Text

Proposed Tech Spec Changes for Clarification,To Correct Errors,To Achieve Consistency,Or for Editorial or Administrative Changes
	ML20153C207
Person / Time
Site:	Limerick
Issue date:	02/13/1986
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20153C202	List: ML20153C198; ML20153C203; ML20153C207;
References
	NUDOCS 8602190036
	Download: ML20153C207 (28)
	v • d • e

-

PLANT SYSTEMS j

. SPRAY AND/OR SPRINKLER SYSTEMS LIMITING CONDITION ' FOR OPERATION

-3.7.6.2 The following spray and sprinkler systems-shall be OPERABLE:

Fire Zone Description Reactor Enclosure Hatchway Water Curtains:

1.

EL 253' 2.

EL 283' 3.

EL 313' Fire Area Separation Water Curtains:

48A 1.

Area 602, EL 313' 45A 2.

Area 402, EL 253' 44 3.

Area 304, EL 217' (2 curtains) 22 Cable Spreading Room, Room 450, EL 254' 27 Control Structure Fan Room, EL 304' 27 CREFAS System Filters, EL 304' 28B SGTS Filters, Compartment 624 and SGTS Access Area 625, EL 332' 33 RCIC Pump Room, Room 108, EL 177' 34 HPCI. Pump Room, Room 109, EL 177 '

41 RECW Area, EL 201' l

42A Safeguard System Access Area 200, EL 201' g

44 Safeguard System Access Area 304, EL 217' (Partial) (2 systems) 45A CRD Hydraulic Equipment Area 402, Reactor Enclosure, j

EL 253' (Partial) 45B Neutron Monitoring System Area 406, EL 253' (Partial)-

47A General Equipment Area 500 and Corridor 506, Reactor Enclosure, EL 283' (Partial) 51A & B Reactor Enclosure Recirculation System Filters, EL 331' 79,80,81,82 Diesel Generator Cells (4 cells),

APPLICABILITY:

Whenever equipment protected by the spray and/or sprinkler systems is required to be OPERABLE.

ACTION:

a.

With one or more of the above required spray and/or sprinkler j

systems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous fire r

watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.

b.

The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

O LIMERICK - UNIT 1 3/4 7-22

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.4 The'following programs shall be established, implemented, and maintained:

a.

Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of 4

systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

The systems include the core spray, high pressure coolant injection, I

reactor core isolation cooling, residual heat removal, post-accident sampling system, safeguard piping fill system, control rod drive scram discharge system, and containment air monitor systems.

The program shall include the following:

1.

Preventive maintenance and periodic visual l

inspection requirements, and I

l 2.

' Integrated leak test requirements for each system at refueling cycle intervals or less.

b.

In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

This program shall include the following:

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

'l c.

Post-accident Sampling f

A program which will ensure the capability to obtain and j

analyze reactor coolant, radioactive iodines and particulates in. plant gaseous effluents, and containment e

atmosphere samples under accident conditions.

The program shall include the following:

1.

Training of personnel, 2.

Procedures for sampling and analysis, and 3.

Provisions for maintenance of sampling and analysis equipment.

I LIMERICK - UNIT 1 6-14

va TAB M 3.3.7.5-1

[

ACCIDINF KNTIDRING INSIMMNTATION 2y MINIM N APPLICABLE g

mw1RsD mura omas OezwrIaar, x

INSTRUMENT OF OmNNEIS OPERABLE OCNDITIONS ACTICN I

1.

Reactor Vessel Pressure 2

1 1,2 80 y

2.

Reactor Vessel Mter Ievel 2

1 1,2 80 3.

92ppression Owrber Wter Ievel 2

1 1,2 80 4.

Suppressicn Ownber Water Tenperature 8,6 locations 6,

1,2 80 1/ location 5.

Suppressicn 01anber Air Tenperature 1

1 1,2 80 6.

Drywell Pressure 2

1 1,2 80 w

7.

Drywell Air W rature 1

1 1,2 80 vi 8.

Drywell Oxygen Cbncentration Analyzer 2

1 1,2 80 9.

Drywell Hydrogen Concentraticn Analyzer 2

1 1,2 80 10.

Safety / Relief Valve Position Indicators 1/ valve 1/ valve 1,2 80

11. Primary Cbntainwnt A,at-IDCA Radiation Mxtitors 4

2 1,2,3 81

12. North Stack Wide Range Accident Monitor **

3*

1,2,3 81

13. Neutrcn Flux 2

1 1,2 80 i'

9 i

._;_ 7

'mBLE 3.3.2-2 (Continued)

C ISCUGTON AC'RRTION INSTRMEN'DGTOi SE'IPOINIS E

E Au m sIm S

'JRIP FUNCTICN

'IRIP SESPOINT VAHE i

g 6.

PRIMARY CON'DW@ENT ISOIATIOi n

a.

Reactor Vessel Water Level

~

1.

Low, Iae-Level 2

> -38 inches *.

[-136 inches

> -45 inches 2.

Low, Iow, Low, Level 1

[-129 inches

b.

Drywell Pressure - High 3 1.68 psig

$ 1.88 psig c.

North Stack Effluent Radiation - High j 2.1 tCi/cc

$ 4.0 uCi/cc d.

Refueling Area Ventilaticn Exhaust Duct - Radiation - High 5 2.0 mR/h

< 2.2 miVh w

e.

