Text

Proposed Tech Specs 3.3/4.3, Reactivity Control
	ML20082V212
Person / Time
Site:	Quad Cities
Issue date:	09/17/1991
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20082V205	List: ML20082V202; ML20082V212;
References
	NUDOCS 9109230163
	Download: ML20082V212 (63)
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F E

I PROPOSED TECH SPEC TS 3.3/4,3

' REACTIVITY CONTROL' 0109 a2016.:. 91o91, PDP CDOCK <Y5 0 00 2".14 4

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS A. SHUTDOWN MARGIN A. SHUTDOWN MARGIN The core loading shall be limited 1.- The SHUTDOWN MARGIN shall to that which can be made suberi- be determined to be equal tical during the OPERATING CYCLE to or greater than that with a SHUTDOWN MARGIN equal to or specified in Specification greater than (0.25) % Ak/k with the 3.3.A at any time during strongest OPERABLE control rod the OPERATING CYCLE:

fully- withdrawn and all other OPERABLE rods fully inserted. a. By mansurement, prior j to or during the j APPLICABILITY: first startup after !

each REFUELING OUTAGE OPERATIONAL MODES 1, 2, 3, 4, and where CORE ALTERA-

5. TIONS were performed.

ACILclu b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after detection of a with-drawn control rod

1. In OPERATIONAL MODES 1 or 2, that is immovable, as ,

restore the required SHUTDOWN a result of excessive MARGIN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be friction or mechani-in at least HOT SHUTDOWN cal interference, or within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , untrippable, except that the required

2. In-OPERATIONAL MODES 3 or 4,. SHUTDOWN MARGIN shall immediately verify all inser- be verified accepta-table control rods to be ble with an increased inserted and . suspend all -allowance for the

-activities that could reduce withdrawn worth of the SHUTDOWN MARGIN. Esta- immovable or untrip-blish SECONDARY CONTAINMENT pable control rod.

INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3. In OPERATIONAL MODE 5, suspend CORE ALTERATIONS and other activities that could reduce the SHUTDOWN MARGIN

, and: insert all insertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Establish SECONDARY CONTAIN-MENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3.3/4.3-1

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

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

.B. Reactivity Anomalies B. Reactivity Anomalies The reactivity equivalent of the The reactivity equivalence of difference between the actual the difference between the critical rod configuration and the actual critical rod configu-expected configuration shall not ration and the expected confi-exceed it Ak/k. guration shall be verified to be less than or equal to 1% Ak/k:

APPLICABILITY:

1. During the first startup OPERATIONAL MODES 1 and 2. following CORE ALTERATIONS, and ACTION:

2. At least every equivalent With the reactivity equivalence full power month, difference exceeding it Ak/k:

1. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months perform an analysis to determine and explain the cause of the reactivity dif ference; opera-tion may continue if the dif-ference- -is explained and corrected.

2. Otherwise, with the provi-sions of ACTION 3.3.B.1 not mot, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

C. Control Rod OPERABILITY C. Control Red OPERABILITY ,

All control rods shall be 1. When above the low power

-OPERABLE. setpoint of the RWM, all withdrawn control rods not APPLICABILITY: required to have their directional control valves OPERATIONAL MODES 1 and 2. disarmed elactricaLly or hydraulically shall be ACTION: demonstrated- OPERABLE by moving each control rod at least one notch:

1. With one control rod inoperable due to being a. At least once per 7 immovable, as a result of days, and excessive friction or b. At least once per 24 mechanical interference, or hourc when any con-known to be untrippable: trol rod is immovable as a result of exces-l 3.3/4.3-2 l

1

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

a. Within one hours sive friction or nochanical interfer-1)- Verify that the ence.

inoperable control rod, if withdrawn, 2. All control rods shall be is separated from demonstrated OPERABLE by all.other inopera- performance of Surveillance ble -control rods Requirements 4.3.D, 4.3.c, by at least two 4.3.H and 4.3.I.

control cells in all directions.

2) Disarm the asso-ciated directional control valves either:

a) Electrically, or b) Hydraulically by cloring the drive wa-ter and ex-haust water isolation valves.

3) Initiate compli-ance with surveil-lance Requirements 4.3.C.1.b and 4.3.A.1.b.

4) Disarmed directio-nal control valves may be rearmed in-termittently, .un-

-der administrative control, to permit testing associated with restoring the control rod (s) to OPERABLE status.

Otherwise, with -the provisions of ACTION 3.3.C.1 not met, be in at least HOT SHUTDOWN 3.3/4.3-3

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

%~

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Restore the inoperable control rod to OPERAD12 status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2. With one or more control rods trippable but inoperable for causes other than addressed in ACTION 3.3.C.1, above:

a. If the inoperable con-trol rod (s) is with-drawn, within one hourt

1) Verify that the inoperable with-drawn control rod (s) is separa-ted from all other inoperable with-drawn control rods by at least two control cells in all directions, and

2) Demonstrate the

! insertion capabi-l lity. of -the in-l operable withdrawn i control rod (s) by

-inserting the con-trol rod (s) at least one notch by drive water pres-sure within the normal operating range. The in-eperable control rod may then be withdrawn to a po-l sition no further l withdrawn than its position when 3.3/4.3-4

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

s 9 QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 found to b9 inoperable.

Otherwise, with the provisions of ACTION 3.3.C.2.a not met, insert the inoperable withdrawn control rod (s) and disarm the associated directional control valves either:

1) Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves.

b. If the inoperable con-trol -rod (s) is inner-

ted, within one hour disarm the associated directional control valves either:

'1) Electrically, or

2) Hydraulically by closing the drive water and exhaust '

water isolation valves.

Otherwise, with the provisions- of ACTION 3.3.C.2.b not met, be in at least HOT SHUT-DOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

c. Disarmed directional control valves may be ,

rearmed intermittently, under administrative control, to permit testing associated with restoring the control 3.3/4.3-5

QUAD CITIES UNIT ~ l & 2 :

DPR-29 & DPR-30 l rod (s) to OPERABLE status,

d. The provisions of Specification 3.0.D are not applicable.

3. With more than 8 control rods inoperable, ce in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D. Control Rod Maximum Scram D. Control Rod Maximum Scram .

Insertion Times Insertion Times The maximum scram insertion cime 1. ThS 2s simum scram insertion to 90%-insertion for any OPERABLE time of the control rods control rod shall not exceed 7 shall be demonstrated seconds, through measurement with reactor coolant pressure APPLICABILITY: greater th5n 800 psig and, during single control rod OPERATIONAL MODES 1 and 2. scram time -tests, the control rod drive pumps 16Tlu isolated f rom the accumula-tors:

With the maximum scram insertion time of one or more control rods a. For all control rods exceeding 7 seconds: prior to thermal po-wer exceeding 30% of

, 1. Declare the control rod (s) RATED THERMAL POWER with the slow insertion time following CORE AL-inoperable,-and TERATIONS or after a reactor- shutdown

2. Perform the Surveillance that is greater than Requirements of Specification 120 days.

4. 3. D.1. c at least once per

.60 days' when ' operation is b. For specifically continued with three or more affected individual control rods with maximum control rods fol-

. scram insertion-times in ex- lowing maintenance'on to cess of 7 seconds. or modification the control rod or

3. 'Otherwise, with the control rod -drive

. provisions of ACTION . 3.3. D system which could not' met,-be in at least HOT affect the scram SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . insertion time of those specific con-4 '. The provisions of Specifi- trol rods. The pro-cation 3.0.0 are not visions of Specifi-3.3/4.3-6 l

l

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 applicable. cation 4.0.D are not applicable for entry into OPERATIONAL MODE 2 provided this sur-veillance is comple-ted prior to 'ntry ,

into OPERATIONAL MODE 1.

c. For at least 10% of the control rods, on a rotating basis, at least once per 120 days of REACTOR POWER OPERATION._

E. Control Rod Average Scram E. Control Rod Average Scram Insertion Times Insertion Times The average scram insertion time The control rod average scram of all OPERABLE control rods from times shall be demonstrated by the fully withdrawn position, scram time testing from the based on .doenergization of the fully withdrawn position as scram pilot _ valve solenoids as required by Surveillance time zero, shall not exceed any of Requirement 4.3.D.

the following:

1. The control rod average Average Scram scram times- whall be

% Inserted From Insertion demonstrated by scram time Fully Withdrawn Times (sec) testing from the fully withdrawn position as 5 0.375 required by Surveillance 20 0.900 Requirement 4.3.D.

50 2.00 90 3.50 2. The cycle cumulative mean scram time for 20% inser-APPLICABILITY: tion will be determined

-immediately following the OPERATIONAL MODES I and 2. testing required in Speci-fication 4.3.D and the ACTION: -MCPR operating limit adjusted,.If necessary, as

1. With the average scram required by Specification insertion time exceeding any 3.5.0 of the above limits, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.3/4.3-7

4 .

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

2. Uith the overall average of the 20% insertion scram time data generated in the current cycle exceeding the limit specified in the CORE OPERA-TING LIMITS REPORT, the MCPR operating limit must be modii' led as required by Specification 3.5.0.

F. Four Control Rod Group Scram F. Four Control Rod Group Scram Insertion Times Insertion Times The average of the scram insertion All control rods shall be demon-times for the three fastest con- strated OPERABLE by scram time trol rods of all groups of four testing from the fully withdrawn control rods in a two-by-two array position as required by shall be no greater than: Surveillance Requirement 4.3.D.

% Inserted From Average Scram Fully Withdrawn Times (soc)

S 0.398 20 0.954 50 2.12 90 3.80 APPLICABILIT,Yi OPERATIONAL MODES 1 and 2.

ACTION:

With the average scram times of control rods exceeding the above limits:

1. Declare the control rods with the slower than average scram insertion times inoperable until an analysis is per-formed to determine that required scram reactivity remains for the slow four control rod group, and

, 2. Perform the Surveillance

! Requirements of Specification

4. 3. D.1. c at least once per 3.3/4.3-8 i

f

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 60 days when operation is continued with an, average scram insertion time (s) in excess of the average scram insertion time limit.

3. Otherwise, with the provisions of ACTION 3.3.F not met, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4, The provisions of Specifi-cation 3.0.D are not applicable.

G. Control Rod Scram Accumulators G. Control Rod Scram Accumulators All control rod scram accumulators Each control rod scram accumu--

shall be~ OPERABLE. lator shall be determined OPER-ABLE at least once per 7 days by APPLICABILITY: verifying that the indicated pressure is greater than or OPERATIONAL MODES- 1, 2 and 5. equal to 950 psig unless the This Specification is applicable control rod is inserted and in OPERATIONAL MODE 5 for the disarmed or scrammed, accumulators associated with each withdrawn control rod and is not applicable to control rods removed per Specification 3.10.D or 3.10.E.

ACTION:

1. In OPERATIONAL MODES 1 or 21

a. With- one control rod scram accumulator inoperable, within 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s:

1) Restore the inop-erable accumulator to OPERABLE status or

2) Declare the con-trol rod associ-ated with the inoperable accumu-3.3/4.3-9

l .

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 lator inoperable.

3) Otherwise, with the provisions of ACTION 3.3.G.1.a not met, be in at .

least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With more than one control rod scram accumulator inoperable, declare the associated control rods inoperable and:

1) If the control rod associated with any inoperablo scram accumulator is withdrawn, im-mediately verify that at least one control rod drive pump is operating by inserting at least one with-drawn control rod at least one notch or place the reac-tor mode switch in the shutdown position.

