Text

Proposed TS 3.3.2.1 to Extend Surveillance Test Intervals & Allowed Outage Times for Ci Actuation Instrumentation
	ML20035F940
Person / Time
Site:	Limerick
Issue date:	04/19/1993
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20035F939	List: ML20035F938; ML20035F940;
References
	NUDOCS 9304230096
	Download: ML20035F940 (26)
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f Attachment.3 LIMERICK GENERATING STATION Docket No. 50-352 50-353 License Nos. NPF-39

)

NPF-85 l

l PROPOSED TECHNICAL SPECIFICATIONS CHANGES l

List of Attached Pages Unit 1 xviii 3/4 3-29 B 3/4 3-1*

xix 3/4 3-30 B 3/4 3-2 3/4 3-9 3/4 3-31 B3/43-3*

3/4 3-16 3/4 3-27 j

3/4 3-28 Unit 2 xviii 3/4 3-29 B 3/4 3-1*

xix 3/4 3-30 B 3/4 3-2

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3/4 3-9 3/4 3-31 B3/43-3*

3/4 3-16 3/4 3-27 3/4 3-28 Page provided for completeness - no hanges have been made to this page.

9304230096 930419 PDR ADOCK 05000352 P

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INDEX BASES PAGE SECTION

' ~ ~

B 3/4 D-1 3/4.0 APPLICABILITY......................................................

3/4.1 REACTIVITY CONTROL SYSTEMS B 3/41-1 3/4.1.1 SHUTDOWN MARGIN...............................................

B 3/4 1-1 3/4.1.2 RE ACT I V I T Y AN0M AL I ES..........................................

B 3/4 1-2 3/4.1.3 CONTROL R0DS..................................................

B 3/4 1-3 3/4.1.4 CONTROL R0D PROGRAM CONTR0LS..................................

B 3/4 1-4 3/4.1.5 ST ANDBY LIQUID CONT ROL SYST EM.................................

3/4.2 POWER DISTRIBUTION LIMITS i

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION B 3/4 2-1 RATE..........................................................

l B 3/4 2-2 l

3/4.2.2 APRM SETP0lNTS................................................

B3/42-3 l

LEFT I NT ENT ION ALLY lB LAN K..................................................

B 3/4 2-4 3/4.2.3 MINIMUM CRITICAL POWER RATI0..................................

B 3/4 2-5 3/4.2.4

! ' NE AR HE AT GEN ERAT ION RAT E...................................

3/4.3 INSTRUMENTATION B 3/4 3-1 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.....................

B3/43-2 3/4.3.2 ISOLAT ION ACTU ATION INSTRUMENTAT ION...........................

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION B 3/4 3-2 I N ST RUMENT AT I ON...............................................

B 3/4 3-3 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.............

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION B 3/4 3-4 I N ST RUM EN T AT I ON...............................................

B 3/4 3-4 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION.............................

3/4.3.7 MONITORING INSTRUMENTATION B 3/4 3-5 Radiation Monitoring Instrumentation..........................

LIMERICK - Unit.}

xviii 9

,-y.-,.

I I

INDEX i

BASES PAGE SECTION d

INSTRUMENTATION (Continued)

B 3/4 3-5 Seismic Monitoring Instrumentation............................

B 3/4 3-5 (Deleted).....................................................

B 3/4 3-5 Remote Shutdown System Instrumentation and Controls...........

B 3/4 3-5 Accident Monitoring Instrumentation...........................

B 3/4 3-5 Source Range Monitors.........................................

B 3/4 3-5 T rav e rsing In-Core P robe System...............................

B 3/4 3-5 Chl orine and Toxi c Gas Detecti on System.......................

B 3/4 3-5 Fi re Detecti on Inst rument ati on................................

B3/43-7 Loose-Part Detection system..................................

B3/43-7 (Deleted).....................................................

B3/43-7 Of f gas Moni tori ng Inst rument ati on.............................

B 3/4 3-7 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM...........................

3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION B 3/4 3-7 I N ST RUMENT AT I ON...............................................

i Bases Figure 8 3/4.3.1 Reactor Vessel Water Level...........................

B 3/4 3-8 3/4.4 REACTOR COOLANT SYSTEM B 3/4 4-1 3/4.4.1 RECI RCU LAT I ON SYST EM.......................................

B 3/4 4-2 3/4.4.2 S AF ETY / R E LI E F V ALU ES.......................................

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE B 3/4 4-3 Le akage Detecti on Systems............................

B 3/4 4-3 Ope rati on al Le ak ag e..................................

B 3/4 4-3 3/4.4.4 CHEMISTRY..................................................

xix LIMERICK - Unit 1 1

DISTRUMENTATION 3/4.3.2. ISOLATION ACTUATION INSTRUMENTATION

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LIMITING CONDITION FOR OPERATION The isolation actuation instrumentation channels shown in Table 3.3.2 OPERABLE with their trip setpoints set consistent with the values shown in the Trip 3.3.2 Setpoint column of Table 3.3.2.-2 and with ISOLATION SYSTEM RESPONSE 3.3.2-3.

APPLICABILITY: As shown in Table 3.3.2-1.

ACTION:

With an isolation actuation instrumentation c.

