Text

Markup of Proposed Operating License, Technical Specifications & Updated Final Safety Analysis Report Pages
	ML100850395
Person / Time
Site:	Limerick
Issue date:	03/25/2010
From:	; Exelon Generation Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML100850379	List: ML100850391; ML100850392; ML100850395; ML100850398; ML100850399; ML100850401; ML100850403; ML100850406; ML100850410; ML100850411; RIS 2002-03, Attachment 4: NRC Regulatory Issue Summary 2002-03 Cross-Reference; RS-10-002, Request for License Amendment Re Measurement Uncertainty Recapture Power Uprate;
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ATTACHMENT 2 Markup of Proposed Operating License, Technical Specifications, and UFSAR Pages Limerick Generating Station, Units 1 and 2 Facility Operating License Nos. NPF-39 and NPF-85 REVISED OPERATING LICENSE AND TECHNICAL SPECIFICATIONS PAGES Operating License, Page 3 (Units 1 and 2)

TS Page 1-6 (Units 1 and 2)

TS Page 2-4 (Units 1 and 2)

TS Page 3/4 1-19 (Units 1 and 2)

TS Page 3/4 3-5 (Units 1 and 2)

TS Page 3/4 3-7 (Units 1 and 2)

TS Page 3/4 3-8 (Units 1 and 2)

TS Page 3/4 3-46 (Units 1 and 2)

TS Page 3/4 3-48 (Units 1 and 2)

TS Page 3/4 3-60 (Units 1 and 2)

TS Page 3/4 4-1 (Units 1 and 2)

TS Page 3/4 4-2 (Units 1 and 2)

REVISED UFSAR PAGES 7.4-12 7.6-43 9.3-23

(3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility, and to receive and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and fuel channels from the Shoreham Nuclear Power Station.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.D. below) and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level 3515 Exelon Generation Company is authorized to operate the facility at reactor core power levels not in excess of 3458 megawatts thermal (100% rated power) in accordance with the conditions specified herein and in Attachment 1 to this license. The items identified in Attachment 1 to this license shall be completed as specified. Attachment 1 is hereby incorporated into this license.

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 198, are hereby incorporated into this license.

Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 62, 106, 131, 147, 198

(4) Pursuant to the Act and 10 CFR Parts 30, 40, 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility, and to receive and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and fuel channels from the Shoreham Nuclear Power Station.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.D. below) and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level 3515 Exelon Generation Company is authorized to operate the facility at reactor core power levels of 3458 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 159, are hereby incorporated into this license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Fire Protection (Section 9.5, SSER-2, -4)*

Exelon Generation Company shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility, and as approved in the NRC Safety Evaluation Report dated August 1983 through Supplement 9, dated August 1989, and Safety Evaluation dated November 20, 1995, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

The parenthetical notation following the title of license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Amendment No. 1, 27, 51, 68, 92, 108, 159

DEFINITIONS PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3458 MWt.

3515 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper, or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.1.

b. All reactor enclosure secondary containment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d. The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.

e. At least one door in each access to the reactor enclosure secondary containment is closed.

f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.

g. The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.1a.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

RECENTLY IRRADIATED FUEL 1.35 RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All refueling floor secondary containment penetrations required to be closed during accident conditions are either:

LIMERICK - UNIT 1 1-6 Amendment No. 33, 66, 105, 106, 185

DEFINITIONS PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3458 MWt.

3515 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.1.

b. All reactor enclosure secondary containment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d. The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.

e. At least one door in each access to the reactor enclosure secondary containment is closed.

f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.

g. The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.1a.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

RECENTLY IRRADIATED FUEL 1.35 RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All refueling floor secondary containment penetrations required to be closed during accident conditions are either:

LIMERICK - UNIT 2 1-6 Amendment No. 48, 51, 69, 146

TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES

1. Intermediate Range Monitor, Neutron Flux-High 120/125 divisions 122/125 divisions of full scale of full scale

2. Average Power Range Monitor:

a. Neutron Flux-Upscale (Setdown) 15.0% of RATED THERMAL POWER 20.0% of RATED THERMAL POWER

b. Simulated Thermal Power - Upscale:

