Text

Limerick Generating Station, Units 1 & 2, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Rev. 2.
	ML17200D096
Person / Time
Site:	Limerick
Issue date:	07/19/2017
From:	Jim Barstow; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML17200D096 (120)
	v • d • e

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90

July 19, 2017

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

Subject:

Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control," Revision 2 In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) requests an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License Nos. NFP-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively.

The proposed changes replace existing TS requirements related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIC) to protect Safety Limit 2.1.4. Safety Limit 2.1.4 requires reactor vessel water level to be greater than the top of active irradiated fuel.

The proposed amendment has been reviewed by the LGS Plant Operations Review Committee in accordance with the requirements of the Exelon Quality Assurance Program.

This amendment request contains no regulatory commitments.

provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides the existing TS Bases pages marked up to show the proposed changes (information only).

Exelon requests approval of the proposed amendment by February 28, 2018 in support of the Spring 2018 Unit 1 refueling outage. Once approved, the amendments shall be implemented within 90 days.

U.S. Nuclear Regulatory Commission Application to Revise TS to Adopt TSTF-542 Docket Nos. 50-352 and 50-353 July 19, 2017 Page 2 In accordance with 1 O CFR 50.91, "Notice for public comment; State consultation," paragraph (b), Exelon is notifying the Commonwealth of Pennsylvania of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

If you have any questions or require additional information, please contact Glenn Stewart at (610) 765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 19 1 h day of July 2017. Respectfully, James Barstow Director, Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments

1. Description and Assessment

2. Markup of Technical Specifications Pages 3. Markup of Technical Specifications Bases Pages (For Information Only) cc: USNRC Region I , Regional Administrator USNRC Senior Resident Inspector, LGS USNRC Project Manager, LGS R. R. Janati, Pennsylvania Bureau of Radiation Protection wl attachments II II II ATTACHMENT 1 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 Docket Nos. 50-352 and 50-353 Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control," Revision 2 Description and Assessment

License Amendment Request Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 1 of 6 Docket Nos. 50-352 and 50-353 Description and Assessment 1.0 DESCRIPTION

Exelon Generation Company, LLC (Exelon), requests an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License Nos. NPF-39 and NPF-

85 for Limerick Generating Station (LGS), Units 1 and 2, respectively.

The proposed changes replace existing TS requirements related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel (RPV) Water Inventory Control (WIC) to protect Safety Limit 2.1.4. Safety Limit 2.1.4 requires reactor vessel water level to be greater than the top of active irradiated fuel.

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation Exelon has reviewed the safety evaluation provided to the Technical Specifications Task Force on December 20, 2016 (Reference 1), as well as the information provided in TSTF-542 (Reference 2). Exelon has concluded that the justifications presented in TSTF-542 and the safety evaluation prepared by the Nuclear Regulatory Commission (NRC) staff are applicable to LGS, Units 1 and 2, and justify this amendment for the incorporation of the changes to the LGS

TS.

The following LGS TS reference or are related to OPDRVs and are affected by the proposed changes: 1.0 Definitions 3.3.2 Isolation Actuation Instrumentation 3.3.3 Emergency Core Cooling System Actuation Instrumentation 3.3.7 Radiation Monitoring Instrumentation3.5.2 ECCS - Shutdown 3.5.3 Suppression Chambe r3.6.5.1.2 Refueling Area Secondary Containment Integrity

3.6.5.2.2 Refueling Area Secondary Containment Automatic Isolation Valves 3.6.5.3 Standby Gas Treatment System

-Common System3.7.2 Control Room Emergency Fresh Air Supply System

-Common System 3.8.1.2 A.C. Sources

- Shutdown3.8.2.2 D.C. Sources

- Shutdown 3.8.3.2 Distribution

- Shutdown

2.2 Variations

Exelon is proposing the following variations from the TS changes described in TSTF-542.

These variations do not affect the applicability of TSTF-542 or the NRC staff's safety evaluation to the proposed license amendment.

LGS TS are based on the previous version of the NRC's Standard TS (NUREG-0123, Revision 2) (Reference 3) and, therefore, the wording and format varies slightly from the NRC Improved Standard Technical Specifications (NUREG-1433) shown in TSTF-542, Revision 2, and the applicable parts of the NRC's safety evaluation. This minor variation is administrative in nature and does not affect the applicability of TSTF-542 to the LGS TS.

License Amendment Request Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 2 of 6 Docket Nos. 50-352 and 50-353 Description and Assessment

In alignment with TSTF-542, Rev. 2, Proposed Safety Basis (Section 3.1.2), the existing LGS TS 3.5.2 requirement to suspend core alterations as an action for Emergency Core Cooling System (ECCS) inoperability is no longer warranted since there are no postulated events associated with core alterations that are prevented or mitigated by the proposed RPV water inventory control requirements. In addition, loss of RPV inventory events are not initiated by core alteration operations. Refueling Limiting Conditions for Operation (LCOs) provide requirements to ensure safe operation during core alterations, including required water level above the RPV flange. Therefore, LGS proposes to delete TS 3.5.2, Action 'b' in its entirety, including the action relating to core alterations.

In alignment with NUREG-1433, Rev. 4, and cons istent with TSTF-542, Rev. 2, LGS proposes to revise TS 3.5.3, "Suppression Chamber," to remove TS requirements associated with OPERATIONAL CONDITIONS (OPCONs) 4 and 5 since they are redundant to the requirements and intent of the newly proposed TS Section 3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control (WIC)." For example, the existing LGS TS LCO 3.5.3.b contains conditions that allow the suppression chamber level to be less than the required 16 feet 0 inches in OPCONs 4 and 5 if the conditions are met, such as maintaining an operable flow path for the Core Spray System to take suction from the Condensate Storage Tank (CST) and ensuring there is sufficient level (29 feet) in the CST. These conditions are satisfied by the proposed LCO 3.5.2.a.2.b). In addition, existing LGS Surveillance Requirement (SR) 4.5.3.1.b

requires verifying that the suppression chamber wate r level is 16 feet 0 inches. This is satisfied by proposed SRs 4.5.2.2 and 4.5.2.3.

Because LGS, Unit 1 and Unit 2 TS are based on NUREG-0123, Revision 2, the current LGS TS in Table 3.3.3-1, "Emergency Core Cooling System Actuation Instrumentation," do not include requirements for the following functions that are listed in TSTF-542: "1b - Core Spray

Pump Discharge Flow-Low (Bypass)" and "2b - Low Pressure Coolant Injection Pump Discharge Flow-Low (Bypass)." Therefore, to align with current LGS instrumentation TS, no requirements were added for these functions as part of the newly proposed TS Table 3.3.3.A-1.

TSTF-542, Table 3.3.5.2-1, "RPV Water Inventory Control Instrumentation," contains Function 2.a, Reactor Steam Dome Pressure - Low (Injection Permissive)," as a permissive for the injection function of the Low Pressure Coolant Injection (LPCI) system in Modes 4 and 5. The current LGS TS Table 3.3.3-1, "Emergency Core Cooling System Actuation Instrumentation," contains a similar Function 2.c, Reactor Vessel Pressure - Low; however, for LGS, this function is only required in OPCONs 1, 2, and 3, and is combined with the Drywell Pressure - High function to provide an automatic initiation signal for LPCI, which is separate from the injection logic. For LGS, the permissive for the injection function of LPCI in OPCONs 4 and 5 from TS Table 3.3.3-1 is Function 2.d, Injection Valve Differential Pressure - Low. This interlock, as determined by monitoring the differential pressure across the injection valve, is to prevent opening the injection valve if reactor pressure is greater than the Residual Heat Removal (RHR) system piping design maximum pressure. Therefore, the new proposed TS Table 3.3.3.A-1, "RPV Water Inventory Control (WIC) Instrumentation," for LGS will include Function 2.a, Injection Valve Differential Pressure - Low (Permissive), for the injection function of the LPCI mode of the RHR system rather than the Reactor Vessel Pressure - Low [Reactor Steam Dome Pressure - Low] function specified in TSTF-542. This variation is consistent with the current LGS TS and operation of the plant, and does not affect the applicability of TSTF-542 to the LGS

TS.