-I trip setps,at ess ble 3.3.2-a) conservative than the value shown in the t'lowable Values colu:mi u chancel is n estored to 04. 3LE 2, declare the channel inoperable until t, status with its trip setp With the number of OPERABLE channels less than required by the Minimum OP b)

Channels per Trip System requirements for one trip system:

If placing the inoperable channel (s) in the tripped condition would cause an isolation, the inoperable channel (s) shall be restored to OPERABLE status 1.

If this cannot be accomplished, the ACTION required by Table within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3.3.2-1 for the affected trip function shall be taken, or the channel shall be placed in the tripped condition.

or If placing the inoperable channel (s) in tha tripped condition would not caus an isolation, the inoperable channel (s) and/or that trip systein shall 2.

be placed in the tripped condition within:

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common

to RPS Instrumentation.

a) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not common

to RPS Instrumentation.

b)

The provisions of Specification 3.0.4 are not applicable.

Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1.

3/4 3-9 LIMERICK - UNIT 1

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS l

f ACTION 20 Be in at least HOT SHUIDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUIDOWN next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 21 Be in at least STARTUP with the associated isolation valves closed withi hours or 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 S within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l J

ACTION 22 Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATION CONDITION 1 or 2, verify the affected system isolation valvesIn are closed within I hour and declare the affected system inoperable.

ACTION 23 OPERATIONAL CONDITION 3, be in at least COLD SHUIDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ACTION 24 close the affected system isolation valves within the next hour and declare the affected system inoperable 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 .

ACTION 25 Establish SECONDARY CONTAINMENT INTEGRITY with the stand by gas treatment i

system operating within I hour.

ACTION 26 Close the affected system isolation valves within I hour.

TABLE. NOTATIONS J

Required when (1) handling irradiated fuel 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.

May be bypassed under administrative control, with all turbine stop valves closed.

During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.

See Specification 3.6.3. Table 3.6.3-1 for primary containment isolation valves which (a) are actuated by these isolation signals.

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (b) for required surveillance without placing the trip system in the tripped condition provided at least one OPEPABLE channel in the same trip system is monitoring that parameter. Trip functions common to RPS Actuation Instrumentation are In addition, for the HPCI system and RCIC system shown in Table 4.3.2.1-1.

isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for that valve is OPEPABLE, one channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for required surveillance without placing the channel or trip system in the tripped condition.

LIMERICK - UNIT 3/4 3-16

TABLE 4.3.2.11-1

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ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CHANNEL CllANNEL FUNCTIONAL CllANNEL CONDITIONS FOR WHitH

CHECK _

TEST CALIBRAT10f!

SURVEILLANCE REQUIRE TRIP FUNCTION 1.

,, MAIN STEAM LINE ISOLATION S

Q R

1,2,3 Reactor Vessel Water Level a.

S Q

R 1, 2, 3 1)

Low, Low, Level 2 2)

Low, Low, Low - Level 1 b.

Main Steam Line S

Q R

1,2,3 l

Radiation ## - High c.

Main Steam Line S

Q R

1 l

Pressure - Low d.

Main Steam Line S

Q R

1,2,3

(

Flow - High S

Q R

1, 2**, 3**

]

g Condenser Vacuum - Low e.

f.

Outboard MSIV Room 5

Q R

1,2,3 Temperature - High Turbine Enclosure - Main Steam 5

Q R

1,2,3 l

g.

Line Tunnel Temperature - High N.A.

R N.A.

1. 2, 3 h.

Manual Initiation RHR SYSTEM SHUTOOWN COOLING MODE ISOLATION 2.

j 5

Q R

1,2,3 I

Reactor Vessel Water Level ##

a.

Low - Level 3 S

Q R

1, 2, 3 b.

Reac+.or Vessel (RHR Cut-In Permissive) Pressure - High N.A.

R A.

1, 2, 3 Manual Initiation c.

i e

T AB LE e.3.PI.Y-F (t@mwnery---- ------- --- --- - -----__

TSOLATf0N ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERAT10NAL CHAf!NEL CHANNEL FUNCTI0f!AL CHANNEL CONDITf0NS FOR WHfCH CHECK _

TEST CALIBRATION SURVEILLANCE REQUIRE e

R 1, 2, 3 l

TRIP FUNCTION REACTOR WATER CLEANUP SYSTEM ISOLATION S

Q 3.

a.

RWCS A Flow - High R

1,2,3 l

S Q

,J.

RWCS Area Temperature - High R

1, 2, 3 l

e RWCS Area Ventilation S

Q c.

A Temperature - High N.A.

R N.A.

1,2,3 d.

SLCS Initiation l

R 1,2,3 Reactor Vessel Water Level S

Q c.

Low, Low, - Level 2 N.A.

R N.A.

1,2,3 f.

Manual Initiation t'.

HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATI0f(

l 4.

HPCI Steam Line S

Q R

1,2,3 Y

a.

Pressure - High 5

A b.

HPCI Steam Supply S

Q R

1, 2, 3 l

Pressure, low S

Q R

1,2,3 l

HPCI Turbine Exhaust Diaphragm c.

Pressure - High d.

HPCI Equipment Room 5

Q R

1, 2, 3 l

Temperature - Higli HPCI Equipment Room S

Q R

1, 1, 3 l

e.

