Text

Proposed Tech Specs Allowing Performance of CRDM & Ni Instrumentation Replacement to Not Be Considered Core Alterations During Operational Condition 5,Refueling,while Fuel Is in Reactor Vessel
	ML20210R358
Person / Time
Site:	LaSalle
Issue date:	08/13/1999
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20210R354	List: ML20210R350; ML20210R358;
References
	NUDOCS 9908170092
	Download: ML20210R358 (18)
	v • d • e

r ATTACHMENT B Proposed Changes to Technical Specifications for LaSalle County Station, Units 1 and 2 MARKED-UP TS PAGES FOR PROPOSED CHANGES REVISED PAGES NPF-11 NPF-18 1-2 1-2 3/4 1-1 3/4 1 3/4 1-6 3/4 1-6 3/434 3/434 3/4 9-3 3/4 9-3 3/4 9-7 3/4 9-7 l

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9908170092 990813 PDR ADOCK 05000373 P PDR ,

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DEFINITIONS CCRE ALTERATION 1.7 hET h ORE ALTERATION shall be the addition, removal, relocation or movement ofm A fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of Lthe movement of a taxat ta a safe amervative position.

CORE OPERATIS LIMITS REPORT

1. 8 The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.6.A.6. Plant operation within these operating limits is addressed in individual specifications.

CRITICAL POWER RATIO 1.9 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the I

assembly which is calculated by application of the GEXL correlation to cause some point in the assembly to experience boiling transition, divided.by the actual assembly operating power.

l DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131, I microcuries/ gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132,1-133,1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E-AVERI.GE DISINTEGRATION ENERGY 1.11 E shall be the average, weighted in proportion to the concentration of I

each radionuclide in the reactor coolant at the time of sampling, of the i sum of the average beta and gamma energies per disintegration, in MeV, i for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. i EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.12interval The EMERGENCY C0RE COOLING SYSTEM (ECCS) RESPONSEl TIME from when the monitored parameter exceeds its ECCS actuatien setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their requ? red positions, pump dischargt pressures reach their required values, etc. l Times shall include diesel generator starting and sequence loading deiays where applicable.

The response time may be measured by any series of -

sequential, is measured.

overlapping or total steps such that the entire. response time

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.13 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE l that time interval to energization of the recirculation pump circuit LA SALLE UNIT 1 1-2 Amendment No. 70

ATTACHMENT H Proposed Changes to Technical Specifications for LaSalle County Station, Units 1 and 2 Insert A CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);

b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

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3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.1.1 The SHUTDOWN MARGIN shall be equal to or greatet than:

a. 0.385 delta k/k with the highest worth rod analytically determined, or

b. 0.28K delta k/k with the highest worth rod determined by test.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5.

ACTION: ,

j With the SHUTDOWN MARGIN less than specified:

a. In OPERATIONAL CONDITION 1 or 2 reestablish the required SHUTDOWN MARGIN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-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 .

b. In OPERATIONAL CONDITION 3 or 4 immediately verify all insertable control rods to be inserted and suspend all activities that could reduce the SHUTDOWN MARGIN. In OPERATIONAL CONDITION 4 ,

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

c. In OPERATIONAL CONDITION 5, suspend CORE ALTERATION nd 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 CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.1 The SHUTDOWN MARGIN shall be detemined to be equal to or greater than specified at any time during the fuel cycle:

a. By measurement, prior to or during the first startup after each refueling,

b. By measurement, within 500 MWD /T prior to the core average exposure at which the predicted SHUTDOWN MARGIN, including uncertainties and calculation biases, is equal to the specified limit. 5

c. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after detection of a withdrawn control rod that is l immovable, as a result of excessive friction or mechanical inter-forence, or is untrippable, except that the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod.

L tacept movement of IIDes, SRMs or special movable detectors LA SALLE - UNIT 1 3/4 1-1 Amendment No. 28

i REACTIVITY CONTROL SYSTEM

. CONTROL ROD MAXIMUM SCRAM INSERTION TIMES j

I LIMITING CONDITION FOR OPERATION i

3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 05, based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

I ACTION:

l With the maximum scram insertion time of one or more control rods exceeding

7.0 seconds

1. Declare the control rod (s) with the slow insertion time inoperable. l and

2. Perform the Surveillance Requirements of Specification 4.1.3.2.c at least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 seconds.

