ML20065N430

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Proposed TS for Extension of Surveillance Intervals & AOTs for Containment Isolation Instrumentation & Other Selected Actuation Instrumentation,Incorporating Info from Recently Issued TS Amends & Correcting Typos
ML20065N430
Person / Time
Site: Limerick  Constellation icon.png
Issue date: 04/18/1994
From:
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20065N404 List:
References
NUDOCS 9404270154
Download: ML20065N430 (27)
v • d • e


Text

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t ATTACHMENT 1 Limerick Generating Station Uiiits 1 and 2 Revised Technical Specifications Pages for TSCR No. 92-10-0 9404270154 940418 PDR ADOCK 05000352 ~

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1

~..; T INDEX BASES.

SECTION EAGE I

1/4.0 APPLICABILITY................................................... B 3/4 0-1

;2/4.1 REACTIVITY CONTROL SYSTEMS ,

'3/4.1.1 SHUTDOWN MARGIN............................................ B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES....................................... B 3/4 1-1 L3/4.1.3 CONTROL R0DS............................................... B 3/4 1-2 ,

a 3/4.1.4. CONTROL R00 PROGRAM CONTR0LS............................... B 3/4 1-3 13/4.1.5 STANDBY LIQUID CONTROL SYSTEM.............................. B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE....................................................... B 3/4 2-1 3/4.2.2 -(DELETED).................................................. B 3/4 2-2 LEFT INTENTIONALLY BLANK.............................................. B 3/4 2-3 l 1

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

3/4.2.4 LINEAR HEAT GENERATION RATE................................ B 3/4 2-5 1 1

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

3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION........................ B 3/4 3-2

-3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION............................................ B 3/4 3-2 3/4.3.4 RECIRCULATION. PUMP TRIP ACTUATION INSTRUMENTATION.......... B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION.~........................................... B 3/4 3-4~ j l

1

3/4.3.6 ' CONTROL R00' BLOCK INSTRUMENTATION.......................... B 3/4'3-4 )

'3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring' Instrumentation.....................- B 3/4 3-5' I 1

I LfMERICK --UNIT 1 xviii

=. , _ _ __ _ _ -.

.. __ _~

..- INDEX BASES SECTION PAGE INSTRUMENTATION (Continued)

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 Traversing In-Core Probe System.......................... B 3/4 3-6 l Chlorine and Toxic Gas Detection Systems................. B 3/4 3-6 Fire Detection Instrumentation........................... B 3/4 3-6 Loose-Part Detection System.............................. B 3/4 3-7 (Deleted)................................................ B 3/4 3-7 Offgas Monitoring Instrumentation........................ B 3/4 3-7 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM........................ B 3/4 3-7 1 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION............................................ B/3/4 3-7 l Bases Figure B 3/4.3-1 Reactor Vessel Water Leve1............................. B 3/4 3-8 )

3/4.4 REACTOR COOLANT SYSTEM  ;

1 3/4.4.1 RECIRCULATION SYSTEM....................................... B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES....................................... B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE i Leakage Detection Systems................................ B 3/4 4-3 I Ope rat i on al Lea kag e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 3/4.4.4 CHEMISTRY.................................................. B 3/4 4-3 LIMERICK - UNIT 1 xix

i s

. INSTRUMENTATION BASES. ,

3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.

Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30851P, Supplement 2, " Technical Specification Improvement Analysis for BWR Instrumentation Common to RPS and 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 Improvement Analysis for BWR Isolation Actuation Instrumentation," as approved by the NRC and documented'in the NRC SER (letter to S.D. Floyd from C. E. Rossi dated June 18,1990).

Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involvad.

Exctot for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors  !

are connected. For D.C. operated valves, a 3 second delay is assumed before the valve starts to move. For A.C. operated valves, it is assumed that the A.C.

power supply is lost and is restored by startup of the emergency diesel generators. In this event, a time of 13 seconds is assumed before the valve starts to move. In addition to the pipe break, the failure of the D.C. operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 10-second diesel startup and the 3 second load center loading delay. The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second delay. It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions.