A Temperature - High f.

HPCI Pipe Routing Area S

0 R

1, 2, 3 Temperature - High N.A.

R N.A.

1, 2, 3 Manual Initiation g.

h.

HPCI Steam Line N.A.

Q R

1,2,3 l

Pressure Timer A

m

.- m

,-.s__2_____

TABLE 4.3.2.1-1 (Continued)

ISOLATTON ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREM OPERATIONAL CHANNEL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH CHECK _

TEST CALIBRATI0f(

SURVEILLANCE REQUIRE TRfP FUNCTION

' ' REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION 5.

i I

R 1, 7.

RCIC Steam Line S

Q a.

a Pressure - High R

1, 2, 3 b.

RCIC Steam Supply 5

0 Pressure - Low R

1,2,3 RCIC Turbine Exhaust Diaphragm S

Q c.

Pressure - High d.

RCIC Equipment Room 5

Q R

1,2,3 l

Temperature - High l

R 1, 2, 3 RCIC Equipment Room S

Q e.

a Temperature - High g

I R

1, 2, 3 f.

RCIC Pipe Routing Area S

Q Temperature - High N.A.

R N.A.

1,2,3 Manual Inftiation g.

R 1,2,3 h.

RCIC Steam Line N.A.

Q a Pressure Timor o

b

TABLE 4.3.2.1-1_ (Continued)

ISOLATION ACTUATf0N INSTRUMENTATION SURVEfLLANCE REQUfREM OPERATIONAL CHANNEL CONDITIONS FOR WHICH CHANNEL FUNCTIONAL CHANNE1 CALIBRATI0t{

SURVEILLANCE REQUIRE.

CIIECK_

TEST TRfP FUNCTION

' RIMARY CONTAINMENT ISOLATION P

6.

R 1, 2, 3 l

Reactor Vessel Water level S

Q 1,2,3 a.

R 1)

Low, Low - Level 2 S

Q 2)

Low, Low, Low - Level 1 R

1,2,3 I

S 0

b.

Drywell PressureH - itigh R

1, 2, 3 North Stack Eff;uent S

Q c.

Radiation - liigh d.

Deleted Reactor Enclosure Ventilation S

Q R

1, 2, 3 e.

Exhaust Duct - Radiation - High f.

Outside Atmosphere to Reactor N.A.

M Q

1,2,3 Enclosure A Pressure - Low Deleted g.

R 1,2,3 l

h.

Drywell Pressure - High/

5 Q

Reactor Pressure - Low f.

Primary Containment Instrument N.A.

M Q

1,2,3 Gas to Drywell A Pressure - Low 1,2,3 N.A.

R N.A.

j.

Manual Initiation

= _.........

.. =.

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CHANNEL CHANNEL FUNCTIONAL CllANNEL CONDITIONS FOR VHICH CHECK _

TEST CALIBRATION SURVEILLANCE REQUIRE TRIP FUNCTION

7.

SECONDARY CONTAINMENT ISOLATION S

Q R

1,2,3 l*

Reactor Vessel Water Level j

a.

j Low, Low - Level 2 S

Q R

1, 2, 3 l

,, h.

Drywell Pressure ## - High i

S Q

R l

c.1.

Refueling Area Unit i Ventilation Exhaust Duct Radiation - High S

Q R

{

2.

Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S

Q R

1,2,3 l

d.

Reactor Enclosure Ventilation Exhaust Duct Radiation - High Outside Atmosphere To Reactor N.A.

M Q

1,2,3 e.

Enclosure A Pressure - Low f.

Outside Atmosphere To Refueling

+

N.A.

M Q

Area A Pressure - Low

~

Reactor Enclosure N.A.

R N.A.

1, 2, 3 g.

Manual Initiation h.

Refueling Area N.A.

R H.A.

Manual Initiation

Required when (1) handling irradiated fuel in the refueling area secondary containment, or (2 l

ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel with the Vesse head removed and fuel in the vessel.

- When not administrative 1y bypassed and/or when any turbine stop valve is open.

During operation of the associated Unit 1 or Unit 2 ventilation exhaust system, ifThese trip functions (Ib, 2a, 6b, and 7b) are common to the RPS actuation trip function.

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3.s.3 INSTRUMENTATION

_ BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION 3/4.3.1 The reactor protection system automatically initiates a reactor scram to:

Preserve the integrity of the fuel cladding.

a.

Preserve the integrity of the reactor coolant system.

b.

Minimize the energy which must be adsorbed following a c.

loss-of-coolant accident, and Prevent inadvertent criticality.

d.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service When necessary, one channel may be made inoperable because of maintenance.

for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels The outputs of the channels in a trip system are combined The tripping of l

each trip system.

in a logic so that either channel will trip that trip system.The system meets the inten both trip systems will produce a reactor scram.

Specified of IEEE-279 for nuclear power plant protection systems.

surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P, " Technical Specification Improvement Analyses for BWR Reactor Protection System," as approved NRC and documented in the NRC Safety Evaluation Report (SER) (letter to T. A.

15, 1987. The bases for the trip settings Pickens from A. Thadani dated July of RPS are discussed in the bases for Specification 2.2.1.

