3.0.C are not applicable .


The water level in the spent fuel storage pool shall be determined to be at least at· its minimum requfred depth at least once per 7 days.


•

• •


3. 10 - LIMITING CONDITIONS. FOR OPERATION

L. Residual Heat Removal and Coolant Circulation - Low Water Level

Two shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE, with each loop consisting of at least:

1. One OPERABLE RHR pump, and

2. One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL MODE 5, when irradiated fuel is in the reactor vessel and the water level is < 23 feet above the top of the reactor pressure vessel flange.

ACTION:

With less than the above required shutdown cooling mode loops of the RHR system OPERABLE, within one hour and at. least once per 24 hours thereafter, demonstrate the OPERABILITY of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown· cooling mode loop .

® /?/

RHR Low Water Level 3~.Y



1 . At least one shutdown cooling mode loop of the RHR system shall be verified to be capable of circulating reactor coolant at least once per 1 2 hours.

2. Monitor the reactor coolant temperature at least once per hour.

QUAD CITIES - UNITS 1 & 2 3/4.10-16 Amendment Nos. 157 & 153

•

• /

• •·


BASES

3/4.10.C Control Rod Position

The requirement that all control rods be inserted during other CORE AL TERATION(s) ensures that fuel will not be loaded into a cell witho.ut an inserted control rod.

3/4.1 O.D

0

time l=ias elapsed ta allo•N the radioactive deeay of the short lived fission proat1ets._ time is consistent with the assYmptions Ysed in the aeeidei:it analyses.


The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status regarding core reactivity conditions during movement of fuel within the reactor pressure vessel.

3/4.1 O.F DELETED

3/4. 1 O.G Water Level - Reactor Vessel


The r.estrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. This minimum water depth is consistent with the assumptions of the accident analysis .



• ADMINISTRATIVE CONTROLS

6.8 PROCEDURES AND PROGRAMS

6.8.A Written procedures shall be established, implemented, and maintained covering the activities referenced below:

6.8.B

1. The applicable procedures recommended in Appendix A, of Regulatory Guide 1.33, ··Revision 2;·February-l978, - .. - · · · · · · - · -· ·

2. The Emergency Operating Procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Section 7. 1 of Generic Letter No. 82-33,



5. PROCESS CONTROL PROGRAM (PCP) implementation,

6. OFFSITE D.OSE CALCULATION MANUAL. (ODCM) implementation, and

7. Fire Protection Program implementation.

Deleted

6.8.C Deleted

6.8.D The following programs shall be established, implemented, and maintained:

1 . Reactor Coolant Sources Outside Primary Containment

This program provides controls to minimi leakage from those portions of systems outside primary containment that could ontain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include CS, HPCI, LPCI, RCIC, process sampling, containment monitoring, and standby gas treatment systems. The program shall include the following:




Procedures and Pr9grams 6.8


••

• •


A program shall be provided conforming with 10 CFR 50. 36a for the control of radioactive effluents and for maintaining the doses t EMBERS OF T

~- ram radioactive effluents as low as reasonably achievable. The ~ program ( 1) shall be contained in the ODCM, (2) shall be implemented by station

a

b

c

d

--procedures;· and· ·(3)· shall· include· remedial ·actions to· be-taken whenever .. the·'program limits are exceeded. The program shall include the following elements:

a. Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM

w<J cA/5

b. Limitations on the instantaneous oncentrations of radioactive material released in liquid effluents to NRESTRICTED AREAS conforming to ten (10) times the concentration values in 10 CFR Part 20, Appendix 8, Table 2, Column 2 to 10 CFR Part 20.1001 - 20.2402,

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, . I"• dfS

d .. Limitations on the annual and quarterly doses to a from radioactive materials in liquid effluents released from each Unit conforming ~o Appendix I to 10 CFR Part 50,

e. Determination of cumulative and projected dose contributions from radioactive ·effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the OOCM at least every 31 days,

E PUBLIC shall be an individual in a CONTROLLED or UNRESTRICTED AREA is not a MEMBER OF UBLIC during any period in which the individual receives a

rea, access to which is limited by the license the purpose of protecting individuals against u risks from exposure to radiation and radioactive materials. RES D AREA(s) do not include area ed as. residential quarters, but separate rooms in a residential building may be apart as a

D AREA.

e SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee_·~-------~----~-------------------.../




•

f. Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose conforming to Appendix I to 10 CFR Part 50,

· · ·· g. · · Limitations-on·the dose ·rate-resulting .. from-radioactivewmaterials .released. in gaseous effluents from the site to areas at or beyond the SIT NDARY shall be limited to the following:

a) For noble gases: less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, .and ·'

b) For lodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: less than or equal to a dose rate of 1500 mrem/yr to any organ.

h. Limitations on the annual and quarterly air doses res~lting from noble gases released in gaseous effluents from each Unit to areas beyond the~

r'liO'<Af'Jf--48 conforming to Appendix I to 10 CFR Part 50, .........--~

i. Limitations on the annual and quarterly doses to a Ht\M_,..E,...,M,.,,.,.B'fl'ErrR.-;:;O;r:F:-:T:r;Hi"'tt='Er:P;,U1rB;iL11rf(j from lodine-131, lodine-133, tritium, ·and all radionuclides in particulate form with halflives greater than 8 days in gaseous effluents released from each Unit" conforming to Appendix I to 10 CFR Part 50, ,

j. Limitations on the annual dose or dose commitment to any t!!!~~"-'-'.......___,_,u.a.o PUBL.: due to releases of radioactivity and to radjation from uranium fuel cycle sources conforming to 40 CFR Part 190 .


•

•



REPORTING REQUIREMENTS



1. Deleted

2. Annual Report

Annual reports covering the activities of the Unit for the previous calendar year, as described in this section shall be submitted prior to of each year .


High Radiation Area 6. 12

•. ADMINISTRATIVE CONTROLS

•

6.12 HIGH RADIATION AREA

6.12.A Pursuant to 10 CFR 20.1601 (c), in lieu of the requirements of paragraph 20.1601 of 10 CFR Part 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr at 30 cm ( 12 in.) shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by

... requiring issuance. of .a Radiation .. Wor!<_ Pe.rm.it_ .IRWP ..:.. o.i:: ~guiy.51!.e_n~ c::f9c~!TI~_r:'l!l.~ ... ~r)_y . individual or group of individuals permitted to ente such areas shall be provided with or accompanied by one or more of the following:

1 . A radiation monitoring device which continuously indicates the radiation dose rate in the area.

2. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or

3. An individual qualified in radiation protection procedures with a radiation dose rate. monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the RWP (or equivalent document).

6.12.B In addition ~o the requirements of 6~ 12.A, above, areas accessible to personnel with radiation levels greater than 1000 mrem/hr at 30 cm ( 12 in.) from the radiation source or from any surface which the radiation penetrates shall require the following:

1 . Doors shall be locked to prevent unauthorized entry and shall not prevent individuals from leaving the area. .In place of locking the door, direct or electronic surveillance that is capable of preventing unauthorized entry may be used. The keys shall be maintained under the administrative control of the Shift Engineer on duty and/or health physics supervision.

2. Personnei access and exposure control requirements of activities being performed within these areas shall be specified by an approved RWP (or equivalent document) .

c:ealth Physics personnel or personnel escorted by health physics personnel shall be exempt from the R wP issuance

•

. requirements during the performance of their assigned radiation protection duties, provided they are otherwise , following plant radiation protection procedures for entry into high radiation areas.




,.

3. Each person entering the area shall be provided with an alarming radiation monitoring device that continuously integrates the radiation dose rate (such as an electroni~ dosimeter.) Surveillance and radiation monitoring by a Radiation Protection Technician may be substituted for an alarming dosimeter.

,c;.·'1····~---------------------------------,,,,,~ 4. During emergency situations whiel=I involve personnel injury or actions taken to~-

.·.;prevent major equipment..damage,.sur;.r.eillanee .~f'ld_.r~dia~ion ~o~!~.~ririg_ c:>F.. ~~e .YY()fk area by a qualified indiltidual may be substituted for the routine RWP J'}roeedt:Jre (or .equivalent doeument).

5. For individual HIGH RADIATION AREAS accessible to personnel with radiation levels of greater than 1000 mrem/h at 30 cm ( 12 in.) that are located within large areas where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual areas, then such individual areas shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device.


PCP 6.13

.• ADMINISTRATIVE CONTROLS

•

•

6.13 PROCESS CONTROL PROGRAM (PCP)


1 . Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

a. Sufficient i~f~r~~ti~~ to --~~pport .th~ cha~g~· together .wfo1.~tti'e appr~pr"iaie ..... analyses or evaluations justifying the change(s) and,

b. A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations. ~~

2. Shall become effective after approval

QUAD CITIES - UNITS 1 & -2 6-21 Amendment Nos.

ODCM 6.14

•• ADMINISTRATIVE CONTROLS

•

• . ; .. ~'.::/ ...

6. 14 OFFSITE DOSE CALCULATION MANUAL (ODCM)


1 . Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

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


b~ A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50. 36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. · ·

2. Shall become effective afte~ approv~Jj

3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Report for the period of the report in which any change to the ODCM was made effective. Each change shall be identified by mark.ings in the margin of the affected pages, clearly · indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

• )-ev;evJ dr1J ae!ce;;f~r'/Ce, f/'\C (u clin<j


•

• • c:lhluplcleanestwpf

ATTACHMENT C

REVISED TSUP PAGES FOR DRESDEN NUCLEAR POWER STATION



·• LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION


3/4.6.A Recirculation Loops .................... ' ........ · ....... . 3/4.6-1

3/4.6.B Jet Pumps .......................................... . 3/4.6-3

3/4.6.C Recirculation Pumps ................................... . 3/4.6-5

3/4.6.D Idle Recirculation Loop Startup · ............ , .............. . 3/4.6-6

. 3/4.6.E Safety Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.6-7

3/4.6.F Relief Valves ........................ · ................ . 3/4.6-8

3/4.6.G Leakage Detection Systems ............................. · .. 3/4.6-10

3/4.6.H Operational Leakage ............... · ......... · ............ . 3/4.6-11·

••• 3/4.6.I ·chemistry ... ;· ....................................... : .

Table 3.6.1-1, Reactor Coolant System· chemistry Limits

3/4.6-13

3/4.6.J Specific Activity ................ , ........... · ......... . 3/4.6-16

3/4.6.K Pressure/Temperature Limits .......................... : ... . . ' . . 3/4.6-19

Figure 3.6.K-1, Minimum Reactor Vessel Metal Temperature vs. Rx.V~ssel Pressure

3/4.6.L

3/4.6.M

3/4.6.N

• 3/4.6.0

3/4.6.P

•

Reactor Steam Dome Pressure .............. -. ............ . 3/4.6-22

Main Steam .Line Isolation Valves ......................... . 3/4.6-23

Structural Integrity ................................... ~ .. 3/4.6-24

Shutdown Cooling - HOT SHUTDOWN ...................... . 3/4.6-25

Shutdown Cooling - COLD SHUTDOWN 3/4.6-27

DRESDEN - UNITS 2 & 3 VIII Amendment Nos.


..• LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION


3/4.10.A Reactor Mode Switch .................................. . 3/4.10-1

3/4.10.8 Instrumentation ...................................... . 3/4.10-3

3/4.1 o.c Control Rod Position ................................... . 3/4.10-5

3/4.10.D DELETED

3/4.10.E Communications 3/4.10-7

3/4.1 O.F . DELETED .......................................... . 3/4.10-8

3/4.10.G Water Level - Reactor Vessel ...................... ; ..... . 3/4.10-9

3/4.10.H Water Level - Spent Fuel Storage Pool ...................... . 3/4.10-10

• 3/4.10.1

3/4.10.J

Single Control Rod Removal ........ , .................... .

Multiple Control Rod Removal· ............................. .

3/4.10-11

3/4.10-13

3/4.10.K Shutdown Cooling and Coolant Circulation - High Water Level 3/4.10-15

3/4.10.L Shutdown Cooling and Coolant Circulation - Low Water Level 3/4.10-16

• •

DRESDEN - UNITS 2 & 3 XII Amendment Nos.

•

•

TABLE OF CONTENTS

BASES,

SECTION


3/4.10.A Reactor Mode Switch .................................. .

3/4.1 O.B Instrumentation ...................................... .

3/4.1 o.c Control Rod Position ................................... .

3/4.1 O.D DELETED


3/4.10.F DELETED ....................... _ ................... .

3/4.10.G Water Level - Reactor Vessel ............................ .

3/4.10.H Water Level - Spent Fuel Storage Pool ...................... .

3/4.10.1 Single Control Rod Removal ............................. .

3/4.10.J Multiple Control Rod Removal ............................ .

3/4.10.K Shutdown Cooling and Coolant Circulation - High Water Level

3/4.10.L Shutdown Cooling and _Coolant Circulation - Low Water Level

B TOC

B 3/4.10-1

B 3/4.10-1

B 3/4.10-2

8 3/4.10-2

8 3/4.10-2

B 3/4.10-2

B 3/4.10-2

B 3/4.10-3

8 3/4.10-3

B 3/4.10-3

8 3/4.10-3

DRESDEN - UNITS 2 & 3 xxv Amendment Nos.

•

••

•

Definitions 1.0

1.0 DEFINITIONS

MINIMUM CRITICAL POWER RATIO <MCPR) The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

OFFSITE DOSE CALCULATION MANUAL (QDCM) The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calCulation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring AlarmfTrip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Specification .6.8 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specification 6.9.

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrum~ntation, controls, normal or emergency electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem,

. train, component or -device to perform its specified safety function(s) are also capable of performing their: related S\.lpport function(s).

OPERATIONAL MODE An OPERATIONAL MODE, i.e., MODE, shall be any orie inclusive combination of mode switch position and average reactor. coolant temperature as specified in Table 1-2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the :reactor core and related instrumentation and 1) described in Chapter 14 of the UFSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission. ·

PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE. shall be leakage throug~ a non-isolable fault in a reactor coolant system component body, pipe wall or vessel wall.