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.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness l of the systems to provide the design protection. Although the instruments are I listed by system, in some cases the same instrument may be used to send the i actuation signal to more than one system at the same time. 1 L

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

1

. INSTRUMENTATION BASES j 3/4.3.3 EMERGENCY' CORE CQDLING ACTUATION INSTRUMENTATION (Continued) l Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30936P, Parts 1 and 2, " Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," as approved by the NRC and documented in the SER (letter to D. N. Grace from A. C. Thadani dated December 9,1988 (Part 1) and letter to D. N. Grace from C. E. Rossi dated December 9, 1988 (Part 2)).

Operation with a tri) set less conservative than its Trip Setpoint but within its specified Allowaale 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.

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971, NED0-24222, dated December 1979, and Section 15.8 of the FSAR.

The end-of-cycle recirculation pump trip (E0C-RPT) system ir a supplement to the reactor trip. During turbine trip and generator load rejection events, the E0C-RPT will reduce the likelihood of reactor vessel level decreasing to level

2. Each E0C-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the E0C-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.

A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second E0C-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one E0C-RPT system; a position switch from each of the other two stop valves provides input to the other EOC-RPT system. For each E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the E0C-RPT system and trip both recirculation pumps.

Each E0C-RPT system may be manually bypassed by use of a keyswitch which is administrative 1y controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room.

The E00-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 i 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 specified Allowable Value is acceptable on the basis that the i difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. I LIMERICK.- UNIT 1 B 3/4 3-3 1

t -INSTRUMENTATION BASES.

3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip <

setpoints and response times for. isolation of the reactor systems. When l necessary, one channel may be inoperable for brief intervals to conduct required  ;

surveillance. l Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30851P, Supplement 2, " Technical Specification Improvement Analysis for BWR Instrumentation Common to RPS and 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 Improvement Analysis for BWR Isolation Actuation Instrumentation," as approved by the NRC and documented in the NRC SER (letter to S.D. Floyd from C. E. Rossi dated June 18, 1990).

Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high of low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.

Except for the MSIVs, the safety analysis does not address individual sensor .

response times or the response times of the logic systems to which the sensors are connected. For D.C. operated valves, a 3 second delay is assumed before the valve starts to move. For A.C. operated valves, it is assumed that the A.C.

power supply is lost and is restored by startup of the emergency diesel generators. In this event, a time of 13 seconds is assumed before the valve starts to move. In addition to the pipe break, the failure of the D.C. operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 10-second diesel startup and the 3 second load center loading delay. The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second delay. It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions.

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.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.

I LIMERICK - UNIT 2 B 3/4 3-2

'INSTRUMENTATIQN BASES 314.3.3 EMERGENCY CORE COOLING ACTVATION INSTRUMENTATION (Continued)

Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30936P, Parts 1 and 2, " Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," as approved by the NRC and documented in the SER (letter to D.

N. Grace from A. C. Thadani dated December 9, 1988 (Part 1) and letter to D. N.

Grace from C. E. Rossi dated December 9, 1988 (Part 2)).

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.

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971, NED0-24222, dated December 1979, and Section 15.8 of the FSAR.

The end-of-cycle the reactor trip. Duringrecirculation turbine trippump and trip (E0C-RPT)d generator loa rejection events, thesystem is a sup E0C-RPT will reduce the likelihood of reactor vessel level decreasing to level

2. Each E0C-RPT system trips both recirculation pumps, reducirg coolant flow in order to reduce the void collapse in the core during two of the most limiting ,

pressurization events. The two events for which the E0C-RPT protective feature '

will function are closure of the turbine stop valves and fast closure of the i turbine control valves.

A fast closure sensor from each of two turbine control valves provides input to the E0C-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second E0C-RPT system. Similarly, a

)osition switch for each of two turbine stop valves provides input to one E0C-1PT system; a position switch from each of the other two stop valves provides input to the other E0C-RPT system. For each E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the E0C-RPT system and trip both recirculation pumps.