2) Insert the in-operable control rods and disarm the associated directional con-trol valves either:

a) Electrically, or b) Hydraulically by closing the drive water and 3.3/4.3-10 L _--.-- --__ - .-. -_--____.---- _ _ _ _ _ _ __

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 exhaust water isolation valves.

3) Otherwise, with the provisions of ACTION 3.3.G.l.b not met, be in at least HOT SHUTDOWN within-12 hours.

2. In OPERATIONAL MODE St

a. With one withdrawn control rod with its associated scram accu-mulator inoperable, insert the affected control rod and disarm the associated direc-tional control valves within one hour, either:

1) Electrically, or

2) Hydraulically- by closing the drive water and exhaust water-. isolation valves. .

b. With more than one withdrawn control rod with the associated scram accumulator inop-erable or no control -

rod drive pump opera-ting, immediately place l

the reactor mode switch

! in the SHUTDOWN position.

3. The provisions of Specification 3.0.D are not applicable.

l 3.3/4.3-11 l

t

i QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 Control Rod Drive Coupling H. Control Rod Drive Coupling H.

All control rods shall be coupled Each affected control rod shall be demonstrated to be coupled to to their drive mechanisms. its drive mechanism by observing APPLICABILITY: any indicated response of the nuclear instrumentation while OPERATIONAL MODES 1, 2, and 5. Withdrawing the control rod to This Specification is applicable the fully withdrawn position and in OPERATIONAL MODE 5 for control then verifying that the control rods withdrawn and is not appli- rod drive does not go to the cable to control rods removed per overtravel position:

Specification 3.10.D or 3.10.E.

1. Prior to reactor critica-ACTION:

lity after completing CORE ALTERATIONS that could have

1. In OPERATIONAL MODES 1 and 2 affected the control rod with one control rod not drive coupling integrity, coupled to its associated drive mechanism, within 2 2. Anytime the control rod is hours: withdrawn to the " Full out" position in subsequent

a. If permitted by the operation, and RWM, insert the control rod drive mechanism to 3. Following maintenance on or accomplish recoupling modification to the control and verify recoupling rod or control rod drive by withdrawing the system which could have control rod, and: affected the control rod drive coupling integrity.

1) Observing any indicated re-sponse of the nuclear instru-mentation, and

2) Demonstrating that the control rod will not go to the overtravel posi-tion.

b. If recoupling is not accomplished on the first attempt or, if not permitted by the RWM, then until permit-ted by the RWM, declare the control rod inop-3.3/4.3-12

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

- erable, insert the con- H

-trol rod and disarm the !

- associated -directional l control-valves either ]

1). Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves.

4 Disarmed _ directional

control valves- may be rearmed, underLadminis- ,

trative' control, to permit testing associa-tad with restoring the control rod to OPERABLE status,

c. Otherwise,- ..i f - the provisions of. ACTIONS

< 3.3.H.1 are not met, be in at least HOT SHUT-DOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. lThe provisions of' Specification 3.0.D are not applicable.

2. In OPERATIONAL MODE 5 with a

. withdrawn control rod not coupled.-to its associated drive mechanism, -within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

. - a.. Insert the - control: rode uto .' accomplish _ re-

. coupling and Lverify recoupling. by - with-drawing'the control rod and _ demonstrating that the control -rod will-not .._go to:.the 'over-travel position,-or

. b. If recoupling is not 3.3/4.3-13 w Wrwa -

i

) l l

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 accomplished,- insert the control rod and disarm the associated directional control valves either:

1) Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves.

Directional control valves may be rearmed intermittently, under administrative control, to permit testing asso-ciated with restoring '

the control rod to OPERABLE status.

I. Control Rod Position Indication I. Control Rod Pos! tion Indication System System All control rod position indica- The control rod position indi-tors shall be OPERABLE. cation system shall be deter-mined OPERABLE by verifying:

APPLICABILITYJ,

1. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERATIONAL MODES 1, 2, -and 5. that the position of each This Specification is applicable control rod is indicated, in OPERATIONAL MODE 5 for with-drewn control rods and is not- 2. That the-indicated control applicable to control rods removed rod position changes during por Specification 3.10.D or the movement of the control

-3.10.E. rod drive when performing Surveillance Requirement ACTION: 4.3.C.1.

1. In OPERATIONAL MODES 1 or 2 3. That the control rod posi-with one or more control rod tion indicator correspond.s positio.. indicators inop- to the control rod position erable, within one hour: indicated by the " Full out" position indicator when

a. Determine the position performing Surveillance of the control rod by Requirement 4.3.H.2.

an alternate method, or 3.3/4.3-14

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

b. Move the control rod to a position with an OPERABLE position indicator, or

c. Declare the control rod inoperable, insert the inoperable withdrawn control rod (s), and disarm the associated directional control >

valves either:

1

1) E'ectrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves.

The control rod direc-tional control valves may be rearmed inter-mittently, under admi-nistrative control, to permit testing associa-ted with restoring the control rod (s) to OPERABLE status,

d. Otherwise, with the provisions of ACTION 3.3.I.1 not met, be in at least HOT SHUT-DOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

e.- The provisions of Specification 3.0.D are-not applicable.

- 2. In OPERATIN.AL MODE 5 with a withdrawn control rod posi-tion indicator inoperable:

a. Move the control rod to a position with an OPERABLE position 3.3/4.3-15

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

b. Insert the control rod.

J. Control Rod Drive Housing Support J. Control Rod Drive Housing Support The control rod drive housing support shall be in place. The control rod drive housing support shall be verified to be APPLICABILITY: in place by a visual inspection prior to startup any time it has OPERATIONAL MODES 1, 2, and 3. been disassembled or when main-tenance has been performed in ACTION: the control rod drive housing support area.

With the control rod drive housing support not in place, be in at least HOT S!!UTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in at least COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

K. Scram Discharge Volume Vent and K. Scram Discharge Volume Vent and Drain valves Drain valves All scram discharge volume vent scram discharge volume vent and and drain valves shall be drain valves shall be demons-OPERABLE. trated OPERABLE:

APPLICABILITY: 1. At least once per 31 days by verifying each valve to OPERATIONAL MODES 1 and 2. be open, and ACTION: 2. At least once per 92 days by cycling each valve

1. With one of the following through at least one com-conditions present, restore plete cycle of travel.

the valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be 3. At least once per REFUELING in at least HOT SHUTDOWN OUTAGE, the scram discharge within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s: volume vent and drain valves shall be demons-

a. One scram discharge trated to:

volume vent valve inoperable and open, or a. Close within 30 seconds after receipt

b. One scram discharge of a signal for con-volume drain valve trol rods to scram, inoperable and open, or and 3.3/4.3-16

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

c. The combination of any b. Open af ter the serain one scram discharge signal is reset, volume vent valve and any one drain valve ,

being inoperable and open.

2. With any scram diccharge volume vent valve (s) and/or drain valve (s) inoperabl6 due to reasons other than those of ACTION 3.3.K.1 above, restore the inoperable r

valve (s) to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

L. ' Rod Worth Minimizer L. Rod Worth Minimizer The rod worth minimizer shall be The RWH shall bo demonstrated OPERABLE. OPERABLE:

APPLICABILITY: 1. In OPERATIONAL MODE- 2 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to OPERATIONAL MODES 1 and 2, when withdrawal of control. rods thermal power is less than 20% of for the= purpose of making RATED THERMAL POWER. The rod the reactor critical, and worth minimizer may be bypassed in OPERATIONAL MODE 1 prior 4

for low power physics testing to to reducing thermal power demonstrate the SHUTDOWN MARGIN below 20% RATED THERMAL requirements of Specification- POWER, by verifying proper 3.3. A : if a technically- qualified indication of the selection individual is prosent and verifies error of at least one-out-the step-by-step rod movements - of of-sequence control rod, the test procedure. Entry into OPERATIONAL MODE 2 and withdrawal 2. In OPERATIONAL MODE 2 of selected control rods is per- within. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to mitted_for the purpose of deter- withdrawal of control rods mining the OPERABILITY of the RWM for the purpose of making prior to withdrawal of control the reactor critical, by rods for the_ purpose of bringing verifying the rod block the reactor _to criticality, function by demonstrating inability to withdraw an ACTION: out-of-sequence control rod.

With the RWM inoperable, verify control rod mover . .t and compli- 3. In OPERATIONAL MODE 1 prior ance with the prescribed control to reducing - thermal power 3.3/4.3-17 l

l

QUAD CITIES UNITS 1 & 2 DPR-29 & DpR-30 rod pattern by a accond licensed below 20% RATED TilERMAL operator or technically qualified POWER by verifying the rod individual who is present at the block function by demons-trating inability to wJth-reector control console. Other- draw an out-of-sequence wise, control kud movement may be only by actuating the manual scram control rod.

cr placing the reactor mode switch verifying that the in the SHUTDOWN position. 4. By control rod patterns and sequence input to the RWM computer are correctly loaded following any loading of the program into the computer.

M. Rod Block Monitor M. Rod Block Monitor Each of the required RDM noth rod block monitor (RBM) CilANNZ13 shall be demonstrated CHANNELS shall be OPERABLE. OPERABLE by performance of at APPLICABIlJ111 1. CilANNEL PUNCTIONAL TEST OPERATIONAL MODE 1, when thermal and CilANNEL CALIl\ RATION at power is greater than or equal to the irequencies and ior the One OPERATIONAL MODES specifled 30% of RATED Ti!ERMAL POWER.

channel may be bypassed above 30% in Table 4.2-1.

RATED T}lERMAL POWER without a time FUNCTIONAL TEST restriction provided that a LIMI- 2. CllANNEL TING CON *ROL ROD PATTERN does not prior to control Tod with-exist and the remaining RBM C11AN- drawal and daily thereaf ter Both RBM C'IAH- when the reactor is opera-NEL is OPERABLE.

NLL5 are automatically bypassed tini on a LIMITING CONTROL when a peripheral control rod is ROD PATTERN.

selected.

hCT10Ni

1. With one RBM CHANNEL inoperable and not bypassed:

a. Verify that the reactor is not operating on a LIMITING CONTROL ROD PATTERN, and

b. Restore the inoperable RBM CHANNEL to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.3/4.3-18

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

c. Othorvise, with the provisions of ACTION 3.3.M.1 not Pet, place the inoperable rod block monitor CilANNEL in the tripped condi-tion within the next hour.

With both RBM CHANNELS in-operable, place at least one inoperable rod block monitor ci:ANNEL it. the tripped condition within one hour.

N. Source Range Monitoring Function N. Source Eango Monitoring Function At least the following sourco Each of the required source range monitor channels shall be range monitor cilANNELS shall be OPERABLE: demonstrated OPERABLE by:

a. In OPERATIONAL MODE 2, three. 1. Verifying, prior to with-

b. In OPERATIONAL MODE 3 and 4, drawal of the control rods, two, that the SRM count rate is at least 3 cps with the AEELICM!1LITY1 detector fully inserted, and OPERATIONAL MODES 2, 3, and 4.

This Specification is applicablo 2. Performance of a CHANNEL in OPERATIONAL MODE 2 with IRM's CllECK at least once port on range 2 or below.

a. 12 hours in OPERA-ACTIOf{1 TIONAL MODE 2 with IRMs on range 2 or

1. In OPERATIONAL MODE 2 with below, and one of the required sourco range monitor CHANNELS in- b. 24 hours in OPERA-operable, at least 3 source TIONAL MODES 3 or 4.

range monitor CHANNEL 3 shall be restored to OPERABLE 3. Performance of a CHANNEL statun within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or the CALIBRATION at least orse

, reactor shall be in at least per 18 months. Neutron i HOT SHUTDOWN within the next detectors may be excluded 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . from the CHANNEL CALIBRA-TION.