Reactor Enclosure Ventilaticn Exhaust h

Duct - Radiaticn - High '

f 1.35 mR/h j 1.5 mEVh f.

Outside Atuosphere 'Ib Reactor Enclosure A Pressure - Low

> 0.1 inches of HE

> 0.0 inches of H2O g.

Outside Atnosphere 'Ib Refueling Area A Pressure - Low

> 0.1 inches of HE

> 0.0 incies of H2O h.

Drywell Pressure - HigV f 1.68 psig/

$ 1.88 psig/

Reactor Pressure - Low

> 455 psig (decreasing)

> 435 psig (decreasing) 1.

Primary Ccntainment Instrument

> 2.0 psig

> 1.9 psig Gas to Drywell A Pressure-Iow j.

Manual Initiaticn N.A.

~ N.A.

i

~

_i l...'__.

i ~. ---

-- ~.-

--~ ~-

~

i TABIE 3.3.2-2 (Continued) e H

ISMATICN ACIUATION INErrRLMENTATION SEIPOINFS k

i AIIDNABLE Q

TRIP FtNCTICN TRIP SE*1POINF VAUE 7.

SEINDARY CDNTAINMENT ISGATICN C

2:

H a.

Reactor Vessel Water Ievel -

i 8

Iow, Iow-Ievel 2

__ -38 inches

__ -45 inches g

b.

Drywell Pressure - High

< 1.68 psig

< 1.88 psig c.

Refueling Area Ventilation Exhaust Ibet Radiation - High

< 2.0 mR/h

< 2.2 mR/h w

N d.

Reactor Ehclosure Ventilation Exhaust Ibet Ihdiation - High

< 1.35 mR/h

__ 1.5 mR/h g

1w e.

Outside Atmosphere 'Ib Reactor Enclosure P

A ressure - Iow

> 0.1 inches of H2O

> 0.0 inches of H2O f.

Outside Abnosphere to Refueling Area a Pressure - Iow

> O.1 inches of H2O

> 0.0 inches of H2O g.

Manual Initiation N.A.

N. A.

See Bases Figure B.3/4 3-1.

ja'

~

- 'Ihe low setpoints are for the BWCU Heat Exchanger Ibczns; the high setpoints are for the punp rooms.

j

'mBLE 3.3.2-3 I

ISMATICN SYS'ID1 INSTRLNENTATICN RESPCNSE TDE f

k 4

l TRIP KNCTICH RESPCNSE TDE (Seconds)#

l 1.

MAIN SEAM LDE ISCIATICN i

a.

Reactor Vessel Water I4 vel

1) Iow, Iow - Imvel 2

< 13(a)**

l

2) Iow, Iow, Iow - Invel 1 11.0*/113(a)**

g b.

Main Steam Line Radiation - Hicjh(b) 1 1.0*/1 13(a)**

l c.

Main Steam Line l

Pressure - Iow 1 1.0*/1 13(a)**

d.

Main Steam Line j

Flow - High 1 0.5*/1 13(a)**

e.

Condenser Vactun - Iow N.A.

i f.

Main Steam Line Tunnel Temperature - High N.A.

i g.

Turbine Enclosure - Main Steam Line Tunnel Temperature - High N.A.

h.

Manual Initiation N.A.

.i 2.

RHR SYSTDi SHUPDOWN CIXLING POIE ISEATICH a.

Reactor Vessel Water Imvel Iow - Ievel 3 1 13(a) b.

Reactor Vessel (RHR Cut-In Permissive) Pressure - High N.A.

c.

Manual Initiation N.A.

3.

REACIOR WAER CIEANJP SYS'IB4 ISEATICN ll l.

a.

PHCS A Flow - High 113f #

i l!

l b.

1NCS Area Temperature - High N.A.

c.

INCS Area Ventilation fTemperature-High N.A.

d.

SICS Initiation N.A.

e.

Reactor Vessel Water Imvel -

Iow, Iow-Imvel 2

< 13(a) 4 f.

Manual Initiation N.A.

LDERICK - INIT 1 3/4 3-23 l

t

._.m

. - -. - - - =. -

f' REACTOR COOLANT SYSTEM

' SURVEILLANCE REQUIREMENTS (Continu2d)

S 4.4.6.1.2 The reactor coolant system temperature and pressure 3

shall be determined to be to the right of the criticality limit I

line of Figure 3.4.6.1-1 curves C and C' within 15 minutes prior lj to the withdrawal of control rods to bring the reactor to criticality and at least once per 30 minutes during system f]

heatup.

4.4.6.1.3 The reactor vessel material surveillance specimens L

shall be removed and examined, to determine changes in reactor pressure vessel material properties, as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table 4.4.6.1.3-1.

The results -of these examinations shall be used to update the curves of Figure 3.4.6.1-1.

4.4.6.1.4 The reactor flux wire specimens shall be removed at ti the first refueling outage and examined to determine reactor pressure vessel fluence as a function of time and power level and ff used to modify Figure B 3/4 4.6-1.

The results of these fluence a

determinations shall be used to adjust the curves of Figure ll 3.4.6.1-1, as required.

1:

4.4.6.1.5 The reactor vessel flange and head flange temperature j

shall be verified to be greater than or equal to 80 degrees

~*

l Fahrenheit I!

a.

In OPERATIONAL CONDITION 4 when reactor coolant system l

ll temperature ist i

l.

< 100 degrees Fahrenheit, at least once per 12 i;

hours.