• • • 0 TABLE 4.1.A-1 :JJ :JJ m m )> en REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS n 0 -i m 0 z :JJ

Applicable CHANNEL "'O c :JJ z OPERATIONAL CHANNEL FUNCTIONAL CHANNEL181 0 =i -i en Functional Unit MODES CHECK TEST CALIBRATION m

n N :::! f20 0 w 1. Intermediate Range Monitor: z

en s;u1c1, w101 -<

a. Neutron Flux - High 2 s1b1 E101 en -f

3,4, 5 s w101 Elo> m s:: b. Inoperative 2,3,4,5 NA w101 NA

2. Average Power Range Monitor10:

w Setdown Neutron Flux - High 2 s1b1 s;uic>, w101 SA101 ..._ a. ~

...... 3, 5lm) s w101 SAio> I

-..J w1d.e>, SA b. Flow Biased Neutron Flux - High 1 S,D w

c. Fixed Neutron Flux - High 1 s w Wld>, SA


3. Reactor Vessel Steam Dome Pressure - High 1, 2rn NA M a

)> 4. Reactor Vessel Water Level - Low 1, 2 D M Elhl

3 CD :J

5. Main Steam Line Isolation Valve - Closure 1, 21P1 NA M E a. 3 CD :J r+

Main Steam Line Radiation - High 1, 2(i) s M Elql :JJ z 6. "ti 0 en en w

7. Drywall Pressure - High 1, 21nl NA M a ..._ ~ .......

)>

•

•




TABLE NOTATION


(b) The IRM and SRM channels shall be determined to overlap for at least (Yi) decades during each startup after entering OPERATIONAL MODE 2 and the IRM and APRM channels shall be determined to overlap for at least (Yi) decades during each controlled shutdown, if not performed within the previous 7 days.


(d) This calibration shall consist of the adjustment of the APRM CHANNEL to conform, within 2% of RATED THERMAL POWER, to the power values calculated by a heat balance during OPERATIONAL MODE 1 when THERMAL POWER is ~25% of RATED THERMAL POWER. This adjustment must be accomplished: a) within 2 hours if the APRM CHANNEL is indicating lower power values than the heat balance, or b) within 12 hours if the APRM CHANNEL is indicating higher power values than the heat balance. Until any required APRM adjustment has been accomplished, notification shall be posted on the reactor control panel.

Any APRM CHANNEL gain adjustment made in compliance with Specification 3.11.B shall not be included in determining the above difference. This calibration is not required when THERMAL POWER is < 25% of RATED THERMAL POWER. The provisions of Specification 4.0.D are not applicable.



(g) DELETED.

(h) Trip units are calibrated at least once per 31 days and transmitters are calibrated at the frequency identified in the table.



(k) This function may be bypassed, provided a control rod block is actuated, for reactor protection system reset in Refuel and Shutdown positions of the reactor mode switch .


•

•

•


TABLE 4.1 .A-1 (Continued)


(I) This function not required to be OPERABLE when THERMAL POWER is less than 45% of RATED THERMAL POWER.



(o) The provisions of Specification 4.0.D are not applicable to the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for a period of 24 hours after entering OPERATIONAL MODE 2 or 3 when shutting down from OPERATIONAL MODE 1.

(p) This function is not required to be OPERABLE when reactor pressure is less than 600 psig.

(q) · A current source provides an instrument channel alignment every 3 months .


• • • TABLE 4.2.B-1 0 z :::0 m ECCS ACTUATION INSTRUMENTATION en en SURVEILLANCE REQUIREMENTS

-f 0 :::0 m c z s::


-f z CHANNEL FUNCTIONAL CHANNEL OPERATIONAL )> =i Functional Unit CHECK TEST CALIBRATION MODE(s) -f en 0 !\.)

.L. CORESPRAY(CS)SYSTEM z

!20 w

a. Reactor Vessel Water Level - Low Low s M Qlfl 1, 2, 3, 4lbl, 5lbl

b. Drywell Pressure - Highldl NA M Q 1, 2, 3

c. Reactor Vessel Pressure - Low (Permissive) NA M Q 1, 2, 3, 4lbl, 5lbl

d. CS Pump Discharge Flow - Low (Bypass) NA Q Qlel 1, 2, 3, 4lb) I 5lb)

.£,. LOW PRESSURE COOLANT INJECTION {LPCI} SUBSYSTEM w -- Reactor Vessel Water Level - Low Low ~ a. s M Qlfl 1 I 2, 3, 4(b) I s(b) N

I

b. Drywell Pressure - Highldl ..... NA M Q 1, 2, 3 <X>

c. Reactor Vessel Pressure - Low (Permissive) NA M Q 1, 2, 3, 4lb) I 5lb)

d. LPCI Pump Discharge Flow - Low (Bypass) NA Q Q(e) 1, 2, 3, 4lb) I 5lb)

~ HIGH PRESSURE COOLANT INJECTION {HPCI} SYSTEM'81

a. Reactor Vessel Water Level - Low Low s M Qlfl 1, 2, 3

b. Drywell Pressure - Highldl NA M Q 1, 2, 3. m )> (") 3 c. Condensate Storage Tank Level - Low NA M NA 1, 2, 3 (") CD en ::I a. d. Suppression Chamber Water Level - High NA M NA 1, 2, 3 )> 3 0 CD

....+

::I e. Reactor Vessel Water Level - High (Trip) NA M Qlfl 1, 2, 3 c ....+

Q) ....+

z f. HPCI Pump Discharge Flow - Low (Bypass) NA Q Q 1, 2, 3

5· 0 ::I en

Manual Initiation NA E NA 1, 2, 3 w

g. --~ N

CD

0 :0 m CJ)

0 m z c z =i CJ)

N

~ w

w --~· N

I ..... lO

)>

3 CD ::J 0.. 3 CD ::J ....+

z 0 en

•·, ,.

Functional Unit

,.' TABLE'4.2.B-1 (Continued)

ECCS ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ·

CHANNEL CHECK

.CHANNEL FUNCTIONAL

TEST

4. AUTOMATIC DEPRESSURIZATION SYSTEMtal

a. Reactor Vessel Water Level - Low Low s M

b. Drywell Pressure - Hightdl NA M

c. Initiation Timer NA E

d. Low Low Level Timer NA E

e. CS Pump Discharge Pressure - High NA M (Permissive)

f. LPCI Pump Discharge Pressure - High NA M (Permissive)

Q... LOSS OF POWER

a. 4. 1 6 kv Emergency Bus Undervoltage NA E (Loss of Voltage~

b. 4. 1 6 kv Emergency Bus Undervoltage NA E (Degraded Voltage)

CHANNEL CALIBRATION

Q(f)

Q

E

E

Q

Q

E.

E


MODE(s)

1, 2, 3

1, 2, 3

1, 2, 3

1, 2, 3

1, 2, 3

1, 2, 3

1, 2, 3, 4tc)' 5tcl

1, 2, 3, 4tc)' 5tcl

z CJ)

-I ::0 c ~ m z -t )> -t 0 z

m n n CJ)

)> 0 ....+ c Q)

!:!'. 0 ::J

w :;;; N OJ

•

•

INSTRUMENTATION ECCS Actuation 3/4.2.B

TABLE 4.2.B-1 (Continued)

ECCS ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS

TABLE NOTATION

(a) Not required to be OPERABLE when reactor steam dome pressure is :!i: 150 psig.

(b) When the system is required to be!OPERABLE per Specification 3.5.B.

(c) Required when the associated diesel generator i~ required to be OPERABLE per SpecificatiQn 3.9.B. .

(d) This function is not required to be OPERABLE when PRl_MARY CONTAINMENT INTEGRITY is not required.

(e) Trip units are .calibrated c;it least once per 31 days and transmitters are calibrated at the . frequency identified in the table.

(f) Unit 2 transmitters are calibrated once per 18 months. Unit 2 trip units ;;ind Unit 3 level switches are calibrated .at t~e frequency .indentified in the table. ·

. ~ '

\


0 :::c m (/)

0 m z c z =i (/)

I\)

!20 w

)> 3 CD ::J c. 3 CD ::J .-+

z 0 (/)

• TABLE 4.2.C-1

ATWS - RPT INSTRUMENTATION SURVEILLANCE REQUIREMENTS

Functional Unit

1 . Reactor Water Level - Low Low

2. Reactor Vessel Pressure - High

CHANNEL CHECK

s s

CHANNEL FUNCTIONAL

TEST

a a

a Trip units are calibrated at least once per 92 days and transmitters are calibrated at the frequency identified in the table, ·

• CHANNEL

CALIBRATION

z (/) -i ::u c ~ m z -i )> -t 0 z

)> -t ~ (/)

0 ::c m (J)

0 m z

c z =i (J)

N

rzo Functional Unit

w 3. SOURCE RANGE MONITORS

a. Detector not ft:ill in1b1

b. Upscale1c1 ·

c. lnoperativelcl

w .._

•• . TABLE. 3.2.E-1 (Continued)

CONTROL ROD .BLOCK INSTRUMENTATION

Trip Set point

NA

::s 1 x 105 cps

NA

Minimum Applicable CHANNEL(s) per OPERATIONAb Trip Function1il MODE(s)

3 2 2 5

3 2 2 5

3 2 2 5

.JI. N

I 4. INTERMEDIATE RANGE MONITORS w 0

)> 3 CD :::J c. 3 CD :::J ,.... z 0 (/)

a. Detector not full in

b. Upscale

c. Inoperative

d. Downscale1d1

NA· 6 2,5

::s 108/125 6 2,5 of full scale

NA 6 2,5

;?: $/125 6 2,5 of full scale

.•' ..

•• z -f :JJ c s: m z -f )>

~

ACTION z

51 51

51 51

51 51

51

51

51

51

n 0 :::J ,.... .... Q. ::c 0 c. to 0 C')

7" (/)

w .._ .JI. N m

/

•• :' ,\ • • TABLE 4.2.E-1 0 z :::0 CONTROL ROD BLOCK INSTRUMENTATION m Cf)

(/) SURVEILLANCE REQUIREMENTS -f

0 :::0 m c z s:


-f z CHANNEL FUNCTIONAL CHANNEL OPERATIONAL )> =i Functional Unit CHECK TEST CALIBRATION181 MODE(s) -f Cf) 0 (\) z flO w .L. ROD BLOCK MONITORS

a. Ups~ale NA S/U(b,c), Mic) Q 1 (d).

b. Inoperative NA S/U(b,c), M(c) NA 1 (d)

c. Downscale NA S/U(b,c), M(c) Q 1 (d)

l.:. AVERAGE POWER RANGE MONITORS

w -- a. Flow Biased Neutron Flux - High ~

Dual Recirculation Loop Operation (\) 1. NA S/Ulbl, M. SA 1 I

w ~ 2. Single Recirculation Loop Operation NA S/U1b1, M SA 1

b. Inoperative NA S/Ulbl, M NA 1, 2, s(j)

c. Downscal,e NA s;u1b1, M Q 1

d. Startup Neutron Flux·_· Hi~h NA S/Ulbl, M SA 2, 5lil

~ SOURCE RANGE MONITORS n 0

)> a .. Detector not full in111 NA S/Ulbl, W E 2,(i) 5 ::J .+

3 .... 0

C1l b. Upscale1g1 NA S/U1b1; W E 2,(i) 5 ::J :::0 c. 0 3 c. lnoperative1g1 NA . S/Ulbl, W NA 2,(i) 5 c. C1l !!! ::J .+ 0

z 0

"" 0 (/) (/)

w ........ ~ (\)

m

·•· al. • t TABLE 4.2.E-1 (Continued) CJ z ::c CONTROL ROD BLOCK INSTRUMENTATION U> m U> SURVEILLANCE REQUIREMENTS

~

CJ :XJ m c z ~

m

c z ~ z CHANNEL Applicable )>

=i CHANNEL FUNCTIONAL CHANNEL OPERATIONAL ~ U> 0 N Functional Unit CHECK TEST CALIBRATION 181 MODE(s) z S20 w


a. Detector not full in NA s1u1b1, w E 2(i), 5

b. Upscale NA s1u1b1, w E 2(i), 5

c. Inoperative NA s1u1b1, w NA 2<il, 5

d. Downscale1h1 NA s1u1b1, w E 2<il, 5

w -. ~ SCRAM DISCHARGE VOLUME ~SDV} ..;:.

N I Water Level - High NA Q NA 1, 2, 5<eJ w a.

(J1

b. SDV Switch in Bypass NA E NA 5<eJ

n 0

)> :::> r+

3 .,...

CD 2..

:::> ::c a. &. 3 CD !!! :::> r+ 0

z 0 '1"

0 (/) (/)

w -. ..;:. N m

•

•


TABLE.4.2.E-1 (Continued)


. TABLE NOTATION



(c) Includes reactor manual control "relay select.matrix'.' systern input.

(d) With THERMAL POWER 2: 30% of RATED THERMAL POWER.

(e) With more than one control rod withdrawn. Not applicable to control rods. removed per Specification 3.10.1 or 3.1 O.J.

(f). This function shall be automatically bypassed if detectqr count rate is. > 100 cps or the IRM channels are on range 3 or higher.

(g) This function shali be automatically bypassed when the associate·d IRM channels are on range 8 or higher.

(h) This function shall be automatically bypassed when the IRM channels are on range 1.

(i) The provisions of Specification 4.0.D are not applicable to the CHANNEL FUNCTIONAL TEST ·and CHANNEL CALIBRATION surveillances for entry into the applicable OPERATIONAL MODE(s) from OPERATIONAL MODE 1 provided the surveillances are performed within .12 hours after such entry ·

(j) Required to be OPERABLE only during SHUTDOWN MARGIN demonstrations performed per Specification 3.12.B .


••

••




ACTION

ACTION 60 - a. With the number of OPERABLE accident monitoring· instrumentation

.ACTION 61-

{

CHANNEL(s) less than the Required CHANNEL(s) shown.in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours.

b. With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Minimum CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours.