Each E0C-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room. l l

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 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.

LIMERICK - UNIT 2 B 3/4 3-3

~

ATTACHMENT 2 Umerick Generating Station Units 1 and 2 Revised Technical Specifications Pages for TSCR No. 92-11-0 i

i

,, I a- INSTRUMENTATIQB 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6. The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.

APPLICABILIIY: As shown in Table 3.3.6-1.

ACTION:

a. With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. '
b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per trip Function requirement, take the ACTION required by Table 3.3.6-1.

SURVEILLANCE REQUIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE

  • oy the performance of I the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.
  • A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition, provided at least one other operable channel in the same trip system is monitoring that parameter.

LIMERICK - UNIT 1 3/4 3-57 .-

l l

.:, l TABLE 3.3.6-1 (Continued)

CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS ,

l ACTION 60 -

Declare the RBM inoperable and take the ACTION required I by Specification 3.1.4.3. J ACTION 61 -

With the number of OPERABLE Channels:

a. One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or place the inoperable channel in the tripped condition.
b. Two or more less than required by the Minimum OPEPABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.

ACTION 62 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l ACTION 63 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, initiate a rod block.

NOTES

  • For OPERATIONAL CONDITION of Specification 3.1.4.3.
  • With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

"o These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs, are in the core.

(a) The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30% of RATED THERMAL POWER.

(b) This function shall be automatically bypassed if detector count rate is

> 100 cps or the IRM channels are on range 3 or higher.

(c) This function is automatically bypassed when the associated IRM channels are on range 8 or higher.

(d) This function is automatically bypassed when the IRM channels are on range 3 or higher.

(e) This function is automatically bypassed when the IRM channels are on range 1.

(f) Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3.

LIMERICK - UNIT 1 3/4 3-59

a '

i

.. TABLE 4.3.9.1-1 l

EEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION EVRVEILLANCE RE0VIREMENTS OPERATIONAL CONDITIONS CHANNEL FOR WHICH CHANNEL FUNCTIONAL CHANNEL SURVEILLANCE TRIP FUNCTION CHECK TEST CALIBRATION RE0VIRED

1. Reactor Vessel Water D Q R 1 Level-High, Level 8 l

l l

4 LIMERICK - UNIT 1 3/4-3-115

REACTOR COOLANT SYSTEM 3/4.4.2 S'AFETY/ RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safet.y valve function of at least 11 of the following reactor coolant system safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*#

4 safety / relief valves 0 1130 psig i 1%

5 safety / relief valves 0 1140 psig i 1%

5 safety / relief valves 01150 psig i 1%

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With the safety valve function of one or more of the above required safety / relief valves inoperable, be in at least H0T SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1 and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b. With one or more safety / relief valves stuck open, provided that suppression pool average water temperature is less than 105 F, close the stuck open safety / relief valve (s); if unable to close the stuck open valve (s) within 2 i minutes or if suppression pool average water temperature is 110*F or greater, place the reactor mode switch in the Shutdown position,
c. With one or more safety / relief valve acoustic monitors inoperable, restore the  ;

inoperable acoustic monitors to OPERABLE status within 7 days or be in at least  ;

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

SURVEILLANCE REQUIREMENTS 4.4.2.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 0.20 of the full open noise leve1## by performance of a:

a. CHANNEL FUNCTIONAL TEST at least once per 92 days, and a j
b. CHANNEL CALIBRATION at least once per 18 months **.

4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure testeo and ,

stored in accordance with manufacturer's recommendations at least once per 24 months, and i they shall be rotated such that all 14 safety relief valves are removed, set pressure l tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 54 months.

o The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

  • The provisions of Specification 4.0.4 are not applicable provided the Surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.
  1. Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refueling.
    1. Initial setting shall be in accordance with the manufacturer's recommendation. Adjustment to the valve full open noise level shall be accomplished during the startup test program.