License Amendment Request Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 3 of 6 Docket Nos. 50-352 and 50-353 Description and Assessment LGS TS include Amendment Nos. 216 for Unit 1 and 178 for Unit 2 (Reference 4) for TSTF-523, "Generic Letter 2008-01, Managing Gas Accumulation." As discussed in the Technical Evaluation of TSTF-542, Rev. 2, the changes in TSTF-523 are also applicable to the proposed SRs 4.5.2.4 and 4.5.2.5. Therefore, the following changes are being made to the proposed SRs 4.5.2.4 and 4.5.2.5 based on the changes made to the corresponding LGS SRs in the above referenced amendments that adopted TSTF-523. The following changes have no effect on the adoption of the TSTF-542 and are an acceptable variation in accordance with Section 3.2.4.4 of TSTF-542:

SR 4.5.2.4 has been modified from "Verify, for the required ECCS injection/spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve," to "Verify, for the required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water."

SR 4.5.2.5 has been modified to retain the note: "Not required to be met for system vent flow paths opened under administrative control." During the development of this LAR to adopt TSTF-542, Rev.2, an administrative error was identified within the LGS Index. As part of LGS Amendment Nos. 174 for Unit 1 and 136 for Unit 2 (ADAMS Accession No. ML043220090), the 'E-AVERAGE DISINTEGRATION ENERGY ' definition was deleted. The LGS TS Index is being revised to reflect this deletion. This change is administrative in nature and does not affect the applicability of TSTF-542 to the LGS TS.

The model application provided in TSTF-542 includes an attachment for typed, camera-ready (revised) TS pages reflecting the proposed changes. LGS is not including such an attachment due to the number of TS pages included in this submittal that have the potential to be affected by other unrelated license amendment requests. Providing only mark-ups of the proposed TS changes satisfies the requirements of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," in that the mark-ups fully describe the changes desired. This is administrative in nature and does not affect the applicability of TSTF-542 or the NRC's safety evaluation to t he proposed license amendment.

The LGS TS contain a Surveillance Frequency Control Program (SFCP). Therefore, the SR frequencies for proposed TS 3.5.2 are "in accordance with the Surveillance Frequency Control Program," and the SR frequencies specified in TSTF-542 will be incorporated into the LGS SFCP upon implementation of the proposed amendment.

LGS TS pages 3/4 3-15, 3/4 3-64, and 3/4 3-66 are provided for information only since the table notations where the reference to operations with the potential for draining the reactor vessel are proposed to be deleted refer to the tables on these pages.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Exelon Generation Company, LLC (Exelon), requests an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License Nos. NPF-39 and NPF-

85 for Limerick Generating Station (LGS), Units 1 and 2, respectively.

License Amendment Request Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 4 of 6 Docket Nos. 50-352 and 50-353 Description and Assessment Exelon requests adoption of TSTF-542, "Reactor Pressure Vessel Water Inventory Control,"

which is an approved change to the Standard Technical Specifications (STS), into the LGS TS. The proposed amendment replaces the existing requirements in the TS related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water (RPV) Inventory Control (WIC) to protect Safety Limit 2.1.4. Safety Limit 2.1.4 requires reactor vessel water level to be greater than the top of active irradiated fuel.

Exelon has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes replace existing TS requi rements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.4. Draining of RPV water inventory in OPERATIONAL CONDITION 4 (i.e., cold shutdown) and OPERATIONAL CONDITION 5 (i.e., refueling), is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in OPERATIONAL CONDITION 4 or 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

The proposed changes reduce the probability of an unexpected draining event (which is not a previously evaluated accident) by impos ing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times.

These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.

The proposed changes reduce the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in OPERATIONAL CONDITIONS 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in OPERATIONAL CONDITION 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in OPERATIONAL CONDITIONS 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is

available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.

License Amendment Request Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 5 of 6 Docket Nos. 50-352 and 50-353 Description and Assessment The proposed changes reduce or eliminate some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in OPERATIONAL CONDITIONS 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes replace existing TS requi rements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.4. The proposed changes will not alter the design function of the equipment involved. Under the proposed changes, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements.

The event of concern under the current requirements and the proposed changes is an unexpected draining event. The proposed changes do not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes replace existing TS requi rements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.4. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the TAF in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant License Amendment Request Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 6 of 6 Docket Nos. 50-352 and 50-353 Description and Assessment configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.

Therefore, the proposed changes do not involve a significant reduction in a margin of

safety. Based on the above, Exelon concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.0 ENVIRONMENTAL CONSIDERATION

Exelon has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation," or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

5.0 REFERENCES

1. Final Safety Evaluation of Technical Specifications Task Force Traveler TSTF-542, Revision 2, "Reactor Pressure Vessel Water Inventory Control," dated December 20, 2016 (TAC No. MF3487). ADAMS Accession No.

ML16343B008.

2. TSTF-542, Revision 2, "Reactor Pressure Vessel Water Inventory Control," dated March 14, 2016. ADAMS Accession No.

ML16074A448.

3. NUREG-0123, Revision 2, "Standard Technical Specifications General Electric Boiling Water Reactors (GE-STS)," dated August 1979.

4. Letter from U.S. NRC (R. B. Ennis) to Ex elon (B. Hanson), "Limerick Generating Station, Units 1 and 2 - Issuance of Amendments Re: Adoption of Technical Specification Task

Force (TSTF) Traveler TSTF-523, 'Generic Letter 2008-01, Managing Gas Accumulation' (TAC Nos. MF4412 and MF4413)," dated May 11, 2015. ADAMS Accession No.

ML15083A403.

ATTACHMENT 2 Proposed Technical Specifications Changes (Mark-ups)

Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos.

NPF-39 and NPF-85 Docket Nos. 50-352 and 50-353 Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control," Revision 2 Revised Proposed Technical Specifications Pages Unit 1 TS Pages i vii xii xviii xx 1-2 1-2a* 3/4 3-15 **

3/4 3-16 3/4 3-31 3/4 3-33 3/4 3-35 3/4 3-36 3/4 3-40 3/4 3-41 3/4 3-41a*

3/4 3-41b*

3/4 3-41c*

3/4 3-41d*

3/4 3-41e*

3/4 3-64**

3/4 3-65 3/4 3-66**

3/4 3-67 3/4 5-6 3/4 5-6a*

3/4 5-7 3/4 5-8 3/4 5-9 3/4 6-47 3/4 6-50 3/4 6-52 3/4 7-6 3/4 7-7 3/4 8-9 3/4 8-14a 3/4 8-20 Unit 2 TS Pages i vii xii xviii xx 1-2 1-2a* 3/4 3-15**

3/4 3-16 3/4 3-31 3/4 3-33 3/4 3-35 3/4 3-36 3/4 3-40 3/4 3-41 3/4 3-41a*

3/4 3-41b*

3/4 3-41c*

3/4 3-41d*

3/4 3-41e*

3/4 3-64**

3/4 3-65 3/4 3-66**

3/4 3-67 3/4 5-6 3/4 5-6a*

3/4 5-7 3/4 5-8 3/4 5-9 3/4 6-47 3/4 6-50 3/4 6-52 3/4 7-6 3/4 7-6a 3/4 8-9 3/4 8-14a 3/4 8-19 3/4 8-20

New TS Page

- Information Only TS Page

INDEX DEFINITIONS

SECTION 1.0 DEFINITIONS PAGE

1.1 ACTION

....................................................... 1-1

1.2 AVERAGE PLANAR EXPOSURE

...................................... 1-1

1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE

................... 1-1

1.4 CHANNEL CALIBRATION

.......................................... 1-1

1.5 CHANNEL CHECK

................................................ 1-1

1.6 CHANNEL FUNCTIONAL TEST

...................................... 1-1

1.7 CORE ALTERATION

.............................................. 1-2

1.7A CORE OPERATING LIMITS REPORT

................................. 1-2

1.8 CRITICAL POWER RATIO

......................................... 1-2

1.9 DOSE EQUIVALENT I-131

........................................ 1-2

1.9a DOWNSCALE TRIP SETPOINT (DTSP)