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are No credit was completed within the time limit assumed in the safety ana Response time may be demonstrated by any series of sequential, overlappin total channel test measurement, provided such tests demonstrate the total Sensor response time verification may be channel response time as defined.

demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

B 3/4 3-1 LIMERICK - UNIT 1

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INS:IRUMENTATIO!!

BASES 1s0LA110N ACTUATION INSTRUMElifAT10N 37 &.3.2 This specification ensures the effectiveness of the instrumentation used to prescribing the OPERABILITY trip mitigate the consequences of accidents by When necessary, setpoints and response times for isolation of the reactor sys 1

surveillance.

Specified surveillance intervals and maintenance outage times have bee 4

determined in accordance with NEDC-30851P, Supplement 2, " Technical Specification Improvement Analysis for BWR Instrumentation Common ECCS Instrumentation," as approved by the NRC and documented in the NRC l

Evaluation Report (SER) (letter to D. N. Grace from C. E. Rossi dated Janu 1989) and NEDC-31677P-A, " Technical Specificati j

J 18,1990).

the NRC SER (letter to S. D. Floyd from C. E. Rossi dated June Some of the trip settings may have tolerances explicitly stated where bo the high and low values are critical and may have a subs of the setting have a direct bearing on safety, are established at a level away safety.

from the normal operating range to prevent inadvertent actuation of the system involved.

Except for the MSIVs, the safety analysis does not address individual se response times or the response times of the logic systems to are connected.

For A.C. operated valves, it is assumed that the A.C.

power supply is lost and is restored by startup of the emer valve starts to move.

generators.

In addition to the pipe break, the f ailure of the D.C. operated valve is assumed; thus the signal delay (sensor response) is concurrent with starts to move.

The safety 10-second diesel startup and the 3 second load center loading delay.

4 analysis considers an allowable inventory loss in esch case which in turn It follows determines the valve speed in conjunction with the 13-second delay.

that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions.

l Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Yalue is an allowance for instrument drift specifically allocated for each trip in the safety analyses.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INS 4

The emergency core cooling system actuation instrumentation is provided t i

initiate actions to mitigate the cor. sequences of accidents that are beyond the

~

This specification provides the OPERABILITY ability of the operator to control.

requirements, trip setpoints and response times that will ensure effectivenes Although the instruments are of the systems to provide the design protection.

1 listed by system, in some cases the same instrument may be used to send the j

actuation. signal to more than one system at the same time.

B 3/4 3-2 LIMERICK - UNIT I

=-

4 f

INSTRUMENTAT10?!

1

~

BASES (Continued)

EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION i

3/4.3.3 i

Actuation Instrumentation)," as approved by the NRC and documented in the S letter to D. N. Grace from C. E. Rossi dated December Operation with a trip set less conservative than its Trip Setpoint but withi its specified Allowable Value is acceptable on the basis that the difference l

between each Trip Setpoint and the Allowable value is an allowance for l

instrumentation drift specifically allocated for each trip in the safety analyses. '

3.4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATIOf The anticipated transient without scram (ATWS) recirculation pump trip l

system provides a means of limiting the consequences of the unlikely occurren

~

l The response of the j

of a failure to scram during an anticipated transient. plant to f

i dated March 1971, NEDO-General Electric Company Topical Report NED0-10349, 24222, dated December 1979, and Section 15.8 of the FSAR.

i system is a supplement to The end-of-cycle recirculation pump trip (EOC-RPT)d rejection events, the During turbine trip and generator loa l

E0C-RPT will reduce the likelihood of reactor vessel lev the eactor trip.

order to reduce the void collapse in the core during two of the most 2.

pressurization events.will function as closure of the turbine stop valves and fast cl turbine control valves.

7 A fast closure sensor from each of two turbine control valves provides input i

to the EOC-RPT system; a fast closure sensor from each of the other two turbin

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8 l

Similarly, a control valves provides input to the second EOC-RPT system.

position switch for'each of two turbine stop valves provides input to one EOC RPT system; a position switch form each of the' other two stop valves pr i

input to the other EOC-RPT system. contacts are arranged to form a 2-out The control valves and a 2-out-of-2 logic for the turbine stop valves.

operation of either logic will actuate the EOC-RPT system and trip both recirculation pumps.

Each EOC-RPT system may be manually bypassed by use of a keyswitch whic l

The manual bypasses and the automatic Operating administratively controlled.

Bypass at less than 30% of P,ATED THERMAL POWER are annunciated in i room.

the E0C-RPT system response time is the time assumed in the analysis betwee initiation of valve motion and complete suppression of the electric arc, i.e.,

175 ms. Included in this time are:

allotted for breaker arc suppression, and the response time of the system logic.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.

B3/43-3 LTMERICK - UNIT I i

t INDEX BASES j

PAGE B3/40-1 f

SECTION 3 / 4. 0 AP P L I C AB I L I T Y................................................

r 3/4.1 REACTIVITY CONTROL SYSTEMS B3/41-1 SHUT DOWN MARGIN...............................................

l 3/4.1.1 B3/41-1 REACT I V ITY AN0 MALI ES..........................................

3/4.1.2 B 3/41-2 C ONT ROL R0D S..................................................