With the number of OPERABLE accident monitoring instrumentation CHANNEL{s) · less than the Minimum CHANNEL(s) shown in Table 3.2.F-1, initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours, and:

a. Either.restore the inoperable CHANNEL(s) to OPERABLE status within 7 days of the event, 9r


, ACTION 62- a. With the number of OPERABLE accident mon.itoring instrumentation·

ACTION 63 -

CHANNEL(s) one less than the Required CHANNEL(s) shown in Table 3~2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 30 days ·or be . . . in at least HOT SHUTDOWN within the next 1 2 hours.

b. With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Minimum CHANNEL(s) shown in Table 3.2.F-1 restore at least one inoperabl~ CHANNEL to OPERABLE status wit_hin 7 days or be in at least HOT SHUTDOWN within the next 12 hours.

a. With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Required CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status prior to startup from. a COLD SHUTDOWN of longer than 72 hours.

b. With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Minimum CHANNEL(s) shown in Table 3.2.F-1,


0 :::0 m en 0 m z

c z =i en N

f20 w

)> 3 CD :> a. 3 CD :> ..... z 0 (/)

• TABLE 4.2.F-1


INSTRUMENTATION CHANNEL CHECK

1. Reactor Vessel Pressure M

2. Reactor Vessel Water Level M

3 Torus Water Level M

4. Torus Water Temperature M

5. Drywall Pressure - Wide Range M

6. Drywall Pressure - Narrow Range M

7. Drywall Air Temperature M

8. Drywall Oxygen Concentration M - Analyzer and Monitor

9. Drywall Hydrogen Concentration M - Analyzer and Monitor

10. Safety/Relief Valve Position Indicators Mlcl

- Acoustic & Temperature

11 . (Source Range) Neutron Monitors M

12. Drywall Radiation Monitors M

13. Torus Pressure M

., z en -I :::0 c ~ m

CHANNEL Applicable z

-I OPERATIONAL )>

CALIBRATION MODE!s} -I 0

SA 1, 2 z

SA1d1 1, 2

A 1, 2

A 1, 2

E 1, 2

Q 1, 2

E 1, 2

Q 1, 2

Q 1, 2

E 1, 2

Q(b) 1 I 2, 3

E1a1 1, 2 )> 0

Q 1, 2 ~. a. CD :> ..... ~ 0 :> ;:t: 0 .., (/)

w ~ N ,,

•

•




TABLE NOTATION

(a) CHANNEL 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.

(b) Neutron detectors may be excluded from the CHANNEL CALIBRATION.

(c) CHANNEL CHECK of the Acoustic Monitors shall consist of verifying the instrument threshold levels.

(d) Analog transmitters are calibrated every 18 months. The control room indicator for the analog transmitter is calibrated at the frequency identified the table .


• · .• I .r

•


BASES


The control rod block functions are provided to prevent excessive control rod withdrawal so that the MINIMUM CRITICAL POWER RATIO (MCPR) does not go below the MCPR fuel Cladding integrity Safety Limit. During shutdown conditions, control rod block instrumentation initiates withdrawal blocks to ensure that all control rods remain inserted to prevent inadvertent criticality.

The trip logic for this function is one-out-of-n; e.g., any trip of-one of the six average power range monitors (APRMs), eight intermediate range monitors (IRMs), or four source range monitors (SRMs), will result in a rod block. The minimum instrument CHANNEL requirements assure sufficient instrumentation to assure that the single failure criterion is met. The minimum instrument CHANNEL requirements for the rod block monitor may be reduced by one for a short period of time to allow for maintenance, testing, or calibration.

The APRM rod block function is flow-biased 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. ·

In the REFUEL and STARTUP/HOT STANDBY OPERATIONAL MODE(s), the APRM rod block function setpoint is sig~ificantly reduced to provide the same type of protection in the R,EFUEL and STARTUP/HOT ·STANDBY OPERATIONAL MODE(s) 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 rod block monitor (RBM) 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. The trip setting is flow-biased. At low 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 the specified power level. The worst-case single control rod withdrawal error is analyzed for each reload to assure that, ~ith the specific trip settings, r.od withdrawal is blocked before the MCPR reaches the fuel cladding integrity Safety Limit.

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 ten above the indicated level. Analysis of the worst-case ac.cident results in rod block action before MCPR approaches the MCPR fuel cladding integrity Safety Limit.

A downscale indication on an APRM is an indication that the instrument has failed or is not sufficiently sensitive. In either case, the instrument will not respond to changes in control rod motion, and the control rod motion is thus prevented. ·

The SRM rod blocks of low count rate and the detector not fully inser~ed assure that the SRMs are not withdrawn from the core prior to commencing rod withdrawal for startup. The scram discharge volume, high water level rod 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.

DRESDEN - UNITS 2 & 3 B 3/4.2-3 Amendment No.

REACTIVITY CONTROL Maximum Scram Times 3/4.3.D





• a


APPLICABILITY:


ACTION:

With the m_aximum scram insertion time of one or more control rods exceeding 7 seconds: · ·

1 . Dedare the control rod(s) e~ceeding the above maximum scram insertion time inoperable, and

2. When operation is continued with three or more control rods with maximum scram insertion times in excess of 7 seconds,. perform Surveillance

· Requirement 4.3.D.3 at least once per 60 days of POWER OPERATION .

. With the .Provisions of the ACTION(s) above not met, be in at least HOT SHUTDOWN within 1 2 hours .


l .· For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER:


b. after a reactor shutdown that is . greater than 120 days,

2. For specifically affected individual contr.ol rods181 following inaintenanc·e · .

. on or modification to the control rod or. control rod drive system which could

; affect the' scram inserti'on time of those specific con~rol .rods; and ·

3. For at least 10% of the control rods, oil a rotating basis, at leas_t once per 1 io days of POWER OPERATION ..

The provisions of Specification 4.0.D are not applicable provided this surveillance is conducted prior to exceeding 40% of RATED THERMAL POWER.


REACTIVITY CONTROL

• , 3.3 - LIMITING CONDITIONS FOR OPERATION ..-

•


The economic generation control (EGC) . system may be in operation with automatic flow control provided:

1. Core flow is within 65% to 100% of rated core flow, and

·2. THERMAL POWER is ~ 20% of RATED THERMAL POWER.

. APPLICABILITY·

OPERATIONAL MODE ·1.

'ACTION:

With core flow less than 65% or greater than 100% of-rated core flow, or THERMAL POWER less thari 20% of RATED THERMAL POWER, restore operation to within the limits within one hour. Otherwise, immediately remove the plant from EGC operation .

EGC 3/4.3.N



The economic generation control system shall be demonstrated OPERABLE by verifying that core flow is within 65% to · 100% of rated core flow and THERMAL POWER is 2!: 20% of RATED THERMAL POWER:

1 . Prior to entry into EGC operation, and

2. At least once per 12 hot.irs while operating in EGC .


•

•• ' . . •• •. ..J·

•


BASES



The SHUTDOWN MARGIN limitation is. a restriction to be applied principally to a new refueling pattern. Satisfaction of the limitation must .be determined at the time of loading and must be such that it will apply to the entire subsequent fuel cycle. This determination is provided by core design

. calculations and administrative control of fuel loading patterns. These procedures include restrictions to allow only· those intermediate fuel assemblX configurations that have been shown to provide the required SHUTDOWN MARGIN.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed during. xenon free conditions and adjusted to 68 ° F to accommodate the current moderator temperature. The generalized form is that the reactivity of the core loading will be limited so the core can be made subcritical by at least R + 0.38% Ak/k or R + 0.28% Ak/k, as appropriate, with the strongest control r_od fully withdrawn and all others fully inserted. Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of determination of the highest control rod worth, either analytically or by test. This is due to the reduced uncertainty in the SHUTDOWN MARGIN test when the highest worth control rod is determined by demonstration. When SHUTDOWN MARGIN is determined by calculations not associated with a test, additional margin must be added to the specified SHUTDOWN MARGIN limit to account for uncertainties in the calculati_on. ·

The value of R in units of % Ak/k is the difference between the calculated beginning-of-life core reactivity, at the beginning of the operating cycle, and the calculated value of maximum core reactivity at any time later in the operating cycle, where it would be greater than at the beginning. The value of R shall include the potential SHUTDOWN MARGIN loss assuming full 840 settling in all inverted poison tubes present in the core. R must be a positive quantity or zero and a new value of R must be determined for each new tuel cycle.

The value of % Ak/k in the above expression is provided as a finite, demonstrable, subcriticality margin. This margin is verified using an in-sequence control rod withdrawal .at the beginning-of-life fuel cycle conditions.. This assures subcriticality with not only the strongest fully withdrawn but at least an R + 0.28% Ak/k (or 0.38% Ak/k) margin beyond this condition. This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated atthe time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion following a scram signal. Any control rod that is immovable as a res.ult of excessive friction or mechanical interference, or is known to be unscrammable, per Specification 3.3.C, is considered to be incapable of insertion following a scram signal. It is important to note that a control rod can be electrically immovable, but scrammable, and no increase in SHUTDOWN MARGIN is required for these control rods.


• ./



APPLICABILITY:

OP~RATIONAL MODE(s) 1, 2tbl and 3tbl.

ACTION:


a. With one CS subsystem inoperable, provided that the LPCI subsystem is OPERABLE, restore the inoperable CS subsystem to OPERABl,.E status within 7 days, or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following 24 hours.

b. With both CS subsystems inoperable; be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.

2. · For the LPCI subsystem:

a. With one LPCI pump inoperable1d1,

provided that both CS subsystems are OPERABLE, restore the inoperable LPCI pump to OPERABLE status within 30 days, or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following 24 hours.



b. Three LPCI pumps together develop a flow of at least 14,500 gpm against a test line pressure corresponding to a reactor vessel pressure of ~ 20 psig.

c. The HPCI pump develops a flow of at least 5000 gpm against a system head corresponding to reactor vessel pressure, when steam is being supplied to the turbine between 920 and 1005 psigtcl.


a. For the CS system, the LPCI subsystem, and the HPCI system, verify. each system/subsystem actuates on an actual or simulated · automatic initiation signal. Actual injection of coolant into the reactor vessel may be excluded from this test.


1 ) The system develops a flow of ~ 5000 gpm against a system. head corresponding to reactor vessel pressure, when steam is being supplied to the turbine between 1 50 and 350 psig 101 •

b The HPCI system and ADS are not required to be OPERABLE when reactor steam dome pressure is s 150 psig.

d The provisions Cif Specification 3.9.A, Actions 4.a or 6.b are applicable to the LPCI subsystem such that with an inoperable diesel generator, for the remaining OPERABLE diesel generator, both LPCI pumps (and their associated flow path) associated with that OPERABLE diesel generator, shall be OPERABLE. Otherwise, enter Specification 3.5.A., Action 2.c .

• c The provisions of Specification 4.0.D are not applicable provided the surveillance is performed within 12 hours after reactor steam pressu_re is adequate to perform the test.


• --

... , , •. :

., I . ../ .

•



b. With the LPCI subsystem otherwise inoperable1d1, provided that both CS subsystems are OPERABLE, restore the LPCI subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the

. following 24 hours.

c. With the LPCI subsystem and one or both CS subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.

3. With the HPCI system inoperable, · provided both CS subsystems, the LPCI

subsystem, the ADS and the Isolation Condenser (IC) system are OPERABLE, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 1 2 hours and reduce reactor steam dome pressure to ::5 1 50 psig within the following 24 hours.

4. For the ADS:

a. With .one of the above required ADS valves inoperable, provided the HPCI system, both CS subsystems and three LPCI pumps are OPERABLE, restore the . inoperable AOS valve to OPERABLE




c. Performing a CHANNEL CALIBRATION of the CS and LPCI system discharge line "keep filled" alarm instrumentation.

d. Deleted.

4. At least once per 18 months for the ADS:

a. Verify the ADS actuates on an actual or simulated automatic ·initiation signal. Actual valve actuation may be excluded from this test .

b. Manually. opening each ADS valve when the reactor steam dome . pressure is ~ 1 50 psig1c1 and observing that either:

1) The turbine control valve or · turbine bypass valve position responds accordingly, or

2) There is a corresponding . change·j'n the measured steam flow.

d . The provi.sions of Specification 3.9.A, Actions 4.a or 6.b are applicable to the LPCI subsystem such that with an inoperable diesel generator, for the remaining OPERABLE 'diesel generator, both LPCI pumps (and their associated flow path) associated with that OPERABLE diesel generator, shall be OPERABLE. Otherwise, enter Specification 3.5.A., Action 2.c .

c The provisions of Speci~ication 4.0.D are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.


• :....-'

•




The specific activity of the reactor coolant shall be limited to s0.2 µCi/gram DOSE EQUIVALENT 1-131.

APPLICABILITY:

OPERATIONAL MODE(s) 1, 2 and 3, with any main steam line not isolated;

ACTION:

1 . With the specific activity of the reactor coolant > 0.2 µCi/gram DOSE EQUIVALENT 1-131 but s4.0 µCi/gram DOSE EQUIVALENT 1-131, determine DOSE EQUIVALENT 1-131 once per 4 hours and restore DOSE EQUIVALENT 1-131 to within limits within 48 181

hours

2. With the specific activity of the reactor coolant > 0.2 µCi/gram DOSE EQUIVALENT 1-131, for greater than 48 hours or with specific activity of the reactor coolant > 4.0 µCi/gram DOSE EQUIVALENT 1-131, determine DOSE EQUIVALENT 1-131 once per 4 hours, and isolate all main steam lines within · 12 hours, or be in HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours .




In OPERATIONAL MODE 1 the specific activity of the reactor coolant shall be verfied to be s 0.2 µCi/gram DOSE EQUIVALENT 1-131 once per 7 days.

a The provisions of Specification 3.0.D are not applicable.


PRIMARY SYSTEM BOUNDARY Specific Activity 3/4.6.J ., ··' 3.6 - LIMITING CONDITION FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENT

THIS PAGE INTENTIONALLY LEFT BLANK

• DRESDEN - UNITS 2 & 3 3/4.6-17 Amendment Nos.