- Two Recirculation Loop Operation 0.66 W + 62.8% and 0.66 W + 63.3% and 0.65 W + 61.7% 116.6% of RATED 0.65 W + 62.2% 117.0% of RATED THERMAL POWER THERMAL POWER

- Single Recirculation Loop Operation*** 0.66 (W-7.6%) + 62.8% and 0.66 (W-7.6%) + 63.3% and 116.6% of RATED 117.0% of RATED 0.65 (W-7.6%) + 61.5% THERMAL POWER 0.65 (W-7.6%) + 62.0%

THERMAL POWER

c. Neutron Flux - Upscale 118.3% of RATED 118.7% of RATED THERMAL POWER THERMAL POWER

d. Inoperative N.A. N.A.

e. 2-Out-Of-4 Voter N.A. N.A.

f. OPRM Upscale **** N.A.

3. Reactor Vessel Steam Dome Pressure - High 1096 psig 1103 psig

4. Reactor Vessel Water Level - Low, Level 3 12.5 inches above instrument 11.0 inches above zero* instrument zero

5. Main Steam Line Isolation Valve - Closure 8% closed 12% closed

6. DELETED DELETED DELETED

7. Drywell Pressure - High 1.68 psig 1.88 psig

8. Scram Discharge Volume Water Level - High

a. Level Transmitter 260' 9 5/8" elevation** 261' 5 5/8" elevation

b. Float Switch 260' 9 5/8" elevation** 261' 5 5/8" elevation

9. Turbine Stop Valve - Closure 5% closed 7% closed

10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 500 psig 465 psig

11. Reactor Mode Switch Shutdown Position N.A. N.A.

12. Manual Scram N.A. N.A.

See Bases Figure B 3/4.3-1.

- Equivalent to 25.45 gallons/scram discharge volume.

- - The 7.6% flow offset for Single Loop Operation (SLO) is applied for W 7.6%. For flows W < 7.6%, the (W-7.6%) term is set equal to zero.

- - - See COLR for OPRM period based detection algorithm trip setpoints. OPRM Upscale trip output auto-enable (not bypassed) setpoints shall be APRM Simulated Thermal Power 30% and recirculation drive flow < 60%.

29.5%

LIMERICK - UNIT 1 2-4 Amendment No. 7, 30, 66, 89, 106, 141, 177

TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES

1. Intermediate Range Monitor, Neutron Flux-High 120/125 divisions 122/125 divisions of full scale of full scale

2. Average Power Range Monitor:

a. Neutron Flux-Upscale (Setdown) 15.0% of RATED THERMAL 20.0% of RATED POWER THERMAL POWER

b. Simulated Thermal Power - Upscale:

- Two Recirculation Loop Operation 0.66 W + 62.8% and 0.66 W + 63.3% and 116.6% of RATED 0.65 W + 62.2% 117.0% of RATED 0.65 W + 61.7% THERMAL POWER THERMAL POWER

- Single Recirculation Loop Operation*** 0.66 (W-7.6%) + 62.8% and 0.66 (W-7.6%) + 63.3% and 116.6% of RATED 117.0% of RATED 0.65 (W-7.6%) + 61.5% THERMAL POWER 0.65 (W-7.6%) + 62.0% THERMAL POWER

c. Neutron Flux - Upscale 118.3% of RATED 118.7% of RATED THERMAL POWER THERMAL POWER

d. Inoperative N.A. N.A.

e. 2-Out-Of-4 Voter N.A. N.A.

f. OPRM Upscale **** N.A.

3. Reactor Vessel Steam Dome Pressure - High 1096 psig 1103 psig

4. Reactor Vessel Water Level - Low, Level 3 12.5 inches above instrument 11.0 inches above zero* instrument zero

5. Main Steam Line Isolation Valve - Closure 8% closed 12% closed

6. DELETED DELETED DELETED

7. Drywell Pressure - High 1.68 psig 1.88 psig

8. Scram Discharge Volume Water Level - High

a. Level Transmitter 261' 1 1/4" elevation** 261' 9 1/4" elevation

b. Float Switch 261' 1 1/4" elevation** 261' 9 1/4" elevation

See Bases Figure B 3/4.3-1.

- Equivalent to 25.58 gallons/scram discharge volume.

- - The 7.6% flow offset for Single Loop Operation (SLO) is applied for W 7.6%. For flows W < 7.6%, the (W-7.6%) term is set equal to zero.