3/4.1.3 B3/41-3 l

CONTROL R0D PROGRAM CONTR0LS..................................

i B3/41-4 l

3/4.1.4 ST ANDBY LIQUID CONTROL SY STEM.................................

t I

3/4.1.5 3/4.2 POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION B 3/4 2-1.

l 3/4.2.1 RATE..........................................................

l B3/42-2 AP RM S ET P0 l NT S................................................

3/4.2.2 B 3/4 2-3 l

LEFT INT ENT IONALLY BLANK.................................................i B 3/4 2-4 MINIMUM CRITICAL POWERRATI0..................................

3/4.2.3 B3/42-5 LINE AR HEAT GENERAT ION RATE...................................

1 3/4.2.4 3/4.3 INSTRUMENTATION B 3/4 3-1 REACTOR PROTECTI ON SYSTEM INST RUMENTATION.................

l 3/4.3.1 B3/43-2 ISOLAT ION ACTUATION INSTRUMENT ATION...........................

3/4.3.2 EMERGENCY CORE COOLING SYSTEM ACTUATION B 3/4 3-2 3/4.3.3 I NST RUM E NT AT I ON...............................................

B 3/4 3-3 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.............

3/4.3.4 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION B 3/4 3-4 3/4.3.5 I N ST RU M E NTAT I ON...............................................

B 3/4 3-4 CONTROL ROD BLOCK INSTRUMENT AT ION.............................

1 3/4.3.6 3/4.3.7 MONITORING INSTRUMENTATION B 3/4 3-5 l

Rad i ati on Moni tori ng Instrumentation..........................

N

.. ~

xviii LIMERICK - UNIT 2

i INDEX pASES' PAGE t

SECTION INSTRUMENTATION (Continued)

B 3/4 3-5 Seismic Monitoring Instrumentation............................

B 3/4 3-5

( D el e t e d ).....................................................

B 3/4 3-5 Remote Shutdown System Instrumentation and Controls...........

B 3/4 3-5 Accident Monitoring Instrumentation...........................

B 3/4 3-5 Source Range Monitors.........................................

B 3/4 3-6 l

Traversing In-Core Probe System...............................

B 3/4 3-6 Chlorine and Toxic Gas Detecti on Systems......................

B 3/4 3-6 Fire Detection Instrumentation................................

B 3/4 3-7 Loose-Part Detection System..................................

B 3/4 3-7

( D el e t e d ).....................................................

B 3/4 3-7 t

Offgas Monitoring Instrumentation.............................

B 3/4 3-7 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM...........................

3/4.3.9 FEEDk'ATER/ MAIN TURBINE TRIP SYSTEM ACTUATION B3/43-7 I NST RUMENT AT ION...............................................

Bases Figure B 3/4.3.1 Reactor Vessel Water Leve1...........................

B 3/4 3-8 3/4.4 REACTOR COOLANT SYSTEM B3/44-1 3/4.4.1 RECIRCULATION SYSTEM..........................................

B 3/4 4-2 3/4.4.2 SAFETY / RELIEF VALVES..........................................

3/4.4.3 REACTOR COOLANT SYSTEM LEAYAGE B 3/4 4-3 Le akage Detection Systems...............................

B3/44-3 Op e r ati on al L e ak ag e.....................................

B 3/4 4-3a 3/4.4.4 C H EM I S T RY...............................................

1 l

LIMERICYs - UNIT 2 xix w---~

INSTRUMENTATION 3/4.3.2.

ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be i

3.3.2 OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2.-2 and with ISOLATION SYSTEM RESPONSE TIME as show 3.3.2-3.

i APPLICABILITY: As shown in Table 3.3.2-1.

ACTION:

With an isolation actuation instrumentation channel trip setpoint less a) conservative than the value shown in the Allowable Values column of Table 3.3 2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

j With the number of OPERABLE channels less than required by the Minimum OPERABLE b)

Channels per Trip System requirements for one trip system:

If placing the inoperable channel (s) in the tripped condition would cause 1.

an isolation, the inoperable channel (s) shall be restored to OPERABLE status If this cannot be accomplished, the ACTION required by Table within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3.3.2-1 for the affected trip function shall be taken, or the channel shall be placed in the tripped condition.

i or l

If placing the inoperable channel (s) in the tripped condition would not cause l

2.

an isolation, the inoperable channel (s) and/or that trip system shall l

be placed in the tripped condition within:

a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common

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

to RPS Instrumentation.

The provisions of Specification 3.0.4 are not applicable.

1 1

Trip functions common to RPS Actuation Instrumentation are shown l

8 l

in Table 4.3.2.1-1.

LIMERICK - UNIT 2 3/4 3-9 i

TABLE 3.3.2-1 (Continued)

JJiOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 Be in at least HOT SHUIDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in CO next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Be in at least STARTUP with the associated is f

ACTION 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months or be in at least l

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

i ACTION 22 Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In OPERATI0h CONDITION 1 or 2, verify the affected system isolation f

In are closed within I hour and declare the affected system inoperable.

ACTION 23 3, be in at least COLD SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(

OPERATIONAL CONDITION l

Restore the manual initiation function to OPERABLE status close the affected system isolation valves within l ACTION 24 l

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 .