............................... 1-2

1.9b DRAIN TIME

................................................... 1-2

1.10 E-AVERAGE DISINTEGRATION ENERGY(DELETED)

..................... 1-2 a 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME

........... 1-2 a

1.12 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME

.... 1-3 1.13 (DELETED)

.................................................... 1-3

1.14 (DELETED)

.................................................... 1-3

1.15 FREQUENCY NOTATION

........................................... 1-3

1.15a HIGH (POWER) TRIP SETPOINT (HTSP)

............................ 1-3

1.16 IDENTIFIED LEAKAGE

........................................... 1-3

1.16a INTERMEDIATE (POWER) TRIP SETPOINT (ITSP)

.................... 1-3

1.17 ISOLATION SYSTEM RESPONSE TIME

............................... 1-3

1.18 LIMITING CONTROL ROD PATTERN

................................. 1-3

1.19 LINEAR HEAT GENERATION RATE

.................................. 1-3

1.20 LOGIC SYSTEM FUNCTIONAL TEST

................................. 1-4

LIMERICK - UNIT 1 i Amendment No. 33, 37, 66 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)

3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION .......................... 3/4 3-9

Table 3.3.2-1 Isolation Actuation Instrumentation .......................... 3/4 3-11 Table 3.3.2-2 Isolation Actuation Instrumentation Setpoints ................ 3/4 3-18 Table 3.3.2-3 Isolation System Instrumen- tation Response Time ..................... 3/4 3-23 Table 4.3.2.1-1 Isolation Actuation Instrumen- tation Surveillance Requirements ........................... 3/4 3-27 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION .............................................. 3/4 3-32

Table 3.3.3-1 Emergency Core Cooling System Actuation Instrumentation ................ 3/4 3-33

Table 3.3.3-2 Emergency Core Cooling System Actuation Instrumentation Setpoints ................................ 3/4 3-37

Table 3.3.3-3 Emergency Core Cooling System Response Times ........................... 3/4 3-39

Table 4.3.3.1-1 Emergency Core Cooling System Actuation Instrumentation Surveillance Requirements .............. 3/4 3-40 3/4.3.3.A REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC) INSTRUMENTATION .............................................. 3/4 3-41a Table 3.3.3.A-1 RPV Water Inventory Control (WIC)

Instrumentation ......................... 3/4 3-41b

Table 3.3.3.A-2 RPV Water Inventory Control (WIC)

Instrumentation Setpoints ............... 3/4 3-41d

Table 4.3.3.A-1 RPV Water Inventory Control (WIC)

Instrumentation Surveillance Requirements 3/4 3-41e

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION

ATWS Recirculation Pump Trip System Instrumentation .......... 3/4 3-42

Table 3.3.4.1-1 ATWS Recirculation Pump Trip System Instrumentation ................. 3/4 3-43

Table 3.3.4.1-2 ATWS Recirculation Pump Trip System Instrumentation Setpoints .............................. 3/4 3-44

Table 4.3.4.1-1 (Deleted) .............................. 3/4 3-45

End-of-Cycle Recirculation Pump Trip System Instrumentation .............................................. 3/4 3-46 LIMERICK - UNIT 1 vii Amendment No. 33, 186 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION PAGE REACTOR COOLANT SYSTEM (Continued)

3/4.4.9 RESIDUAL HEAT REMOVAL

Hot Shutdown ............................................... 3/4 4-25

Cold Shutdown .............................................. 3/4 4-26

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING ........................................... 3/4 5-1

3/4.5.2 ECCS --- SHUTDOWNREACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC) ............................. 3/4 5-6

3/4.5.3 SUPPRESSION CHAMBER ........................................ 3/4 5-8

3/4.6 CONTAINMENT SYSTEMS

3/4.6.1 PRIMARY CONTAINMENT

Primary Containment Integrity .............................. 3/4 6-1

Primary Containment Leakage ................................ 3/4 6-2

Primary Containment Air Lock ............................... 3/4 6-5

MSIV Leakage Alternate Drain Pathway ....................... 3/4 6-7

Primary Containment Structural Integrity ................... 3/4 6-8

Drywell and Suppression Chamber Internal Pressure .......... 3/4 6-9

Drywell Average Air Temperature ............................ 3/4 6-10

Drywell and Suppression Chamber Purge System ............... 3/4 6-11

3/4.6.2 DEPRESSURIZATION SYSTEMS

Suppression Chamber ........................................ 3/4 6-12

Suppression Pool Spray ..................................... 3/4 6-15

Suppression Pool Cooling ................................... 3/4 6-16

3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES ....................... 3/4 6-17

LIMERICK - UNIT 1 xii Amendment No. 33, 107, 146 INDEX BASES

SECTION PAGE 3/4.0 APPLICABILITY .................................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS

3/4.1.1 SHUTDOWN MARGIN............................................. B 3/4 1-1

3/4.1.2 REACTIVITY ANOMALIES........................................ B 3/4 1-1

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

3/4.1.4 CONTROL ROD PROGRAM CONTROLS................................ B 3/4 1-3

3/4.1.5 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

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

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

3/4.3 INSTRUMENTATION

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

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

3/4.3.3.A REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC) INSTRUMENTATION

............................

B 3/4 3-2a

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

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

3/4.3.7 MONITORING INSTRUMENTATION

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

LIMERICK - UNIT 1 xviii Amendment No. 7, 33, 66, 69 INDEX BASES

SECTION PAGE REACTOR COOLANT SYSTEM (Continued)

3/4.4.5 SPECIFIC ACTIVITY ............................................ B 3/4 4-4

3/4.4.6 PRESSURE/TEMPERATURE LIMITS .................................. B 3/4 4-4

Bases Table B 3/4.4.6-1 Reactor Vessel Toughness ..................... B 3/4 4-7

Bases Figure B 3/4.4.6-1 Fast Neutron Fluence (E>1 MeV) At 1/4 T As A Function of Service Life .......................... B 3/4 4-8

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES ............................. B 3/4 4-6

3/4.4.8 (DELETED) .................................................... B 3/4 4-6

3/4.4.9 RESIDUAL HEAT REMOVAL ........................................ B 3/4 4-6

3/4.5 EMERGENCY CORE COOLING SYSTEMS

3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN

.................... B 3/4 5-1 3/4.5.2 REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC) ................................................ B 3/4 5-3a

3/4.5.3 SUPPRESSION CHAMBER .......................................... B 3/4 5-4

3/4.6 CONTAINMENT SYSTEMS

3/4.6.1 PRIMARY CONTAINMENT

Primary Containment Integrity ................................ B 3/4 6-1

Primary Containment Leakage .................................. B 3/4 6-1

Primary Containment Air Lock ................................. B 3/4 6-1

MSIV Leakage Control System .................................. B 3/4 6-1

Primary Containment Structural Integrity ..................... B 3/4 6-2

Drywell and Suppression Chamber Internal Pressure ................................................... B 3/4 6-2

Drywell Average Air Temperature .............................. B 3/4 6-2

Drywell and Suppression Chamber Purge System ................. B 3/4 6-2

3/4.6.2 DEPRESSURIZATION SYSTEMS ..................................... B 3/4 6-3

LIMERICK - UNIT 1 xx Amendment No. 33, 199 Associated with Amendment 216 DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed

and fuel in the vessel. The following exceptions are not considered to be

CORE ALTERATIONS:

a) Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special

moveable detectors (including undervessel replacement); and b) Control rod movement, provided there are no fuel assemblies in the associated core cell.

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

CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides the core operating limits for the current operating reload

cycle. These cycle-specific core operating limits shall be determined

for each reload cycle in accordance with Specifications 6.9.1.9 thru

6.9.1.12. Plant operation within these limits is addressed in

individual specifications.