LIMERICK - UNIT 1 3/4 4-7

l falLTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. By verifying at least 8 suppress 9n pool water temperature indicators in I at least 8 locations, OPERABLE by performance of a:
1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
3. CHANNEL CALIBRATION at least once per 18 months, i with the temperature alarm setpoint for: l
1. High water temperature: l a) First setpoint s 95 F b) Second setpoint s 105 F c) Third setpoint s 110*F d) Fourth setpoint s 120*F
d. By verifying at least two suppression chamber water level indicators OPERABLE by performance of a:
1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
2. CHANNEL fur 1CTIONAL TEST at least once per 92 days, and
3. CHANNEL CALIBRATION at least once per 18 months, with the water level alarm setpoint for high water level s 24'l-1/2"
e. Drywell-to-suppression chamber bypass leak tests shall be conducted at I 40 +/- 10 month intervals to coincide with the ILRT at an initial differential pressure of 4 psi and verifying that the A/(k calculated from the measured leakage is within the specified limit. If any drywell-to-suppression chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 24 months until two consecutive tests meet the specified limit, at which time the test schedule may be resumed. l
f. By conducting a leakage test on the drywell-to-suppression chamber vacuum I breakers at a differential pressure of at least 4.0 psi and verifying that l the total leakage area A//k contributed by all vacuum breakers is less l than or equal to 24% of the specified limit and the leakage area for an l individual set of vacuum breakers is less than or equal to 12% of the specified limit. The vacuum breaker leakage test shall be conducted during each refueling outage for which the drywell-to-suppression chamber bypass leak tt t in Specification 4.6.2.1.d is not conducted.

LIMERICK - UNIT 1 3/4 6-14

Q.

c . INSTRUMENTATION

.pASES' 3/4.3.3 EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION (Continued)

Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30936P, Parts 1 and 2, " Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," as approved by the NRC and documented in the SER (letter to D. N. Grace from A. C. Thadani dated December 9,1988 (Part 1) and letter to D. N. Grace from C. E. Rossi dated December 9, 1988 (Part 2)).

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.

3/4.3.4 RECIRCVLATION PVMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971, NED0-24222, dated December 1979, and Section 15.8 of the FSAR.

The end-of-cycle recirculation pump trip (E0C-RPT) system is a supplement to the reactor trip. During turbine trip and generator load rejection events, the E0C-RPT will reduce the likelihood of reactor vessel level decreasing to level

2. Each E0C-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.

A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second E0C-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one E0C-RPT system; a position switch from each of the other two stop valves provides input to the other E0C-RPT system. For e&ch E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the E0C-RPl system and hip both recirculation pumps.

Each E0C-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room.

The E0C-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.

LIMERICK - UNIT 1 B 3/4 3-3

~

INSTRUMENTATI0ff BASES 3/4.3.4 RECIRCULATION PVMP TRIP ACTUATION INSTRUMENTATION (Continued)

Specified surveillance intervals and maintenance outage times have been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D.

Binz, IV, from C.E. Rossi dated July 21, 1992).

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.

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel. This instrumentation does not provide actuation of any of the emergency core cooling equipment.

Specified surveillance intervals and maintenance outage times have been specified in accordance with recommendations made by GE in their letter to the BWR Owner's Group dated August 7, 1989,

SUBJECT:

" Clarification of Technical Specification changes given in ECCS Actuation Instrumentation Analysis."

Operation with a trip set less conservative than its Trip Setpoint but eithin 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 1 analyses.  !

1 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION 1 The control rod block functions are provided consistent with the I requirements of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits and-Section 3/4.3 The trip logic is arranged so that a trip in any one of the Instrumentation. l inputs will result in a control rod block.

Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30851P, Supplement 1, " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"

as approved by the NRC and documented in the SER (letter to D. N. Grace from C.

E. Rossi dated September 22,1988). 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 Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.