2. In OPERATIONAL MODES 3 or 4 with one or more of the above 4. Performance of a CllANNEL required source range monitor FUNCTIONAL TEGT CHANNELS inoperable, verify 3.3/4.3-19 l

l

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

all insertable control rods a. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to be inserted in the core to moving the reactor and lock the reactor mode mode switch from the switch in the shutdown posi- SHUTDOWN position, if tion within one hour. not performed within the previouc 7 days, and

b. At least once por 31 :

days. ;

0. Economic Generation Control System O. Economic Generation Control !

System operation of the unit with the economic generation control (EGC) prior to entering EGC and once system with automatic flow control per shift while operating in shall be permissible only in the EGC, the ECC operating parame-range of 65% to 100% of rated core ters shall be reviewed for flow with reactor power above 20%. acceptability.

APPLICABILITY!

OPERATIONAL MODE 1. [

ACTION!

With plant operation in EGC outside the above limits, restore operation to within the limits within one hour or remove the +

plant from EGC operation.

l l

3.3/4.3-20

, -*-s- ~ m,-- + + w : vr v., , -.n,.,r,- -+r-- , -- 4 - - - - - - - - - - - - -

QUAD CITIES UllITS 1 & 2 DPR-29 & DPR-30 3.3/4.3 REACTIVITY C0!iTROL A. SilUTDOW!l MARGIli The SliUTDOW!l MARGIll limitation is a routriction to be applied principally to the design of new fuel which may be loaded in the core or into a particular refueling pattorn. Satisfaction of the limitation can only be demonstrated at the timo of loading and must be such that it will apply to the entire subsequent fuel cycle. The generalized form is that the reactivity of the coro loading will be limited so the coro can be made subcritical by at least R + 0.25% Ak/k, in the most reactive condition during the operating cycle, with the strongest control rod fully withdrawn and all other fully inserted. The value of R in t Ak/k is the amount by which the core reactivity, at any time in the operating cycle, o1d is calculated to bo greater than at the time of the check, i.e., the initial loading. R must be a positivo quantity or zero. A core which contains temporary control or other burnable neutron absorbers may have a reactivity characteristic which increases with eore lifetimo, reachos a maximum, and decreases thereafter.

The value ot R is the difference between the calculated core reactivity, at the beginning of the operating cycj o, and the calculated value of the core reactivity at any time later in the cycle, where it would be greater than at the beginning. The value of R shall include the potential shutdown margin loss assuming full B 4 C settling in all inverted poison tubes present in the core. A new value of R must be determined for each new fuel cycle.

The 0.25% Ak/k in the expression R + 0.25% Ak/k is provided as a finito, demonstrable, suberiticality margin. This margin is demonr.trated by full w* :hdrawal of the strongest rod and partial withdrawal of an adjacent rod to a position calculated to insert at least R + 0.251, Ak/k in reactivity. Observation of suberiticality in this condition assures suberiticality with not only the strongest rod fully withdrawn but with a margin of at least R + 0.25% Ak/k beyond this condition.

B 3.3/4.3-1

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( .

T QUAD CITIM 9Qff 1 & .

DPR-29 & .ut-30 B. Reactivity Anomalies During each fuel cycle, excess operating reactivity varies as fuel depletes and as any burnable poison in supplementary-control is burned. The magnitudo of this excess reactivity c.ay be inferred from the critical rod .

configuration. As fuel burnup progresses, anomalous i behavior in the excess reactivity may be detected by comparison of the critical rod pattern selected base states to the predicted rod inventory at that stato.

Power operating base conditions provide the most sensitive and directly interpretable data relative to ,

core reactivity. Furthermore, using power operating base conditions permit frequent reactivity comparisons.

Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% Ak/k. Deviations in core reactivity greator than 1% Ak/k are not expected and require thorough evaluation. A 1% reactivity limit is considered safe, since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.

C. Control Rod OPERABILITY ACTIO!1 statement 3.3.C.1 requires that a rod be taken out-of-service if it cannot be moved with drive pressure.

If the rod is fully inserted and then disarmed electrically or hydraulically, it is in a safo position of maximum contribution to shutdown reactivity. (11oto r To disarm the drive electrically, four amphenol-typo plug connectors are removed from the drivo- insert and withdrawal solenoids, rondering the drive immovable.

This procedure is equivalent to valving out the drive and is preferred, as drive water cools and minimizes crud accumulation in the drivo. ) . If it is disarmed in a non-fully inserted position, that position shall be consistent with the SHUTDOW11 MARGIt4 limitation stated in Specification 3.3.A. This assures that the core can be shut down at all times with the remaining control rods, l

ssnuming the strongest operable control rod does not insert. An allowable pattern for control rods valved out-of-service, which shall moet the specification, will be available to the operator. The number of rods permitted to be inoperable could be many more than the eight allowed by the specification, particularly late in the operation cycles however, the occurrence of more than eight could be indicative of a generic control rod drive problem and the reactor will be shut down.

B 3.3/4.3-2

l I

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

D. Control Rod Maximum Scram Insertion Timest l E. Control Rod Averago Scram Insertion Times; and l T. Four Control Rod Group Scram Insertion Times l l

The control rod system in analyzed to bring the reactor i subcritical at a rato fast enough to prevent fuel damago, 1 1.e., to prevent the MCPR from becoming less than the fuel cladding integrity SAFETY LIMIT.

Analysis of the limiting power transient shows that tbo negativo reactivity rates, resulting from the scram with the average responso of all the drives, as given in the above specification, provide the required protection, and MCPR remains greator than the fuol cladding integrity SAFETY LIMIT. It is necessary to raise the McPR operating limit (por Specification 3.5.0) when the average 20% scram insertion timo reacher. the limit specified in the CORE OPERATING LIMITS REPORT on a cycle cumulative basis (overall average of survoillanco data to dato) in older to comply with assumptions in the implementation procedure for the ODYN transient analysis computer code. The basis for choosing this 20% scram insortion timo limit is discussed further in the bases for Specification 3.5.0. In the analytical treatment of the transients, 290 milliseconds are allowed betwoon a neutron sensor reaching the scram point and the start of motion of the control rods. This is adequate and conservative when compared to the typically observed timo delay of about 210 milliseconds. Approximately 90 milliseconds after neutron flux reachos the trip point, the pilot scram valvo solenoid doenergizes and 120 milliseconds later the control rod motion is estimated to actually begin. However, 200 milliseconds rather than =

l

?O milliseconds is conservatively assumed for this timo interval in the transient analyses and is also included in the allowable scram insertion times specified in Sp'ici fications 3. 3. D, 3.3.E, and 3.3.F.

The scram times for all control rods will be determined prior to thermal power exceeding 30% of RATED THERMAL POWER following CORE ALTERATIONS or after a reactor shutdown that is greater than 120 days. Scram times will be determined for specifically affected control rods following maintenance or modification to the control rod drive system which could af fect the scram insortion time of those specific control rods. Also, at least 10% of the contr ol rods are scram tine tested, on a rotating basis, at least once per 120 days of REACTOR POWER j

OPERATION.

B 3.3/4.3-3

i . .

QUAD CITIES UllITS 1 &2 DPR-29 & DPR-30 Scram times of new drives are approximately 2.5 to 3 seconds; however, lower rates _of change in scram times following initial plant operation at power are expected.

The test schedule providos reasonable assurance of detection of slow drives before system deterioration beyond the limits of Specification 3.3.C. The program was developed on the basis of the statistical approach outlined below and judgement.

The histcry of drive performance accumulated to dato indicates that the 90% insertion times of new and overhauled drives approximate a normal distribution about the mean which tends to become skewed toward longer scram times as operating timo is accumulated. The probability of a drive not exceeding the mean 90% insertion time by 0.75 . seconds is greator than 0.999 for a normal distribution. The measurement of the scram performance of the drives surrounding a drivo, which exceeds the expected range of scram perfcrmanco, will detect local variations and also provido assurance that local scram time limits are not exceeded. Continued monitoring of other drives exceeding the expected range of scram times provides surveillance of possible anomalous performance.

The numerical values assigned to the predicted scram performance are based on the analysia of the Dresden 2 startup data and of data from other BWRs, such as Nine Mile Point and Oyster Creek.

The occurrence of scram times within the limita, but significantly longer than average, should be viewed as an indication of a systematic problem with control rod drives, especially if the number of drives exhibiting such scram times exceeds eight, which is the allowable number of inoperable rods.

G. Control Rod Scram Accumulators Control rods with inoperable accumulators are declared inoperable and specification.3.3.C then applies. This prevents a pattern of inoperat',lo accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be insorted with normal drive water pressure. OPERABILITY of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

B 3.3/4.3-4

QUAD CITIES Ul1ITS 1 & 2 DPR-29 & DPR-30 H. Control Rod Drive coupling control rod dropout accidents, as discussed in Reference 1, can load to significant core damago. If coupling integrity is maintained, the possibility of a rod drop ,

accident is eliminated. The overtravel position feature providos a positive check, as only uncoupled drives may reach this position. . Houtron instrumentation responso to rod movement provides verification that a rod is following its drive. Absence of such response to drivo movement would indicato an uncoupled condition.

I. Control Rod Position Indication System (RPIS) 14ormal control rod position is displayed by two-digit indication to the operator from position 00 to 48, Each even number is a latching position, whereas each odd number provides information wnile the rod is in-motion and inputs for rod drift annunciation. The ACTION statement provides for the condition where no positive information is displayed for a large portion or all of the rod's travcl. In this case, for OPERATIollAL MODES 1 and 2, the position of the control rod is determined by an alternate method, or the control red is moved to a position with an OPERABLE position indicator, or the control rod is declared inoperablo, fully inserted and disarmed. If those ACTIO!1 statomonts are not met, the reactor is placed in at least HOT SHUTDOWil within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With a withdrawn control rod position indicator inoperable in OPERATIO!1AL MODE 5, the control rod is required to be moved to a position with an OPERABLE pocition indicator, or the control rod is inserted. Usually, only one digit of one or two of a rod's positions is unavailable with a faulty RPIS, and the control rod may be located in a known position, j J. Control Rod Drive Housing Support The control rod housing support restricts the outward '

movement of a control rod to less than 3 inches in the extremely remoto event of a housing failuro.--The amount of reactivity which could be added by this small amount ,

of rod withdrawal, which is less than a normal single withdrawal incremont, will not contribute to any-damage to the primary coolant system. The design basis is given in Section 6.6.3 of ths SAR. This support is not required if the reactor coolant system is at atmospheric pressure, since there would then be no driving force to l rapidly eject a drive housing.

B 3.3/4.3-5

9 QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 K. Scram Discharge volume Vent and Drain Valves The operability of the scram discharge volume vont and drain valves assures the proper venting and draining of the volume, so that water accumulation in the volume does not occur. These specifications provide for the per! ic verification that the valves are open, and for ao testing of these valves under reactor scram conditions during each REFUELING OUTAGE.