2.

< 90 degrees Fahrenheit, at least once per 30 minutes.

il l{

b.

Within 30 minutes prior to and at least once per 30

!i minutes during tensioning of the reactor vessel head I

bolting studs.

b it i:

i d

u.

i ff LIMERICK - UNIT 1 3/4 4-19 l

...-,.-__.---_:L*.7,-....

]

REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES l

LIMITING CONDITION FOR OPERATION l

3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shall be OPERABLE with closing times greater than or equal to 3 and less than or equal to,5 seconds.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

i ACTION:

a.

With one or more MSIVs inoperable:

1.

Maintain at least one MSIV OPERABLE in each af fected main steam line that is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either:

a)

Restore the inoperable valve (s) to OPERABLE status, or b)

Isolate the affected main steam line by use of a deactivated MSIV in the closed position.

j 2.

Otherwise, be in at least HOT SHUTDOWN within the i

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

t b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

'l 1

1 4.4.7 Eac'i of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 and 5 seconds when tested pursuant to Specification 4.0.5.

i i

LIMERICK - UNIT 1 3/4 4-23

i EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.1 The emergency core cooling systems shall be demonstrated OPERABLE by:-

a.

At least once per 31 days:

1.

For the CSS,.the LPCI system, and the HPCI systems a)

Verifying by venting at the liigh point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.

b)

Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct

position.

i 2..

For the LPCI system, verifying that both LPCI system subsystem cross-tie valves (HV-51-182 A, B) are closed with power removed from the valve operators.

3.

For the HPCI system, verifying that the HPCI pump flow controller is in the correct position.

4.

For the CSS and LPCI system, performance of a CHANNEL FUNCTIONAL TEST of the injection header AP instrumentation.

b.

Verifying that, when tested pursuant to Specification 4.0.5:

1.

Each CSS pump in each subsystem develope a flow of at least 3175 gpm against a test line pressure t

corresponding to a reactor vessel to primary containment differential pressure of 1105 paid plus head and line losses.

2.

Each LPCI pump in each subsystem develops a fibw'of at least 10,000 gpm against a test line pressure corresponding to a reactor vessel to primary containment dif ferential pressure of 120 paid plus head and line losses.

i' 3.

The HPCI pump develops a flow of at least 5600 gpm against a test line pressure which corresponds to a

?

reactor vessel pressure of 1000 psig plus head and line losses when steam is being supplied to the turbine at 1000, +20, -80 psig.**

c.

At least once per 18 months:

1.

For the CSS, the LPCI system, and the NPCI system, performing a system functional test which includes simulated automatic actuation of the system thro'aghout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position.

Actual injection of coolant into the reactor vessel may be excluded from this test.

9 h

Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another l,

mode of operation.

- The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

If operability is not successfully demonstrated within the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, reduce reactor steam dome pressure to less than 200 psig.

LIMERICK - UNIT 1 3/4 5-4

j

/

4 i

EMERGENCY CORE COOLING SYSTEMS I'.

$URVEILLANCE REQUIREMENTS (Continued) i i

\\

2.

For the HPCI system, verifying [that:

a)

The system develops a flod of at least 5600'gpm against a test line pressure corresponding to a reactor vessel pressure of > 200 psig plus head and line losses, When steam is Eoing supplied to the turbine at 200 +15,

-0 psig.**

4 b)

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

3.

Performing a CHANNEL CALIBRATION of the CSS, LPCI, and HPCI system discharge line " keep filled" alarm instrumentation.

4.

Performing a' CHNRNEL CALIBRATION of the CSS header AP

.i instrumentation and verifying the setpoint to be < he allowable value of 4.4 paid.

j 5.

Performing a CHANNEL PALIBRATION of the LPCI header AP jj instrumentation and verifying the setpoint to be < the

];

allowable value of 3.0 psid.

i fi d.

For the ADS:

j 1.

At least once per 31 days, performing a CHANNEL p

FUNCTIONAL TEST of the accumulator backup compr.ess,ed gas system low pressure alarm system.

i-2.

At least once per 18 months:

a)

Performing a system functional ttst Which includes ii simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.

b)

Manually opening each ADS valve When the reactor

)

steam dome pressure is greater than or equal to 100 psig** and observing that either:

9 D 1)

The control valve or bypass valve position i!!

responds accordingly, or

'l 2)

There is a corresponding change in the measured steam flow.

e c)

Performing a CHANNEL CALIBRATION of the accumulator backup compressed gas system low pressure alarm system and verifying an alarm setpoint of 90 + 2 psig on decreasing pressure.

i 4

4

- The provisions of Specification 4.0,4 are not applicable provided the surveillance is perforned within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

If HPCI or ADS operability is not 6

successfully demonstrated within the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, reduce reactor i

steam dome pressure to less than 200 psig or 100 psig, respectively.

j i

LIMERICK - UNIT 1 3/4 5-5

[

4

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..l i

. CONTAINMENT SYSTEMS l..

i.

DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE l

LIMITING CONDITION FOR OPERATION 3.6.1.6 Drywell and suppression chamber internal pressure shall be a

maintained between -1.0 and +2.0 psig.

g APPLICABILITY:

OPERATIONAL CONDITIOAS 1, 2, and 3.

i ACTION:

With the drywell and/or suppression chamber internal pressure outside.

of the specified limits, restore the internal pressure to within the

+

limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within the next 12 i

hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I i

l

'?.