LIHERICK - UNIT I B 3/4 3-4

i ~

INSTRUMENTATION BASE 1 1/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that:

(1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64.

The specified surveillance interval for the Main Control Room Normal Fresh Air Supply Radiation Monitor has been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D. Binz, IV, from C.E. Rossi dated July 21, 1992).

3/4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This capability is required to permit comparison of the measured response to that used in the design basis for the unit.

3/4.3.7.3 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.

3/4.3.7.4 REMOTE SHUTOOWN SYSTEM INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown system instrumentation and controls ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the unit from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50, Appendix A.

3/4.3.7.LACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident,"

December 1975 and NUREG-0737, " Clarification of TMI Action Plan Requirements,"

November 1980.

3/4.3.7.6 SOURCE RANGE MONITORS The source range monit A* provide the operator with information of the status of the neutron level in the c.are at very low power levels during startup and shutdown.

At these power levels, rractivity additions shall not be made without this flux level information available tc the operator. When the intermediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.

LIMERICK - UNIT 1 B 3/4 3-5

\

INSTRUMENTATION BASES 3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor Core.

The TIP system OPERABILITY is demonstrated by normalizing all probes (i.e.,

detectors) prior to performing an LPRM calibration function. Monitoring core thermal limits may involve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be required to be OPERABLE. The OPERABILITY of individual detectors to be used for monitoring is demonstrated by comparing the detector (s) output in the resultant heat balance calculation (P-1) with data obtained during a previous heat balance calculation (P-1).

3/4.3.7.8 CHLORINE AND T0XIC GAS DETECTION SYSTEMS The OPERABILITY of the chlorine and toxic gas detection systems ensures that an accidental chlorine and/or toxic gas release will be detected promptly and the necessary protective actions will be automatically initiated for chlo-rine and manually initiated for toxic gas to provide protection for control room personnel. Upon detection of a high concentration of chlorine, the c" trol room emergency ventilation system will automatically be placed in the chlor ane isolation mode of operation to provide the required protection. Upon detection of a high concentration of toxic gas, the control room emergency ventilation system will manually be placed in the chlorine isolation mode of operation to provide the required protection. The detection systems required by this speci-fication are consistent with the recommendations of Regulatory Guide 1.95, " Pro-tection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release," February 1975.

Specified surveillance intervals and maintenance outage times have been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical l Specifications," as approved by the NRC and documented in the SER (letter to R.D.

Binz, IV, from C.E. Rossi dated July 21,1992).

3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERABILITY of the detection instrumentation enrures that both adequate .

warning capability is available for prompt detection of fires and that fire l suppression systems, that are actuated by fire detectors, will discharge extin-quishing agent in a timely manner. Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an j integral element in the overall facility fire protection program.  !

Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification.

Consequently, the minimum number of OPERABLE fire detectors must be greater.  !

LIMERICK - UNIT 1 B 3/4 3-6

v

' INS-TRUMENTATION BASES 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION (Continued)

The loss of detection capability for fire suppression systems, actuated by fire detectors, represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initi-ated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

The surveillance requirements for demonstrating the OPERABILITY of the fire detectors are based on the recommendations of NFPA 72E - 1990 Edition.

3/4.3.7.10 LOOSE PART DETECTION SYSTEM The OPERABILITY of the loose-part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system and avoid or mitigate damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the recem-mendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981.

3/4.3.7.11 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.

3/4.3.7.12 0FFGAS MONITORING INSTRVMENTATION l

This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures and noble gases in the off-gas system.

3/4.3.8. TURBINE OVERSPEED PROTECTION SYSTEM This specification is provided to ensure that the turbine overspeed protection system instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures.