L. Rod Worth Minimiter control rod withdrawal and insertion sequences are established to nosure that the maximum, in-sequence individual control rod or control rod segments which are withdrawn could not be worth enough to cause the rod drop accident design limit of 280 cal /gm to be exceeded, if they were to drop out of the core in the manner defined for the rod drop accident. These sequences are developed prior to initial operation of the unit following any refueling outage and the requirement that an operator follow these sequences is supervised by the RWM or a second qualified station employee. These sequencen are developed to limit reactivity worths of control rods and, together with the integral rod velocity limiters and the action of the control rod drive system, limit potential reactivity insertion auch that the results of a control rod drop accident will not exceed a maximum fuel energy content of 280 cal /gn. The peak fuel enthalpy of 230 cal /gm is below the energy content at which rapid fuel dispersal and primary system damage have been found to occur based on experimental data, as discussed in Reference 2.

The analysis of the control rod drop accident was originally presented in Sections 7.9.3, 14.2.1.2, and 14.2.1.4 of the SAR. Improvements in analytical capability have allowed a more refined analysis of the control rod drop accident.

These techniques are described in a topical report (Reference 2) and two supplements (References 3 and 4).

In addition, a banked position withdrawal sequence described in Reference 5 has been developed to further reduce incremental rod worths. Method and basis for the rod drop accident analyses are documented in Reference 1.

By using analytical models described in those reports coupled with conservative or worst-case input parameters, B 3.3/4.3-6 i

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 it has been determined that for power levels less than 20% of rated power, the specified limit on in-sequence control rod or control rod segment worths will limit the peak fuel enthalpy to less than 280 cal /gm. Above 20%

power even single operator errors cannot result in out-of-sequence control rod worths which are sufficient to reach a peak fuel enthalpy of 280 cal /gm, should a postulated control rod drop accident occur. i 1

The following parameters and worst-case assumptions have l been utilized in the analysis to determine compliance ;

with the 280 cal /gm peak fuel enthalpy. Each core reload i vill be analyzed to show conformance to the limiting ;

parameters.

a. An interassembly local peaking factor. (To include the power spike effect caused by gaps between fuel-pellets,

b. The delayed neutron fraction chosen for the bounding reactivity curve.

c. A beginning-o'-life Doppler reactivity feedback.

d. Scram times slowar than the Technical Specificaticens rod scram insertion rate (Section 3.3.C.1).

e. The maximum possible rod drop velocity of 3.11 feet por second.

f. The design accident and scram reactivity shape function.

g. The moderator temperature at which criticality occurn.

In tuost cases, the worth of -in-sequence rods or rod segments, in conjunction with the actual values of the other important accident analysis parameters described above, would most likely result in a peak fuel enthalpy substantially less than 280 cal /gm design limit.

Should a c.ontrol rod drop accident result in a peak fuel energy convent of 280 cal /gm, fewer than 660 (7x7) fuel rods are ceriservatively estimated to perforate. This L would result in an offsite dose well below the guideline value of 10 CPR 100. For 8x8 fuel, fewer than 850 rods are conservatively estimated to perforate, with nearly l 3 3.3/4.3-7 l

l.

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 the same consequences as for the 7x7 fuel case because of the rod power differences.

The rod worth minimizer provides automatic supervision to assure that out-of-sequence control rods will not be withdrawn or inserted, i.e., it limits operator deviations from planned withdrawal sequencen (reference SAR Section 7.9).

It . serves as -a backup to procedural control of control

. rod worth. In the event that the rod worth minimizer is out-of-service when required, a second licensed operator or other technically qualified individual can manually fulfill the control rod pattern conformance function of the rod worth minimizer. In this case, the normal procedural- controls are backed up by independent procedural controls to assure conformance.

M. Rod Block-Monitor The rod' block monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density - during high power operation. Two channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the operator, who withdraws control rods according to a written sequence.

The specified restrictions with one channel out-of-service conservatively assure that fuel damage will not occur due to rod withdrawal errors unen this condition exists.- During reactor operation with certain LIMITING CONTROL ROD PATTERNS, the withdrawal of a designated I single control rod could result in one or more fuel rods j with MCPRs less than the MCPR fuel cladding integrity 7

SAFETY LIMIT. During use of such patterns, it is. judged that testing of the RBM system, to assure its operability prior to withdrawal of such rods, will assure that .

improper withdrawal does not occur. It is the responsibilit.y of the Nuclear Engineer to identify these limiting rod patterns and the designated rods, either l when the patterns are initially established or as they I develop due to the occurrence of inoperable control rods in other than limiting patterns.

If a peripheral control rod is selected, the neutron leakage is sufficiently high such that withdrawal of this rod will not violate the fuel cladding integrity SAFETY B 3.3/4.3-8

t QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 LIMIT. Thus, the RBM function is not required for withdrawal of peripheral control rods.

N. Source Rango Monitoring Function The source range monitor (SRM) system performs no automatic safety system function, i.e., it has no scram function. It does provide the operator with a visual indication of neutron lovel. This is nooded for knowledgeabic and officient reactor startup at low neutron levels. The conocquences of reactivity accidents are functions of the initial neutron flux. The requi'ement of at least 3 counts por second assures that any transient, should it occur, begins at or above the initial value of 10*" of rated power used in the analysos of transients from cold conditions. Two operable SRM channels would bo adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of throo oporable SRMs is provided as an added conservatism.

O. Economic Conoration Control System operation of the facility with the economic generation control system (EGC) (automatic flow control) is limited to the range of 65% to 100% of rated core flow. In this flow range and with reactor power above 20%, the reactor could safely tolerato a rate of change of load of a MWo/ soc (reference SAR Section 7.3.6).

Limits within the EGC and the flow control system prevent ratos of change greater than approximately 4 MWo/ soc.

When EGC ir, in operation, this fact will bo indicated on the main control room consolo.

D 3.3/4.3-9

, -. - -- , - - . - - , - . , , , - , , - - - -- , , - - -. . . , - . ---m -n,, -,n ,, ,--

.. ._ ________._.________________m _... _ _ _ ___

b QUAD CITIES Ul41TS 1 & 2 DPR-29 & DPR-30 References

1. " Generic Reload Puol Applications", NEDE-24011-P-A*

2. C.J. Paone, R.C. Stirn, and J.A. Wooley, " Rod Drop Accident Analysis for Large BWRs", GE Topical Report NEDO-10527, March 1972.

3. C.J. Paone, R.C. Stirn, and R.M. Young, " Rod Drop Accident Analysis for Largo BWRs", Supplement 1. GE Topical Report NEDO-10527, July 1972.

i

4. J.M. ifaun, C.J. Paono, and R.C. Stirn, " Rod Drop Accident Analysis for Large BWRs, Addendum 2. Exposed Cores",

supplement 2. GE Topical Report NEDo-10527, January 1973. -

5. C.J. Paane, " Banked Position Withdrawal Sequenco",

Licensing Topical Report NEDO-21231, January 1977.

Approved revision number at_ time reload fuel analyses are -

performed.

l

(

l

! B 3.3/4.3-10

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

e 4 EXISTING TECH SPEC TS 3.3/4,3

' REACTIVITY CONTROL' l

QUAD CITIES DPR-29 3.3/4.3 REACTIVITY CONTROL LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMEN15 Applicability: Applicability:

Applies to the operational status of the Applies to the surveillance requirements control rod system. of the control rod system.

Objective: Objective:

To assure the ability of the control rod To verify the ability of the control rod system to control reactivity. system to control reactivity.

SPECIFICATIONS A. Reactivity Limitations A. Reactivity Limitations

1. Reactivity margin - core loading 1. Reactivity margin - core loading The core loading shall be lim' Sufficient control rods shall be ited to that which can be made withdrawn following a refueling subtritical in the most reactive outage when core al'.erati n s condition during the operating were performed to demonstrate cycle with the strongest op- with a margin of 0.25% ak erable control rod in its full- that the core can be made out position and all other op- subtritical at any time in the erable rods fully inserted. subsequent fuel cycle with the strongest operable control rod fully withdrawn and all other operable rods fully inserted.

2. Reactivity margin - inoperable 2. Reactivity margin - inoperable control rods control rods

a. Control rod drives which Each partially or fully with-cannot be moved with control drawn operable control rod shall

o. rod drive pressure shall be be exercised one notch at least considered inoperable except once each week. This test shall as in c below. If a be performed at least once per partially or fully withdrawn 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in the event power opera-control rod drive cannot be tion is continuing with three or moved with drive or scram more inoperable control rods or pressure the reactor shall in the event power operation is be brought to a shutdown continuing with one fully or condition within 3.3/4.3-1 Amendment No. 114 )

l , .

QUAD CITIES DPR-29 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months unless investiga* partially withdrawn rod which tion demonstrates that the tennot be moved and for which cause of the failure is i.ot control rod drive mechanism dam-due to a f ailed control rod age has not been ruled out. The drive mechanism collet surveillance need not be com-housing. pleted within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the number of inoperable rods has been reduced to less than three and if it has been demonstrated that control rod drive rechanism collet housing failure is not the cause of an immovable control rod.

b. The control rod directional control valves for inoper-able controls rods shall be disarmed electrically and the control rods shall be in such positions that Specification 3.3.A.1 is met except as in d below,

c. Control rod drives which are fully inserted and electrically disarmed shall not be considered inoperable.

. d. Control rods with scram '

times greater than those permitted by Specification 3.3.C are inoperable, but if they can be moved with control rod drive pressure they need not be disarmed electrically if Specifica-tion 3.3.A.1 is met for each position of these rods,

e. During reactor power operation, the number of inoperable control rods shall not exceed eight.

3.3/4.3-2 Amendment No. 114 I l

1- . !

l QUAD-CITIES l DPR-29 l

3. Rod Position Indication System 3. Rod Position Ind: cation Syftem '

i

a. The position of a control a. Once per shift during power i rod shall be determined from operation and during con-the rod position indication trol rod withdrawal the 1

system (RPIS). control rod display shall be, observed for centrol rod position indication, i

b. If the position of a con- b. All control rods that have trol rod cannot be deter- been fully inserted and mined from the RPIS, such scrammed shall be given an control rod shall be movea insert signal once per shiftc to a known position or fully inserted, scrammed, and considered inoperable.

B. Control Rods B. Control Rods

1. Each control rod shall be 1. The coupling integrity shall be coupled to its drive or com- verified for each withdrawn pletely inserted and the control control rod as follows:

rod directional or control valves disarmed electrically, a. When the rod is withdren the first time subsequent to each refueling outage or after maintenance, observe discernibic response of the nuclear instrumentation; however, for initial rods when response is not dis-cernible, subsequent exer-cising of these rods after the reactor is critical shall be performed to veri-fy instrumentation response.

b. Wnen the rod is fully with-drawn the first time subse-quent to each refueling outage or after mair,te-nance, observe that the drive does not go to the overtravel position.

l 3.3/4.3-3 Amendment No. 114 L

QUAD-LITIES DPR-29

2. This requirerent does not apply 2. The control rod drive housing in the refuel condition when the support system shall be inspec-reactor is vented. Two control ted after reassembly and the re-rod drives may be removed as long sults of the inspection recorded.

as Specification 3.3.A.1 is net.

3. The control rod drive housing 3. The correctness of the control support system shall be in place rod withdrawal sequence input to during reactor power operation the RWH computer shall be veri-and when the reactor coolant fied after loading the sequence.

system is pressurized above atmospheric pressure with fuel Prior to the start of control in the reactor vessel, unless rod eithdrawal towards critical-all control rods are fully ity, the capability of the rod inserted and Specification worth minimizer to properly ful-3.3.A.1 is met. Till its function shall be veri-f;ed by the following checks;

a. Control rou withdrawal a. The RW ccmputer online sequences shall be estab- diagnostic test shall be lished so that maximum succassfully performed.

reactivity that could be added by dropout of any increment of any one con-trol blade would be such e

that the rod drop accident design limit of 280 cal /gm is not exceeded.

b. Whenever the reactor is in b. Proper annunciation of the the Startup/ Hot Standby or selection error of one out-Run mode below 20% rated of-sequence control rod thermal power, the rod worth shall oe verified.

minimizcr shall be operable.