SURVEILLANCE REQUIREMENTS 4.6.1.6 The drywell and suppression chamber internal pressure shall be determined to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 9

d i

al LIMERICK - UNIT 1 3/4 6-9

i -

CONTAINMENT SYSTEMS

' BASES 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL' INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the ij life of the unit.

Structural integrity is required to ensure that the ll containment will withstand the maximym pressure of 44.02 psig in the i,

event of a LOCA.

A visual inspection in conjunction with Type A leakage test a is sufficient to demonstrate this capability.

3/4.6.1.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The limitations of drywell and suppression chamber internal pressure ensure that the containment peak pressure of 44.02 psig does not exceed the design pressure of 55 psig during LOCA conditions or that.the external pressure dif ferential does not exceed the design maximu~m external pressure di f ferential.of 5.0 psid.

The limit of -1.0 to + 2.0 psig for initial containment pressure will limit the total pressure to 44.02 Which is less than the design pressure and is consistent with the safety analysis.

3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures tha't' the containment peak air temperature does not exceed the design temperature of 340 degrees Fahrenheit during steam line break conditio,ns and is consistent with the safety analysis, i!

!!ij 3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM o

The drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required for inerting, deinerting and pressure control.

The 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months j:

per 365 day limit on purge valve operation.is imposed to protect the integrity of the SGTS filters.

Analysis indicates that should a LOCA occur While this pathway is being utilized, the associated pressure surge through the (18 or 24") purge lines will adversely af fect the lI integrity of SGTS.

This limit is not imposed, however, on the subject valves When pressure control is being performed through the 2-inch bypass line, since a pressure surge through this line does not threaten the OPERABILITY of SGTS.

I i

i

j LIMERICK - UNIT 1 B 3/4 6-2 l

L-

a l}

. CONTAINMENT SYSTEMS I-DRYWELL AVERAGE AIk TEMPERATURE LIMITING CONDITION FOR OPERATION t

3.6.1.7 Drywell average air temperature shall not exceed 135 degrees Fahrenheit.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With the drywell average air temperature greater than 135 degrees Fahrenheit, reduce the average air temperature to within the limit 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 and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS The drywell average air temperature shall be the vo hmetric 4.6.1.7 average of the temperatures at the following locations and shall be determined to be within the limit at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

Approximate Number of Elevation Installed Sensors

a.

330' 3

b.

320' 3

l c.

.260' 3-l l

d.

248' 6

l l

At least one reading from each elevation is required for a volumetric l

average calculation.

i t t-l LIMERICK - UNIT 1 3/4 6-10

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION SYSTEMS SUPPRESSION CHAMBER LIMITING CONDITION FOR OPERATION 3.6.2,1 The suppression chamber shall be OPERABLE with:

a.

The pool water:

1.

Volume

between 122,120 cubic feet and 134,600 cubic fe et, equivalent to a level between 22' 0" and 24' 3",

and a 2.

Maximum average temperature of 95 degrees Fahrenheit except that the maximum average temperature may be permitted to increase'to:

a) 105 degrees Fahrenheit during testing which adds heat to the suppression chamber.

b) 110 degrees Fahrenheit with THERMAL POWER less than.

or equal to 1% of RATED THERMAL POWER.

c) 120 degrees Fahrenheit with the main steam line isolation valves closed following a scram.

Drywell-to-suppression chamber by/ 6/E~ design value of 0.0500 b.

pass leakage less than or equal to 10% of the acceptable A square feet.

c.

At least eight suppression pool water temperature instrumentation indicators, one in each of the eight i

locations.

I t

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With the suppression chamber water level outside th5* above

~

a.

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

b.

With the suppression chamber average water temperature greater than 95 degrees Fahrenheit, restore the average

i temperature to less than or equal to 95 degrees Fahrenheit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , except, as permitted above:

1.

With the suppression chamber average water temperature l'

greater than 105 degrees Fahrenheit during testing which adds heat to the suppression chamber, stop all testing which adds heat to the suppression chamber and restore the average temperature to less than 95 degrees Fahrenheit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 and in COLD SHUTDOWN l

within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With the suppression chamber average water temperature greater thans a) 95 degrees Fahrenheit for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and THERMAL POWER greater than 1% of RATED THERMAL ll POWER, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the neyt 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

'l b) 110 degrees Fahrenheit, place the reactor mode switch in the Shutdown position and operate at ll least one residual heat removal loop in the suppression pool cooling mode.

Includes the volume inside the pedestal.

LIMERICK - UNIT 1 3/4 6-12

~

8 CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT REFUELING AREA SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

OPERATIONAL CONDITION *.

ACTION:

l Without REFUELING AREA SECONDARY CONTAINMENT INTEGRITY, suspend handling of irradiated fuel in the secondary containment, CORE

!j ALTERATIONS and operations with a potential for draining the reactor ll vsasel.

The provisions of Specification 3.0.3 are not applicable.

}l SURVEILLANCE REQUIREMENTS t

4.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:

a.

Verifying at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the pressure within the refueling area secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.

b.

Verifying at least once per 31 days that:

l.

All refueling area secondary containment equipmen " '

hatches and blowout panels are closed and sealed.

2.

At least one door in each access to the refueling area

~

secondary containment is closed.

I 3.