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the (GEXL) correlation to

cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same inhalation committed effective dose

equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-

133, I-134, and I-135 actually present. The inhalation committed effective

dose equivalent (CEDE) conversion factors used for this calculation shall be

those listed in Table 2.1 of Federal Guidelines Report 11, "Limiting Values

of Radionuclide Intake and Air Concentration and Dose Conversion Factors for

Inhalation, Submersion, and Ingestion," ORNL, 1989, as described in

Regulatory Guide 1.183. The factors in the column headed "effective" yield

doses corresponding to the CEDE.

DOWNSCALE TRIP SETPOINT (DTSP) 1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.

DRAIN TIME 1.9b The DRAIN TIME is the time it would take for the water inventory in and above the Reactor Pressure Vessel (RPV) to drain to the top of the

active fuel (TAF) seated in the RPV assuming:

a) The water inventory above the TAF is divided by the limiting drain rate; b) The limiting drain rate is the larger of the drain rate through a single penetration flow path with the highest flow rate, or the

sum of the drain rates through multiple penetration flow paths

LIMERICK - UNIT 1 1-2 Amendment No. 37, 66, 87, 174, 185 DEFINITIONS DRAIN TIME (Continued) susceptible to a common mode failure (e.g., seismic event, loss of normal power, single human error), for all penetration flow paths

below the TAF except:

1. Penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank

flanges, or other devices that prevent flow of reactor

coolant through the penetration flow paths;

2. Penetration flow paths capable of being isolated by valves that will close automatically without offsite power prior to

the RPV water level being equal to the TAF when actuated by

RPV water level isolation instrumentation; or

3. Penetration flow paths with isolation devices that can be closed prior to the RPV water level being equal to the TAF

by a dedicated operator trained in the task, who is in

continuous communication with the control room, is stationed

at the controls, and is capable of closing the penetration

flow path isolation device without offsite power.

c) The penetration flow paths required to be evaluated per paragraph b) are assumed to open instantaneously and are not subsequently

isolated, and no water is assumed to be subsequently added to the

RPV water inventory;

d) No additional draining events occur; and

e) Realistic cross-sectional areas and drain rates are used.

A bounding DRAIN TIME may be used in lieu of a calculated value.

1.10 (Deleted)

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-

point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall

include diesel generator starting and sequence loading delays where

applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMERICK - UNIT 1 1-2 a Amendment No. 37, 66, 87, 174, 185

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION MINIMUM APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a),(c) PER TRIP SYSTEM (b) CONDITION ACTION

7. SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level Low, Low - Level 2 B 2 1, 2, 3 25

b. Drywell Pressure - High H 2 1, 2, 3 25

c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High R 2 *# 25

2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High R 2 *# 25

d. Reactor Enclosure Ventilation Exhaust Duct Radiation - High S 2 1, 2, 3 25

e. Deleted

f. Deleted

g. Reactor Enclosure Manual Initiation NA 1 1, 2, 3 24

h. Refueling Area Manual Initiation NA 1

25

LIMERICK - UNIT 1 3/4 3-15 Amendment No. 6, 23, 33, 40, 112

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 - Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24-hours.

ACTION 21 - Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within 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 .

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

ACTION 23 - In OPERATIONAL CONDITION 1 or 2, verify the affected system isolation valves are closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable. In

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

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

the affected system inoperable or be in at least HOT SHUTDOWN within the next

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

ACTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 26 - Close the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

TABLE NOTATIONS

Required when (1) handling RECENTLY IRRADIATED FUEL in the secondary containment , or (2) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

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

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

(a) DELETED

(b) 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. Trip functions common to RPS Actuation Instrumentation are

shown in Table 4.3.2.1-1. In addition, for the HPCI system and RCIC system

isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for

that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8

hours for required surveillance without placing the channel or trip system in the

tripped condition.

LIMERICK - UNIT 1 3/4 3-16 Amendment No. 23,40,53,69,146, 185

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK (a) TEST (a) CALIBRATION(a) SURVEILLANCE REQUIRED

7. SECONDARY CONTAINMENT ISOLATION a. Reactor Vessel Water Level Low, Low - Level 2 1, 2, 3

b. Drywell Pressure## - High 1, 2, 3

c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High *#

2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High *#

d. Reactor Enclosure Ventilation Exhaust Duct Radiation - High 1, 2, 3

e. Deleted

f. Deleted

g. Reactor Enclosure Manual Initiation N.A. N.A. 1, 2, 3

h. Refueling Area Manual Initiation N.A. N.A. *

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

Required when (1) handling RECENTLY IRRADIATED FUEL in the secondary containment , or (2) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel. **When not administratively bypassed and/or when any turbine stop valve is open.

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

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

LIMERICK - UNIT 1 3/4 3-31 Amendment No. 23, 40, 53, 69, 89, 112, 185, 186

TABLE 3.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER APPLICABLE TRIP OPERATIONAL TRIP FUNCTION FUNCTION (a) CONDITIONS ACTION

1. CORE SPRAY SYSTEM***

a. Reactor Vessel Water Level - Low Low Low, Level 1 2/pump (b) 1, 2, 3 , 4*, 5* 30 b. Drywell Pressure - High 2/pump (b) 1, 2, 3 30

c. Reactor Vessel Pressure - Low (Permissive) 6 (b) 1, 2, 3 31 4*, 5* 32 d. Manual Initiation 2 (e) 1, 2, 3 , 4*, 5* 33

2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM***

a. Reactor Vessel Water Level - Low Low Low, Level 1 2 1, 2, 3 , 4*, 5* 30 b. Drywell Pressure - High 2 1, 2, 3 30 c. Reactor Vessel Pressure - Low (Permissive) 2 1, 2, 3 31 d. Injection Valve Differential Pressure-Low 1/valve 1, 2, 3 , 4*, 5* 31 (Permissive)

e. Manual Initiation 1 1, 2, 3 , 4*, 5* 33

3. HIGH PRESSURE COOLANT INJECTION SYSTEM

a. Reactor Vessel Water Level - Low Low Level 2 4 1, 2, 3 34

b. Drywell Pressure - High

1. 1. 4 1, 2, 3 34 c. condensate Storage Tank Level - Low 2 (c) 1, 2, 3 35 d. Suppression Pool Water Level - High 2 1, 2, 3 35 e. Reactor Vessel Water Level - High, Level 8 4 (d) 1, 2, 3 31 f. Manual Initiation

1. 1. 1/system 1, 2, 3 33

LIMERICK - UNIT 1 3/4 3-33 Amendment No. 224 TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION 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) Also provides input to actuation logic for the associated emergency diesel generators.

(c) One trip system. Provides signal to HPCI pump suction valves only.

(d) On 1 out of 2 taken twice logic, provides a signal to trip the HPCI pump turbine only.

(e) The manual initiation push buttons start the respective core spray pump and diesel generator. The "A" and "B" logic manual push buttons also

actuate an initiation permissive in the injection valve opening logic.

(f) A channel as used here is defined as the 127 bus relay for Item 1 and the 127, 127Y, and 127Z feeder relays with their associated time delay

relays taken together for Item 2.

When the system is required to be OPERABLE per Specification 3.5.2.DELETED

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

- Required when ESF equipment is required to be OPERABLE.

1. Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.

1. 1. The injection functions of Drywell Pressure - High and Manual Initiation are not required to be OPERABLE with reactor steam dome

pressure less than 550 psig.

LIMERICK - UNIT 1 3/4 3-35 Amendment No. 53, 224

TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION STATEMENTS

ACTION 30 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the associated

system inoperable.

b. With more than one channel inoperable, declare the associated system inoperable.

ACTION 31 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the

associated ECCS inoperable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 32 -

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

ACTION 33 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the

inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the

associated ECCS inoperable.

ACTION 34 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a. For one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the HPCI system

inoperable.

b. With more than one channel inoperable, declare the HPCI system inoperable.