A second operator or quali-fied technic 11 person may t

be used as a substitute for an inoperable rod worth minimizer which fails after withdrawal of at least 12 control rods to +.he fully withdrawn position. The rod worth minimizer may alsn be bypassed for low power physics testir.g to demon-strate tue shutdown margin requirements of Specifica-tion 3.3.A if a nuclear engineer is present and verifies the step-by-step rod movements of the test procedure.

3.3/4.3-4 Amendment No. 114 l

l

QUAD-CITIES DPR-29

c. The rod block function of the RWM shall be verified by withdrawing the first rod as an out-of-sequence control rod no more than to the block point.

4. Control rods shall not be with- 4. Prior to control rod withdrawal drawn for startup or refueling for startup or during refueling, unless at least two source range verify that at least two source channels have an observed count range channels have an observed rate equal to or greater than count rate of at least three three counts per second and counts per second.

these SRM's are fully inserted.

5. During operation with limitirg 5. When a limiting control rod pat-control rod patterns, as deter- tern exists, an instrument func-mined by the nuclear engineer. tional test of the RBM shall be either: performed prior to withdrawal of the designated rod (s) and daily thereafter.

a. both RBM channels shall be operable, b, control rod withdrawal shall be blocked; or

c. the operating power level shall be limited so that the MCPR will remain above the MCPR fuel cladding in-tcegrity safety limit as-su:..ing a single error that results in complete with-drawal of any single op-erable control rod.

6. The scram discharge volume vent and drain valves shall be veri-fied open at least once per 31 days. These valves may be closed intermittently for test-ing under administrative control and at least once per 92 days, each valve shall be cycled through at least one complete cycle of full travel. At least once each Refuelir,g Outage, the scram discharge volume vent and drain valves will be demonstrated to:

3.3/4.3-5 Amendment No. 114

i QUAD-CITIES DPR-29

a. Close within 30 seconds af-ter receipt of a signal for control rods to scram, and

b. Open when the scram signal is reset. I C. Scram Insertion Times C. Scram Insertion Times

1. The average scram insertion 1. Af ter refueling outage and prior _ ,

time, based on the deenergiza- to operation above 30% power, tion of the scram pilot valve with reactor pressure above 800 solenoids et time zero, of all psig, all control rods shall be t

i operable control rods in the subject to scram-time reactor power operation measurements from the fully condition shall be no greater withdrawn position.The scram- t than: . times for single rod scram testing shall be measured Average -Scram without reliance on the control

% Inserted From Insertion rod drive pumps.

Fully Withdrawn Times (sec)_

5 0.375 20 0.900 50 2.00 90 3.50 '

The average of the scram inser- .

tion times for the three fastest -

control rods of all groups of f our- control rods in a two by two array shall be no greater than: ._

% Inserted from Average Scram Fully Withdrawn Times (sec) 5 0.398 20 0.954

.- 50 2.12 90 3.80

2. The maximum scram insertion time 2. All-control rod drives shall for 90%-of any operable control have experienced scram test :

rod shall not exceed 7 seconds. -measurements each year, Also,

50% of the control rod drives in each quadrant of the reactor- !

core shall be measured for the scram times specifled in Specification 3.3.C during the i

i 3.3/4.3 6 Amendment No. 114- !

~

l QUAD-CITIES DPR-29 interval not more frequently than 16 weeks nor less frequently than 32 weeks. These tests shall be performed with a reactor pressure above 800 psig and may be measured during a reactor scram. Whenever all of the control rod drive scram times have been measured, an evaluation shall be made to pro-vide reasonable assurance that proper control rod drive per-formance is being maintained.

The results of measurements per-formed on the control rod drives shall be submitted in the annual operating report to the NRC.

3. If Specification 3.3.C.1 cannot 3. The cycle cumulative mean scram be met, the reactor shall not be time for 20% insertion will be ,

made supercritical; if op- determined immediately following erating, the reactor shall be the testing required in Specifi-shut down immediately upon de- cations 4.3.C.1 and 4.3.C.2 and termination that average scram the MCPR_ operating limit ad-time is deficient. justed, if necessary, as re-quired by Specification 3.5.K.

4. If Specification 3.3.C.2 cannot be met, the deficient control rod shall be considered inop-erable, fully inserted into the core, and electrically disarmed. .

5. If the overall average of the 20% insertion scram time data generated to date in the current cycle exceeds the limit specifin' in the CORE OPERATING L! HITS l REPORT, the MCPR operating limit must be modified as required by Specification 3.5.K.

D. Control Rod Accumulators D. Control Rod Arcumulators At all reactor operating pressures, a Once a shift, check the status of the rod accumulator may be inoperable pressure and level alarms for each provided that no other control rod in accumulator.

the nine-rod square array around that rod has: i

1. an inoperable accumulator, 3.3/4.3-7 Amendment No. 320

e e QUAD-CITIES DPR-29 ,

1

2. a directional control valve electrically disarmed while in a nonfully inserted position, or )

3. a scram insertion greater than maximum permissible insertien time.

If a control rod with an inoperable accumulator is inserted full-in and its directional control valves are electrically disarmed, it shall not be considered to have an inoperable accumulhtor, and the rod block asso-ciated with that inoperable accumu-lator may be bypawsed.

E. Reactivity Anomalies E. Reactivity Anomalies lhe reactivity equivalent of the dif- During the startup test program and ference between the actual critical startups following refueling outages, rod configuration and the expected the critical rod configurations will configuration during power operation be compared to the expected configur-shall not exceed 1*. ok. If this ations at selected operating condi-limit is exceeded, the reactor shall tions. These comparisons will be be shutdown until the cause has been used as base data for reactivity determined and corrective actions monitoring during subsequent power have been taken. In accordance with operation throughout the fuel cycle.

. Specification 6.6, the NRC shall be At specific power operating condi ,

notified of this reportable occur- tions, the critical rod cnnfiguration rence within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 1 will be compared to the configuration r/ expected based upon appropriately

'y corrected past data. This comparison

, will be made at least every equiva-a lent full power month, F. Economic Generation Control System F. Economic Generation Control System Operation of the unit with the eco- Prior to entering EGC and once per nomic generation control system with shift while operating in EGC, the EGC automatic flow control shall be per- operating parameters will be reviewed missible only in the range of 65% to for acceptability.

100% of rated core flow, with reactor power above 20%.

G. If Specifications 3.3.A through D above are not met, an orderly shut-down shall be initiated and the reac-tor shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.3/4.3-8 Amendment No. 114

QUAD-CITIES

. DPR-29 3-.-HH!'!NG 00M4HMS F0p opranon nam s.3/ws generivery co^ tract A. Reactivity-timitatitmr slurtetuN magid

1. Reactivity mergin cerdoading-The N N Nt N t h imitation is a restriction to be applied principally to the design of new fuel which may be loaded in the core or into a particular refueling pattern. Satisfaction of the limitation can only be demonstrated at the time of loading and must be such that it will apply to the entire subsequent fuel cycle. The generalized form is that the reactivity of the cote loading will be limitedsothecorecanbemadesubcriticalbyatleastR+0.25%ag in the mor,t reactive condition during the operating cycle, with the strongest control rod fully withdrawn and all others f ully inserted. The value of R in % Ak is the amountly which the core h reactivity, at any time in the operatigg cycle fis calculated to be greaterthanatthetimeofthechecgi.e.,theinitialloading.

R must be a positive quantity or zero. A core which contains temporary control orwhich reactivity characteristic otherincreases burnablewith neutron absorbers may h core lifetime, ~

(h oucA a maximum, and decreases thereafte a.

The value of R is the difference betwee the calculated core reactivit s t the beginning of the oper ting cycigpnd the calculated"value of the core reactivity any time later in the cyc1g where it would be greater than at the beginning. The value of R shall include the potential shutoown margin loss assuming f ull 840 settling in all inverted poison tubes present in the core. A new value of R must be determined for each new fuel cycle.

The 0.25% AL in the expression R + 0.25% ok is provided as a finite, demonstrable, subtriticality margin. This margin is

, demonstrated by full withdrawal of the strongest rod and partial withdrawal of an adjacent rod to a position calculated to insert at least R 4 0.25% Ak in reactivity. Observation of subtriticality in this condition assures subeviticality with not only tha stongest rod fully withdrawn but a - in of at least R + 0.25% ok beyond thi sMcna+ ion 3 with

d. Control Red CPERnBILtry C. Reee t444y-margift---stw4-eonttol-coes ActicN sidemenf c. Cor- hadm uliyQ Etec+f-ieet4en- 3.3. A4 requires that a r be taken out3ofsservice if it cannot be moved with drive pressure. .If the rod is fully inserted and then disarmed electricall MNote: 10disarmthedrivel plectrically, four_ amphenoEtype plug connectors are removed from the drive insert and withdrawal solenoids, rendering the drive j t immovable. This procedure is equivalent to valving out the drive move +o af ter- comple{lon Of %e senlence .

i 3.3/4.3-9 Amendment No. 114

QUAD-CITIES DPR 29

~

, neve % %Is c nok: lelion cf

~ and is preferred, as drive water cools and minimizes crudl

_ accumulation in the drive.) t is in a safe positTon of maximum contribution to shutd .

If it it disarmed eleeteica14yinano@ownreactviully. ried position that positicn shall

@ARCri% beconsistentwiththe)MMp( eactM4y limitation stated in Specification 3.3. A.X This assures that the core can be shut down at all times with the remaining control rods, assuming the strongest operable control rod does not insert. An allowable pattern for control rods valved outefservice, which shall meet the specification, will be available to the operator. The number of rods permitted to be inoperable could be many more than the eight allowed by the specification, particularly late in the operation cycle; however, the occurrence of more than eight could be indicative of a generic control rod drive problem and the reactor will be shut down.

Mqif damage within the control rod drive mechanism and in f partTtta r, cracks in drive internal housings, cannot be ifd out, then a gen roblem affecting a number of drivej c inot be ruled out. Circumfere 41 cracks resulting from ysefs assisted intergranular corrosfBn@ve occurredpi tM collet housin0 of drives at several BWRs. liih f' cracking could occur in a number of drives and if the cra pagated until severance of the collet housing occyr d scram co dAe prevented in the affected rods. L,.imiting the period of operitQwith a potentially severed co1 lance after detec J i @ Jet'Tiousing ne stuck rod and will requiring assure that increased the su reactor w n t be operated with a large number of rods with failed collet hous'ngt.

tenirol

.T. 'l(. , Rod Position Indication System (RPIS)

Normal control rod position is displayed by two-digit indication to the operator from position 00 to 48. Each even number is a latching position, whereas each odd number provides information lile the rod is irGhotion and inpu for rod drift annunciation.

ALnoN Th?tM provides for the condition w@here no positive information is 8 **' M C' N displayed for a large portion or all of the rod's travel. In-thb nserf eete-the-cod-4-9 hen-aW4-hert-+4 nabr-4ndW-idu14y-uramed A 9

and-t+eated-*-an-4noperable-we. Usuallgonly one digit of one or two of a rod't positions is unavailable with a faulty RPIS, and the control rod may be located in a known position.