All refueling area secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic isolation dampers / valves and required to be closed during accident conditions are

'j closed by valves, blind flanges, or deactivated

}.

automatic dampers / valves secured in position.

lI l

c.

At least once per 18 months:

Operating one standby gas treatment subsystem for one hour and maintaining greater than or equal to 0.25 inch of vacuum l,-

water gauge in the refueling area secondary containment at a flow rate not exceeding 764 cfm.

Rsquired when (1) irradiated fuel is being handled in the refueling crea secondary containment, or (2) during CORE ALTERATIONS, or (3) lp, during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel.

{a l

l LIMERICK - UNIT 1 3/4 6-47

CONTAINMENT SYSTEMS STANDBY GAS TREA'INENT SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3-Two independent standby gas treatment subsystems shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDTIONS 1, 2,

3, and *.

I ACTION:

a.

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

In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

In Operational Condition *, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

The provisions of Specification 3.0.3 are not applicable.

b.

With both standby gas treatment subsystems inoperable in Operational Condition *, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS or operations with a potential for draining the reactor vessel.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS I

4.6.5.3 Each standby gas treatment subsystem shall be demonstrated

{

OPERABLE:

a.

At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal absorbers and verifying that the subsystem operates with the heaters OPERABLE.

Required when (1) irradiated fuel is being handled.in the refueling area secondary containment, or (2) during CORE ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

1

, LIMERICK - UNIT 1 3/4 6-52

~

J l

l TAatt 1.A.1-1 (Continued)

I ti PART A - PRIMART CONTAIMMkMT ISOLATION VALVE $

H h

IM30ARD OUfa0ARS ISOL.

W F EMr.T R ATION rDMCTION ISOLATION ISOLATION MAX 1501.

SIGMALCS).

MOTES FCID H

gggggg ggggggg ggggggg gggg,gy gyy, gy gyy, Oy (SEC)(26)

(20)

P

,l 0283 DRTWELL MS/02 SAAFLE SV57-133 5

3.N.R.S 11 57 C

cg SV57-143 5

3.M.R.S 11 F4 5V57-195 5

3.M.R.S 11 i s 8

g 030s-1 DRTMELL PRESSURE MV42-147A 45 10 42 INSTRUMENTATION 0353 TIF FURet 59-1956(CK)

MA 59 (DOUBLE "0" RING)

MY59-131 7

B.M.S 16 035C-8 TIF DRIVES XY59-141A-E MA 3.M 11.16621 59 LJ (DOUBLE "0" RIMS)

)

XV59-140A-E F9 11.16 c4 037A-D CRD INSERY LINES BALL CHECK MA 12 47

[

MCU MA 1,3 034A-D CED WITMDRAW LINES MCU MA 12 47 3DY YENTS C DRAIMS XV47-1F010 25 3e XV47-triso se se XV47-tr811 25 3e XV47-17101 39 30 039A(5)

DETWELL SPRAT MY51-tr0211(3)

160 4.11 51 MV51-tr016A(B) 169 11 040E DATWELL PRE 55URE MV42-1479 45 to 42 INSTRUMENTATION 040F-2 CONTAIMMENT IMSTRUMENT Mv59-101 45

' C.M.S 5

59 a

GA5-5UCTION MV59-103 7

C.M s l'

e

+

O e

annun o.u.a-A PRIMARY CONTAINMENT ISOLATION VALVES NOTATION NOTES (Continued) 15.

Check valve used instead of flow orifice.

16.

Penetration is sealed by 'a flange with double 0-ring seals.

These seals are leakage rate tested by pressurizing between g

the O-rings. ~ Both the TIP Purge Supply (Penetration 35B)

,l and the TIP Drive Tubes (Penetration 35C-G) are welded to i

their respective flanges.

Leakage through these seals is

.}

included in the Type C leakage, rate total for this penetration.

The ball valves (XV-141A-E) are Type C tested.

i It is not practicable to leak test the shear valves (XV-140A-E) because squib firing is required for closure.

Shear valves (XV-140A-G) are normally open.-

17.

Instrument line isolation provisions consist of an excess flow check valve.

Because the instrument line is connected

,l to a closed cooling water system inside containment, no flow lt orifice is provided.

The excess flow check valves are subject to operability testing, but no Type C test is performed nor required.

The line does not isolate during a LOCA and can leak only if the line or instrument should rupture.

Leaktightness of the line is verified during the integrated leak rate test (Type A test).

18.

In addition to double "O" ring seals, this penetration is tested by pressurizing volume between doors per j

Specification 4.6.1.3.

19.

The RHR system safety pressure relief valves will be exempted from the initial LLRT.

The relief valves in these lines will be exposed to containment pressure during thi ~

initial ILRT and all subsequent ILRTs.

In addition, modifications will be performed at the first refueling to-facilitate local testing or removal and bench testing of the relief valves during subsequent LLRTs.

Those relief valves which are flanged to facilitate removal will be equipped with double O-ring seal assemblies'on the flange closest to primary containment by the end of the first refueling outage.

These seals will be leak rate tested by pressurizing between the 0-rings, and the results added into the Type C total for this penetration.

20.

See Specification 3.3.2, Table 3.3.2-1, for. a description of the PCRVICS isolation signal (s) that initiate closure of each automatic isolation valve.