ACTION 35 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at

least one inoperable channel in the tripped condition within 24

hours or declare the HPCI system inoperable.

ACTION 36 - With the number of OPERABLE channels less than the Total Number of Channels, declare the associated emergency diesel generator

and the associated offsite source breaker that is not supplying

the bus inoperable and take the ACTION required by Specification

3.8.1.1 or 3.8.1.2, as appropriate.

LIMERICK - UNIT 1 3/4 3-36 Amendment No. 11, 53, 158

TABLE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS

TRIP FUNCTION CHANNEL CHECK(a) CHANNEL FUNCTIONAL TEST(a) CHANNEL CALIBRATION(a)

OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REQUIRED

1. CORE SPRAY SYSTEM

a. Reactor Vessel Water Level -

Low Low Low, Level 1 1, 2, 3 , 4*, 5* b. Drywell Pressure - High 1, 2, 3 c. Reactor Vessel Pressure -Low 1, 2, 3 , 4*, 5* d. Manual Initiation N.A.N.A.1, 2, 3 , 4*, 5*

2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM

a. Reactor Vessel Water Level -

Low Low Low, Level 1 1, 2, 3 , 4*, 5* b. Drywell Pressure - High 1, 2, 3 c. Reactor Vessel Pressure -Low 1, 2, 3 d. Injection Valve Differential Pressure - Low (Permissive) 1, 2, 3 , 4*, 5* e. Manual Initiation N.A.N.A.1, 2, 3 , 4*, 5*

3. HIGH PRESSURE COOLANT INJECTION SYSTEM***

a. Reactor Vessel Water Level -

Low Low, Level 2 1, 2, 3

b. Drywell Pressure - High 1, 2, 3 c. Condensate Storage Tank Level -

Low 1, 2, 3

d. Suppression Pool Water Level -

High 1, 2, 3

e. Reactor Vessel Water Level -

High, Level 8 1, 2, 3

f. Manual Initiation N.A.N.A.1, 2, 3

LIMERICK - UNIT 1 3/4 3-40 Amendment No. 53, 71, 186 TABLE 4.3.3.1-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS

TRIP FUNCTION CHANNEL CHECK(a) CHANNEL FUNCTIONAL TEST(a) CHANNEL CALIBRATION(a)

OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REQUIRED

4. AUTOMATIC DEPRESSURIZATION SYSTEM

a. Reactor Vessel Water Level -

Low Low Low, Level 1 1, 2, 3

b. Drywell Pressure - High 1, 2, 3 c. ADS Timer N.A.1, 2, 3

d. Core Spray Pump Discharge Pressure - High 1, 2, 3

e. RHR LPCI Mode Pump Discharge Pressure - High 1, 2, 3

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

g. Manual Initiation N.A.N.A.1, 2, 3

h. ADS Drywell Pressure Bypass TimerN.A.1, 2, 3

5. LOSS OF POWER

a. 4.16 kV Emergency Bus Under-voltage (Loss of Voltage)

1. N.A. N.A. 1, 2, 3, 4**, 5**

b. 4.16 kV Emergency Bus Under -

voltage (Degraded Voltage) 1, 2, 3, 4**, 5**

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

When the system is required to be OPERABLE per Specification 3.5.2.DELETED

- Required OPERABLE when ESF equipment is required to be OPERABLE.

- - Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

1. Loss of Voltage Relay 127-11X is not field setable.

LIMERICK - UNIT 1 3/4 3-41 Amendment No. 53, 186

INSTRUMENTATION 3/4.3.3.A REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC)

INSTRUMENTATION

LIMITING CONDITION FOR OPERATION

3.3.3.A The RPV Water Inventory Control (WIC) instrumentation channels shown in

Table 3.3.3.A-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3.3.A-1

ACTION:

a. With one or more channels inoperable in a trip system, take the ACTION referenced in Table 3.3.3.A-1 for the trip system.

SURVEILLANCE REQUIREMENTS

4.3.3.1.A Each RPV Water Inventory Control (WIC) instrumentation channel shall be

demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST

and LOGIC SYSTEM FUNCTIONAL TEST as shown in Table 4.3.3.A-1 and at the frequencies

specified in the Surveillance Frequency Control Program unless otherwise noted in

Table 4.3.3.A-1.

LIMERICK - UNIT 1 3/4 3-41a Amendment No.

TABLE 3.3.3.A-1 RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER APPLICABLE TRIP OPERATIONAL TRIP FUNCTION FUNCTION CONDITIONS ACTION

1. CORE SPRAY SYSTEM

a. Reactor Vessel Pressure - Low (Permissive) 6 (a) 4, 5 39 b. Manual Initiation 2 (a) 4, 5 40

2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM

a. Injection Valve Differential Pressure - Low (Permissive) 1/valve (a) 4, 5 39 b. Manual Initiation 1 (a) 4, 5 40

3. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water Level -

Low - Level 3 2 in one (b) 38 trip system

4. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. Reactor Vessel Water Level -

Low, Low - Level 2 2 in one (b) 38 trip system

___________________________

(a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, "REACTOR PRESSURE VESSEL (RPV) WATER

INVENTORY CONTROL (WIC)."

(b) When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME.

LIMERICK - UNIT 1 3/4 3-41b Amendment No.

TABLE 3.3.3.A-1 (Continued)

RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION ACTION STATEMENTS

ACTION 38 - Declare the associated trip system for the penetration flow path(s) incapable of automatic isolation and calculate DRAIN TIME.

ACTION 39 - Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , place channel in trip. Otherwise, declare associated low pressure ECCS injection/spray subsystem inoperable.

ACTION 40 - Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , restore channel to OPERABLE status. Otherwise, declare associated low pressure ECCS injection/spray subsystem

inoperable.

LIMERICK - UNIT 1 3/4 3-41c Amendment No.

TABLE 3.3.3.A-2 RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION SETPOINTS TRIP FUNCTION ALLOWABLE VALUE 1. CORE SPRAY SYSTEM

a. Reactor Vessel Pressure -Low(Permissive)>435psig (decreasing)

b. Manual Initiation N.A. 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM

a. Injection Valve Differential Pressure -Low(Permissive)<84psid b. Manual Initiation N.A. 3. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water Level- Low - Level 3 11.0 inches

4. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. Reactor Vessel Water Level - Low, Low - Level 2

-45 inches

LIMERICK - UNIT 1 3/4 3-41d Amendment No.

TABLE 4.3.3.A-1 RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION SURVEILLANCE REQUIREMENTS

TRIP FUNCTION CHANNEL CHECK(a) CHANNEL FUNCTIONAL TEST(a) LOGIC SYSTEM FUNCTIONAL TEST(a) OPERATIONAL CONDITIONS FOR WHICH SURVEILLANCE REQUIRED

1. CORE SPRAY SYSTEM

a. Reactor Vessel Pressure - Low (Permissive)N.A.

4, 5 b. Manual Initiation N.A.N.A.4, 5

2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM

a. Injection Valve Differential Pressure

-Low (Permissive)

N.A.4, 5 b. Manual Initiation N.A.N.A.4, 5

3. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water Level -

Low - Level 3 N.A. (b)

4. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. Reactor Vessel Water Level -

Low, Low - Level 2 N.A. (b)

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

(b) When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME.

LIMERICK - UNIT 1 3/4 3-41e Amendment No.

TABLE 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP INSTRUMENTATION OPERABLE CONDITIONS SETPOINT ACTION

1. Main Control Room Normal 4 1,2,3, 1 x 10

-5 µCi/cc 70 Fresh Air Supply Radiation and

Monitor

2. Area Monitors

a. Criticality Monitors

1) Spent Fuel 2 (a) 5 mR/h and 20mR/h (b) 71 Storage Pool

b. Control Room Direct1 At All Times N.A.(b) 73 Radiation Monitor

3. Reactor Enclosure Cooling Water Radiation Monitor 1 At All Times 3 x Background (b) 72

LIMERICK - UNIT 1 3/4 3-64 Amendment No. 185

TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION TABLE NOTATIONS

When RECENTLY IRRADIATED FUEL is being handled in the secondary containment or during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

(a) With fuel in the spent fuel storage pool.