PRIMARY SYSTEM BOUNDARY Specific Activity 3/4.6.J

., 3.6 - LIMITING CONDITION FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENT

THIS PAGE INTENTIONALLY LEFT BLANK

•• DRESDEN - UNITS 2 & 3 3/4.6-18 Amendment Nos.

I

•'

'

•




Two1°1 shutdown cooling (SOC) loops shall be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling loop shall be in operation1bHc1, with each loop consisting of at least:

1. One OPERABLE SOC pump, and

2. One OPERABLE SOC heat exchanger.

APPLICABILITY:

OPERATIONAL MODE 3, with reactor vessel coolant temperature less than the SOC cut-in permissive setpoint.

ACTION:

1 . With less than the above required SOC loops OPERABLE, immediately initiate corrective action to return the required

. loops to OPERABLE status as soon as possible. Within 1 hour and at least once per 24 hours thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable SOC loop. Be in at least COLD SHUTDOWN within 24 hours1d1,

SOC- HOT SHUTDOWN 3/4.6.0



At least one SOC loop, one recirculation pump or alternate method shall be verified to be in operation and circulating reactor coolant at least once per 1 2 hours.

a One shutdown cooling loop may be inoperable for up to 2 hours for surveillance testing provided the other loop is OPERABLE and in operation.

b A shutdown cooling pump may be removed from operation for up to 2 hours per 8 hour period provided the other loop is OPERABLE.

c The shutdown cooling loop may be removed from operation during hydrostatic testing .

d Whenever two or more SDC loops are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.


•

•

CONTAINMENT SYSTEMS





Otherwise, be in at least HOT SHUTDOWN within the next 12 h.ours

· and in COLD SHUTDOWN within the following 24 hours.

PCIVs 3/4. 7 .D



b. At least once per 18 months by removing at least one explosive squib from an explosive valve such that each explosive squib will be tested at least once per 90 months, and initiating the removed explosive squib(s). The replacement charge for the exploded squib(s) shall be from the same manufactured batcti as the. one fired or from another batch which has been certified by having at least one of that batch successfully fired. No squib shall remain in use beyond the expiration of its shel.f-life or operating life, as applicable.

6. At the frequency specified by the Primary Containment L~akage Rate. Testing Program, verify leakage for any one main steam line isolation valve when tested at P1 (25 psig) is s 11.5 scfh .


•

•

•

CONTAINMENT SYSTEMS




APPLICABILITY:


ACTION:

1. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:


b. In OPERATIONAL MODE *, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATION(s), and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.C are not applicable.

2. With both standby gas treatment subsystems inoperable in OPERATIONAL MODE(s) 1,2 or 3, restore at least one subsystem to OPERABLE status within one hour, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours .

SBGT 3/4.7.P




1. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours with the heaters operating.

2. At least once per 18 months or (l)

after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem by:

a. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of < 1 % and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 4000 cfm ±10%.

b. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-D-3803-89, for a methyl iodide penetration of < 10%, when tested at 30°C and 70% relative humidity; and

* When handling irradiated fuel in the secondary containment, during CORE ALTERATION(s), and operations with a potential for draining the reactor vessel.


•

•

CONTAINMENT SYSTEMS


3. With both standby gas treatment subsystems inoperable in OPERATIONAL MODE *, suspend handling of irradiated fuel in the secondary containment, CORE AL TERATION(s), and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.C are not applicable.

SBGT 3/4. 7 .P


c. Verifying a subsystem flow rate of 4000 cfm ± 10% during system operation when tested in accordance with ANSI N510-1980.

3. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-D-3803-89, for a methyl iodide penetration of < 10%, when tested at 30°C and 70% relative humidity.


a. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches water gauge while operating the filter train at a flow rate of 4000 cfm :± 10%.


1 ) Manual initiation from the control room, and


c. Verifying that the heaters dissipate 30 ± 3 kw when tested in accordance with ANSI N510-1989. This reading shall include the appropriate correction for variations from 480 volts at the bus .

* When handling irradiated fuel in the secondary containment, during CORE Al TERATION(s), and operations with a potential for draining the reactor vessel.

DRESDEN - UNITS 2 & 3 3/4. 7-24 Amendment No.

•

•

CONTAINMENT SYSTEMS SBGT 3/4.7.P

3. 7 - LIMITING CONDITIONS FOR OPERATION 4. 7 - SURVEILLANCE REQUIREMENTS

DRESDEN - UNITS 2 & 3 3/4. 7-25

5. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the in-place penetration and leakage testing acceptance criteria of < 1 % in accordance with ANSI N 510-1980 while operating the system at a flow rate of 4000 cfm ± 10%.

6. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the in-place penetration and leakage testing acceptance criteria of < 1 % in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 4000 cfm ± 10% .

Amendment No.

•

• .



2. In OPERATIONAL MODE *,with the control room emergency filtration system or the RCU inoperable, immediately suspend CORE AL TERATION(s), handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.


DRESDEN - UNITS 2 & 3 3/4.8-7

b. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-D-3803-89, for a methyl iodide penetration of < 0.50%, when tested at 30°C and 70% relative humidity; and

c. Verifying a ~ystem flow rate of 2000 scfm ± 10% during system operation when tested in accordance with ANSI N510-1980.

4. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978·, meets the laboratory testing criteria of

c:; ASTM~D-3803-89, for a methyl iodide penetration of < 0.50%, when te_sted at 30°C and 70% relative humidity.


a. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches water gauge while operating the filter train at. a flow rate of 2000 scfm ± 1 0 % .

b. Verifying that the filter train starts and isolation dampers close on manual initiation from the control room .

Amendment Nos.


. • 3.8 - LIMITING CONDITIONS FOR OPERATION 4.8 - SURVEILLANCE REQUIREMENTS

-· DRESDEN - UNITS 2 & 3 3/4.8-8

c. Verifying that during the pressurization mode of operation, control room positive pressure is maintained at <?: 1 /8 inch water gauge relative to adjacent areas during system operation at a flow rate ~ 2000 scfm.

d. Verifying that the heaters dissipate 12 ± 1.2 kw when tested in accordance with ANSI N510-1989. This reaqing shall include the appr9priate correction for variations

· ' from 480 volts at the bus.

6. After each complete or partial replacement of an HEPA filter bank by verifying that the HEPA filter bank satisfies the in-place penetration arid leakage testing acceptance criteria of <0.05% in.accordance with ANSI N510-1980 while operating the system at' a flow rate of 2000 scfm ± 10%.

7. After each complete or partial replacement of an charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the in-place penetration and leakage testing acceptanc:e criteria of < 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant , test gas while operating the system a

1t

flow rate of 2000 scfm ± 10%.

Amendment Nos.

·~ .

•

PLANT SYSTEMS B 3/4.8

BASES

3/4.8.G Sealed Source Contamination

The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources, including startup sources and fission detectors, are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring or boron measuring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.8.H Explosive Gas Mixture

This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the offgas holdup system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 1 OCFR Part 50.

3/4.8.I Main Condenser Offaas Activity

Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 1 OCFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 1 OCFR Part 50. The release rates are determined at a more expeditious frequency following the determination of an increase of greater than 50%, as indicated by the air ejector noble gas monitor, after factoring out increases due to changes in THERMAL POWER level and off-gas flow in the nominal steady-state fission gas release from the primary coolant.

3/4.8.J Liquid Holdup Tanks

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of Appendix B, Table 2, Column 2, in unrestricted areas. Recirculation of the tank contents for the purpose of reducing the radioactive content is not considered to be an addition of radioactive material to the tank .


-·






a. Both Unit engineered safety features 41 60 volt buses:

1) For Unit 2, Nos. 23-1 and 24-1,

2) For Unit 3, Nos. 33-1 and 34-1.



2) For Unit 3, Nos. 38 and 39.

c. -The Unit 120 volt Essential Service Bus and Instrument Bus .

. 2. 250 volt D.C. power distribution, consisting of:

a. For Unit 2, TB MCC 2 and RB MCC 2,

b. For Unit 3, TB MCC 3 and RB MCC 3.

3. For Unit -2, 125 volt D.C. power distribution, consisting of:

a. TB Main Bus Nos. 2A-1 and 3A,

b. TB Res. Bus Nos. 2B and 2B-1 ,

c. Reserve Bus No. 2, and

d._ RB Distribution Panel No. 2.






•


3.9 - LIMITING CONDITIONS FOR. OPERATION


a


APPLICABILITY:

OPERATIONAL MODE(s) 1, 2, 3, 4 181 and 5lal

ACTION:

1 . With one RPS electric power monitoring CHANNEL for an inservice RPS MG set or alternate power supply inoperable, restore th~ inoperable power monitoring CHANNEL to OPERABLE status within 72 hours or remove. the associated RPS MG set or alternate power supply· from service.

2, With both RPS electric power monitoring CHANNEL(s) for an inservice RPS MG set or alternate power supply inoperable, restore at least one electric power monitoring CHANNEL t.o OPERABLE status within 30 minutes or remove the associated RPS MG set or alternate power supply from service .





The specifif;ld RPS electric power monitoring CHANNEL(s) shall be determined OPERABLE:

1. By performance of a CHANNEL FUNCTIONAL TEST1b1 each time the plant is in COLD SHUTDOWN for a period of more than 24 hours, unless performed in the previous 6 months.

2. At least once per 18 months by demonstrating the OPERABILITY of overvoltage, undervoltage, and underfrequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic, and output circuit breakers, and .verifying the following setpoints:

a; Overvoltage :s 129.6 volts AC

b. Undervoltage ;::::: 105.3 volts AC

c. Underfrequency ~ 55.4 Hz

b Only required to be performed prior to entering MODE 2 or 3 from MODE 4.


•


in components and features not available, the plant must be placed in a condition for which the Limiting Condition for Operation does not apply.

The term verify as used toward A.C. electrical power sources means to administratively check by examining logs or other information to determine if certain components are out-of-service for preplanned preventative maintenance, testing, or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component.

With one offsite circuit and one diesel generator inoperable, individual redundancy is lost in both the offsite and onsite electrical power system. Therefore, the allowable outage time is more limited. The time limit takes into account the capacity and capability of the remaining sources, reasonable time for repairs, and the low probability of a design basis event occurring during this period.

With both of the required offsite circuits inoperable, sufficient onsite A.C. sources are available to maintain the unit in a safe shutdown condition in the event of. a design basis transient or accident. In fact, a simultaneous loss of offsite A.C. sources, a loss-of-coolant accident, and a worst-case single failure were postulated as a part of the design basis in the safety analysis. Thus, the allowable outage time provides a period of time to effect restoration of all or all but one of the offsite circuits commensurate with the importance of maintaining an A.C. electrical power system capable of meeting its design intent.

With two diesel generators inoperable there are no remaining standby A.C. sources. Thus, with an assumed loss of offsite electrical power, insufficient standby A.C. sources are available to power the minimum required ESF functions. Since the offsite electrical power system is the only source of A.C. power for this level of degradation, the risk associated with continued operation for a very short time could be less than that associated with an immediate controlled shutdown, which could result in grid instability and possibly a loss of total A.C. power. The allowable time to repair is severely restricted during this condition. The intent here is to avoid the risk associated' with an immediate controlled shutdown and to minimize the risk associated with this level of degradation.

Surveillance Requirements are provided which assure proper circuit continuity for the offsite A.C. electrical power supply to the onsite distribution network and availability of offsite A.C. electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure distribution buses and loads are connected to their preferred power source. The frequency is adequate since breaker position is not likely to change without the operator being aware of it and because status is displayed in the control room. Should the action provisions of this specification require an increase in frequency, this Surveillance Requirement assures proper circuit continuity for the available offsite A.C. sources during periods of degradation and potential information on common cause failures that would otherwise go undiscovered .


•·

. ELECTRICAL POWER SYSTEMS B 3/4.9

BASES

Each battery of the D.C. electrical power systems is sized to start and carry the normal D.C. loads plus all D.C. loads required for safe shutdown on one unit and operations required to limit the consequences of a design basis event on the other unit for a period of 4 hours following loss of all A.C. sources. The battery chargers are sized to restore the battery to full charge under normal (non-emergency) load conditions. A normally disconnected alternate 1 25 volt battery is also provided as a backup for each normal battery. If both units are operating , the normal 125 volt battery must be returned to service within the specified time frame since the design configuration of the alternate battery circuit is susceptible to single failure and, hence, is not as reliable as the normal station circuit. During times when the other unit is in a Cold Shutdown or Refuel condition, an alternate 125 volt battery is available .to replace a normal station 125 volt battery on a continuous basis to provide a second available power source.

With one of the required D.C. electrical power subsy~tems inoperable the remaining system has the capacity to support a safe s_hutdown and to mitigate an accident condition. However, a subsequent worst-case single failure would result in complete loss of ESF .functions. Therefore, an allowed outage time is provided based on a reasonable time to _assess plant status as a function of the inoperable D.C. electrical power subsystem and, if the D.C: electrical power subsystem is not .

, restored to OPERABLE status, prepare to eff.ect an orderly and sate plant shutdown. ·'

Inoperable chargers do not necessarily indicate that the D.C. systems are not capable .of performing · their' post-accident functions as long as the batteries ~re within .their specified. parameter limits. · With both the required charger inoperable ~nd the battery degraded, proi:npt action is required to assure an adequate D .. C. power supply. . .

ACTION(s) are provided to delineate the measurements _and .time frames nee_ded to continue to assure OPERABILITY of the Station batteries when battery parameters are outside their identified limits.