HK Control Rod Wtht5t ewn4 Drive. Coupling X Control rod dropout accidentwas discussed in Reference Mgan lead to significant core damage. If coupling integrity is mal'6tained, the possibility of a rod dropwt accident is eliminated. The

overtravel position feature provides a positive check, as only uncoupled drives may reach this position.

t 3.3/4.3-30 Amendment No. 114

QUAD-CITIES DPR-29 Neutron instrumentation response to rod movement proviees a verification that the rod is following its drive. Absence of such response to drive movement would indicate an uncoupled condition.

J. Control Rod Drive Housing &Lppori JL The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system. The design basis is given in Section 6.6.1, and the design evaluation is given in Section 6.6.3 of the SAR. This support is not required if the reactor coolant system is at atmospheric pressure, since there would then be no driving force torapidlyejectadrivehousing. Additione44yT-the-support-is-not-requi red-i f-e Mnent Nd -rod s-a re-f uHy-i nsected-oMf-a n-adequa te shutdown-margin-with-onu;ontrel- red withdrawn has been demonstruedr44 nc e-the-eeastee-veruld-cew4navbeel t4 eal-e ven--4n the event-of-completeMeetion-of-the-sttengest-contee4-cod L. . Rod IDorth ininimiecr*

J. Controlrodwithdrawalan(insertionsequencesareestablishedto assure that the maximungiMequence individual control rod or control rod segments which are withdrawn could not be worth enough to cause the rod drop accident de61gn limit of 280 cal /gm to be exceede %if they were to drop out of the core in the manner defined for the Yod drop accident. These sequences are developed prior to initial operation of the unit fu lowing ar,y refueling outage and the requirement that an operator follow these sequences is

. supervised by the RWM or a second qualified station employee.

These sequences are developed to limit reactivity worths of control rods and,together with the integral rod velocity limiters and the action of the control rod drive system, limi@ potential reactivity insertion such that the results of a control rod drop accident will not exceed a maximum fuel energy content of 280 cal /gm. The peak fuel enthalpy of 280 cal /gm is below the energy content at which rapid fuel dispersal and primary system damage have been found to occur based on experimental datgs 46 discussed in Reference 2.

The analysis of the control rod drop accident was originally presented in Sections 7.9.3, 14.2.1.2 and 14.2.1.4 of the SAR.

Improvements in analytical capability have allowed a more refined analysis of the control rod drop accident.

These techniques are described in a topical report (Reference 2) and two supplements (References 3 and 4). In addition, a banked position withdrawal sequence described in Reference 5 has been developed to further reduce incremental rod worths. Method and basis for the rod drop accident analyses are documented in Reference 1, 3.3/4.3-11 Amendment No. 114 i

QUAD-CITIES DPR-29 By using the analytical models described in thosr reports coupled with conservative or worst-case input parameters, it has been determined that for power levels less than 20% of rated power, the specified limit on i$equence control rod or control rod segment worths will limit the peak fuel enthalpy to less than 280 cal /p;gm.

Above 20% power oven single operator errors cannot result in out-of-sequence control rod worths which are suf ficient to reach a peak fuel enthalpy of 280 cal /g should a postulated control rod drop eee44e 4 accident occur. 0 9 m, The following parameters and worst-case assumptions have been utilized in the analysis to determine compliance with the 280 cal /gm peak fuel enthalpy. Each core reload will be analyzed to show conformance to the limiting parameters.

4

a. An interassembly local peaking factor.*

b. The delayed neutron fraction chosen for the bounding reactivity curve.

c. A beginning of-life Doppler reactivity feedback,

d. Scram times slower than the Technical Specification rod scram insertion rate (Section 3.3.C.1).

e. The maximum possible rod drop velocity of 3.11 fps.

f. The design accident and scram reac'tivity shape function,

g. The moderator temperature at which criticality occurs.

Inmostcasegtheworthofikequencerodsorrodsegmentgin conjunction with the actual values of the other important accident .

analysis parameters described above, would most likely result in a peak fuel enthalpy substantially less than 280 cal /g design limit.

9m Should a control to rop accident result in a peak fuel energy content of 280 cal /y.() fewer than 66(M7 x 7) fuel rods are conserva-tively estimated to parforate. This would result in an offsite dose well below the guideline value of 10 CFR 100, for 8 x 8 fuel, fewer than 850 rods are conservatively estimated to perforate, with nearly the same consequences as for the 7 x 7 fuel case because of the rod power differences.

To include the power spike ef fect caused by gaos between fuel pellets 3.3/4.3-12 Amendment No. 114

QUAD-CITIES OPR-29 The rod worth minimizer provides automatic supervision to assure ,

that outGofesequence control rods will not be withdrawn or insertedy i.e. , it limits operator deviations f rom planned withdrawal sequences (referenco SAR Section 7.9).

procedural control of control rod worth. It In serves as a backup the eveDLth#t the3to h worth minimizer is outGofOservice when required, aflicensed operator or other%M0le44chMul-esNoye6 can manually fulfill the technkoll)ua indivi lunti! control led ) rod pattern conformance function of the rod worth

/ minimizer. In this case, the normal procedural controls are backed up by independent procedural controls to assure conformance.

H. . source. Kanye monitoriev a functoco,

% The source range monitor (SRM) system performs no automatic safety system function, i.e. it has no scram function, it does provide theoperatorwithav\sualindicationofneutronlevel. This is needed for knowledgeable and efficient reactor startup at low neutron levels. The consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per secund assures that any transient, should it occur, begins at or above the initial value -f 10 8 of rated power used in

%e the analyses of transients f rom cold conditions. One opereble SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRMs is provided as an added conservatism.

H. Roct BIc>ck. mondey

% The rod block monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.

This system backs up the operator, who withdraws control rods according to a written sequence. The specified restrictions with one channel outGnf6 service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists. During reactor operation with certain ////////16//W ///

/s/ff////, the withdrawal of a designated single control rod could result in one or more fuel rods with MCPRs less than the MCPR fuel

o. cladding integrity ///#f ff/lL During use of such pattern @it is judgedthattestingoftheRBMsystegoassureitsoperabiliTy prior to withdrawal of such rod gwill assure that improper withdrawal does not occur. It iT the responsibility of the Nuclear Engineertoidentifytheselimitingpatternsandthedesignatedrodg either when the patterns are initially established or at they develop due to the occurrence of inoperable control rodt. in other than limiting patterns.

(Insert B) 3.3/4.3-13 Amendment No. 114 l

L QUAD-CITIES l DPR-29 K. Scram Discharge Volume Vent and bmin Valycs l

% The operability of the / cram fischarge /olume vent and drain valves i assurestheproperventinganddrainingofthe/olume,sothatwater accumulationinthefolumedoesnotoccur. These specificationb provide for the periodic verification that the valves are open, and for the testing of these valves under reactor scram conditions during each J/f///fM h!//#.

D &nflol Rod Hanimum Scram inser+ien Times } G. Confrcl Rcd Avemge. Scmm :

C. Scr= != :rt!"" * :: Insertien %e.s ; anet., F. Four lonfrcl Reci ;

Grou.p Seram Inser/hn nmes '

The control rod system is analyzed to bring the reactor subtritical at a rate f ast enough to prevent fuel damage, i.e. , to prevent the MCPR from becoming le.s than the fuel cladding integrity /Al/// f/@. ,

Analysis of the limiting power transient shows that the negative reactivity rate esulting from the scram with the average response of all the drive ,a given in the above specification, provide the required protection, an- CPR remains greater than the fuel cladding integrity ,

///f//

Specification 3.5.

////f. It is necessary tr raise the MCPR operating limit (per when the average 20% scram insertion time rerahes the limit specified in the CORE OPERA 11NG LIMITS REPORT on a cycle l cumulativebasis(overallaverageofsurveillancedatatodate)in order to comply with assumptions in the implementation procedure for ,

the 00YN transient analysis computer code. The basis for choosing this 20% scram insertio

i -

for specification 3.5.Atime limitanalytical M.>In the is discussed treatmentfurtherofinthe thetransients, bases 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods. This is adequate and conservative when compared to the typically observed time delay of about 210 milliseconds. Approximately 90 milliseconds after ,

neutron flux reaches the trip point, the pilot scram valve solenoid deenergizes and 120 milliseconds later the control rod motion is estimated to actually begin. However, 200 milliseconds rather than 120 milliseconds '

-is conservatively assumed for this time interval in the transient analyses and is also included in the allowable.* cram insertion times specified in Specification iHh6. J 8,0,2.8.6 anot A 8.a 1trverse-timee-fw-e44-eent+el-rd, uf444e-deteem4*ed-at the tis of eeeh refueling-outcge. f. repeesent-at4ve-semp4e-ef-centeM-red <,-wth se=am-tested-4w4ng-the intewa4-of-9reetw-4han-16 eecks but nM =^"e then-32-weeks, (Inseet C ) yow,,,(t i Scram times of new drives are approximately 2.5 to 3 seconds; ower

! rates of change in scram times following initial plant operation at power are expected. The test schedule provides reasonable assurance of detection of slow drives before system deterioration beyond the limits '

[ of Specification 3.3.C. The program was developed on the basis of the statistical approacn outlined below and judgment. ,

The history of drive performance accumulated to date indicates that the 2 90% insertion times of new and overhauled drives approximate a normal

^

distribution about the mean which tends to become skewed toward longer 3.3/4.3-14 Amendment No. 120

,.n~ ~,r # . ym,.o -

,..a,. . - ,-,.,_-,..m,,.

, . . ,,,,.,_.,9..n ._wu,,_ ,,mm,. , . - , 3 y , , . , , ,

l QUAD m ill; DPk29 jd O' scram times as operating time is a umulated. The probability of a drive not exceeding the mean 90% inserti n time by 0.75 seconds is greater than 0.999 for a normal distri'oution. The measurement of the scram performance of the drives surrounding a driv exceed 449 the expected range of scram performancq7 pill detect local var ations and also provide assurance that local scrartime limits are not exceeded. Continued monitoring of other drives exceeding the expected range of scram times provides surveillance of possible anomalous performance.

The numerical values assigned to the predicted scram performance are based on a analysis of the Dresden 2 startup data and of data from other BW uch as Nine Mile Point and Oyster Creek.

The occurrence of scram times within the limits, but significantly longer than average, should be viewed as an indication of a systematic problem with control rod drives, especially if tne number of drives exhibiting such scram times exceeds eight,ithe allowable number of inoperable rods. (EFhich is) 4 Scrtzm

% Control Rod 4Accumulators

- basis ~ r this specification was not described in the SAR and i there sresented in its entirety. Requiring no more tt m inoperab.e% imulator in any nine-rod square at ray is aw on a series of XY P0Q-4 quar core calculations of a cold cl core. The worst ca:e in a nine-rod w d nak 1.0. Other repeating rod sequences witwal sequence re rods resu pl4tfidrawr. resulb <in k> 1.0.

At reactor pressures in excess o- psig,eventhosecontrol'Ndswith inoperable accumulators will able et required scram insertion times due to the actiop # reactor pressure. Q addition, they may be normally inserteDrs/Tng the control rod drive hyd ic system.

Procedara 01 will assure that control rods with rable a rs will be spaced in a one-in-nine array rather t rouped (Inser-+ D) i 3.3/4.3-15 Am:ndment No. 114

QUAD-CITIES DPR-29 B.

[. Reactivity Anomalies During each fuel cycle, excess operating reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned.

The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod patterr selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most sensitive and directly interpretable data relative to core rea:tivity, hrthermcre, using power operating base conditions permi@ frequent reactivity cumparisons.- Requiring a reactivity comparison at the tpec1?ied frequency assures that a comparison will be made before the core mactivity change exceeds 1% ok. Deviations in core reactivity greater than 1% Ak are not expected and require thorough evaluation.