(b) Alarm only.

ACTION STATEMENTS

ACTION 70 - With one monitor inoperable, restore the inoperable monitor to the OPERABLE status within 7 days or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , initiate and maintain operation of the control room emergency

filtration system in the radiation isolation mode of operation.

With two or more of the monitors inoperable, within one hour, initiate and maintain operation of the control room emergency

filtration system in the radiation mode of operation.

ACTION 71 - With one of the required monitor inoperable, assure a portable continuous monitor with the same alarm setpoint is OPERABLE in

the vicinity of the installed monitor during any fuel movement.

If no fuel movement is being made, perform area surveys of the

monitored area with portable monitoring instrumentation at least

once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 72 - With the required monitor inoperable, obtain and analyze at least one grab sample of the monitored parameter at least once

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

ACTION 73 - With the required monitor inoperable, assure a portable alarming monitor is OPERABLE in the vicinity of the installed monitor or

perform area surveys of the monitored area with portable monitor-

ing instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

LIMERICK - UNIT 1 3/4 3-65 Amendment No. 185

TABLE 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL WHICH SURVEILLANCE INSTRUMENTATION CHECK(c) TEST (c) CALIBRATION(c) IS REQUIRED

1. Main Control Room Normal Fresh Air Supply Radiation Monitor 1, 2, 3, and *

2. Area Monitors

a. Criticality Monitors

1) Spent Fuel Storage (a)

Pool

b. Control Room Direct At All Times Radiation Monitor

3. Reactor Enclosure Cooling Water Radiation Monitor (b) At All Times

LIMERICK - UNIT 1 3/4 3-66 Amendment No. 70, 185, 186

TABLE 4.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS

When RECENTLY IRRADIATED FUEL is being handled in the secondary containment or during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

(a) With fuel in the spent fuel storage pool.

(b) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that

participate in measurement assurance activities with NBS. These

standards shall permit calibrating the system over its intended range of

energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

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

LIMERICK - UNIT 1 3/4 3-67 Amendment No. 185, 186

EMERGENCY CORE COOLING SYSTEMS 3/4 5.2 ECCS -- SHUTDOWN REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC)

LIMITING CONDITION FOR OPERATION

3.5.2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be > 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months AND At least two one of the following shall be OPERABLE:

a. Core spray system (CSS) subsystem s with a subsystem comprised of:

1. Two OPERABLE CSS pumps, and

2. An OPERABLE flow path capable of taking suction from at least one of the following water sources and transferring the water

through the spray sparger to the reactor vessel:

a) From the suppression chamber, or

b) When the suppression chamber water level is less than the limit or is drained, from the condensate storage tank

containing at least 135,000 available gallons of water, equivalent to a level of 29 feet.

b. Low pressure coolant injection (LPCI) system subsystem s with a subsystem comprised of:

1. One OPERABLE LPCI pump, and

2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor

vessel.**

APPLICABILITY: OPERATIONAL CONDITIONS 4 and 5

. ACTION:

a. With one none of the above required subsystems inoperableOPERABLE , restore at least twoone subsystem s to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Otherwise, initiate action to establish a method of water injection capable of operating without offsite electrical power or suspend all operations with a potential for draining the reactor vessel. b. With both of the above required subsystems inoperable, suspend CORE ALTERATIONS and all operations with a potential for draining the reactor vessel. Restore at least one subsystem to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or establish SECONDARY CONTAINMENT INTEGRITY within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months DELETED. _

The ECCS is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded, the spent fuel pool gates are removed, and water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

- One LPCI subsystem may be considered OPERABLE during alignment and operation

for decay heat removal if capable of being manually realigned and not otherwise

inoperable.

LIMERICK - UNIT 1 3/4 5-6 Amendment No. 95

EMERGENCY CORE COOLING SYSTEMS

LIMITING CONDITION FOR OPERATION (Continued)

ACTION: c. With DRAIN TIME less than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and greater than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

1. Verify SECONDARY CONTAINMENT INTEGRITY is capable of being established in less than the DRAIN TIME, 2. Verify each secondary containment penetration flow path is capable of being isolated in less than the DRAIN TIME, and

3. Verify one standby gas treatment subsystem is capable of being placed in operation in less than the DRAIN TIME.

d. With DRAIN TIME less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , immediately:

1. Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV

water level greater than TAF for greater than or equal to

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ,***

2. Initiate action to establish SECONDARY CONTAINMENT INTEGRITY, 3. Initiate action to isolate each secondary containment penetration flow path or verify it can be manually isolated

from the control room, and

4. Initiate action to verify one standby gas treatment subsystem is capable of being placed in operation.

e. With required ACTION and associated allowed outage time for ACTIONs c. or d. not met, or DRAIN TIME less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , initiate action to restore DRAIN TIME to greater than or equal

to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

___________________________

- - The required injection/spray subsystem or an additional method of water injection shall be capable of operating without offsite electrical power.

LIMERICK - UNIT 1 3/4 5-6a Amendment No.

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS

4.5.2.1 Verify DRAIN TIME is greater than or equal to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months in accordance with the Surveillance Frequency Control Program At least the above required ECCS shall be demonstrated OPERABLE per Surveillance Requirement 4.5.1

.* 4.5.2.2 Verify, for a required LPCI subsystem, the suppression pool water level is greater than or equal to 16 feet 0 inches in accordance with the Surveillance

Frequency Control Program The core spray system shall be determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the condensate storage tank required volume when the condensate storage tank is required to be OPERABLE per Specification 3.5.2a.2.b). 4.5.2.3 Verify, for a required CSS subsystem, that the suppression pool water level is greater than or equal to 16 feet 0 inches or the condensate storage tank

water level is greater than or equal to 29 feet 0 inches in accordance with the

Surveillance Frequency Control Program.

4.5.2.4 Verify, for the required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.

4.5.2.5 Verify, for the required ECCS injection/spray subsystem, each manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position in accordance with

the Surveillance Frequency Control Program.

^

4.5.2.6 Operate the required ECCS injection/spray subsystem through the recirculation line for greater than or equal to 10 minutes in accordance with the

Surveillance Frequency Control Program.

4.5.2.7 Verify each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated isolation

signal in accordance with the Surveillance Frequency Control Program.

4.5.2.8 Verify the required ECCS injection/spray subsystem actuates on a manual initiation signal in accordance with the Surveillance Frequency Control

Program.##

___________________________

One LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperableDELETED

. #Not required to be met for system vent flow paths open under administrative control.

^Except that an automatic valve capable of automatic return to its ECCS position

when an ECCS signal is present may be in position for another mode of operation.

1. Vessel injection/spray may be excluded.

LIMERICK - UNIT 1 3/4 5-7 Amendment No. 95, 186

EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 SUPPRESSION CHAMBER

LIMITING CONDITION FOR OPERATION

3.5.3 The suppression chamber shall be OPERABLE:

a. In OPERATIONAL CONDITIONS 1, 2, and 3 with a contained water volume of at least 122,120 ft 3 , equivalent to a level of 22'0".

b. DELETEDIn OPERATIONAL CONDITION 4 and 5* with a contained water volume of at least 88,825 ft 3 , equivalent to a level of 16'0", except that the suppression chamber level may be less than the limit or may be drained provided that:

1. No operations are performed that have a potential for draining the reactor vessel,

2. The reactor mode switch is locked in the Shutdown or Refuel position,

3. The condensate storage tank contains at least 135,000 available gallons of water, equivalent to a level of 29 feet, and

4. The core spray system is OPERABLE per Specification 3.5.2 with an OPERABLE flow path capable of taking suction from the condensate storage tank and transferring the water through the spray sparger to the reactor vessel.