Battery sur"veillance requirements are based on the defined battery· cell parameter values. Category A defines the normal parameter limit for each designated pilot cell in each battery. The pilot cells · are the average cells in the batt~ry based· on previous test results. These cells are monitored

. closely as an indication of batt~ry performance. Category B .defines the normal. parameter limits for each connected cell. The term "connected ce.11" excludes any battery cell that may be jumpered .out because of a degraded condition or for any other reason. Category B also defines allowable values for each connected cell. These values, although reduced, provide assurance that sufficient capacity exists to perform the intended function and maintain a margin of safety. When any battery p_arameter is outside the Category B allowable value, the assurance of sufficient capacity as described above no longer exists. and the battery must be·declared inoperable.

Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform .their intended function. The voltage requirements .are based on the nominal design voltage of the battery and are consistent vvith the initial voltages assumed in the battery sizing calculations.


REFUELING OPERATIONS Decay Time 3/4.1 O.D

\. 3.10 - LIMITING CONDITIONS FOR OPERATION 4.10 - SURVEILLANCE REQUIREMENTS

D. Deleted D. Deleted

. j .J THIS PAGE INTENTIONALLY LEFT BLANK

•• DRESDEN - UNITS 2 & 3 3/4.10-6 Amendment Nos.

REFUELING OPERATIONS Pool Water Level 3/4.10.H



The pool water level shall be maintained at a level of ~33 feet.

APPLICABILITY:


ACTION:

With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the spent fuel storage pool area after placing the fuel assemblies and crane load in a safe condition. The provisions of Specification 3.0.C are not applicable.





.'.,.~ BASES

. I

•

3/4.10.C Control Rod Position


3/4.1 O.D Deleted

3/4.10.E Communications

The requirement for communications capability ensures that refueling ~tation personnel can be promptly informed of significant changes in the facility status regarding core reactivity conditions during movement of fuel within the reactor pressure vessel.

3/4.10.F DELETED

3/4.10.G Water Level - Reactor Vessel

3/4.10.H Water Level - Spent Fuel Storage Pool

The restrictions on minimum water level ensure that sufficient water depth is available to remove ~9% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. This minimum water depth is consistent with the assumptions of the accident analysis .




PROCEDURES AND PROGRAMS


1. The applicable procedures recommended in Appendix A, of Regulatory Guide 1 .33, Revision 2, February 1978,

2. The Emergency Operating Procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Section 7.1 of Generic Letter No. 82-33,





,. 7. Fire Protection Program implementation.

•

6.8.B Deleted.

6.8.C Deleted.

6.8.D The following programs shall be established, implemented, and maintained:

1 . Reactor Coolant Sources Outside Primary Containment

This program provides controls to minimize leakage from those portions of systems outside primary containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include CS, HPCI, LPCI, IC, process sampling (post accident sampling of reactor coolant and containment atmosphere), containment monitoring, and standby gas treatment systems. The program shall include the following:



DRESDEN - UNITS 2 & 3 6-9 Amendment No.

•

•



2. In-Plant Radiation Monitoring

This program provides controls which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

a. Training of personnel,

b. Procedures for monitoring, and

c. Provisions for maintenance of sampling and analysis equipment.

3. Post Accident Sampling

This program provides controls which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and primary containment atmosphere samples under accident conditions. The program shall include the following:

a. Training of personnel,

b. Procedures for sampling and analysis,

c. Provisions for maintenance of sampling and analysis equipment .


•

•




A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The p'rogram ( 1) shall be contained in the ODCM, (2) shall be implemented by station procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

b. Limitations on the instantaneous concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to ten (10) times the concentration values in 10 CFR Part 20, Appendix B, Table 2, Column 2 to 10 CFR Part 20.1001 - 20.2402,


d. Limitations on the annual and quarterly doses to a member of the public from radioactive materials in liquid effluents released from each Unit conforming to Appendix I to 10 CFR Part 50,

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days,


••

•

•



f. Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose conforming to Appendix I to 10 CFR Part 50,

g. Limitations on the dose rate resulting from radioactive materials released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:

a) For noble gases: less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, and

b) For lodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: less than or equal to a dose rate of 1500 mrem/yr to any organ.

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each Unit to areas beyond the site boundary conforming to Appendix I to 10 CFR Part 50,

i. Limitations on the annual and quarterly doses to a member of the public from lodine-131, lodine~133, tritium, and all radionuclides in particulate form with halflives greater than 8 days in gaseous effluents released from each Unit conforming to Appendix I to 10 CFR Part 50,

j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

5. Primary Containment Leakage Rate Testing Program

A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemption. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Testing Program," dated September 1995.

The peak calculated primary containment internal pressure for the design basis loss of coolant accident, Pa' is 48 psig.

The maximum allowable primary containment leakage rate, La, at Pa' is 1.6% of primary containment air weight per day.




•

•

6.9 REPORTING REQUIREMENTS



1. Deleted

2. Annual Report

Annual reports covering the activities of the Unit for the previous calendar year, as described in this section shall be submitted prior to May 1 of each year.

The reports required shall include:

a. Tabulation of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem/year and their associated person rem exposure according to work and job functions, e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter or TLD. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions.

b. The results of specific activity analysis in which the reactor coolant exceeded the limits of Specification 3.6.J. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded; (2) results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the reactor coolant exceeded the radioiodine limit .


•

•



3. Annual Radiological Environmental Operating Report

The Annual Radiological Environmental Operating Report covering the operation of the Unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.8.2, IV.8.3, and IV.C of Appendix I to 10 CFR Part 50.

4. Radioactive Effluent Release Report

The Radioactive Effluent Release Report covering the operation of the facility during the previous calendar year shall be submitted prior to April 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

5. Monthly Operating Report

Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety valves or safety/relief valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report.

6. CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

(1) The Control Rod Withdrawal Block Instrumentation for Table 3.2.E-1 of Specification 3.2.E.

(2) The Average Planar Linear Heat Generation Rate (APLHGR) Limit for Specification 3. 11.A.

(3) The Steady State Linear Heat Generation Rate (SLHGR) for Specification 3.11.D.

(4) The Minimum Critical Power Operating Limit (including 20% scram insertion time) for Specification 3.11.C. This includes rated and off-rated flow conditions.


•

•



HIGH RADIATION AREA

6.12.A Pursuant to 10 CFR 20.1601 (c), in lieu of the requirements of paragraph 20.1601 of 10 CFR Part 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr at 30 cm (12 in.) shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)1•1 (or equivalent document). Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

1. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

2. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been 'established and personnel have been made knowledgeable of them; or


a Health Physics personnel or personnel escorted by health physics personnel shall be exempt from the RWP issuance requirem.ents during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.


•

••

•



6.12.B In addition to the requirements of 6.12.A, areas accessible to personnel with radiation levels greater than 1000 mrem/hr at 30 cm ( 12 in.) from the radiat.ion source or from any surface which the radiation penetrates shall require the following:

1. Doors shall be locked to prevent unauthorized entry and shall not prevent individuals from leaving the area. In place of locking the door, direct or electronic surveillance that is capable of preventing unauthorized entry may be used. The keys shall be maintained under the administrative control of the Shift Manager on duty and/or health physics supervision.

2. Personnel access and exposure control requirements of activities being performed within these areas shall be specified by an approved RWP(or equivalent document).

3. Each person entering the area shall be provided with an alarming radiation monitoring device that continuously integrates the radiation dose rate (such as an electronic dosimeter.) Surveillance and radiation monitoring by a Radiation Protection Technician may be substituted for an alarming dosimeter.

4. Deleted .

5. For individual HIGH RADIATION AREAS accessible to personnel with radiation levels of greater than 1000 mrem/h at 30 cm ( 12 in.) that are located within large areas where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual areas, then such individual areas shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device .


•

•

PCP 6.13


PROCESS CONTROL PROGRAM !PCP)




b. A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

2. Shall become effective after review and acceptance, including approval by the Station Manager .


•

•

ODCM 6.14


OFFSITE DOSE CALCULATION MANUAL (QDCMl

6.14.A Changes to the ODCM:


a. Sufficient information to support the change .together with the appropriate analyses or evaluations justifying the change(s) and,

b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

2. Shall become effective after review and acceptance, including approval by the Station Manager.

3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Report for the period of the report in which any change to the ODCM was made effective. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented .


•

• c:lllup\cleaneslwpf

ATTACHMENT C (continued)

REVISED TSUP PAGES FOR QUAD CITIES NUCLEAR POWER STATION


•

••

•

TABLE OF CONTENTS

DEFINITIONS

TOC

SECTION PAGE

Section 1 DEFINITIONS

ACTION .......................................... .

AVERAGE PLANAR EXPOSURE (APE) ..................... .

AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) .. .

CHANNEL ........................................ .

CHANNEL CALIBRATION ............................ · .. .

CHANNEL CHECK ................................... .

CHANNEL FUNCTIONAL TEST .......................... .

CORE ALTERATION .................................. .

CORE OPERATING LIMITS REPORT (COLR) ........... · ...... .

CRITICAL POWER RA TIO (CPR) ......................... .

DOSE EQUIVALENT 1-131 ............................. .

FRACTION OF LIMITING POWER DENSITY (FLPD) ............. .

FRACTION OF RA TED THERMAL POWER (FRTP) .............. .

FREQUENCY NOTATION .............................. .

IDENTIFIED LEAKAGE ................................ .

LIMITING CONTROL ROD PATTERN (LCRP) ................. .

LINEAR HEAT GENERATION RATE (LHGR) .................. .

LOGIC SYSTEM FUNCTIONAL TEST (LSFT) ................. .

MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) ... .

MINIMUM CRITICAL POWER RATIO (MCPR) ................. .

OFFSITE DOSE CALCULATION MANUAL (ODCM)

QUAD CITIES - UNITS 1 & 2 Amendment Nos.

1-1

1-1

1-1

1-1

1-1

1-1

1-2

1-2

1-2

1-2

1-2

1-3

1-3

1-3

1-3

1-3

1-3

1-3

1-3

1-3

1-4

•

~·

•

TABLE OF CONTENTS


SECTION


3/4.6.A Recirculation Loops .................................. .

3/4.6.8 Jet Pumps ........................................ .

3/4.6.C Recirculation Pumps ................................. .

3/4.6.D Idle Recirculation Loop Startup .......................... .

3/4.6.E Safety Valves

3/4.6.F Relief Valves

3/4.6.G Leakage Detection Systems ............................ .

3/4.6.H Operational Leakage ................................. .

3/4.6.I Chemistry ........................................ .


3/4.6.J Specific Activity .................................... .

3/4.6.K Pressure/Temperature Limits ............................ .

TOC

PAGE

3/4.6-1

3/4.6-3

3/4.6-5

3/4.6-6

3/4.6-7

3/4.6-8

3/4.6-10

3/4 .. 6-11

3/4.6-13

3/4.6-16

3/4.6-19


3/4.6.L

3/4.6.M

3/4.6.N

3/4.6.0

3/4.6.P

Reactor Steam Dome Pressure .......................... .

Main Steam Line Isolation Valves .................. ~ ..... .

Structural Integrity .................... · .............. .

Residual Heat Removal - HOT SHUTDOWN .................. .

Residual Heat Removal - COLD SHUTDOWN ................. .

3/4.6-22

3/4.6-23

3/4.6-24

3/4.6-25

3/4.6-27

QUAD CITIES - UNITS 1 & 2 VIII Amendment Nos.



SECTION PAGE


3/4.10.A Reactor Mode Switch ................................ . 3/4.10-1

3/4.10.8 Instrumentation .................................... . 3/4.10-3

3/4.1 o.c Control Rod Position ~ ........... : .................... . 3/4.10-5

3/4.1 O.D DELETED ......................................... . 3/4.10-6

3/4.1 O.E Communications ................................. · .... . 3/4.10-7

3/4.10.F DELETED ................. · ........................ . 3/4.10-8

3/4.10.G Water Level - Reactor Vessel ....... · .................... . 3/4.10-9

3/4.1 O.H Water Level - Spent Fuel Storage Poo.1 ..................... . 3/4.10-10

:• 3/4.10.1

. /'

Single Control Rod Removal .............. ·~ ........... . 3/4.10-11

3/4.1 O.J Multiple Control Rod Removal ........................... . 3/4.10-13

3/4.10.K Residual Heat Removal and Coolant Circulation - High Water Level ... · 3/4.10-15

3/4.10.L Residual Heat Remoyal and Coolant Circulation - Low Water Level ... 3/4.10-16

• QUAD CITIES - UNITS 1 & 2 XII Amendment Nos.

TABLE OF CONTENTS B TOC

• BASES

SECTION PAGE


3/4.10.A Reactor Mode Switch ................................. . B 3/4.10-1

3/4.10.8 Instrumentation .................................... . B 3/4.10-1

3/4.1 o.c Control Rod Position ................................. . B 3/4.10-2

3/4.1 O.D DELETED ......................................... . B 3/4.10-2

3/4.10.E Communications ...................................... . B 3/4.10-2

3/4.1 O.F DELETED B 3/4.10-2

3/4.1 O.G Water Level - Reactor Vessel ........................... . B 3/4.10-2

3/4.1 O.H Water Level - Spent Fuel Storage Pool ..................... . B 3/4.10-2 I. 3/4.10..1 Single Control Rod Removal ............................ . B 3/4.10-3

3/4'. 1 O.J Multiple Control Rod Removal .............. · ............. . B 3/4.10-3

3/4.10.K Shutdown Cooling and Coolant Circulation - High Water Level ..... . B 3/4.10-3

3/4.1 O.L Shutdown Cooling and Coolant Circulation - Low Water Level . B 3/4.10-3

• QUAD CITIES - UNITS 1 & 2 xxv Amendment Nos.

•

,.

-~-•

Definitions 1 .0

1.0 DEFINITIONS

OFFSITE DOSE CALCULATION MANUAL (ODCM) The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain ( 1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Specification 6.8 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specification 6.9.


OPERATIONAL MODE An OPERATIONAL MODE, i.e., MODE, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1 ~2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the UFSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system component body, pipe wall or vessel wall.