- - - See COLR for OPRM period based detection algorithm trip setpoints. OPRM Upscale trip output auto-enable (not bypassed) setpoints shall be APRM Simulated Thermal Power 30% and recirculation drive flow < 60%.

29.5%

LIMERICK - UNIT 2 2-4 Amendment No. 48, 51, 52, 109, 139

REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE and consist of a minimum of the following:

a. In OPERATIONAL CONDITIONS 1 and 2, two pumps and corresponding flow paths,

b. In OPERATIONAL CONDITION 3, one pump and corresponding flow path.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 ACTION:

a. With only one pump and corresponding explosive valve OPERABLE, in OPERATIONAL CONDITION 1 or 2, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With standby liquid control system otherwise inoperable, in OPERATIONAL CONDITION 1, 2, or 3, restore the system to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that:

1. The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2. The available volume of sodium pentaborate solution is at least 3160 gallons.

3. The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis.

LIMERICK - UNIT 1 3/4 1-19 Amendment No. 59,66,87,185, 186

REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE and consist of a minimum of the following:

a. In OPERATIONAL CONDITIONS 1 and 2, two pumps and corresponding flow paths,

b. In OPERATIONAL CONDITION 3, one pump and corresponding flow path.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3 ACTION:

a. With only one pump and corresponding explosive valve OPERABLE, in OPERATIONAL CONDITION 1 or 2, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With standby liquid control system otherwise inoperable, in OPERATIONAL CONDITION 1, 2, or 3, restore the system to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that:

1. The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2. The available volume of sodium pentaborate solution is at least 3160 gallons.

3. The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis.

LIMERICK - UNIT 2 3/4 1-19 Amendment No. 24,48,49,146, 147

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) 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 condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b) This function shall be automatically bypassed when the reactor mode switch is in the Run position.

(c) DELETED (d) The noncoincident NMS reactor trip function logic is such that all channels go to both trip systems. Therefore, when the "shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 6 IRMs.

(e) An APRM channel is inoperable if there are less than 3 LPRM inputs per level or less than 20 LPRM inputs to an APRM channel, or if more than 9 LPRM inputs to the APRM channel have been bypassed since the last APRM calibration (weekly gain calibration).

(f) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

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

(j) This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER.

29.5%

(k) Also actuates the EOC-RPT system.

(l) DELETED (m) Each APRM channel provides inputs to both trip systems.

(n) DELETED (o) With THERMAL POWER 25% RATED THERMAL POWER. The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is 30% and recirculation drive flow is < 60%. The OPRM trip output may be automatically bypassed when APRM Simulated Thermal Power is < 30% or recirculation drive flow is 60%. 29.5%

(p) A minimum of 23 cells, each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for an OPRM channel to be OPERABLE.

LIMERICK - UNIT 1 3/4 3-5 Amendment No. 41, 53, 141, 177

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) 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 condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b) This function shall automatically be bypassed when the reactor mode switch is in the Run position.

(c) DELETED (d) The noncoincident NMS reactor trip function logic is such that all channels go to both trip systems. Therefore, when the "shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 6 IRMs.

(e) An APRM channel is inoperable if there are less than 3 LPRM inputs per level or less than 20 LPRM inputs to an APRM channel, or if more than 9 LPRM inputs to the APRM channel have been bypassed since the last APRM calibration (weekly gain calibration).

(f) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

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

(j) This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER.

29.5%

(k) Also actuates the EOC-RPT system.

(l) DELETED (m) Each APRM channel provides inputs to both trip systems.

(n) DELETED (o) With THERMAL POWER 25% RATED THERMAL POWER. The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is 30% and recirculation drive flow is < 60%. The OPRM trip output may be automatically bypassed when APRM Simulated Thermal Power is < 30% or recirculation drive flow is 60%.

29.5%

(p) A minimum of 23 cells, each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for an OPRM channel to be OPERABLE.