3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/ main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system / main turbine trip system l in the event of failure of feedwater controller under maximum demand. l LIMERICK - UNIT 1 B 3/4 3-7

c, INSTRUMENTATION 3/4.3.6 CONTROL R00 B1QCK INSTRUMENTATION

-LIMITING CONDITION FOR OPERATION 3.3.6. The control rod block instrumentation channels shown in Table 3.3.6-1 i shall be OPERABLE with their trip setpoints set consistent with the values j shown in the Trip Setpoint column of Table 3.3.6-2.

APPLICABILITY: As shown in Table 3.3.6-1.

J ACTION:

a. With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.E-2, declare the channel inoperable until the channel is-restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.  ;
b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per trip Function requirement, take the ACTION required by Table 3.3.6-1.

SURVEILLANCE REQUIREMENTS 4.3.6 Each of the above required control rod block trip systems and 1 instrumentation channels.shall be demonstrated OPERABLE

  • by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations

[ :l i

for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.  !

I

  • A channel may be placed_in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition, provided at least one other operable channel in the same trip system is monitoring that parameter.

LIMERICK - UNIT 2 3/4 3-57

U-

S . '. TABLE 4.3.9.l-1 ji .

ffE0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION 4, SURVEILLANCE RE0VIREMENTS OPERATIONAL, CONDITIONS.

CHANNEL. FOR:WHICH-CHANNEL FUNCTIONAL CHANNEL .

' SURVEILLANCE TRIP FUNCTION ' CHECK TEST- CALIBRATION RE0VIRED

1. Reactor' Vessel Water D Q R 1 Level-High,. Level 8 i

2 l

i l

LIMERICK - UNIT 2 3/4 3-115 .

s. m~-- ~ , v-r , ,-,wo- _ - , ,--,r

-e 4 REhCTCRCOOLANTSYSTEM 3/4.4.2 ' SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of at least 11 of the following reactor coolant system safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*#

4 safety relief valves @ 1130 ps g i 1%

5 safety relief valves 0 1140 ps g i 1%

5 safety relief valves 9 1150 ps g i 1%

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With the safety valve function of one or more of the above required safet / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b. With one or more safety / relief valves stuck open pool average water tem t islessthan105f,providedthatsuppression close the stuck open safety / relief valve ; if unable to close the stuck open valve s within 2 minutes or if su) pre ssion (s)pera pool average ure water temperature is 110(F)or greater, place tie reactor mode switch in the Shutdown position.
c. With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable acoustic monitors to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE RE0VIREMENTS 4.4.2.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 0.20 of the full open noise level ## by performance of a:

a. CHANNEL FUNCTIONAL TEST at least once per 92 days, and a l
b. CHANNEL CALIBRATION at least once per 18 months **.

4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and ,

stored in accordance with manufacturer's recommendations at least once per 24 months, and I they shall be rotated such that all 14 safety relief valves are removed, set pressure i tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 54 months.

  • The lift setting pressure shall correspond to ambient conditions of the l valves at nominal operating temperatures and pressures, i
    • The provisions of Specification 4.0.4 are not applicable provided the I Surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is I adequate to perform the test.  !
  1. Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refueling.
    1. Initial setting shall be in accordance with the manufacturer's recommendation. Adjustment to the valve full open noise level shall be accomplished during the startup test program.

LIMERICK - UNIT 2 3/4 4-7

6'.. l

  • CONTAINMENT SYSTEMS-SURVEILLANCE RE0VIREMENTS (Continued)
c. By verifying at least 8 suppression pool water temperature indicators in l at least 8 locations, OPERABLE by performance of .
1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
3. CHANNEL CALIBRATION at least once per 18 months, with the temperature alarm setpoint for:  !
1. High water temperature: l a) First setpoint 5 95'F b) Second setpoint s 105 F c) Third setpoint s 110 F d) Fourth setpoint s 120 F
d. By verifying at least two suppression chamber water level indicators OPERABLE by performance of a:
1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
2. CHANNEL FUNCTIONAL TEST at least once per 92 days, and
3. CHANNEL CALIBRATION at least once per 18 months, with the water level alarm setpoint for high water level s 24'l-1/2"
e. Drywell-to-suppression chamber bypass leak tests shall be conducted at l 40 +/- 10 month intervals to coincide with the ILRT at an initial differential pressure of 4 psi and verifying that the A/(k calculated from the measured leakage is within the specified limit. If any drywell-to-suppression chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 24 months until two consecutive tests meet the specified limit, at which time the test schedule may be resumed.
f. By conducting a leakage test on the drywell-to-suppression chamber l vacuum breakers at a differential pressure of at least 4.0 psi and verifying that the total leakage area A//k contributed by all vacuum  ;