J ACTION 25 Establish SECONDARY CONTAINMENT INTEGRITY with the stand b system operating within I hour.

i ACTION 26 Close the affected system isolation valves within I hour.

TABLE NOTATIONS l

Required when (1) handling irradiated fuel in the refueling area secondary containment, or (2) during CORE ALTEPATIONS, or (3) during operations with a potential for draining the reactor vessel with the vessel head removed an the vessel.

Hay be bypassed under administrative control, with all turbine stop valves j

During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.

See Specification 3.6.3. Table 3.6.3-1 for primary containment isolation (a) are actuated by these isolation signals.

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillance without placing the trip system in the tripped conditio (b) provided at least one OPERABLE channel in the same trip system is mo Trip functions common to RPS Actuation Instrumentation are In addition, for the HPCI system and RCIC system paraneter.

shown in Table 4.3.2.1-1.

isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all require hours for required surveillance without placing the channel or trip system in th, tripped condition.

l s

'TdBLE 4.3.2.FC -

~

~~

~ '

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.-

OPERATIONAL CHANNEL CHANNEL FUNCTIONAL CHANNEt.

CONDITIONS FOR WHICH CHECK _

TEST CALIBRATIO4 SURVEILLANCE REQUIRE TRIP FUNCTION 1.

, MAIN STEAM LINE ISOLATION 5

0 R

1, 2,.1 Reactor Vessel Water Level a.

S Q

R 1,2,3 1)

Low, Low, Le"el 2 2)

Low, Low, Low - Level 1 b.

Main Steam Line S

Q R

1,2,3 l

Radiatior, ## - High Main Steam Line S

Q R

1 l

c.

Pressure - Low d.

Main Steam Line S

Q R

1,2,3 l

Flow - High S

Q R

1, 2'*, 3**

l Ccndenser Vacuum - Low c.

4 f.

Outboard MSIV Room S

Q R

1,2,3 l

me Temperaturc - High 5

0 R

1, 2, 3 l

Turbine Enclosure - Main Steam g.

Line Tunnel Temperature - High N.A.

R N.A.

1. 2, 3 h.

Manual Initiation RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION 2.

l 3

Reactor Vessel Water Level ##

S 0

R 1,2,3 a.

Low - Level 3 S

Q R

1, 2, 3 b.

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

R N.A 1, 2, 3 Manual Initiation c.

~

(([

TABLE 4.3.2.1-1_ (Continued)

ISOLATf0N ACTUATION INSTRUMENTATf0N SURVEfLLANCE REQUfREMENTS OPERATIONAL CHANNEL CHANNEL FUNCTIONAL CHANNEL CONDfTIONS FOR WHfCH ~,

CHECK _

TEST CALIBRATION SURVETLLANCE REQUTRE S

Q R

1,2,3 l

TRfP FUNCTION REACTOR WATER CLEANUP SYSTEM ISOLATION 3

RWCS a Flow - High l

a.

S Q

R 1, 2, 3

[

b RWCS Area Temperature - High

.p.

S Q

R 1, 2, 3 l

m RWCS Arca Ventilation M

c.

a Temperature - High N.A.

R N.A.

1, 2, 3 d.

SLCS Initiation S

Q R

1,2,3 l

Reactor Vessel Water level c.

Low, Low, - Level 2 N.A.

R N.A.

1,2,3 f.

Manual Initiation HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATI0f{

HPCI Steam Line S

Q R

1, 2, 3 l

4.

a.

A Pressure - High g

b.

HPCI Steam Supply S

Q R

1, 2, 3 l

o.

u E

Pressure, Low co S

Q R

1,2,3 l

HPCI Turbine Exhaust Diaphragm c.

Pressure - High d.

HPCI Equipment Room S

Q R

1,2,3 l

Temperature - High HPCI Equipment Room S

Q R

1, 2, 3 l

c.

Temperature - High A

f.

IIPCI Pipe Routing Arca S

0 R

1, 2, 3 l

Temperature - liigh N.A.

R N.A.

1, 2, 3 Manual Initiation g.

h.

HPCI Steam Line N.A.

Q R

1,2,3 l

Pressure Timer A

TABLE 4.3.2.1-1 (Continued)

ISOLATf0N ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL i

CHANNEL CllANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH CllECK_

TEST CALIBRATION SURVEILLANCE REQUIRE TRfP FUNCTION

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION 5.

R 1, 2, 3 l

RCIC Steam Line S

Q

>o a.

a Pressure - High R

1,2,3 b.

RCIC Steam Supply 5

Q Pressure - Low S

Q R

1, 2, 3 l

RCIC Turbine Exhaust Diaphragm c.

Pressure - High 2

d.

RCIC Equipment R*cm 5

Q R

1,2,3 l

w Temperature - High y

RCIC Equipment Room S

Q R

1,2,3 l

w e.

A Iemperature - High f.

RCIC Pipe Routing Area S

Q R

1,2,3 l

Temperature - High N.A.

R N.A.

1, 2, 3 -

Manual Initiation g.

h.

RCIC Steam Line N.A.

Q R

1, 2, 3 l

Pressure Timer A

4 ge*

e 4

e 6

-.e

TABLE 4.3.2.1-1_ (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS _

OPERATIONAL CHANNEL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH CHECK _

TEST CALIBRATIO!!