A 1% reactivity limit is considered safe, since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.

O f. Economic Generation Control System Operation of the facility with the economic generation control system (EGC) (automatic flow control) is limited to the range of 65% to 100% of rcted core flow. In this flow range and with reactor 1.'wer above 20%,

the reactor coula safely tolerate a rate of change of load of 8 FNe/sec (referenceSARSection7.3.)).

Limits within the EGC and th flow control system prevent rates of change greater than approximately 4 FNe/sec. When EGC is in operation, this fact will be indicated on the main control room console. 4tte--

rewit: cf i-itial testing ""' be provided te the-NRC-before the en:;et of--feutinc operation-*Rh-EGC.

3.3/4.3-16 Amendment No. 114

a .

. _ QUAD-CITIES DPR-29 1

References

-1.." Generic-ReloadFuelApplicatioh!"NEDE-24011-P-A* *

2. C. J. . Paone, R. C. Stirn, and J. A. Wooley, " Rod Drop Accident Analysis for Large BWRsg'f( GE Topical Repor(@NEDO-10527, March 1972.

3. C. J. Paone, R. C._Stirn and R. M. Young, " Rod Drop Ac:ident Analysis for Large BWRg3'f dS upplement 1. GE Topical Report NEDO-10527, July 1972.

4. J. M. Haun, C. J. Paone, and R. C. Stirn, " Red Drop Accident Analysis for Large BWRs,fAddendum 2. Exposed Coregg', j Supplement 2 GE Topical Report NED0-10527, January 1973.

, 5. C. J. Paone, " Banked NED0-21231, January1977.

,$osition /ithdrawal fequeneggjiLicensing Topical Repor

Approved' revision number at time reload fuel analyses are performed, i

t b

L i.

l 3.3/4.3-17 Amendment No- 114

TECHNICAL SPECIFICATION BASES 3.3/4.3 REACTIVITY CONTROL Insert "A" In this case, for OPERATIONAL MODES 1 and 2, the position of the control rod is determined by an alternate method, or the control rod is moved to a position with an OPERABLE pcsition indicator, or the control rod is declared inoperable, fully inserted and disarmed, if these ACTION statements are not met, the reactor is placed in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With a withdrawn control rod position Indicator inoperable in OPERATIONAL MODE 5, the control rod is required to be moved to a position with an OPERABLE position indicator, or the control rod is inserted, insert *B" If a peripheral control rod is selected, the neutron leakage is sufficiently high such that withdrawal of this rod will not violate the fuel cladding integrity SAFETY LI ViiT. Thus, the RBM function is not required for withdrawal of peripheral control rods, insert "C" The scram times for all control rods will be determined prior to thermal power exceeding 30% of RATED THERMAL POWER following CORE ALTERNATIONS or after a roactor shutdown that is greater than 120 days, Scram times will be determined for specifically affected control rods following maintenance or modification to the control rod drive system which could affect the scram insertion time of those specific control rods. Also, at least 10% of the controt rods are scram time tested, on a rotating basis, at least once per 120 days of REACTOR POWER OPERATION, insert "D" Control rods with inoperable accumulators are declared inoperable and Specification 3.3.C then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with ino aerable accumulators may still be inserted with normal drive water 3ressure. OPERAB:LITY of the accumulator ensures that there is a means available to nsert the control rods even under the most unfavorable depressurization of the reacior.

ch/967/1

. z.-.

SIGNIFICANT HAZARDS CONSIDERATIONS.

AND-ENVIRONMENTAL ASSESSMENT EVALUATION PROPOSED TS 3.3/4.3

' REACTIVITY CONTROL'

}

.. 1.

i

-_. EYALUATION ZQB #19FIFICART IULIARDA QQRSIDEIULTIQM i

Proposed Specification 3.3/4.3 Reactivity Control The proposed changes provided in this amendment request are made in order-to provide a more user friendly document, incorporate desired. technical improvements,'and to incorporate-some improvements from later operating BWRs. 'These changes have been reviewed by Commonwealth Edison and we believe that they donnot present a-Significant Hazards Consideration. The basis for our determination is documented as follows:

BASIS EQB HQ SIGNIFICANT HAZARQS CONSIDERATION !

Commonwealth Edison has evaluated this proposed-amendment and ,

-determined.that it involves no-significant hazards consideration.

In accordance with the criteria'of 10 CFR 50.92[c) a proposed amendment toEan operating license involves no significant hazards consideration'if operation of the facility, in accordance with the proposed amendment,-would not:

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

a.- The' Generic changes.to the-technical specifications involve administrative changes to format and arrangement of the material. As such, these changes cannot involve a

- significant increase in the probability-or consequences of ,

an accident previously evaluated.

'b.- Proposed' Changes to Shutdown Margin Requirements The proposed changes to the-requirements-for Shutdown Margin technical specifications implement proven STS i guidelines that are applicable to: Quad Cities. Proposed '

- Applicability of Operational-Modes 1, 2, 3, 4 and 5 ensures that Shutdown Margin limitations are enforced when fuelsis in.the vessel and a potential exists for reactivity-excursions. Present Shutdown Margin-limitations are-retainedLand: STS Actions are-added to

- avoid using the provisions of proposed Specification 3.0.C

- and tocimplement proven Action-allowances. Since present-limitations are retained and proven STS provisions are-

-implemented, the changes do not involve a significant increase-in-the. probability or consequences.of an accident previously: evaluated,

c. Proposed-Changes for control Rod Operability-Proposed Specification 3.3.C/4.3.C on Control Rod Operability utilizes present provisions,'STS guidelines and later operating plants' allowances for testing

- -- w

s .

l disarmed control rod directional control valves. The combination of these requirements are used to form acceptable requirements for Quad Cities. The present reference to collet housing failures is deleted since this determination is made during CRD disassembly and inspection while the plant is shutdown and is not determined during plant operation. Proposed Operability is in operational Modes 1 and 2 when control rod insertion rates are considered in the accident analysis. Proposed Actions for Control Rod Operability implement proven STS guidelines that are applicable at Quad Cities. The proposed Surveillance Requirements are based on present and STS provisions that will provide necessary demonstration of control rod operability. Since the present level of control rod operability is maintained, the proposed changes do not involve a significant increane in the probability or consequences of an accident previously evaluated.

d. Proposed Changes for the Control Rod Position Indication system Proposed Specification 3.3.I/4.3.I for the Control Rod Position Indication System is based on present and STS provisions. Operability of the Control Rod Position Indication System is required when control rod movement is allowed in Operational Modes 1, 2 and S The present Specifications do not cor.tain Actkon prov.isions that separately address the needs of the Control Rod Position Indication System in each of the Operational Modes in which the system is required operable. Proposed Actions from the STS are used to address the different requirements for Operational Modes 1 and 2 versus operational Mode 5. The present specifications require control rods that have an inoperable position irdication to be moved to a known position or be inserted, scrammed and considered inoperable. The insertion and disarming of the control rod is sufficient without the additional requirement to scram the control rod, and as such, this requirement to alco scram the control rod is deleted. The intent of present Surveillance Requirement 4.3.A.3.a to verify position indication of each control rod is retained in proposed SR 4.3.I.1. SRs are added to verify position indication during weekly notching of control rods per SR 4.3.C.1 and 4.3.H.2. The proposed changes to the Control Rod Position Indication System specifications improve present requirements and, as such, do not involve a significant increase in the probability or consequences of an accident previously evaluated.

c. Proposed Changes to the Control Rod Drive Coupling Requirements The proposed rewrite of the Control Rod Drive Coupling I

I

specifications combines present requirements and STS guidelines in a manner that improves usability and operational flexibility while retaining necessary provisions. The proposed Applicability of Operational Modes 1, 2 and 5 (for withdrawn control rods) addresses all reactor modes of operation where control rod drive coupling is necessary. Present Action provisions do not allow an attempt to recouple the control rod and this provision is added from STS guidelines. As with present requirements the control rod is inserted and disarmed, but like the STS, this is after the recoupling attempt has not proven successful. The intent of present Surveillance Requirements are maintained with the rewrite of proposed SR 4.3 H. The present level of operability for Control Rod Drive Coupling is maintained by the proposed changes with operational flexibility added to allow one attempt at recoupling. Since necessary control rod coupling requirements are maintainen in the proposed changes, there is no significant increase in the probability or consequences of an accident previously evaluated,

f. Proposed Changes to the Roi Worth Minimizer Requirements The proposed Specification for the Pod Worth Minimizer, 3 . 3 . L/ 4 . 3 . L, is based mainly on present provisions with some STS guidelines. The Applicability comes from present requirements in Operational Modes 1 and 2 below 20% Rated Thermal Power. Exceptions in the Applicability allows the RWM to be bypassed for Low Power Physics tests and allows entry into Operational Mode 2 to allow determination of operabi13ty prior to withdrawing control rods for the purpose of bringing the reactor critical. The intent of present Actions is retained with the addition of the STS guideline to require that without use of the second licensed operator or technically qualified individual, control rod movement is only allowed by initiating a reactor scram. Present SRs are retained by using STS wording and guidelines. The proposed changes to the technical specifications maintains necessary operability requirements for the RWM when required to perform the design function below 20% Rated Thermal Power. Since the present level of operability for the RWM is retained, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated,

g. Proposed Changes to the Source Range Monitoring Requirements The proposed Specification for Source Range Monitoring, 3.3.N/4.3.N, is written using present provisions and STS guidelines. The present LCO for the Source Range Monitoring Function requiring at least two SRM channels to be operable in Operational Modes 2, 3, and 4 is expanded

to require three operable SRMs for Operational Mode 2.

Proposed Applicability is based on STS provisions of Operational Modes 2, 3, and 4 with operability in Operational Mode 2 only when IRMs are on range 2 or below.

Present operability requirements for Operational Mode 5 are contained in Section 3.10/4.10 and are not needed in Section 3.3/4.3. Proposed Action provisions with SRMs inoperable adopt STS guidelines by requiring a plant shutdown if SRMs are not returned to operable status while operating in Operational Mode 2 and rewiring insertion of rods and the reactor mode switch to being locked in the shutdown position with the LCO not met in Operational Nodes 3 or 4. Present SRs are retained and Channel Checks, Channel Calibrations, and Channel Functional Tests are added from STS guidelines. Since with the proposed changes, the present level of SRM operability is maintained for Operational Modes 3 and 4, and increased for Operational Mode 2, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

h. Proposed Changes for Rod Block Monitor Requirements The proposed Specification for RBM, 3.3.M/4.3.M, is written from present provisions, present system design requirements, and STS guidelines. Both RBM channels are required operable in Operational Mode 1 above 30% Rated Thermal Power in accordance with present provisions and with present allowance that one channel may be bypassed above 30% Rated Thermal Power without a time restriction provided that a limiting control rod pattern does not exist. Added to the Applicability is a clarifying statement on system design to state that both RBM channels are automatically bypassed when a peripheral rod is selected. This proposed clarifying statement is necessary to prevent misinterpretation of RBM operability requirements when peripheral. rods are selected above 30%

Rated Thermal Power. STS guidelines are used to rewrite present Action statements to address the conditions where either one RBM channel is inoperable and not bypassed or where both RBM channels are inoperable. Safo conditions are established by the proposed Actions by inhibiting control rod movement if the remedial measures are not met.