.,) ". •• p TABLE 4.1.A-1 JJ c m ):> ):>

Cl REACTOR PROTECTION SYSTEM INSTRUMENTATION. SURVEILLANCE REQUIREMENTS (") -i

("') 0 =i JJ m Applicable CHANNEL ""O (J) JJ

OPERATIONAL CHANNEL FUNCTIONAL CHANNEL181 0 -i c Functional Unit MODES CHECK TEST CALIBRATION m z (")

=i -i CfJ 0 ..... 1. Intermediate Range Monitor: z R° (J)

N -< a. Neutron Flux - High 2 s1b1 s;u1c1, w101 E101 CfJ

-i 3,4, 5 s w101 E101 m

~

b. Inoperative 2, 3,4, 5 NA w101 NA

2. Average Power Range Monitor1'1:

w Setdown Neutron Flux - High 2 s1b1 S/UIC) f WIO) SA101 -- a. ~

..... 3, 51m) s w101 SA 101 I

-...J b. Flow Biased Neutron Flux - High 1 S,'D w w1d,e1, SA

c. Fixed Neutron Flux - High s w w1d1, SA


3. Reactor Vessel Steam Dome Pressure - High 1, 2(ii NA M Q

):> 4. Reactor Vessel Water Level - Low 1, 2 D M Elhl

3 Cl)

::J a. 3 5. Main Steam Line Isolation Valve - Closure 1 NA M E Cl)

::J ...+

z 6. Main Steam Line Radiation - High 1, 2(ii s M EIPI JJ 0 ""O ~ CfJ

w 7. Drywell Pressure - High 1 2(nl NA M Q --

~ ~

):>




TABLE NOTATION




(d) This calibration shall consist of the adjustment of the APRM CHANNEL to conform, within 2% of RA TED THERMAL POWER, to the power values calculated by a heat balance during OPERATIONAL MODE 1 when THERMAL POWER is ~25% of RATED THERMAL POWER. This adjustment must be accomplished: a) within 2 hours if the APRM CHANNEL is indicating lower power values than the heat balance, or b) within 12 hours if the APRM CHANNEL is. indicating higher power values than the heat balance. Until any required APRM adjustment has been accomplished, notification shall be posted on the reactor control panel.·.

Any APRM CHANNEL gain adjustment made in compliance with Specification 3.11.B shall not. be included in determining the above difference. This calibration is not required when THERMAL POWER is < 25% of RATED THERMAL POWER. The provisions of Specification 4.0.D are not applicable.



(g) Deleted.

(h) Trip units are calibrated at least once per 31 days and transmitters are calibrated at the frequency identified in the table.


(j) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.10.1 or 3.10.J.

(k) This function may be bypassed, provided a control rod block is actuated, for reactor protection system reset in Refuel and Shutdown positions of the reactor mode switch.


•

••

•







(o) The provisions of Specification 4.0.D are not applicable to the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for a period of 24 hours after entering OPERATIONAL MODE 2 or 3 when shutting down from OPERATIONAL MODE 1 .

(p) A current source provides an instrument channel alignment every 3 months .


.) \./ • 0 TABLE 3.2.A-1 z c CJ) )> -i 0 ISOLATION ACTUATION INSTRUMENTATION jJ

() c ~ =i m

m z CJ) -i


c Trip CHANNEL(s) per OPERATIONAL -i

z 0 -i Functional Unit Setpointrn TRIP SYSTEMal MODE(s) ACTION z CJ)

..... 1 . PRIMARY CONTAINMENT ISOLATION ~ N a. Reactor Vessel Water Level - Low ~144 inches 2 1, 2, 3 20

b. Drywell Pressure -- High1dl $2.5 psig 2 1, 2, 3 20

c. Drywell Radiation - High $100 R/hr 1 1, 2, 3 23

2. SECONDARY CONTAINMENT ISOLATION

VJ a. Reactor Vessel Water Level - Low!c,kl ~144 inches 2 1, 2, 3 & * 24 --+>- b. Drywell Pressure - High!c,d,kl $2.5 psig 2 1, 2, 3 24 N

I

w c. Reactor Building Ventilation Exhaust $5 mR/hr 2 1, 2, 3 & * * 24 Radiation - High!c,kl

d. Refueling Floor Radiation - High!c.kl $100 m.R/hr 2 1, 2, 3 & * * 24

3. MAIN STEAM LINE (MSL) ISOLATION

a. Reactor Vessel Water Level ~84 inches 2 1, 2, 3 21 - Low Low (/)

0

)> Q)

b. MSL Tunnel Radiation - High1b1 $151h1 x normal 2 1, 2, 3 21 ...+ 3 ()" ct> background ::J ::J

)> a. 3 c. MSL Pressure - Low ~825 psig 2 1 22

(") ...+

ct> c ::J Q) ...+ d. MSL Flow - High1k1 $140% of rated 2/line 1, 2, 3 21 ...+

z 0 0 ::J

~ e. MSL Tunnel Temperature - High $200°F 2 of 4 in 1, 2, 3 21 VJ

each of 2 sets --+>-N )>

p c )> 0 (')

=i m CJ)

c z -i CJ)

!20 N

w ~ N

I

w

)>

3 CD :::J a. 3 CD :::J ....+

z .Q (/)

.i TABLE 3.2.E-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION

Functional Unit

3. SOURCE RANGE MONITORS

a. Detector not full in1b1

b. Upscale1cl

c. lnoperative1c1


a. Detector not full in

b. Upscale

c. Inoperative

d. Downscale1d1

Trip Setpoint

NA

NA

NA

~108/125

of full scale

NA

~3/125

of full scale

Minimum CHANNEL(s) per Trip Function°1

3 2

3 2

3 2

6

6

6

6


MODE(s)

2 5

2 5

2 5

2, 5

2, 5

2, 5

2, 5

••

ACTION

51 51

51 51

51 51

51

51

51

51

z CJ) -i :::0 c ~ m z -i )> -i 0 z

() 0 :::J ....+ ..., 0

:::0 0 a. ~ 0 C"l 7'" (/)

w --.+:>-N m

• / • • TABLE 4.2.E-1

0 z c CONTROL ROD BLOCK INSTRUMENTATION en )> -; 0 SURVEILLANCE REQUIREMENTS JJ () c

~ -; m m CHANNEL Applicable z en -;

CHANNEL FUNCTIONAL CHANNEL OPERATIONAL )>

c Functional Unit . CHECK TEST CALIBRATION1"1 MODE(s) -;

z 0 ~ z en

1 . ROD BLOCK MONITORS Re>

Upscale NA N a. S/Ulb,cl 1

Mic) Q ·1 Id)

b. Inoperative NA S/Ulb,c), Mic) NA 1 (d)

c. Downscale NA S/Ulb,c) 1

Mic) Q 1 ldl

2. AVERAGE POWER RANGE MONITORS

w -- a. Flow Biased Neutron .Flux - High ~

N 1. Dual Recirculation Loop Operation NA s;u1b11

M SA I

w (J1 2. Single Recirculation Loop Operation NA s;u1b1

1 M SA

b. . Inoperative NA S/Ulbl, M NA 1, 2, 5rn

c. Downscale NA s;u1b1, M SA

d. Startup Neutron Flux - High NA s;u1b11

M SA 2, 5rn

3. SOURCE RANGE MONITORS () 0

)> a. Detector not full in'1' NA s;u1b1; w E 21i1, 5 :J ...+

3 .... 0

CD b. Upscale191 NA s;u1b11

W E 21i1, 5 :J JJ a. 0 3 lnoperative191 NA s;u1b1

1 w NA 21i1, 5 a.

CD c. OJ :J

...+ 0

z (') .,,,. 0 (/)

~ w --~ N· m

• • ··• TABLE 4.2.E-1 (Continued) p z c CONTROL ROD BLOCK INSTRUMENTATION CJ) )> --l Cl SURVEILLANCE REQUIREMENTS :0 () c ~

s: m

m z CJ)

CHANNEL Applicable --l )>

c CHANNEL FUNCTIONAL CHANNEL OPERATIONAL ::::! z

Functional Unit CHECK TEST CALIBRATION1"1 MODE(s) 0 =i z CJ)

...... (2cl 4. INTERMEDIATE RANGE MONITORS N

a. Detector not full in NA s;u1b1, W E 2(i1, 5

b. Upscale NA s;u1b1, W E 2(i1, 5 r,

Inoperative NA S/U!bl, W NA 2(i1, 5 c.

d. Downscale1h1 NA S/Ulbl, W E 2(i1, 5

w -- 5. SCRAM. DISCHARGE VOLUME (SDV) +:-N

I Water Level - High NA Q NA 1, 2, 51"1 w a. O>

b. SDV Switch in Bypass NA E NA 5101

() 0

)> ::J ...... 3

..., 0

CD ::J :0 a. 0 3 a. CD OJ ::J 0 ...... z (")

7" 0 en ~ w --+:-

N m

•

/. f

•




TABLE NOTATION




(d) With THERMAL POWER 2'.30% of RATED THERMAL POWER.

(e) With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.10.1 or 3.10.J.

(f)

(g)

(h)

(i)


This function shall be automatically bypassed when the associated IRM channels are on range 8 or higher.

This function shall be automatically bypassed when the IRM channels are on range 1. ·

The provisions of Specification 4.0.D are not applicable to the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for entry into the applicable OPERATIONAL· MODE(s) from OPERATIONAL MODE 1 provided the surveillances are performed within 12 hours after such entry.

(j) Required to be OPERABLE only during SHUTDOWN MARGIN demonstrations performed per Specification 3. 12~B. ·


•

• ' ' -· ,

•


ACTION 60 -

ACTION 61-

ACTION 62-



ACTION

a. With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Required CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 1 2 hours.

b. With the number. of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Minimum CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 1 2 hours.

With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Minimum CHANNEL(s) shown in Table 3.2.F-1, initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours, and:

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


a. With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) one less than the Required CHANNEL(s) shown in Table 3.2.F-1, restore the inoperable CHANNEL(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 1 2 hours.

b. With the number of OPERABLE accident monitoring instrumentation CHANNEL(s) less than the Minimum CHANNEL(s) shown in Table 3.2.F-1; restore at least one inoperable CHANNEL to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 1 2 hours .


• ·,· ., TABLE 4.2.F-1

p z c ACCIDENT MONITORING INSTRUMENTATION (J) :t> -i 0 SURVEILLANCE REQUIREMENTS ::JJ (') c

~ -i m m Applicable z (J) -i

CHANNEL OPERATIONAL :t> c INSTRUMENTATION CHANNEL CHECK CALIBRATION MODE(s) -i z 0 =i 1 . Reactor Vessel Pressure M E 1, 2

z (J) _.

Rel 2. Reactor Vessel Water Level M E 1, 2

N 3 Torus Water Level M E 1, 2

4. Torus Water Temperature M E 1, 2

5. Drywell Pressure - Wide Range M E 1, 2

6. Drywell Pressure - Narrow Range M E 1, 2

7. Drywell Air Temperature M E 1, 2 w --+:> 8. Drywell Oxygen Concentration M Q 1, 2 N - Analyzer and Monitor I

~ N

9. Drywell Hydrogen Concentration M Q 1, 2 - Analyzer and Monitor

10. Safety & Relief Valve Position Indicators M E 1, 2 - Acoustic & Temperature

11. (Source Range) Neutron Monitors M Elbl 1, 2

12. Drywell Radiation Monitors M Ei•l 1, 2, 3 :t> (")

:t> 13. Torus Air Temperature M E 1, 2 (")

3 a. CD

CD 14. Torus Pressure M E 1, 2 :::J :::J .-+ a.

~ 3 CD 0 :::J :::J

. .-+ ;::;· .

z 0 .., 0 en ~

w --+:> N "Tl

••

•


BASES


The control rod block functions are provided to prevent excessive control rod withdrawal so that the MINIMUM CRITICAL POWER RATIO (MCPR) does not go below the MCPR fuel cladding integrity Safety Limit. During shutdown conditions, control rod block instrumentation i'nitiates withdrawal blocks to ensure that all control rods remain inserted to prevent inadvertent criticality.

The trip logic for this function is one-out-of-n; e.g., any trip on one of the six average power range monitors (APRMs), eight intermediate range monitors (IRMs), or four source range monitors (SRMs), will result in a rod block. The minimum instrument CHANNEL requirements assure sufficient instrumentation to assure that the single failure criterion is met. The minimum instrument CHANNEL requirements for the rod block monitor may be reduced by one for a short period of time to allow for maintenance, testing, or calibration.


In the REFUEL MODE during SHUTDOWN MARGIN demonstrations and the STARTUP/HOT STANDBY OPERATIONAL MODE, the APRM rod block function setpoint is significantly reduced to provide the same type of protection in the REFUEL and STARTUP/HOT STANDBY OPERATIONAL MODE(s) 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 rod block monitor (RBM) 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. The trip setting is flow-biased. At low 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 the specified power level. 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.

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 ten above the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the MCPR fuel cladding integrity Safety Limit.

A downscale indication on an APRM is an indication that the instrument has failed or is not sufficiently sensitive. In either case, the instrument will not respond to changes in control rod motion, and the control rod motion is thus prevented.

The SRM rod blocks of low count rate and the detector not fully inserted assure that the SRMs are not withdrawn from the core prior to commencing rod withdrawal for startup. The scram


•• REACTIVITY CONTROL


D. Maximum Scram Insertion Times ·


APPLICABILITY:


ACTION:


1. Declare the control rod(s) exceeding the above maximum scram insertion time inoperable, and

2. When operation is continued with three or more control rods with maximum scram insertion times in excess of 7 seconds, perform Surveillance Requirement 4.3.D.3 at least once per 60 days of POWER OPERATION.

With the provisions of the ACTION above not met, be in at least HOT SHUTDOWN within 1 2 hours.

Maximum Scram Times 3/4.3.D




1 . For all control rods prior to THERMAL POWER exceeding 40% of RA TED THERMAL POWER:

a. following CORE AL TERATION(s), or

b. after a reactor shutdown that is greater than 1 20 days,

2. For specifically affected individual control rods!•I following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods, and

3. For at least 10% of the control rods, on a rotating basis, at least once per 1 20 days of POWER OPERATION.

a The provisions of Specification 4.0.D are not applicable provided this surveillance is conducted prior to exceeding 40% of RATED THERMAL POWER.