SURVEILLANCE REQUIRE

TRIP FUNCTION

6. ' PRIMARY CONTAINMENT ISOLATION S

Q R

1, 2, 3 l

'evel Reactor Vessel c.

a.

5 Q

R 1, 2, 3 1)

Low, Low - Level 2 2)

Low, Low, low - Level 1 S

Q R

1,2,3 l

b.

Drywell Pressurc 11 - High S

Q R

1, 2, 3 North Stack Effluent c.

Radiation - High d.

Deleted Reactor Enclosure Ventilation S

Q R

1, 2, 3 c.

Exhaust Duct - Radiation - High j

N.A.

M Q

1,2,3 f.

Outside Atmosphere to Reactor Enclosure A Pressure - Low g.

Deleted h.

Drywell Pressure - High/

S Q

R 1, 2, 3 l

Reactor Pressure - Low 1.

Primary Containment Instrument N.A.

M Q

1, 2, 3 Gas to Drywell A Pressure - Low N.A.

R N.A.

1, 2, 3 j.

Manual Initiation t

G 6

w w

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE RE OPERATIONAL CHANNEL CONDITIONS FOR WHICH,

CHANNEL FUNCTIONAL CHANNEL SURVEILLANCE REQUIRE CALIBRATION TEST __

CHECK _

TRIP FUNCTION SECONDARY CONTAINMENT ISOLATION 1, 2, 3 7.

R Reactor Vessel Water Level S

Q a.

Low, low - Level 2 1,2,3 R

S Q

Drywell Pressurc ## - High b.

1 l

Refueling Area Unit i Ventilation R

S Q

c.1.

+xhaust Duct Radiation - High l

+#

Refueling Arca Unit 2 Ventilation S

Q R

2.

Exhaust Duct Radiation - High 1,2,3 l

Reactor Enclosure Ventilation S

Q R

d.

Duct Radiation - High w

3:

Exhaust Outside Atmosphere To Reactor N.A.

M Q

1, 2, 3 c.

Enclosure A Pressure - Low Outside Atmosphere To Refueling N.A.

M Q

f.

Area a Pressure - Low N.A.

1,2,3 Reactor Enclosure N.A.

R g.

Manual Initiation N.A.

h.

Refueling Area N.A.

R Manual Initiation dary containment, or (2) during CORE

Required when (1) handling irradiated fuel in the refueling area seconALTE l with the vessel head removed and fuel in the vessel.

- When not administratively bypassed and/or when any turbine stop valve is open.

During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.

function.

1. These trip functions (Ib, 2a, 6b, and 7b) are common to the RPS actuation trip e

e

i i

'e e

3.4.7 INSTRUMENTATION BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION 3/4.3.1 The reactor protection system automatically initiates a reactor scram to:

\\

Preserve the integrity of the fuel cladding.

a.

Preserve the integrity of the reactor coolant system.

b.

Minimize the energy which must be adsorbed following a c.

loss-of-coolant accident, and Prevent inadvertent criticality.

d.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be because of maintenance.

for brief intervals to conduct required surveillance.

l The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels The outputs of the channels in a trip system are combined The tripping of each trip system.

in a logic so that either channel will trip that trip system.The system meets the inte both trip systems will produce a reactor scram.

Specified of IEEE-279 for nuclear power plant protection systems.

surveillance intervals and surveillance and maintenance outage times have bee determined in accordance with NEDC-30851P, " Technical Specification Improvement Analyses for BWR Reactor Protection System," as approved NRC'and documented in the NRC Safety Evaluation Report (SER) (letter to T. A 15, 1987. The bases for the trip settings Pickens from A. Thadani dated July of RPS are discussed in the bases for Specification 2.2.1.

I The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are No credit was completed within the time limit assumed in the safety an Response time may be demonstrated by any series of sequential, overlapp total channel test measurement, provided such tests demonstrate the total Sensor response time verification may be channel response time as defined.

demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

l 8 3/4 3-1 LIMERICK ~- UNIT 2 u

e

-,n

. INSTRUMENTATION g

BASES A

_/.4.3.2 ISOLATION ACTUAT10N INSTRUMENInfl0N 3

This specification ensures tbc effectiveness of the instrumentation used to I

prescribing the OPERABILITY trip mitigate the consequences of accidents by When setpoints and response times for isolation of the reactor syst i

surveillance.

Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-3085IP, Supplement 2, " Technical Specification Improvement Analysis for BWR Instrumentation Comon to RPS a ECCS Instrumentation," as approved by the NRC and documented in the NRC Safety Evaluation Report (SER) (letter to D. N. Grace from C. E. Rossi dated January 6, 1989) and NEDC-31677P-A, " Technical Specification Im 18,1990).

the NRC SER (letter to S. D. Floyd from C. E. Rossi dated June Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substan of the setting have a direct bearing on safety, are established at a level away safety.

from the normal operating range to prevent inadvertent actuation of the systems i

(

involved.

Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to wh

[

l are connected.

For A.C. operated valves, it is assumed that the A.C.

valve starts to move.

power supply is lost and is restored by startup of the emergency diesel In this event, a time of 13 seconds is assumed before the valve In addition to the pipe break, the failure of the D.C. operated generators.

valve is assumed; thus the signal delay (sensor response) is concurrent with the starts to move.