breakers is less than or equal to 24% of the specified limit and the leakage l area for an individual set of vacuum breakers is less than or equal to 12% of  !

the specified limit. The vacuum breaker leakage test shall be conducted during i each refueling outege for which the drywell-to-suppression chamber bypass leak l test in Specification 4.6.2.1.d is not conducted.  !

l LIMERICK - UNIT 2 3/4 6-14

s .

INSTRUMENTATIM i

. BASE $

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION (Continued)  ;

Specified surveillance intervals and maintenance outage times have been  !

determined in accordance with NEDC-30936P, Parts 1 and 2, " Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," as approved by the NRC and documented in the SER (letter to D.

N. Grace from A. C. Thadani dated December 9,1988 (Part 1) and letter to D. N.

Grace from C. E. Rossi dated December 9, 1988 (Part 2)).

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.

3/4.3.4 RECIRCVLATION PVMP TRIP ACTVATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the i plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971, NED0-24222, dated December 1979, and Section 15.8 of the FSAR.

The end-of-cycle recirculation pump trip (E0C-RPT) system is a supplement to  ;

the reactor trip. During turbine trip and generator load rejection events, the E0C-RPT will reduce the likelihood of reactor vessel level decreasing to level

2. Each EOC-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the E0C-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control val %.

A fast closure sensor from each of two turbine control valves provides input to the E0C-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second E0C-RPT system. Simil'.rly, a position switch for each of two turbine stop valves provides input to one E0C-RPT system; a position switch from each of the other two stop valves provides input to the other E0C-RPT system. For each E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the E0C-RPT system and trip both recirculation pumps.

Each E0C-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room.

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.

LIMERICK - UNIT 2 B 3/4 3-3

MSTRUMENTATION I l

BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (continued)

Specified surveillance intervals and maintenance outage times have been I determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance i Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D. Binz, IV, from C.E. Rossi dated July 21,1992).

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.

3/4.3.5 REACTOR CORE ISOLATION CQQlING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel. This instrumentation does not provide actuation of any of the emergency core cooling equipment. l Specified surveillance intervals and maintenance outage times have been specified in accordance with recommendations made by GE in their letter to the BWR Owner's Group dated August 7, 1989,

SUBJECT:

Clarification of Technical Specification changes given in ECCS Actuation Instrumentation Analysis."

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.

)

3/4.3.6 CONTROL R0D BLOCK INSTRUMENTATION i

The control rod block functions are provided consistent with the ,

requirements of the specifications in Section 3/4.1.4, Control Rod Program l Controls and Section 3/4.2 Power Distribution Limits and Section 3/4.3 Instrumentation. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

Specified surveillance intervals and maintenance outage times have been l determined in accordance with NEDC-30851P, Supplement 1, " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"

as approved by the NRC and documented in the SER (letter to D. N. Grace from C. ,

E. Rossi dated September 22,1988).

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.

LIMERICK - UNIT 2 B 3/4 3-4 e-- n e -

1 INSTRUMENTATLOR BASES 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATIQH The OPERABILITY of the radiation monitoring instrumentation ensures that:

(1) the radiation levels are continually measured in the areas served by the 1 individual channels, and (2) the alarm or automatic action is initiated when the  !

radiation level trip setpoint is exceeded; and (3) sufficient informatier. is available on selected plant parameters to monitor and assess th m variable following an accident. This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64. ]

The specified surveillance interval for the Main Control Room Normal Fresh Air Supply Radiation Monitor has been determined in accordance with GENE-770-06-1,

" Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specification," as approved by the NRC and documented in the SER (letter to R. D. Binz, IV, from C. E. Rossi dated July 21,1992).