•

•

REACTIVITY CONTROL



The economic generation control (EGC) system may be in operation with automatic flow control provided:

Core flow is within 65% to 100% of rated core flow, and

2. THERMAL POWER is ~20% of RA TED THERMAL POWER.

APPLICABILITY

OPERATIONAL MODE 1.

ACTION:

With core flow less than 65% or greater than 100% of rated core flow, or THERMAL POWER less than 20% of RA TED THERMAL POWER, restore operation to within the limits within one hour. Otherwise; immediately remove the plant from EGC operation .

EGC 3/4.3.N



The economic generation control system shall be demonstrated OPERABLE by verifying that core flow is within 65% to 100% of rated core flow and THERMAL POWER is ~20% of RA TED THERMAL POWER:

1. Prior to entry into EGC operation, and

2. At least once per 1 2 hours while operating in EGC.


••

•• -./

•


BASES ==================================================================================.L



The SHUTDOWN MARGIN limitation .is a restriction to be applied principally to a new refueling pattern. Satisfaction of the limitation must be determined at the time of loading and must be such that it will apply to the entire subsequent fuel cycle. This determination is provided by core design calculations and administrative control of fuel loading patterns. These procedures include restrictions to allow only those intermediate fuel assembly configurations that have been shown to provide the required SHUTDOWN MARGIN.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of S_HUTDOWN MARGIN will be performed during xenon free conditions and adjusted to 68 ° F to accommodate the current moderator temperature. The generalized form is that the reactivity of the core loading will be limited so the core can b~ made subcritical by at least R + 0.38%. ~k/k or R + 0.28% ~k/k, as appropriate, with th~ strongest control rod fully withdrawn a11d all others fully inserted. Two different values are supplied in the Limiting Condition • for Operation to provide for the different methods of determination of the highest control rod worth, either analytically or by test. This is due to the reduced uncertainty in the SHUTDOWN MARGIN test when the highest worth control rod is determined by demonstration. When SHUTDOWN MARGIN is determined by calculations not associated with a test, additional margin must be added. to the specified SHUTDOWN MARGIN limit to account for uncertainties in the calculation.

The value of R in units of % ~k/k is the difference between the calculated beginning-of-life core reactivity, at the beginning of the operating cycle, and the Cqlculated value of maximum core reactivity at any time later in the operating cycle, where it would be greater than at the beginning. The value of R shall include the potential SHUTDOWN MARGIN loss assuming full B4C settling in all inverted poison tubes present in the core. R must be a positive quantity or zero and a new value of R m.ust be determined for each new fuel cycle. ·

The value of % ~k/k in the above expression is provided as a finite, demonstrable; subcriticality margin. This margin is verified using an in-sequence control rod withdrawal at the beginning-of-life fuel cycle conditions. This assures subcriticality with not only the strongest fully withdrawn but at least an R + 0.28% ~k/k (or 0.38% ~k/k) margin beyond this condition. This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of if)sertion following a scram signal. Any control rod that is immovable as a result of excessive friction or mechanical interference, or is known to be unscrammable, per Specification 3.3.C, is considered to be incapable of insertion following a scram signal. It is important to note that a control rod can be electrically immovable, but scrammable, and no increase in SHUTDOWN MARGIN is required for these control rods.


••

~ ~t. _,

•



4. The automatic depressurization system (ADS) with at least 5 OPERABLE ADS valves.

APPLICABILITY:

OPERATIONAL MODE(s) 1, 21b1 and 31b1.

ACTION:


a. With one CS subsystem inoperable, provided that the LPCI subsystem is OPERABLE, restore the inoperable CS subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following 24 hours.

b. With both CS subsystems inoperable, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.

2. For the LPCI subsystem:

a. With one LPCI pump inoperable10, provided that both CS subsystems are OPERABLE, restore the inoperable LPCI pump to OPERABLE status within 30 days,



2. Verifying that, when tested pursuant to Specification 4.0.E:

a. The CS pump in each subsystem develop a flow of at least 4500 gpm against a test line pressure corresponding to a reactor vessel pressure of 2:90 psig.

b. Two LPCI pumps together develop a flow of at least 9,000 gpm against a test line pressure corresponding to a reactor vessel pressure of 2:20 psig.

c. The HPCI pump develops a flow of at least 5000 gpm against a system head corresponding to reactor vessel pressure, when steam is being supplied to the .turbine between 920 and 1005 psig1c1.


a. For the CS system, the LPCI subsystem, and the HPCI system, verify each system/subsystem actuates on an actual or simulated automatic initiation signal. 'Actual injection of coolant into the reactor vessel may be excluded from this test.

b The HPCI system and ADS are not required to be OPERABLE when reactor steam dome pressure is :;;150 psig.

f The provisions of Specification 3.9.A, Actions 4.a or 6.b are applicable to the LPCI subsystem such that with an inoperable diesel generator, for the remaining OPERABLE diesel generator, both LPCI pumps (and their associated flow path) associated with that OPERABLE diesel generator, shall be OPERABLE. Otherwise, enter Specification 3.5.A, Action 2.c. ·

c The provisions of Specification 4.0.D are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.


,_.. \ I •. ,

•



or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following 24 hour's.

b. With the LPCI subsystem otherwise inoperable(fl, provided that both CS subsystems are OPERABLE, restore the LPCI subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 1 2 hours and in COLD SHUTDOWN within the following 24 hours<d>.

c. With the LPCI subsystem and one or both cs subsystems inoperable, be in at least HOT SHUTDOWN within 1 2 hours and in COLD SHUTDOWN within the next 24 hours<d>.

3. With the HPCI system inoperable, provided both CS subsystems, the LPCI subsystem, the ADS and the Reactor Core Isolation Cooling (RCIC) system are OPERABLE, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to ::;150 psig within the following 24 hours.




1) The system develops a flow of ~5000 gpm against a system head corresponding to reactor vessel pressure, when steam is being supplied to the turbine between 150 and 180 psig!cl.


c. PerformiQg a CHANNEL CALIBRATION of the ECCS discharge line "keep filled" alarm instrumentation.

d. Deleted.

· The provisions of Specification 3.9.A Actions 4.a or 6.b are applicable to the LPCI subsystem such that with an inoperable diesel generator, for the remaining OPERABLE diesel generator, both LPCI pumps (and their associated flow path) associated with that OPERABLE diesel generator, shall be OPERABLE. Otherwise, enter Specification 3.5.A, Action 2.c.

d Whenever the two required RHR SOC mode subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practi,cal by use of alternate heat removal methods.

c The provisions of Specification 4.0.D are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test .


•

•




The specific activity of the reactor coolant shall be limited to ~0.2 µCi/gram DOSE EQUIVALENT 1-131.

APPLICABILITY:

OPERATIONAL MODE(s) 1, 2 and 3, with any main steam line not isolated.

ACTION:

1 . With the specific activity of the reactor coolant >0.2 µCi/gram· DOSE EQUIVALENT 1-1 31 but ~4.0 µCi/gram DOSE EQUIVALENT 1-131, determine DOSE EQUIVALENT 1-1 31 once per 4 hours and restore DOSE EQUl.VALENT 1-131 to within limits within 48 hours1a1•

2. With the specific activity of the reactor coolant >0.2 µCi/gram DOSE EQUIVALENT 1-1 31, for greater than 48 hours, or with the specific activity of the reactor coolant > 4.0 µCi/gram DOSE EQUIVALENT 1-131, determine DOSE EQUIVALENT 1-131 once per 4 hours, a'1d isolate all main steam lines within 1 2 hours, or be in at least HOT SHUTDOWN within the next

. 12 hours and iri COLD SHUTDOWN within· the following .24 hours .




In OPERATIONAL MODE 1, the specific activity of the reactor coolant shall be verified to be ~0.2 µCi/gram DOSE EQUIVALENT 1-131 once per 7 days.

a The provisions of Specification 3.0.D are not applicable.



•· 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS

• THIS PAGE LEFT INTENTIONALLY LEFT BLANK .

• QUAD CITIES - UNITS 1 & 2 3/4.6-17 Amendment Nos.



THIS PAGE LEFT INTENTIONALLY LEFT BLANK .


•

•

•

CONTAINMENT SYSTEMS






PCIVs 3/4.7.D



b. At least once per 18 months by removing at least one explosive squib from an explosive valve such that each explosive squib will be tested at least once per 90 months, and initiating the removed explosive squib(s). The replacement charge for the exploded squib(s) shall be. from the. same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch s.uccessfully fired. No ~ql,lib shall remain in use beyond the expiration of its shelf-life or operating life, as applicable .

6. At the frequency specified by the Primary Containment Leakage Rate Testing Program, verify leakage for any one main steam line isolation valve when tested at P1 (25 psig) is ::;11.5 scfh .


CONTAINMENT SYSTEMS SBGT 3/4.7.P

• 3. 7 - LIMITING CONDITIONS FOR OPERATION 4. 7 - SURVEILLANCE REQUIREMENTS

•

P. Standby Gas Treatment System P. Standby Gas Treatment System

*


APPLICABILITY:


ACTION:

1. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:

a. In OPERATIONAL MODE(s) 1,2 or 3, be in at least HOT SHUTDOWN within the next 12 hours and in

, COLD SHUTDOWN within the following 24- hours.

b. In OPERATIONAL MODE *, · suspend handling of irradiated fuel

in the secondary containment, CORE AL TERA TION(s), and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.C are not applicable.

2. With both standby gas treatment subsystems inoperable in OPERATIONAL MODE(s) 1,2 or 3, restore at least one subsystem to OPERABLE status within one hour, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.


1. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours with the heaters operat_ing.

2. At least once per 18 months or ( 1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsy~tem by:

a. Verifying that the subsystem satisfies the in-place penetration and bypass le~kage testing acceptance criteria of < 1 % and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 4000 cfm ±10%.

b. Verifying within 31 days after removal·that a laboratory analysis .of a representative carbon sample _ obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,I meets the laboratc;>ry testing criteria of ASTM-0-3803-89, for a methyl iodide penetration of < 10%, when tested at 30°C and 70% relative humidity; and

When handling irradiated fuel in the secondary containment, during CORE AL TERA TION(s), and operations with a potenti~I for draining the reactor vessel.

QUAD CITIES - UNITS 1 & 2 3/4.7-24 ~mendment Nos.

CONTAINMENT SYSTEMS

. · ,. 3.7 - LIMITING CONDITIONS FOR OPERATION

3. With both standby gas treatment subsystems inoperable in . OPERATIONAL MODE *, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATION(s), and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.C are -not appliCab1e:. - - . .. . - . . ,.

SBGT 3/4. 7. P


c. Verifying a subsystem flow rate of 4000 cfm ± 10% during system . operation when tested in accordance with ANSI N510-1980.

3. After every 1440 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory

·· ·- ·-· ··-· ·analysis··c:wa·repres·entative'·carbon · sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-D-3803-89, for a methyl iodide penetration of < 10%, when tested at 30°C and 70% relative humidity.


a. Verifying that the pressure drop across the combined HEPA filters and ·charcoal adsorber banks is < 6 inches water gauge while operating the filter train at a flow rate of 4000 cfm ± 10%.




c. Verifying that the heaters dissipate 30 ± 3 kw when tested in accordance with ANSI N510-1989. This reading shall include the

· appropriate correction for variations from 480 volts at the bus.

• When handling irradiated fuel in the secondary containment, during CORE AL TERA TION(s), and operations with a potential for draining the reactor vessel.


•

;. '· 7)) · .. i·d'

, CONTAINMENT SYSTEMS


. .. . . ... .. .. -.. -.... ·- ..... ,., ~-·- . ··~- .

SBGT 3/4.7.P



. .... - ..... -.. ~ ~--·-···· ... _, .............. ~ ........ ·-·····- .. , ........ ___ ~:. ......... :.. .. ....;. ... -.,;.,._ . ..;:.,·.~, .. · ........ , ...... '· .. -:

6. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the in-place penetration and leakage testing acceptance criteria of < 1 % in

· accordance with ANSI N510-1980 for a halogenated hydrocar~on refrigerant test gas while operating the system at a flow· rate of 4000 cfm ± 10%.


•

•



2. In OPERATIONAL MODE *,·with the control room emergency filtration system or the RCU inoperable, immediately suspend CORE AL TERATION(s), handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.



b. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratorytesting criteria of ASTMD-3803-89, for a methyl iodide. penetration of <0.50%, when tested at 30°C and 70% relative humidity; and

c. Verifying a system flow rate of 2000 scfm ± 10% during·system operation when tested in accordance with ANSI N510-1980.

4. After every 720 hours of charcoal adsorber operation by verifyin!;1 within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM-D~3803-89, for a methyl iodide penetration of < 0.50%, ·when t~sted at 30°C and 70% relative humidity:


a. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks. is < 6 inches water gauge while operating the filter train at a flow rate of 2000 scfm ± 10% .

Amendment Nos.

•




b. Verifying that the isolation dampers close on each of the following signals:


2) Simulated automatic isolation signal.

c. Verifying that during the pressurization mode of operation, control room positive pressure is maintained at ~1 /8 inch water gauge relative to adjacent areas during system operation at a flow rate ~2000 scf m.

d. Verifying that the heaters dissipate 12 ± 1.2 kw when tested in accordance with ANSI N510-1989. This reading shall include the appropriate correction for variations from 480 volts at the bus.

6. After each complete or partial replacement of an HEPA filter bank by verifying that the HEPA filter bank satisfies the in-place penetration and leakage testing acceptance criteria of <0.05% in accordanc!=l with ANSI N510-1980 while operating the system at a flow rate of 2000 scfm ± 10%.

7. After each complete or partial replacement of an charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the in-place penetration and leakage testing acceptance criteria of < 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating ttie system at flow rate of 2000 scfm ± 10% .

Amendment Nos.