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

SURVEILLANCE REQUIREMENTS I

4.1.3.2 The maximum scram insertion time of the control rods shall be demonstrated through measurement with reactor coolant pressure greater than or equal to 950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators:

a. For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER following CORE ALTERATIONS

or after a reactor shutdown that is greater,than 120 days,

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

specific control rods, and

c. For at least 1M of the control rods, on a rotating basis, at least once per 120 days of operation. ,-

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Except @vement of SRM. IRM or metal movable detectors ajnormalcontrol rod movement.

- c r . o r ! 3/4 1-6 Amendment No. 94 y

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TABLE 3.31-1 (Continued)

.. .- REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION 1 - Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 - Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within one hour.

ACTION 3 - Suspend all operations invohnng CORE ALTERATION insert allinsertable control rods within one hour.

ACTION 4 - Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

. ACTION C - Deleted ACTION 6 - Inibste a reduchon in THERMAL POWER within 15 minutes and reduce THERMAL POWER to less than 25% of RATED THERMAL POWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACTION 7 - Verify all insertable control rods to be inserted within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 8 - Lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 9 - Suspend all operations involving CORE ALTERATIONS.knd insert all insertable control rods and iock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

capt movement of IRM, SRM or special movable detectors, or replacement of PRM stnngs provM ERM instruma% is OPERABLE per S_- '4 o 9 LA SALLE , UNIT 1 ,

3/434 Amendment No.130 1

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RdFUELINGOPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 At least 2 source range monitor * (SRM) channels shall be OPERABLE # and inserted to the normal operating level with:

a. Continuous visual indication in the control room,

b. One of the required SRM detectors located in the quadrant where CORE.

ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant, and

c. The " shorting links" removed from the RPS ciguitry prior to and j during the time any control rod is withdrawn and shutdown margin j demonstrations.

APPLICABILITY: OPERATIONAL-CONDITION 5, unless the following conditions are i met:

a. No more than four (4) fuel assemblies are present in each core quadrant associated with an SRM;

b. While in core, these four fuel assemblies are in locations adjacent to the SRM; and

c. In the case of movable detectors, detector location shall be selected !

such that each group of fuel assemblies is separated by at least two l (2) fuel cell locations from any other fuel assemblies. .

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATION and insert all insertable control rods.

SURVEILLANCE REQUIREMENTS 4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:

a. At least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s: ..

I

1. Performance of a CHANNEL CHECK,

2. Verifying the detectors are inserted to the normal operating I level, and J

3. During CORE AL1ERATIONS, verifying that the detector of an OPERABLE SRM channel is located in the core quadrant where CORE ALTERATIONS are being performed and another is located in an ,

adjacent quadrant.

"TheuseofspecialmovabhedetectorsduringCOREALTERATIONSinplaceof the normal SRM nuclear detectors is permissible as long as these special detectors are ennn rtad en the nnemal SRM circuits.

bxceptmovementofIRM,SRMorspecialmovabledetectors

. The normal or emergency power source may be inoperable. l U Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

LA SALLE - UNIT 1 3/4 9-3 Amendment No. 32

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. . . . - .-. ----.:---. ..a. - - . . .. .a. . .~

REFUELING OPERATIONS ,

3/4.9.5 COMMUNICATIONS ,

LIMITING CONDITION FOR OPERATION 3.9.5~ Direct communication shall be maintained between the control room and refueling platform personnel.

APPLICABILITY: OPERATIONALCONDITION5,duringCOREALTERATIONS.[

ACTION: -

When direct communication between the control room and refueling platform personnel cannot be maintained, immediately suspend CORE ALTERATIONS. %

. SURVEILLANCE REQUIREMENTS 4.3.5 Direct communication between the control room and refueling platform personnel shall be demonstrated within one hour grior to the start of and at

=least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS. "

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1 rtxcepT. mov...n6 vi mcore instrumentation and control rods with their normal I birive syst==

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LA'SALLE - UNIT 1 3/4 9 !