Surveillance Requirements for the RBM implement present Instrument Functional Test frequencies when operating on a limiting control rod pattern and reference Table 4.2-1 for regular surveillance testing. The proposed changes to the specifications for the RBM add to and improve the usability of the present requirements while maintaining at least the present level of operability. Therefore, the proposed changes do not involve a significant increane in the probability or consequences of an accident previously evaluated.

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1. Proposed Changes for Scram Discharge Volume' Vent and' Drain Valves Requirements Proposed Specification 3.3.K/4.3.K- for the Scram Discharge Volume Vent and Drain-Valves uses later operating BWR plants' -specifications to develop LCO, Applicability and Action requirements. Present Surveillance Requirements follow later operating plants' SRs and are retained. All Scram Discharge volume Vent and Drain Valves are required operable in Operational Modes 1 and 2 in order to provide a means to isclate the-Scram Discharge Volume when noeded for reactor scram purposes involving multiple control rods. The proposed Actions consider the-design of two isolation valves ca each drain and vent line by allowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to return a valve to operable status if one of the vent or drain valves on a line is inoperable or one vent-valve and one drain valve on a SDV is inoperable.

Only 8. hours is allowed for continued operation-if the SDV Vent and Drain Valves are inoperable for reasons other than those-just described. The proposed changes improve the present technical-specification requirements for the SDV Vent and Drain Applicability Valvesrequirements and Action by providing while neededmaintaLCO,ining at least the present level of operability. Therefore, the proposed changes do not-involve a significant increase in the probability or consequences of an accident previously evaluated,

j. Proposed Changes for Control Rod Maximum Scram Insortion Times Proposed specifications for the-Control Rod Maximum Scram Insertion Times, 3.3.D/4.3.D, retains the;present-7 second 4

-limitation while adding STS Applicability:and. Action provisions. Proposed Applicability of Operational Modes 1 and 2. covers the conditions where scram insertion times must be-met for analysis considerations.- Present Action provisions are included in the proposed Actions and-additinnal STS guidelines'are added to ensure that proper control rod operability-is maintained when operation is continued-with three or more control rods with maximum scram insertion-times exceeding 7 seconds. With the maximum scram insertion time of one or more control rods exceeding 7 seconds,-the control rods are declared

. inoperable and with three or more control rods exceeding-

'the limit, scram insertion time tests are performed on an ,

accelerated basis of at least once per 60 days. Proposed Surveillance Requirements for control rod scram time testing follows STS guidelines and retains the present requirement to determine cycle cumulative mean scram time.

The STS guidelines-for control rod scram time testing have been' utilized at otho operating facilities with control

, rod drive systems similar to that at Quad Cities and will maintain necessary assurances of control rod operebility.

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The proposed changes maintain necessary operability of the Control Rod Maximum Scram Insertion Time criteria, and; therefore, do not involve a-significant increase in the probability or consequences of an accident previously evaluated,

k. Proposed Changes for the Control Rod Average Scram Insertion Times Proposed changes to the Specifications for the control Rod Average Scram Insertion Times, 3.3.E/4.3.E, involve retention of present requirements or the intent of present requirements in an STS format. Present Control Rod Average Scram Insertion Times are retained in Operational Modes 1 and 2 using present requirements and STS guidelines. Operational Modes 1 and 2 cover the reactor operating conditions where control rod scram times are required by analysis. The proposed Action allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to reach Hot Shutdown and implements present intent to shutdown the reactor. Proposed Surveillance Requirements reference the SRs of 4.4.D which are discussed above. The proposed changes maintain at least the present level of operability and as such do not involve a significant increase in the probability or consequences of an accident previously evaluated.

1. Proposed Changes for the Four Control Rod Group Scram Insertion Times Requirements Proposed changes to the Specifications for the Four Control Rod Group Scram Insertion Times, 3 '. 3 . F/ 4 . 3 . F ,

involve retention of present requirements or implementation of STS guidelines. Present Four Control Rod Group Scram Insertion Times are retained in Operational Modes 1 and 2 using present requirements and STS guidelinen.- Operational Modes 1 and 2 cover the reactor-operating conditions where control rod scram times are required by analysis. The proposed Actions are based on STS guidelines and requires the slow control rods to be declared inoperable until an analysis is performed to determine that required scram reactivity remains for the slow four control rod group. If continued operation is allowed, accelerated scram time testing is imposed on a frequency of at least once per 60 days. Proposed Surveillance Requirements reference the SRs of 4.4.0 which are discussed above. The proposed changes maintain at least-the present level of operability and as such do not-involve a significant increase in the probability or consequences of an accident previously evaluated.

m. Proposed Changes for the Control Rod Scram Accumulators Requirements Proposed Specifications for the Control Rod Scram

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Accumulators, 3.3.G/4.3.G, are based on present-provisions and STS guidelines. The proposed changes require all Control Rod Scram Accumulators to be operable in Operational Modes 1, 2, and 5 with applicability in Operational Mode 5 being only for withdrawn control rods.

The proposed applicability covers all reactor Operational Modes where control rods can be withdrawn and insertion using accumulators may be required. The proposed Actions implement STS guidelines and separately address operational Modes 1 and 2 and Operational Mode 5. The proposed Actions require the plant to be shutdown while in operational Modes 1 or 2 with remedial steps not met or requires the reactor mode switch to be placed in shutdown if remedial steps are not met in Operational Mode 5.

Present Surveillance Requirements are replaced with STS guidelines to check the pressure at least once per week of each control rod scram accumulator. Operability of the Control Rod Scram Accumulators is maintained at least at the present level of operability by the proposed changes, and as such, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated,

n. Proposed Changes to the Reactivity Anomalies Requirements Proposed changes for Specifications for Reactivity Anomalies, 3.3.B/4.3.B include using present provisions to write the LCo, Applicability and Surveillance Requirements and STS guidelines to develop the Action provisions. Reactivity anomalies of concern are while operating in operational Modes 1 or 2 and if the limit of it delta k/k is exceeded. Present and proposed provisions retain these operational Modes and reactivity difference limitation. Proposed Actions are taken from STS guidelines and allow 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to perform an analysis to determine and explain the cause of the reactivity difference or the plant must be in at least Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The proposed AOT is a reasonable time frame to evaluate core conditions before requiring plant shutdown steps. Present Surveillance Requirements are retained but are rewritten using STS guidelines. The proposed changes adopt a proven A0T of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from later operating plants and with other present requirements retained, the proposed changes maintain necessary restrictions on reactivity anomalies. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated,

o. Proposed Changes for the Economic Generation Control System Requirements Proposed Specifications for the Economic Generation

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Control System, 3.3.0/4.3.0, are based on present provisions rewritten using an STS format. Since the present level of operability for the EGC System is retained by the proposed changes, the changes do not involve a significant increase in the probability or consoquences of an accident previously evalua& d.

2) Create the possibility of a new or different kind of accident from any previously evaluated because:

The proposed changes for Quad Cities Technical Specification Section 3.3/4.3 are based on present provisions and STS guidelines or later operating BWR plants' NRC accepted changes. These proposed changes have been reviewed for acceptability at the Quad Cities Nuclear Station considering similarity of system or component design versus the STS or later operating BWRs. No new modes of operation are introduced by the proposed changes, considering the acceptable operational Modes in present specifications, the STS, or later operating BWRs. The proposed changes do not modify existing setpoints or design assumptions for system or component operation. Surveillance requirements are changed to reflect improvements in technique, frequency of performance or operating experience at later plants. Proposed changes to Action statenents in many places add requirements that are not in the present technical specifications or adopt require.,ents that have been used successfull designs similar to Quad Cities.y atThe other operating proposed BWRs changes with either maintain at least the present level of operability or adopt relaxations to present requirements which still provide a proven acceptable level of operability. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

The proposed changes to Technical Specification Section

3. 3/4. 3 impleraent present requirements, the intent of present requirements, or provisions that have been found acceptable for use on other operating BWRs with system designs similar to that at Quad Cities. The proposed changes are intended to improve readability, usability, and the understanding of technical specification requirements while maintaining acceptable levels of safe operation. The proposed changes have boen evaluated and found to be acceptable for use at Quad Cities based on system design, safety analysis requirements and operational performance. Since the proposed changes are based on NRC accepted provisions at other operating plants that are applicable at Quad Cities and maintain necessary levels of system, component or parameter operability, the proposed changes do not involve a significant reduction in the margin of safety.

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N ENVIRO *91ENIAL ABRaplEET EYAkU21TJ211 PROPOSED SPECIFICATION SECTION 3.3/4.3 REACTIVITY CONTROLS Commonwealth Edison has evaluated the proposed amendment in accordance with the requirements of 10 CFR 51.21 and has determined that the amendment meets the requirements for categorical exclusion as specified by 10 CFR 51.22 (c) (9) .

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

The proposed amendment does not modify the existing facility and does not involve any new operation of the plant.

As such, the proposed amendment does not involve any change in the type or significant increases in effluents, or increase individual or cumulative occupational radiation exposure. The proposed amendment to Section 3.3/4.3,

" Reactivity Controls" contains cdministrative changes such as including appropriate applicability statements within the specifications to define the applicability during operating modo and the required actions to be implemented in the event that specification cannot be met. The information is consistent with the Standard Technical Specifications or later operating plants. In addition some existing requirements have been updated and new requirements added to reflect the Standard Technical Specifications or later operating plant requirements.

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e QC-1 / QC-2 DIFFERENCES TS 3.3/4.3

' REACTIVITY CONTROL' o

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COMPARISON OF UNIT 1 AND UNIT 2 TECHNICAL DPECIFICATIONS FOR THE IDENTIFICATION OF TECHNICAL DIFFERENCES SECTION 3.3/4.3 REACTIVITY CONTROL Commonwealth Edison has conducted a comparison review of the Unit 1 and Unit 2 Technical Specifications to identify any technical differences in support of combining the Technical Specifications into one document. The intent of the review was not to identify any differences in presentation style (e.g. table formats, use of capital letters, etc.) or punctuation but rather to identify areas which the Technical Specifications are technically different.

The review of Section 3.3/4.3, " Reactivity Controls" did not reveal any technical differences. Several administrative differences were identified as follows:

Pace 3.3/4.3-2 3.3.A.2.b Unit 1: inoperable controls rods Unit 2: inoperable control rods Pace 3.3/4.3-5 4.3.B.4 Unit 1: ... verify that at least two source Unit 2: ... verify that a least two source Pace 3.3/4.3-6 3.3.C.2 Unit 1: ...for 90% of any operable control rod...

Unit 2: ...for 90% insertion of any operable control rods...

Pace 3.3/4.3-9 Paragraph A.1, Unit 1: ...the strongest control rod fully Paragraph 1 withdrawn and all oth1rs fully...

Unit 2: ...the strongest control rod fully withdrawn and all other fully... t Paragraph A.1 Unit 1: ... withdrawal of an adjacent rod to a Paragraph 3 position calculated to insert...

Unit 2: ... withdrawal of an adjacent rod to a position calculated to inset...

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E a g #. 3 . 3 / 4.. 3 - 1.0 Paragraph 1 Unit 1: ... rods permitted to be inoperable Unit 2: ... rods permitted to the inoperable Paragraph 2 Unit 1: ...have occurred in the collet housing Unit 2: ...have occurred in the collect housing Engg 3.3/4.3-12 -

Paragraph 2 Unic 1: a. An interassembly local peaking factor.

Unit 2: a. In interascetably local peaking factor (Reference 6).

Paragraph 2 Unit 1: f. The design accident and scram reactivity shape function.

Unit 2: 1. The design accident and scram reactivity shape function, and 4

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ML20082V212

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