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

a. In OPERATIONAL CONDITION 1, 2 or 3 with the suppression chamber water level less than the above limit, restore the water level to within

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

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

b. DELETEDIn OPERATIONAL CONDITION 4 or 5* with the suppression chamber water level less than the above limit or drained and the above required conditions not satisfied, suspend CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel and lock the reactor mode switch in the Shutdown position. Establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The suppression chamber is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded or being flooded from the suppression pool, the spent fuel pool gates are removed when the cavity is flooded, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

LIMERICK - UNIT 1 3/4 5-8 Amendment No.

EMERGENCY CORE COOLING SYSTEMS

SURVEILLANCE REQUIREMENTS

4.5.3.1 The suppression chamber shall be determined OPERABLE by verifying

the water level to be greater than or equal to, as applicable:

a. 22'0" in accordance with the Surveillance Frequency Control Program.

b. DELETED16'0" in accordance with the Surveillance Frequency Control Program. 4.5.3.2 DELETEDWith the suppression chamber level less than the above limit or drained in OPERATIONAL CONDITION 4 or 5*, in accordance with the Surveillance Frequency Control Program:

a. Verify the required conditions of Specification 3.5.3b. to be satisfied, or

b. Verify footnote conditions

to be satisfied.

The suppression chamber is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded or being flooded from the suppression pool, the spent fuel pool gates are removed when the cavity is flooded, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

LIMERICK - UNIT 1 3/4 5-9 Amendment No. 186 CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT

REFUELING AREA SECONDARY CONTAINMENT INTEGRITY

LIMITING CONDITION FOR OPERATION

3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: When RECENTLY IRRADIATED FUEL is being handled in the secondary containment , or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel. ACTION:

Without REFUELING AREA SECONDARY CONTAINMENT INTEGRITY, suspend handling of RECENTLY

IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

4.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be demonstrated

by:

a. Verifying in accordance with the Surveillance Frequency Control Program that the pressure within the refueling area secondary containment is

greater than or equal to 0.25 inch of vacuum water gauge.

b. Verifying in accordance with the Surveillance Frequency Control Program that:

1. All refueling area secondary containment equipment hatches and blowout panels are closed and sealed.

2. At least one door in each access to the refueling area secondary containment is closed, except when the access opening is being used

for entry and exit.

3. All refueling area secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic iso-

lation dampers/valves and required to be closed during accident

conditions are closed by valves, blind flanges, slide gate

dampers or deactivated automatic dampers/valves secured in

position.

c. In accordance with the Surveillance Frequency Control Program:

Operating one standby gas treatment subsystem for one hour and main-taining greater than or equal to 0.25 inch of vacuum water gauge

in the refueling area secondary containment at a flow rate not exceeding

764 cfm.

LIMERICK - UNIT 1 3/4 6-47 Amendment No. 29,71,185,186, 220

CONTAINMENT SYSTEMS REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES

LIMITING CONDITION FOR OPERATION

3.6.5.2.2 The refueling area secondary containment ventilation system automatic

isolation valves shall be OPERABLE.

APPLICABILITY: When RECENTLY IRRADIATED FUEL is being handled in the secondary containment , or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel. ACTION:

With one or more of the refueling area secondary containment ventilation system

automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE

in each affected penetration that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

a. Restore the inoperable valves to OPERABLE status, or

b. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or

c. Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper.

Otherwise, suspend handling of RECENTLY IRRADIATED FUEL in the refueling area secondary containment , and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

4.6.5.2.2 Each refueling area secondary containment ventilation system auto-

matic isolation valve shall be demonstrated OPERABLE:

a. Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one

complete cycle of full travel and verifying the specified isolation

time.

b. In accordance with the Surveillance Frequency Control Program by verifying that on a containment isolation test signal each isolation valve actuates to

its isolation position.

c. By verifying the isolation time to be within its limit in accordance with the Surveillance Frequency Control Program.

LIMERICK - UNIT 1 3/4 6-50 Amendment No. 6,40,71,105,185, 186

CONTAINMENT SYSTEMS

STANDBY GAS TREATMENT SYSTEM - COMMON SYSTEM

LIMITING CONDITION FOR OPERATION

3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and when (1) irradiated fuel is being handled in the refueling area secondary containment, or (2) during CORE

ALTERATIONS , or (3) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel. ACTION: a. In OPERATIONAL CONDITION 1, 2, or 3:

1. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem 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 and in COLD SHUTDOWN

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

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

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

handling of irradiated fuel in the secondary containment , and CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

2. With both standby gas treatment subsystems inoperable, if in progress, suspend handling of irradiated fuel in the secondary

containment and, CORE ALTERATIONS or operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3. are not applicable.

SURVEILLANCE REQUIREMENTS

4.6.5.3 Each standby gas treatment subsystem shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and

charcoal adsorbers and verifying that the subsystem operates with the

heaters OPERABLE.

LIMERICK - UNIT 1 3/4 6-52 Amendment No. 29,40,185,186, 200

PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION

3.7.2 Two independent control room emergency fresh air supply system subsystems

shall be OPERABLE.

NOTE: The main control room envelope (CRE) boundary may be opened intermittently under administrative control

APPLICABILITY: All OPERATIONAL CONDITIONS and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment , or during operations with a potential for draining the reactor vessel. ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3:

1. With one control room emergency fresh air supply subsystem inoperable for reasons other than Condition a.2, restore the

inoperable subsystem 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 and in COLD SHUTDOWN

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

2. With one or more control room emergency fresh air supply subsystems inoperable due to an inoperable CRE boundary, a. Initiate action to implement mitigating actions immediately or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in

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

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , verify mitigating actions ensure CRE occupant exposures to radiological and chemical hazards

will not exceed limits and actions to mitigate exposure

to smoke hazards are taken or be in HOT SHUTDOWN within

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

hours; and

c. Restore CRE boundary to operable status within 90 days or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD

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

b. In OPERATIONAL CONDITION 4, 5, or when RECENTLY IRRADIATED FUEL is being handled in the secondary containment , or during operations with a potential for draining the reactor vessel

1. With one control room emergency fresh air supply subsystems inoperable for reasons other than Condition b.3, restore the

inoperable subsystem to OPERABLE status within 7 days or initiate

and maintain operation of the OPERABLE subsystem in the radiation

isolation mode of operation.

2. With both control room emergency fresh air supply subsystems inoperable for reasons other than Condition b.3, immediately

suspend handling of RECENTLY IRRADIATED FUEL in the secondary

containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

LIMERICK - UNIT 1 3/4 7-6 Amendment No. 40,71,185,186, 188

PLANT SYSTEMS

LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

3. With one or more control room emergency fresh air subsystems inoperable due to an inoperable CRE boundary, immediately suspend

handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

4.7.2.1 Each control room emergency fresh air supply subsystem shall be

demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying the control room air temperature to be less than or equal to

85°F effective temperature.

b. In accordance with the Surveillance Frequency Control Program on a STAGGERED TEST BASIS by initiating, from the control room, flow through

the HEPA filters and charcoal adsorbers and verifying that the subsystem

operates with the heaters OPERABLE.

c. In accordance with the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal

adsorber housings, or (2) following painting, fire, or chemical release

in any ventilation zone communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than

0.05% and uses the test procedure guidance in Regulatory Positions

C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 3000 cfm +/- 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with

Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 2.5%

when tested in accordance with ASTM D3803-1989 at a temperature

of 30°C (86°F) and a relative humidity of 70%.

3. Verifying a subsystem flow rate of 3000 cfm +/- 10% during subsystem operation when tested in accordance with ANSI

N510-1980.

d. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-

sentative carbon sample obtained in accordance with Regulatory

Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%.