In addition, the following non-PCRVICS isolation signals also initiate closure of selected valves EA Main steam line high pressure, high steam line leakage flow, low MSIV-LCS dilution air flow LFHP With HPCI pumps running, opens on low flow in associated pipe, closes When flow is above setpoint LFRC With RCIC pump running, opens on low flow in associated pipe, closes When flow is above setpoint LFCH With CSS pump running, opens on low flow in associated pipe, closes When flow is above setpoint LFCC, Steam supply valve fully closed or RCIC turbine stop valve fully closed All power operated isolation valves may be opened or closed remote manually.

J' f

LIMERICK - UNIT 1 3/4 6-42 7T

CONTAINMENT SYSTEMS e

i SURVEILLANCE REQUIREMENTS 4.6.3.1 Each primary containment isolation valve shown in Table 3.6.3-1 shall be demonstrated OPERABLE prior to returning the valve.to service after maintenance,, repair or replacement work,is performed on the valve or its associated actuator, 4

control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.

.l 4.6.3.2 Each primary containment automatic isolation valve shown in

~

j Table 3.6.3-1 shall be demonst rated OPERABLE during COLD

)'

SHUTDOWN or REFUELING at least once per 18 months by verifying that en a containment isolation test signal each automatic isolation valve actuates to its isolation position.

4.6.3.3 The isolation time of each primary containment power. operated or automatic valve shown in Table 3.6.3-1 shall be determined if to be within its limit when tested pursuant to Specification l;

4.0.5.

ll 4.6.3.4 Each instrumentation line excess flow check valve shown in l

Table 3.6.3-1 shall be demonstrated 'PERABLE at least once g

per 18 months by verifying that the v.. ' e checks flow..

4.6.3.5 Each traversing in-core probe sys'.sm explosive isolation valve shall be demonstrated OPERABLE:

a.

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

,l b.

At least once per 18 months by removing the explosive squib from the explosive valve, such that each explosive squib in each explosive valve will be. tested at least once per 90 il months, and initiating the explosive squib.

The replacement lj charge for the exploded squib shall be from the same 1

manufactured batch as the one fired or from another batch which as been certified by having at least one of that batch successfully fired.

No squib shall remain in use beyond the expiration of its shelf-life and/or operating life, as applicable.

i..'

I; i

LIMERICK - UNIT 1 3/4 6-18 a

a

Li j

CONTAINMENT SYSTEMS 3 /4.6. 3 PRIMARY CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The primary containment isolation valves and the l

instrumentation line excess flow check valves shown in Table 3.6.3-1 y

shall be OPERABLE with isolation times less than or equal to those shown in Table 3.6.3-1.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

With one or more of the primary containment isolation valves shown in Table 3.6.3-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1.

Restore the inoperable valve (s) to OPERABLE status, or 2.

Isolate each affected penetration by use of at least one deactivated automatic valve secured in the isolated position,* or t

3.

Isolate each affected penetration by use of at least one t>

j closed manual valve or blind flange.*

9.

4.

The provisions of Specification 3.0.4 are not applicable provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the af fected penetration is i

isolated-in accordance with ACTION a.2 or a.3 above, and provided that the associated system, if applicable, is declared inoperable and the appropriate ACTION statements for that system are performed.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With one or more of the instrumentation line excess flow-check valves shown in Table 3.6.3-1 inoperable, operation may continue and the provisions of Specificttione 3.0.3 and 3.0.4 l

are not applicable provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1.

The inoperable valve is returned to OPERABLE status, or 2.

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

i.

l Otherwise, be in at least HOT SHUTDOWN within the next 12 l

hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ll

'f i

Isolation valves closed to satisfy these requirements may be reopened on an intermittent basis under adminstrative control.

LIMERICK - UNIT 1 3/4 6-17

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

f PLANT SYSTEMS 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM l*

LIMITING CONDITION FOR OPERATION 1,i 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an. OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel.

I APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome

{>

pressure greater than 150 psig.

ACTION:

i With the RCIC system inoperable, operation may continue a.

provided the HPCI system is OPERABLE, restore the RCIC system to OPERABLE status within ' 14 days.

Otherwise, be in at least i

HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

In the event the RCIC system is actuated and injects water

,'I into the reactor coolant system, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

SURVEILLANCE REQUIREMENTS y

4.7.3 The RCIC' system shall be demonstrated CPERABLE:

j i

a.

At least once per 31 days by:

1.

Verifying by venting at the high point vents that the system piping from the pump discharge valve to the i

system isolation valve is filled with water.

4 2.

Verifying that each valve (manual, power-operated, or a utomatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

i' 3.

Verifying that the pump flow controller is in the correct position.

b.

At least once per 92 days by verifying that the RCIC pump develops a flow of greater than or equal to 600 gpm in the test flow path with a system head corresponding to reactor r

vessel operating pressure when steam is being supplied to the l

turbine at 1000 +20, -80 psig.*

T l

fhe provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is

. adequate to perform the test.

If operability is not successfully demonstrated within the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, reduce reactor steam pressure to less than 150 psig.

I l

LIMERICK - UNIT 1 3/4 7-9 l

l..

-..~ -- l - --...-- -.----

?--

~

t PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) c.

At least once per 18 months by:

1.

Performing a system functional test which includes simulated automatic actuation and restart and verifying that each automatic valve in the flow path actuates to its correct position.

Actual injection of coolant into the reactor vessel may be excluded.

2.

Verifying that the system will develop a flow of greater than or equal to 600 gpm in the test flow path when steam is supplied to the turbine at a pressure of 150 +

1 15,

-0 psig.*

t f

3.