The safety 10-second diesel startup and the 3 second load center loading delay.

analysis considers an allowable inventory loss in each case which in turn It follows determines the valve speed in conjunction with the 13-second delay.

that checking the valve speeds and the 13-second time for en,ergency power l

establishment will establish the response time for the isolation functions.

Operation with a trip set less conservative than its Trip Setpoint but I

Value is acceptable on the basis that the within its specified AllowaP difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drif t specifically allocated for each trip in the safety analyses.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATIOu INSTRUME The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the This specification provides the OPERABILITY ability of the operator to control.

requirements, trip setpoints and response times that will ensure effectiveness Although the instruments are of the systems to provide the design protection.

listed by system, in some cases the same instrument may be used to send the actuation 1 signal to more than one system at the same time.

B 3/4 3-2 LIMERICK - UNIT 2

)

~

6 =i.'

INSTRUMENTATION 8ASES i

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the AlloQable value is an allowance

~

for instrument drif t specifically allocated for each trip in the safety analyses.

RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATI0li 3/4.3.4 The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a f ailure to scram during an anticipated transient. The response of the

. plant to this pos.tulated event f alls,within the' envelope of study events in l

General Electric Company Topical Report HED0-10349, dated March 1971, NEDO-24222, dated December 1979, and Section 15.8 of the FSAR.

i

~

The end-of-cycle recirculation pump trip (EOC.RPT) system is a supplement to i

During turbine trip and generator load rejection events, the the reactor trip.

EOC-RPT will reduce the likelihood of reactor vessel level decreasing to level j

Each EOC-RPT system trips both recirculation pumps, reducing coolant flow in I

l 2.order to reduce the void collapse in the core during two of the most limiting l

pressurization events. The two Eve'nts for which the EOC-RPT protective feature l

will function are closure of the turbine stop valves and f ast closure of the j

turbine control valves.

A fast closure sensor from each of two turbine control valves provides input i

to the EOC-RPT system; a f ast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT system. Similarly, a i

~

~ position switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch from each of the other two stop valves provides 2

input to the other EOC-RPT system. For each EOC-RPT system, the sensor relay contacts are arrariged to form a 2-out-of-2 logic for the fast closure of turbine i

l control valves and a 2-out-of-2 logic for the turbine stop vahes. The operati6n of either logic will actuate the EOC-RPT system and trip both recirculation pumps.

[I.

Each EOC-RPT system may be manually bypassed by use of a keysvitch which is i

administrative 1y controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control l

roome The EOC-RPT system response time is the time assumed in the analysis'between initiation.of valve motion and complete suppression of the electric arc, i.e.,

175 ms.

Included in this time are: the response time of the sensor, the time

' allotted for breaker arc suppression, and the response time of the system logic.

' Operation with a trip set less conservative than its Trip Setpoint but I

within its specif.ied Allowable Value is acceptable on the basis that the.

difference between each Trip Setpoint and the Allowable Value is an allowante for instrument drif t specifically allocated for each trip in the safety analyses.

B 3/4 3-3 LIMERICK - UNIT 2

\\

{

l a.

3 J

l 4

TSCR No. 92-10-0 Docket Nos. 50-352 and 50-353 1

Limerick Generating Station, Unit 2 Instrumentation Drift Data for Containment Isolation Actuation Instrumentation i

There are 58 channels of trip instrumentation for the Containment Isolation Attuation Instrumentation which are currently tested monthly as required by 4

Technical Specifications (TS), Tables 4.3. 2.1-1, 4.3.4.1-1, 4.3.4. 2.1-1, 4.3.7.1-1, 4.3.9.1-1 and TS Sections 4.1.3.1.4, 4.3.7.8.1, 4.4.2.1.

A review of surveillance data sheets was performed for a random sample of six (6) of the trip units to determine how much the trip setting changes over a period of three (3) consecutive surveillance intervals from July,1992 to October,1992.

The results are provided below.

CHANNEL DESCRIPTION NETSEWOINTCHANGE(3 MOS) 1)

PIS-01-2N675A MAIN STEMA UNE FLOW - HIGH 0.01 MA 2)

PIS-01-2N675B MAIN STEAM UNE FLOW - HIGH 0.00 MA l

j 3)

PIS-01-2N675C MAIN STENA UNE FLOW-HIGH 0.00 MA

]

4)

PIS-01-2N675D MAIN STENA UNE FLOW-HIGH 0.00 MA 5)

PDIS49-2N657A NSSSS-RCIC STEAM UNE DIFF. PRESS-HIGH 0.00 MA 1

6)

PIS-49-2N658C NSSSS-RCIC STEMA SUPPLY PRESS.-LOW 0.00 k%

The worst. case as-found drift for the trip units in this sample for a three month period, was i0.01MA. The applicable instrumentation setpoint calculations allow drift of 10.02MA.

If all of the change in the trip unit setting reported above is attributed to drift, the worst case actual drift is 50% of the allowable drift as determined from setpoint methodology. We j

have therefore confirmed that the as-found drift is within that allowed by setpoint calculations.

3 i

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