LIMERICK - UNIT 2 3/4 3-5 Amendment No. 7, 17, 109, 139

TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK(n) TEST(n) CALIBRATION(a)(n) SURVEILLANCE REQUIRED

1. Intermediate Range Monitors:

a. Neutron Flux - High (b) 2 (j) 3(i), 4(i), 5(i)

b. Inoperative N.A. (j) N.A. 2, 3(i), 4(i), 5(i)

2. Average Power Range Monitor(f): , (o), (p)

a. Neutron Flux - Upscale (Setdown) (b) (l) 2

b. Simulated Thermal Power - Upscale (e) (d), (g) 1

c. Neutron Flux - Upscale (d) 1

d. Inoperative N.A. N.A. 1, 2

e. 2-Out-Of-4 Voter N.A. 1, 2

f. OPRM Upscale (e) (c)(g) 1(m)

3. Reactor Vessel Steam Dome Pressure - High 1, 2(h)

4. Reactor Vessel Water Level-Low, Level 3 1, 2

5. Main Steam Line Isolation Valve - Closure N.A. 1

6. DELETED

7. Drywell Pressure - High 1, 2

8. Scram Discharge Volume Water Level - High

a. Level Transmitter 1, 2, 5(i)

b. Float Switch N.A. 1, 2, 5(i)

LIMERICK - UNIT 1 3/4 3-7 Amendment No. 41, 53, 89, 149, 177, 186

TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK (n) TEST (n) CALIBRATION(a)(n) SURVEILLANCE REQUIRED

1. Intermediate Range Monitors:

a. Neutron Flux - High (b) 2 (j) 3(i), 4(i), 5(i)

b. Inoperative N.A. (j) N.A. 2, 3(i), 4(i), 5(i)

2. Average Power Range Monitor(f):

a. Neutron Flux - Upscale (Setdown) (b) (l) , (o), (p) 2

b. Simulated Thermal Power - Upscale (e) (d), (g) 1

c. Neutron Flux - Upscale (d) 1

d. Inoperative N.A. N.A. 1, 2

e. 2-Out-Of-4 Voter N.A. 1, 2

f. OPRM Upscale (e) (c)(g) 1(m)

3. Reactor Vessel Steam Dome Pressure - High 1, 2(h)

4. Reactor Vessel Water Level-Low, Level 3 1, 2

5. Main Steam Line Isolation Valve - Closure N.A. 1

6. DELETED

7. Drywell Pressure - High 1, 2

8. Scram Discharge Volume Water Level - High

a. Level Transmitter 1, 2, 5(i)

b. Float Switch N.A. 1, 2, 5(i)

LIMERICK - UNIT 2 3/4 3-7 Amendment No. 52, 75, 109, 112, 139, 147

TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK(n) TEST(n) CALIBRATION(a)(n) SURVEILLANCE REQUIRED

9. Turbine Stop Valve - Closure N.A. 1

10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low N.A. 1

11. Reactor Mode Switch Shutdown Position N.A. N.A. 1, 2, 3, 4, 5

12. Manual Scram N.A. N.A. 1, 2, 3, 4, 5 29.5%

(a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Calibration includes verification that the OPRM Upscale trip auto-enable (not-bypass) setpoint for APRM Simulated Thermal Power is 30% and for recirculation drive flow is < 60%.

(d) The more frequent calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 25% of RATED THERMAL POWER.

Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.

(e) CHANNEL FUNCTIONAL TEST shall include the flow input function, excluding the flow transmitter.

(f) The LPRMs shall be calibrated at least once per 2000 effective full power hours (EFPH).

(g) The less frequent calibration includes the flow input function.

(h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

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

(j) If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position.

(k) DELETED (l) Not required to be performed when entering OPERATIONAL CONDITION 2 from OPERATIONAL CONDITION 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering OPERATIONAL CONDITION 2.

(m) With THERMAL POWER 25% of RATED THERMAL POWER.

(n) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

Insert 1 LIMERICK - UNIT 1 3/4 3-8 Amendment No. 29, 41, 53, 66, 113, 117, 131, 141, 177, 186, 195

TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK (n) TEST (n) CALIBRATION(a)(n) SURVEILLANCE REQUIRED

9. Turbine Stop Valve - Closure N.A. 1

10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low N.A. 1

11. Reactor Mode Switch Shutdown Position N.A. N.A. 1, 2, 3, 4, 5

12. Manual Scram N.A. N.A. 1, 2, 3, 4, 5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION. 29.5%

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for a least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.