Veri fying that the suction for the RCIC system is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank water level-low signal.

4.

Performing a CHANNEL CALIBRATION of the RCIC system discharge line " keep filled" level alarm instrumentation.

l E.

1

The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12. hours after reactor steam pressure is adequate to perform the tests.

If operability is not successfully demonstrated within the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, reduce the i

reactor steam pressure to less than 150 psig.

LIMERICK - UNIT 1 3/4 7-10 i

J-

.=

L:

ii b',

3/4.8

~

ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C.

SOURCES A.C.- foURCES - OPERATING q

LIMITING CONDITION FOR OPERATION l

3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

i a.

Two physically independent circuits between the of fsite transmission network and the onsite Class lE distribution j

system, and b.

Four separate and independent diesel generators, each with:

1.

A separate day tank containing a minimum of 200 gallons it of fuel, lT 2.

A separate fuel storage system containing a minimum of tj 33,500 gallons of fuel, and j

3.

A separate fuel transfer pump.

i APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

4.

ACTION:

a.

With one diesel generator of the above required A.C.

il electrical power sources inoperable, demonstrate the

l OPERABILITY of the remaining A.C.

sources by performing

{

Surveillance Requirements 4.8.1.1.la.and 4.8.1.1.2a.4., for one diesel generator at a time, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 7 days thereaftert restore the inoperable diesel generator to OPERABLE status within 92 days or be intat least HCYP SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With two diesel generators of the above required A.C.

electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.

sources by performing i

Surveillance Requirements 4.8.1.1.la. and 4.8.1.1.2a.4., for L

one diesel generator at a time, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereaf tert restore at least one of the inoperable diesel generators to OPERABLE status within' 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-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 and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Also see l

action e of 3.8.1.1.

p c.

With three diesel generators of the above required A.C.

electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.

sources by performing Surveillance Requirements 4.8.1.1.1.a. and 4.8.1.1.2a.4., for one diesel generator at a time, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafterr restore at leasr one of the inoperable-diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

With one of fsite circuit and one diesel generator of the above required A.C.

electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.

sources by performing Surveillance Requirements 4.8.1.1.la. and 4.8.1.1.2a.4. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months th ereaf t er.

Restore at least two offsite circuits and at d

f LIMERICK - UNIT _1 3/4 8-1

l;'

ADMINISTRATIVE CONTROLS

~

il l'*

ANNUAL REPORTS (Continuad)

Ji dosimeter, thermoluminescent dosimeter (TLD), or film badge me asur ements.

Small exposures totalling less than 20% of the individual total dose need not be accounted for.

In the

]

aggregate, at least 80% of the total whole-body dose received from external sources should be assigned to specific major y

work functions; j

b.

Documentation of all challenges to safety / relief valves; and i.

y c.

Any other unit unique reports required on an annual basis.

i!-

MONTHLY OPERATING REPORTS j

6.9.1.6 Routine reports of operating statistics and shutdown i;

expe rience, including documentation of all challenges to the main steam system safety / relief valves, shall be submitted on a monthly a

basis to the Director, Office of Inspection and Enforcement, U.S.

l

=

Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to j

the Regional Administrator of the Regional Office of the NRC no later than the 15th of each month following the calendar month covered by the report.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

6.9.1.7 Routine Annual Radiological Environmental Operating Reports j

covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The initial report shall be submitted prior to May 1 of the year following initial criticality.

The Annual Radiological Environmental Operating Reports shall include l, i summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison (as appropriate), with preoperational studies, operational controls and previous environmental. surveillance reports and an assessment of the' observed' impacts of the plant operation on the environment.

The reports shall il also include the results of land use censuses required by l

Specification 3.12.2.

The Annual Radiological Environmental Operating Reports shall include the results of all radiological environmental samples and of all is environmental radiation measurements taken during the report period pursuant to the locations specified in the tables and figures in the i

OFFSITE DOSE CAICULATION MANUAL, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Branch Technical Position, Revision 1, November 1979.

In the event that some individual results are not available for

)

inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following:

a summary description of the radiological environmental monitoring program; at least two I

legible maps.**

A single submittal may be made for a multiple unit station.

4 j

- One map shall cover stations near the SITE BOUNDARY; a second H

shall include the more distant stations.

i 1

LIMERICK.- UNIT 1 6-16

.--.-.-- :L==---...- -

f REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITON FOR OPERATION 3.1. 2 - The reactivity equivalence of the dif ference between the actual ROD DENSITY and the predicted ROD DENSITY shall not exceed 1% 6 k/k.

i APPLICABILITY:

OPERATIONAL CONDITION 1 and 2.

., 3 ACTION:

With the reactivity equivalence difference exceeding 1% A k/k:

a.

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; operation may continue if the difference is explained and corrected.

b.

Otherwise, 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 .

SURVEILLANCE REQUIREMENTS

l i

y 4.1.2 The reactivity equivalence of the difference between the'

'I actual ROD DENSITY and the predicted ROD DENSITY shall be verified to be less than or equal to 1% 6 k/k:

a.

During the first startup following CORE ALTERATIONS, and

j b.

At least once per 31 effective full power days during i

POWER OPERATION.

c.

The provisions of Specification 4.0.4 are not applicable.