DEFINITIONS

, CORE ALTERATION EF h3 1.7 kALTERADON shall be the addition, removal, relocation or movement of rfuel, sources, incore instruments or reactivity controls within the reactor vessel. pressure vessel Suspension with the of CORE vessel headshall ALTERATIONS removed and fuel not preclude in the of completion movementh nent to a_ safe conservative position.

CORE OPERATING LIMITS REPORT 1.8 The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cple in accordance with Specification 6.6.A.6. Plant operation -

within these operating limits is addressed in individual specifications, gitTICAL POWER RATIO 1.9 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the approved CPR \

i l

correlation to cause some point in the assembly to experience boiling -

transition, divided by ,the actual assembly operating power.

DOSE EQUIVALENT I-131 1.10 D0SE EQUIVALENT I-131 shall be that concentration of I-131, microcuries/ gram, which alone would produce the same thyroid dose as the l quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

f-AVERAGE DISINTEGRATION ENERGY 1.11 i shall be the average, weighted in proportion to the concentration of ^

each radionuclide in the reactor coolant at the time of sampling, of the :

sum of the average beta and gamma energies per disintegration, in MeV, for !

isotopes, with half lives greater than 15 minutes, making up at least 95% i of the total non-iodine activity in the coolant. !

EMERRENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.12 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required ,

positions, pump discharge pressures reach their required values, etc. .

l Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

END-OF-CYCLF RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.13 The DD-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to energization of the recirculation pump circuit LA SALLE - UNIT 2 1-2 Amendment No.101

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ATTACHMENT B Proposed Changes to Technical Specifications for LaSalle County Station, Units 1 and 2 Insert A l

l CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be

! CORE ALTERATIONS:

l l a. Movement of source range monitors, local power range monitors, l intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);

l b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

l Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

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3/4.I REACTIVITY CONTROL SYSTEMS ,

s. 3/4.1.1 SHUTDOW MARGIN ,

LIMITINE C0W m 0N FOR OPERATION 3.1.1 The SHUT 00WN MARGIN shall be equal to or greater than:

a. 0.35 delta k/k with the highest worth rod analytically detemined, or -

b. 0.235 delta k/k with the highest worth rod detemined by test.

APPLICAg!LITY: OPERATIONAL C0W m 0NS 1, 2, 3, 4, and 5. 1 AEIl9!!!.

With the SWTDOW MARGIN less 'than specified: ,

s. In OPERATIONAL COWmolt 1.er 2 reestablish the required SHUTDOW M4 REIN within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months er be in at least NOT SHUT 00W within the next-12 hours. .

6. In OPERATIONAL COWm0N 3 or 4, lamediately verify all insertable control rods to be inserted and suspend all activities that -

could mduce the SWTDOW MAREIN. In OPERATIONAL Comm0N 4, .

establish SEcom4RY CONTAll0Gli INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. In OPERATIONAL COWm0N 5, suspend CORE ALTERA and other activities that sould reduce the SHUTDOW MANEIN, and insert all insertable control rods witMa 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Establish SECOWAirl CONTAll0ENT INTEERITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREBENTS a*

4.1.1 The SWTDOW M4 REIN shall be detamined to be equal to or greater than specified at asy ties during the fuel cycle:

. a. By measurement, prior to or during the first startup after each.

mfueling.

h. W asesumment, witMn 500 IRSR prior to the core average exposure at which the predicted SWTDOW M4NEIN, including uncertainties and

.Jealemistion biases, is equal to the specified limit.

c.i1EftEa 12.hesirs after detection of a withdrawn control red that is

. ~ insevable, as a result of excessive friction er eschanical inter-

.~

forence, or is untrippable, except that the above required SHUTOD W

,M4 MIN shall*be verified acceptable with an increased allowance for the withdrom worth of the imeevable or untrippeble centrol rod.

o .

fExcept movement of Imr, SEs or special movable detectojrs i

LA SALLE - UNIT 2 3/4 1-1 .

Amendment No. 53

REACTIVITY CONTROL SYSTEM CONTROL ROD MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 05, based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

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

7.0 seconds

1. Declare the control rod (s) with the slow insertion time inoperable, and

2. Perform the Surveillance Requirements of Specification 4.1.3.2.c at least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 seconds.