3/4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This capability is required to permit comparison of the measured response to that used in the design basis for the unit.

3/4.3.7.3 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.

3/4.3.7.4 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown system instrumentation and controls I ensures that sufficient capability is available to permit shutdown and I maintena..se of HOT SHUTDOWN of the unit from locations outside of the control room. This capability is required in the event control room habitability is 4 lost and is consistent with General Design Criterion 19 of 10 CFR Part 50, l Appendix A. The Unit 1 RHR transfer switches are included only due to their potential impact on the RHRSW system, which is common to both units.

3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0737, " Clarification of THI Action Plan Requirements," November 1980.

3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown. At these power levels, reactivity additions shall not be made without this flux level information available to the operator. When the intermediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.

l LIMERICK - UNIT 2 B 3/4 3-5

l

  • INSTRUMENTATION I

BASES y 3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core.

The TIP system OPERABILITY is demonstrated by normalizing all probes (i.e.,

detectors) prior to performing an LPRM calibration function. Monitoring core thermal limits may involve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be required to be OPERABLE. The OPERABILITY of individual detectors to be used for monitoring is demonstrated by comparing the detector (s) output in the resultant heat balance calculation (P-1) with data obtained during a previous heat balance calculation (P-1).  ;

3/4.3.7.8 CHLORINE AND T0XIC GAS DETECTION SYSTEMS The OPERABILITY of the chlorine and toxic gas detection systems ensures I that an accidental chlorine and/or toxic gas release will be detected promptly l and the necessary protective actions will be automatically initiated for chlo- '

rine and manually initiated for toxic gas to provide protection for control room personnel. Upon detection of a high concentration of chlorine, the control room emergency ventilating system will automatically be placed in the chlorine isolation mode of operation to provide the required protection. Upon detection of a high concentration of toxic gas, the control room emergency ventilation l system will manually be placed in the chlorine isolation mode of operation to J provide the required protection. The detection systems required by this speci- d fication are consistent with the recommendations of Regulatory Guide 1.95 " Pro-tection of Nuclear Power Plant Control Room Operators against an Accidental 1 Chlorine Release," February 1975.

Specified surveillance intervals and maintenance outage times have been I determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER -l (letter to R.D. Binz, IV, from C.E. Rossi dated July 21,1992). l 1

3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERABILITY of the detection instrumentation ensures that both adequate warning capability is available for prompt detection of fires and that fire ,

suppression systems, that are actuated by fire detectors, will discharge extin- '

quishing agent in a timely manner. Prompt detection and suppression of fires 4 will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.

Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification.

Consequently, the minimum number of OPERABLE fire detectors must be greater.

LIMERICK - UNIT 2 B 3/4 3-6

S INSTRUMENTAT' ION BASES-l 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION (Continued)

The loss of detection capability for fire suppression systems, actuated by fire detectors, represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initi-ated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

The surveillance requirements for demonstrating the OPERABILITY of the fire detectors are based on the recommendations of NFPA 72E - 1990 Edition.

3/4.3.7.10 LOOSE PART DETECTION SYSTEM The OPERABILITY of the loose-part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system and avoid or mitigate damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the recom-mendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981.

3/4.3.7.11 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.

3/4.3.7.12 0FFGAS MONITORING INSTRUMENTATION This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures and noble gases in the off-gas system.

3/4.3.8. TURBINE OVERSPEED PROTECTION SYSTEM This specification is provided to ensure that the turbine overspeed protection system instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures.

2/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTVATION INSTRUMENTATION The feedwater/ main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system / main turbine trip system in the event of failure of feedwater controller under maximum demand.

LIMERICK - UNIT 2 B 3/4 3-7 .

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