I L

l!

o I

1' 1

i o

I LIMERICK.- UNIT 1 3/4 1-2 i

e a

TAntt 4.3.1.1-1 (Continued) b F4 REACTOR FROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE RE991REMENTS k

CNANNEL OPERETIONAL pq CNANNEL FUNCTIONAL CNANNEL CONDITICd5 TOR WHICM O

FUNCTIONAL UNIT CHECK TEST CALTnRATION EURVEILL.tE7E RE9UIRED M

0.

Sesam Discharge Volume Water g

Level - Nigh s.

Level Transmitter S

.R R

1, 2.

5(1) l C

b.

Float Switch N.A.

M R

1.

2.

Sti) 1 Z

h 9.

Turbine stop Valve - Closure M.A.

M R

1 F'

10. Turbine control valve Test

?

Closure. Triy 011 l

Pressure - Lou N.A.

M R

1 j

11. Reactor Mode suitch j

Shutdown position M.A.

A N.A.

1 2.

3. 4 5

12. Manual Scram M.A.

M M.A.

1.

2.

3.

4.

5 (a) Neutron detesters may be excluded from CNANNEL CALIRRATION.

LJ (b) The IRM and SRM channels shall be determined to overlay for at least 1/2 decades during each startup j

)[

after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlay j

for et least 1/2 decades during each controlled shutdown. if not performed within the previous 7 days.

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

4 j

a(

(d) This calibration shall sonsist of the adjustment of the APRM channel to conform to the power values i

calculated by a heat balance di Ang OPERATIONAL CONDITION 1 when THERMAL POWER 2 25E of RATED THERMAL 70WER.

Adjust the APRM channel if the absolute difference is greater than IX of RATED 1

THERMAL F0W:R.

Any APRM channel gain adjustment made in compliance with Specification 3.2.2 shall not be i

included in determining the absolute difference.

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

I (f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPN) using the TIP system.

(g) verify measured core flow (total core flow) to be greater than or equal to established core flow at the existing looy flou (APRM N flow).

During the startup test program, data shall be recorded for the patameters listed to provide a basis for establishing the specified relationships. Comparisons of the actual j

data in accordance with the criteria listed shall commence upon the conclusion of the startup test program.

th) Thia function is not required to be OPERABLE when the reactor pressure vessel head is removed per specification 3.10.1.

(i) With any control rod withdrawn. Not ayylicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

ri i

(j) If the RP5 shorting links are required to be removed yer Specification 3.9.2. they may be reinstalled for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance. During this time.

CORE ALTERATION 5 shall be suspended, and no control rod shall be moved from its existing position.

1 1

TABLE 3.3.7.11-1 (Continusd) i,%

ACTION STATEMENTS

,a ACTION 100-With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef fluent releases may continue for up to 14 days provided that prior to initiating a release s a.

At least two independent samples are analyzed in

j accordance with Specification 4.11.1.1.1, and b.

At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive ef fluents via this pathway.

ACTION 101-With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef fluent releases via this pathway may continue for up to 30 days provided that, at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analyzed for gross radioactivity (beta or gamma).

Beta is analyzed at a

.I limit of detection of at least 1N7 microcurie /mL..

Gamma is Analyzed at a limit of detection of at 14ast SN7 microcurie /mL.

j ACTION 102-With the number of channels OPERABLE less than.

required by the Minimum Channels OPERABLE requirement, ef fluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Pump curves generated in situ may be used to estimate flow.

4 ll

't LIMERICK - UNIT 1 3/4 3-100

'i

TABLE 3.3.7.9-1 (Continued)

FIRE DETECTION INSTRWENTATICN y

INSTRWENT IA:ATION WTAL NUmER OF INSTRWENTS FIRE ZCNE STRUCIURE ELEV.

AREA IEAT SEKE FLAME (x/y)

(x/y)

(x/y) 1 j

25 Ccntrol 289' Auxiliary Equipment Focm 542 0/112 57/0 NA (PGCC (Ceilirx3)

Floor) 56/0 (PG:r Floor) 0/13 14/0 (Non-(Non-PGCC PGOC Floor) Floor) 32/0 (Terminal Cabinets) 26 Control 289' Remote Shutdown Panel Area 540 0/4 3/0 NA (Non (Ceiling PGT Level)

Floor) 2/0 (Non-PGCC Floor) 27 Ccntrol 304' Ccntrol Structure 0/23 10/0 NA Fan Rocm 619 4/0 (inside plenum) 28A Ccntrol 332' SGTS Access Area 625 (SGrS 4/0 NA NA Poom Ventilation Exhaust)

(inside plenum) 28B Ccntrol 332' SGrS Filter Ccmpartment 624 4/0 NA NA l

(inside plenum) 28C Ccntrol 332' Ccntrol Room Fresh Air NA 3/0 NA Intake Plenum 31 Unit 1 177' RHR Heat Exchanger &

NA 6/0

.NA Reactcr Pump Rocm 103 (B&D) 32 Unit 1 177' RHR Heat Exchanger &

NA 5/O NA Reactor Pump Roca 102 (A&C) 33 Unit 1 177' RCIC Pump Roca 108 0/3 2/0 NA Reactor 34 Unit 1 177' HPCI Pump Rcom 109 0/4 3/O NA Reactor 35 Unit 1 177'

'A' Core Spray Pump NR 2/0 NA Reactor Rcm 110

!t LIFERICK - LNIT 1 3/4 3-94

ML20153C207

Text

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