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

SURVEILLANCE REOUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demonstrated through measurement with reactor coolant pressure greater than or equal to 950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators: i

a. For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER following CORE ALTERATIONS

or after a reactor shutdown that is greater than 120 days,

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

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

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'Except1 movement of SRM. IRM or special movable detectors j@) normal control rod movement.

LA SALLE - UNIT 2 3/4 1-6 Amendment No. 78 e

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 -

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

ACTION 2 -

Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position.

within I hour.

ACTION 3 -

Suspend all operations involving CORE ALTERATION insert all insertable control rods within one hour.

ACTION 4 -

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

ACTION 5 -

DELETED ACTION 6 -

Initiate a reduction in THERMAL POWER within 15 [ninutes and reduce THERMAL ~ POWER to less than 25% of RATED THERMAL POWER.

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . "

ACTION 7 -

Verify all insertable control rods to be inserted within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 8 -

Lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 9 -

Suspend all operations involving CORE ALTERATIONS. nd insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . -

3 I *Except movement of IRM. SRM. or special movable detectors, or replacement I I

l of LPRM strings provided SRM instrumentation is OPERABLE per

@cification 3.9.2. j -

LA SALLE - UNIT 2- 3/4 3-4 Amendment Nd.114

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RfFUELING OPERATIONS 3/4.9.2 INSTRUNENTATION LINITING CONDITION FOR OPERATION 3.9.2 At least 2 source range monitor **(SRM) channels shall be OPERABLE # and inserted to the normal operating level with:

a. Continuous visual indication in the control room,

b. One of the required SM detectis located in the quadrant where CORE ALTERATIONS are being performed and the other required SM detector located in an adjacent quadrant, and

c. The " shorting links" removed from the RPS ci g uitry prior to and during the time any control rod is withdrawn and shutdown margin demonstrations.

,APPLICA81LITY: OPERATIONAL. CONDITION 5. unless the following conditions are '

a. No more than four (4) fuel assemblies'are present in each core .

guadrant associated with an S M; ,

b. While in core, these four fuel assemblies are in locations adjacent :

to the SM; and l

c. In the case of movable detectors, detector location shall be selected ,

such that each group of fuel assemblies is separated by at least two l

.. (2) fuel cell' locations from any other fuel assemblies.

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.Agjg: ,

With the requirements of the above specificati satisfied, immediately suspend all operations involving CORE ALTERATI and insert all insertable REQUIREPENTS l

4.9.2 Each of the above required SM channels shall be demonstrated OPERABLE by: j o a. 'At least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s: .

1 I

1. Performance of a CHANNEL CHECK, 2c Verifying the detectors are inserted to the normal operating level, and O 3. During CORE ALTERATIONS, verifying that the detector of an

- - OPERAtkE SM channel is located in the core quadrant where CORE ALTERATIONS are being performed and another is located in an adjacent quadrant. ,

"The use of special movable detectors during CORE ALTERATIONS in place of the noresi SM nuclear detectors is permissible as long as these special em+=-tars are connected to the normal 52 circuits. -

@Except movement of IM, SM or special movable detectors 7 '

'The notus) or amergency power source may be inoperable.

"Not required for control rods

removed per Specification 3.9.10.1 or 3.9.10.2.

LA SALLE - UNIT 2 " ' 3/4 9-3 Amendment No. ,1R

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REFUELING OPERATI0ft$

3/4.9,5 C0091UNICATIONS l Luuuni MITION POR OPERATION

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3.9.5 Direct communication ohell be meintained betaseen the centrol rees and refueling platfore pereennel.

A1PLICAELITY: OPDATIONAL CONDIT10ft 5, during ColtE ALTERATIONS.D M: 9 then direct camousication betamen the control reen and refueling platform ,

pereennel connet be esistained, immediately suspend CollE ALTBtATI0fts.M l

s SURVEft 8 M m'IRElerTT -

4.9.5 Direct communication betmoen the centrol roes and refueling platfore personnel shell be demonstrated within one hour prior to the start of and at

. least. ones per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTBtATI0fts.A' .