•






a. Both Unit engineered safety features 4160 volt buses:

1) For Unit 1, Nos. 13-1 and 14-1,

2) For Unit 2, Nos. 23-1 and 24-1.



2) For Unit 2, Nos. 28 and 29, and

c. The Unit 120 volt Essential Service Bus and Instrument Bus.

2. 250 volt D.C. power distribution, consisting of:

a. 1) For Unit 1, TB MCC No. 1,

2) For Unit 2, TB MCC No. 2.

b. 1 i For Unit 1, RB MCC Nos. 1 A and 1 B,

2) For Unit 2, RB MCC Nos. 2A and 2B.

3. For Unit 1, 125 volt D.C. power distribution, consisting of:

a. TB Main Bus Nos. 1A, 1A-1 and 2A,

b. TB Reserve Bus Nos. 1 B and 1 B-1, and

c. RB Distribution Panel No. 1 .






•





APPLICABILITY:

OPERATIONAL MODE(s) 1, 2, 3, 41a1 and 51a1.

ACTION:

1. With one RPS electric power monitoring CHANNEL for an inservice RPS. MG set or alternate power supply inoperable, restore the inoperable power.monitoring CHANNEL to OPERABLE status within 72 hours or remove the associated RPS MG set or alternate power supply from service.

2. With both RPS electric power monitoring CHANNEL(s) for an inservice RPS MG set or alternate power supply inoperable, restore at least one electric power monitoring CHANNEL to OPERABLE status within 30 minutes or remove the associated RPS MG set or alternate power supply from service .

a With any control rod withdrawn.




The specified RPS electric power monitoring CHANNEL(s) shall be determined OPERABLE:

1. By performance of a CHANNEL FUNCTIONAL TEST1b1 each time the plant is in COLD SHUTDOWN for a period of more than 24 hours, unless performed in the previous 6 months.

2. At least once per 18 months by demonstrating the OPERABILITY of overvoltage, undervoltage, and underfrequency protective instrumentation by performance of a CHANNEL CALIBRATION including simwlated ·automatic actuation of the protective relays, tripping logic, and output circuit breakers, and verifying· the following setpoints:

a. Overvoltage ~129.6 volts AC

b. Undervoltage ~105.3 volts AC .

c. Underfrequency ~55.4 Hz

b Only required to be performed prior to entering MODE 2 or 3 from MODE 4.


" • ) . .

•

ELECTRICAL POWER SYSTEMS 8 3/4. 9

BASES

With suitable redundancy in components and features not available, the plant must be placed in a condition for which the Limiting Condition for Operation does not apply.

The term verify as used toward A.C. electrical power sources means to administratively check by examining logs or other information to determine if certain components are out-of-service for preplanned preventative maintenance, testing, or other reasons. It does not mean to. perform the

· surveillance--requirements·-ne-eded .. 10-demonstrate-the·OPERABll:l=rY-of·the-·component. - - ·

With one offsite circuit and one diesel generator inoperable, individual redundancy is lost in both the offsite and onsite electrical power system. Therefore, the allowable outage time is more limited. The time limit takes into account the capacity and capability of the remaining sources, reasonable time for repairs, and the low probability of a design basis event occurring during this · period.

With both of the required offsite circuits inoperable, sufficient onsite A.C. sources are available to maintain the unit in a safe shutdown condition in the event of a design basis transient or accident. In fact, a simultaneous loss of offsite A.C. sources, a loss-of-coolant accident, and a worst-case single failure were postulated as a part of the design basis in the safety analysis. Thus, the allowable outage time provides a period of time to effect restoration of all or all but one of the offsite circuits commensurate with the importance of maintaining an A.C. electrical power system capable of meeting its design intent .

With two diesel generators inoperable there are no remaining standby A.C. sources. Thus, with an assumed loss of offsite electrical power, insufficient standby A.C. sources are available to power the minimum required ESF functions. Since the offsite electrical power system is the only source of A.C. power for this level of degradation, the risk associated with continued operation for a very short time could be less than that associated with an immediate controlled shutdown, which could result in grid instability and possibly a loss of total A.C. power. The allowable time to repair is severely restricted during this condition. The intent here is to avoid the risk associated with an immediate controlled shutdown and to minimize the risk associated with this level of degradation.

Surveillance Requirements are provided which assure proper circuit continuity for the offsite A.C. electrical power supply to the onsite distribution network and availability of offsite A.C. electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure distribution buses and loads are connected to their preferred power source. The frequency is adequate since breaker position is not likely to change without the operator being aware of it and because status is displayed in the control room. Should the action provisions of this specification require an increase in frequency, this Surveillance Requirement assures proper circuit continuity for the available offsite A.C. sources during periods of degradation and potential information on common cause failures that would otherwise go undiscovered .

QUAD CITIES - UNITS 1 & 2 8 3/4.9-2 Amendment Nos.

••

•) '

•


BASES

Each battery of the D.C. electric.al power systems is sized to start and carry the. normal D.C. loads plus all D.C. loads required for safe shutdown on one unit and operations required to limit the consequences of a design basis event on the other unit for a period of 4 hours following loss of all A.C. sources. The battery chargers are sized to restore the battery to full charge under normal (non-emergency) load conditions. A normally disconnected alternate 125 volt battery is also provided as a backup for each normal battery. If both units are operating , the normal 125 volt battery must be returned to service within the specified time frame since the design configuration of the alternate battery circuit is susceptible to single failure and, hence, is not as reliable as. the normal station circuit. During times when the other unit is in a Cold Shutdown or Refuel condition, an alternate 125 volt battery is available to replace a normal station 125 volt battery on a continuous basis to provide a second available power source.

With one of the required D.C. electrical power subsystems inoperable the remaining system has · the capacity to support a safe shutdown and to mitigate an accident condition. However, a subsequent worst-case single failure would result in complete loss of ESF functions. Therefore, an allowed outage time is provided based on a reasonable time to assess plant status as a function of_ the inoperable D.C. electrical power subsystem and, if the D.C. electrical power subsystem is not restored to OPERABLE status, prepare to effect an orderly and safe plant shutdown._.

Inoperable chargers do not necessarily indicate that the D.C. systems are not capable of performing their post-accident functions as long as the batteries are within their specified parameter limits. With both the required charger inoperable and the battery degraded, prompt action is required to assure an adequate D.C. power supply.


Battery surveillance requirements are based on the defined battery cell parameter values. Category . A defines the normal parameter limit for each designated pilot cell in each battery. The pilot cells are the average cells in the battery based on previous test results: These cells are monitored closely as an indication of battery performance. Category B defines the normal parameter limits for each connected cell. The term "connected cell" excludes any battery cell that may be jumpered out because of a degraded condition or for any other reason. Category B also defines allowable values for each connected cell. These values, although reduced, provide assurance that sufficient capacity exists to perform the intended function and maintain a margin of safety. When any battery parameter is outside the Category B allowable value,. the assurance of sufficient capacity as described above no longer exists and the battery must be declared inoperable.

Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations .


-REFUELING OPERATIONS DELETED 3/4.10.D

•• 3.10 - LIMITING CONDITIONS FOR OPERATION 4.10 - SURVEILLANCE REQUIREMENTS

D. DELETED D. DELETED

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.H. Water Level - Spent Fuel Storage Pool

The pool water level shall be maintained at a level of ~33 feet.

APPLICABILITY:

Whenever irradiafea· fu"er asse·mbffes'are in the spent fuel storage pool.

ACTION:

With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the spent fuel storage pool area after placing the fuel assemblies and crane load in a safe condition. The provisions of Specification 3.0.C are not applicable .

Pool Water Level 3/4.1 O.H




••• ~ ....... ••••~, ....... L ........ _ -~ •• ~ ..... _ .... -;;;. ••• A'O


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REFUELING OPERATIONS RHR Low Water Level 3/4.1 O.L

3. 10 - LIMITING CONDITIONS FOR OPERATION 4.10 - SURVEILLANCE REQUIREMENTS


Two shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE, with each loop consisting of at least:

1. One OPERABLE RHR pump, and

2. One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL MODE 5, when irradiated fuel is in the reactor vessel and the water level is < 23 feet above the top of the reactor pressure vessel flange.

ACTION:

With less than the above required shutdown cooling mode loops of the RHR system OPERABLE, within one hour and at least once per 24 hours thereafter, demonstrate the OPERABILITY of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop .


1. At least one shutdown cooling mode loop of the RHR system shall be verified to be capable of circulating reactor coolant at least once per 1 2 hours.

2. Monitor the reactor coolant temperature at least once per hour.


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BASES

3/4.1 o.c Control Rod Position


3/4.1 O.D DELETED


The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status regarding core reactivity conditions during movement of fu.el within the reactor pressure vessel.

3/4.10.F DELETED

3/4.1 O.G Water Level - Reactor Vessel


The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of th.e assumed 10% iodine gap activity released from the· rupture of an irradiated fuel assembly. This minimum water depth is consistent with the assumptions of the accident analysis.

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\ll ADMINISTRATIVE CONTROLS

•

6.8 PROCEDURES AND PROGRAMS


1. The applicable procedures recommended in Appendix A, of Regulatory Guide 1.33, ·Revisio·n .. 2·;··r-ebmary-·l978·, ·· ·- · ·· · · ·· · · -·· .......... ~" -~·-···"'··· .. ...; ..... •·

2. The Emergency Operating Procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Section 7. 1 of Generic Letter No. 82-33,





· 7. Fire Protection Program implementation.

6.8.B Deleted

6.8.C Deleted

6.8.D The fellowing programs shall be established, implemented, and maintained:

1. Reactor Coolant Sources Outside Primary Containment

This program provides controls to minimize leakage from those portions of systems outside primary containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include CS, HPCI, LPCI, RCIC, process sampling (post accident sampling of reactor coolant and containment atmosphere), containment monitoring, and standby gas treatment systems. The program shall include the following:


b. Leak test requirements for each system at a frequency of at least once per operating cycle .


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4. · Radioactive Effluent Controls Program

A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program ( 1) shall be contained in the ODCM, (2) shall be implemented by station procedures, and (3) shall

·include· remedial-actions·to· be··taken·whenever-the ·program·limits··are··exceeded .-· The · program shall include the following elements:

a. Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance .tests and setpoint determination in accordance with the methodology in the ODCM,

b. Limitations on the instantaneous concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to ten ( 10) times the concentration values in 10 CFR Part 20, Appendix B, Table 2, Column 2 to 10 CFR Part 20.1001 - 20.2402,


d. Limitations on the annual and quarterly doses to a member of the public from radioactive materials in liquid effluents released from each Unit conforming to Appendix I to 10 CFR Part 50,

e. Determination of cumulative and projected dose contributions from radioactive · effluents for .the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days,


•



f. Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose confor_ming to Appendix I to 10 CFR Part 50,

g. Limitations on the dose rate resulting from radioactive materials released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:

a) For noble gases: less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the . skin, and

b) For lodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: · less than or equal to a dose rat~ of 1500 mrem/yr-to any organ.

h. Limitations on the· annual and quarterly air doses resulting from noble gases released in gaseous effluents from each Unit to areas beyond the site boundary conforming to Appendix I to 10 CFR Part 50,

i. l-imitations on the annual and quarterly doses to a member of the public from lodine-131, lodine-133, tritium, and all radionuclides in particulate form with halflives greater than 8 days in gaseous effluents released from each Unit conforming to Appendix I to 10 CFR Part 50,

j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190 .



,f j ADMINISTRATIVE CONTROLS

•

6.9 REPORTING REQUIREMENTS


6.9.A. Routine Repo.rts

1. Deleted

2. Annual Report

Annual reports covering the activities of the Unit for the previous calendar year, as described in this section shall be submitted prior to May 1 of each year .


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High Radiation Area 6. 1 2


6.12 HIGH RADIATION AREA

6.12.A Pursuant to 10 CFR 20.1601 (c), in lieu of the requirements of paragraph 20.1601 of 10 CFR Part 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr at 30 cm ( 12 in.) shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP) 181 (or equivalent document). Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

6.12.B

1. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

2. A radiation monitoring device which continuously integrates the radiation dose rate in . the area and alarms when a preset integrated dos.e is received. Entry into such areas

with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or


In addition to the requirements of 6. 12.A, above, areas accessible to personnel with radiation levels greater than 1000 mrem/hr at 30 cm ( 12 in.) from the radiation source or from any surface which the radiation penetrates shall require the following:

1. Doors shall be locked to prevent unauthorized entry and shall not prevent individuals . from leaving the area. In place of locking the door, direct or electronic surveillance that is capable of preventing unauthorized entry may be used. The keys shall be maintained under the administrative control of the Shift Engineer on duty and/or health physics supervision.

2. Personnel access and exposure control requirements of activities being performed within these areas shall be specified by an approved RWP (or equivalent document).

a Health Physics personnel or personnel escorted by health physics personnel shall be exempt from the RWP issuance requirements during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.


~) ·y;

• )

High Radiation Area 6. 12


3. Each person entering the area shall be provided with an alarming radiation monitoring device that continuously integrates the radiation dose rate (such as an electronic dosimeter.) Surveillance and radiation monitoring by a Radiation Protection Technician may be substituted for an alarming dosimeter.

4. Deleted.

5. For individual HIGH RADIATION AREAS accessible to personnel with radiation levels of greater than 1000 mrem/h at 30 cm (12 in_.) that are located within large areas where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual areas, then such individual areas shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device .


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•

PCP 6.13


6.13 PROCESS CONTROL PROGRAM (PCP)

6.13.A Changes to the PCP:


••\. •' ... , ..... ; . .,,.:_ :


b. A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

2. Shall become effective after review and acceptance, including approval by the Station Manager .


ODCM 6.14

-~ ..,.,.~ ===============================================================================


• ' I

••

6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)




b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

2. Shall become effective after review and acceptance, including approval by the Station Manager.

3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Report for the period of the report in which any change to the ODCM was made effective. Each change- shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented .


proposed tech specs for ts upgrade program. · 3,4, 5 s w101 elorl m s:: b. inoperative 2, 3,4, 5...

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