LIMERICK - UNIT 1 3/4 7-7 Amendment No. 5,40,71,144,185, 186, 188 ELECTRICAL POWER SYSTEMS

A.C. SOURCES - SHUTDOWN

LIMITING CONDITION FOR OPERATION

3.8.1.2 As a minimum, the following A.C. electrical power sources shall be

OPERABLE:

a. One circuit between the offsite transmission network and the onsite Class 1E distribution system, and

b. Two diesel generators each with:

1. A day fuel tank containing a minimum of 250 gallons of fuel.

2. A fuel storage system containing a minimum of 33,500 gallons of fuel.

3. A fuel transfer pump.

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and *.

ACTION:

a. With less than the above required A.C. electrical power sources OPERABLE, suspend CORE ALTERATIONS, handling of irradiated fuel in

the secondary containment, operations with a potential for draining the reactor vessel and crane operations over the spent fuel storage pool when fuel assemblies are stored therein. In addition, when in OPERATIONAL CONDITION 5 with the water level less than 22 feet above

the reactor pressure vessel flange, immediately initiate corrective

action to restore the required power sources to OPERABLE status as

soon as practical.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

4.8.1.2 At least the above required A.C. electrical power sources shall be

demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1 and 4.8.1.1.2.

When handling irradiated fuel in the secondary containment.

LIMERICK - UNIT 1 3/4 8-9 Amendment No. 32, 192, 193

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

2. Division 1 or 2 with float current > 2 amps, or with Division 3 or 4 with float current > 1 amp, perform 4.8.2.1.a.2 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for affected battery(s) and restore battery float current to within limits within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

3. One or two batteries on one division with one or more cells electrolyte level less than minimum established design limits, if

electrolyte level was below the top of the plates restore electrolyte

level to above top of plates within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and verify no evidence of

leakage(*) within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In all cases, restore electrolyte level

to greater than or equal to minimum established design limits within

31 days.

4. One or two batteries on one division with pilot cell electrolyte temperature less than minimum established design limits, restore

battery pilot cell temperature to greater than or equal to minimum

established design limits within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

5. Batteries in more than one division affected, restore battery parameters for all batteries in one division to within limits within

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

6. (i) Any battery having both (Action b.1) one or more battery cells float voltage < 2.07 volts and (Action b.2) float current not

within limits, and/or

(ii) Any battery not meeting any Action b.1 through b.5,

Restore the battery parameters to within limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

c. 1. With the requirements of Action a. and/or Action b. not met, or

2. With less than two divisions of the above required D.C. electrical power sources OPERABLE for reasons other than Actions a. and/or b.,

Suspend CORE ALTERATIONS and, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel. d. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

4.8.2.2 At least the above required battery and charger shall be demonstrated

OPERABLE per Surveillance Requirement 4.8.2.1.

______________________

(*) Contrary to the provisions of Specification 3.0.2, if electrolyte level was below the top of the plates, the verification that there is no evidence of

leakage is required to be completed regardless of when electrolyte level is

restored.

LIMERICK - UNIT 1 3/4 8-14a Amendment No. 164

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and *.

ACTION:

a. With less than two divisions of the above required Unit 1 A.C. dis-tribution systems energized, suspend CORE ALTERATIONS and, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel. b. With less than two divisions of the above required Unit 1 D.C. dis-tribution systems energized, suspend CORE ALTERATIONS and, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel. c. With any of the above required Unit 2 and common AC and/or DC distribution system divisions not energized, declare the associated

common equipment inoperable, and take the appropriate ACTION for that

system.

d. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

4.8.3.2 At least the above required power distribution system divisions shall

be determined energized in accordance with the Surveillance Frequency Control

Program by verifying correct breaker alignment and voltage on the

busses/MCCs/panels.

When handling irradiated fuel in the secondary containment.

LIMERICK - UNIT 1 3/4 8-20 Amendment No. 24, 186

INDEX DEFINITIONS

SECTION 1.0 DEFINITIONS PAGE

1.1 ACTION

....................................................... 1-1

1.2 AVERAGE PLANAR EXPOSURE

...................................... 1-1

1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE

................... 1-1

1.4 CHANNEL CALIBRATION

.......................................... 1-1

1.5 CHANNEL CHECK

................................................ 1-1

1.6 CHANNEL FUNCTIONAL TEST

...................................... 1-1

1.7 CORE ALTERATION

.............................................. 1-2

1.7A CORE OPERATING LIMITS REPORT

................................. 1-2

1.8 CRITICAL POWER RATIO

......................................... 1-2

1.9 DOSE EQUIVALENT I-131

........................................ 1-2

1.9a DOWNSCALE TRIP SETPOINT (DTSP)

............................... 1-2

1.9b DRAIN TIME

................................................... 1-2

1.10 E-AVERAGE DISINTEGRATION ENERGY (DELETED) ................... 1-2 a 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME

........... 1-2 a

1.12 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME

.... 1-3 1.13 (DELETED)

.................................................... 1-3

1.14 (DELETED)

.................................................... 1-3

1.15 FREQUENCY NOTATION

........................................... 1-3

1.15a HIGH (POWER) TRIP SETPOINT (HTSP)

............................ 1-3

1.16 IDENTIFIED LEAKAGE

........................................... 1-3

1.16a INTERMEDIATE (POWER) TRIP SETPOINT (ITSP)

.................... 1-3

1.17 ISOLATION SYSTEM RESPONSE TIME

............................... 1-3

1.18 LIMITING CONTROL ROD PATTERN

................................. 1-3

1.19 LINEAR HEAT GENERATION RATE

.................................. 1-3

1.20 LOGIC SYSTEM FUNCTIONAL TEST

................................. 1-4

LIMERICK - UNIT 2 i Amendment No. 4, 48 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)

3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION .......................... 3/4 3-9

Table 3.3.2-1 Isolation Actuation Instrumentation ...... 3/4 3-11 Table 3.3.2-2 Isolation Actuation Instrumentation Setpoints ................ 3/4 3-18 Table 3.3.2-3 Isolation System Instrumentation Response Time ............................ 3/4 3-23 Table 4.3.2.1-1 Isolation Actuation Instrumentation Surveillance Requirements .............. 3/4 3-27 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION .............................................. 3/4 3-32

Table 3.3.3-1 Emergency Core Cooling System Actuation Instrumentation ................ 3/4 3-33

Table 3.3.3-2 Emergency Core Cooling System Actuation Instrumentation Setpoints ...... 3/4 3-37

Table 3.3.3-3 Emergency Core Cooling System Response Times ........................... 3/4 3-39

Table 4.3.3.1-1 Emergency Core Cooling System Actuation Instrumentation Surveillance Requirements .............. 3/4 3-40 3/4.3.3.A REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC)

INSTRUMENTATION .............................................. 3/4 3-41a Table 3.3.3.A-1 RPV Water Inventory Control (WIC)

Instrumentation ......................... 3/4 3-41b

Table 3.3.3.A-2 RPV Water Inventory Control (WIC)

Instrumentation Setpoints ............... 3/4 3-41d

Table 4.3.3.A-1 RPV Water Inventory Control (WIC)

Instrumentation Surveillance Requirements 3/4 3-41e 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION

ATWS Recirculation Pump Trip System Instrumentation .......... 3/4 3-42

Table 3.3.4.1-1 ATWS Recirculation Pump Trip System Instrumentation ................. 3/4 3-43

Table 3.3.4.1-2 ATWS Recirculation Pump Trip System Instrumentation Setpoints .............................. 3/4 3-44

Table 4.3.4.1-1 (Deleted) .............................. 3/4 3-45

End-of-Cycle Recirculation Pump Trip System Instrumentation .............................................. 3/4 3-46

LIMERICK - UNIT 2 vii Amendment No. 147 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION PAGE REACTOR COOLANT SYSTEM (Continued)

3/4.4.9 RESIDUAL HEAT REMOVAL

Hot Shutdown ............................................... 3/4 4-25

Cold Shutdown .............................................. 3/4 4-26

3/4.5 EMERGENCY CORE COOLING SYSTEMS

3/4.5.1 ECCS - OPERATING ........................................... 3/4 5-1

3/4.5.2 ECCS --- SHUTDOWNREACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC)

.............................. 3/4 5-6