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". e- ; _.. Instr o entation and centrol rede with their norma ve system.

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e LA 5ALLE - LNtIT 1 3/4 9-7 ,

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ATTACHMENT C I Proposed Changes to Technical Specification For LaSalle County Station, Units I and 2 1of2 1

I INFORMATION SUPPORTING A FINDING OF l

NO SIGNIFICANT HAZARDS CONSIDERATION

))

Comed has evaluated the proposed changes and determined that it does not involve a significant hazards consideration. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

Involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated; Create the possibility of a new or different kind of accident from any previously analyzed; or Involve a significant reduction in a margin of safety.

Comed proposes to revise Definition 1.7," Core Alteration," and TS sections 3/4.1,3/4.3, and 3/4.9, to be consistent with the requirements provided in NUREG-1433, Revision 1

" Standard Technical Specifications, General Electric Plants, BWR/4."

The determination that the criteria set forth in 10 CFR 50.92 is met for this amendment request is indicated below.

Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed changes incorporate a definition contained in NUREG-1433 Revision 1,

" Standard Technical Specifications, General Electric Plants, BWR/4." There are no modifications to plant equipment or systems and there is no direct effect on plant operation. The proposed changes do not affect any accident initiators or precursors and do not change or alter the design assumptions for systems or components used to mitigate the consequences of an accident. The proposed changes do not affect the design or operation of any system, structure, or component in the plant. The proposed changes do not impact the requirements for refueling evolutions associated with shutdown margin, core monitoring, and reactor protection system operability. There are no changes to parameters goveming plant operation, and no new or different types of equipment will be installed. These changes do not impact any accident previously evaluated in the Updated

ATTACilMENT C Proposed Changes to Technical Specification For LaSalle County Station, Units I and 2 2 of 2 INFORMATION SUPPORTING A FINDING OF NO SIGNIFICANT HAZARDS CONSIDERATION i

Final Safety Analysis Report (UFSAR). Therefore, no increases in the probability of an accident or consequences will result due to this change. j Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes do not affect the design or operation of any plant system, structure, or component. There are no changes to parameters governing plant operation, and no new or different type of equipment will be installed. There is no change in any method by which a safety related system performs its function. No new equipment is being l introduced, and installed equipment is not being operated in a new or different manner. '

There are no setpoints affected by this proposed action. This proposed action will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. As such, no new failure modes are being introduced.

There are no changes to assumptions in accident analysis. Therefore, the proposed I changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Do the proposed changes involve a.significant reduction in a margin of safety?

The proposed changes are consistent with NUREG-1433, Revision 1 " Standard Technical Speci0 cations, General Electric Plants, BWR/4." The proposed changes do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. The initial conditions and methodologies used in the accident analyses remain unchanged. Therefore, accident analyses results are not impacted. There are no resulting effects on plant safety parameters or setpoints. The proposal does not involve a signi0 cant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings, or a tignincant relaxation of the bases for the limiting conditions for operations. Therefore, thes proposed changes do not cause a reduction in the margin of safety.

Therefore, based upon the above evaluation, Comed has concluded that these changes involve no signi0 cant hazards consideration.

t

l ATTACHMENT D I

Pr: posed Ch3cige to Technic:l Specific: tion For LaSalle County Station, Units 1 and 2

, 1 of1 INFORMATION SUPPORTING AN ENVIRONMENTAL ASSESSMENT Comed has evaluated the proposed changes against the criteria for identincation oflicensing and regulatory actions :equiring environmental assessment in accordance with 10 CFR 51.21.

Comed has determ%ed that this proposed license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9) and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92(b). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as denned in 10 CFR 20, or that changes n inspection or a surveillance requirement, and the amendment meets the following speciGc criteria:

(i) The proposed change involves no significant hazards consideration.

As demonstrated in Attachment C, the proposed changes do not involve a significant hazards consideration. 1 1

(ii) There is no significant change in the types or significant increase in the amounts of :

any effluent that may be released offsite.

l There will be no change in the types or significant increase in the amounts of any effluents released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

)

The proposed changes will not result in changes in the operation or configuration of the facility. There will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels within the plant.

Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from the proposed changes.

.

ML20210R358

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