ML053550555

License Amendment Request, Relocation of the Reactor Coolant System Pressure Isolation Valve Table from Technical Specifications to the Technical Requirements Manual

ML053550555
Person / Time
Site:	Limerick
Issue date:	12/21/2005
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML053550555 (35)

Text

Exelon Nuclear 200 Exelon Way www.exe1oncorp corn Nuclear Kennett Square, PA 19348 10 CFR 50.90 December 21,2005 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Limerick Generating Station, Units 1 and 2 Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

SUBJECT:

License Amendment Request Relocation of the Reactor Coolant System Pressure Isolation Valve Table from Technical Specifications to the Technical Requirements Manual Pursuant to 10 CFR 50.90, Exelon Generation Company, LLC (Exelon) hereby requests a change to the Technical Specifications (TS), Appendix A, of Facility Operating License Nos.

NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively.

The amendment revises the Appendix A TS relating to reactor coolant system leakage.

Specifically, the amendment deletes Table 3.4.3.2-1, Reactor Coolant System Pressure Isolation Valves from the LGS TS, Units 1 and 2. The information contained in Table 3.4.3.2-1 is to be relocated to the Limerick Technical Requirements Manual (TRM). In addition, the references to Table 3.4.3.2-1 are being removed from TS Limiting Condition for Operation 3.4.3.2.e and TS ACTION 3.4.3.2.d, as well as from Surveillance Requirement 4.4.3.2.2 and 4.4.3.2.3.

The proposed change is consistent with Generic Letter 91-08, Removal of Component Lists from Technical Specifications, which provides guidance to remove component lists from the Technical Specifications. This request meets all conditions outlined in the Generic Letter.

Additionally, the revision does not alter the requirements for pressure isolation valve and alarm instrumentation operability currently in the Technical Specifications. The LCO and Surveillance Requirements will be retained in the revised Technical Specifications and the proposed change will not affect the meaning, application, and function of the current Technical Specification requirements for the Pressure Isolation Valves in Table 3.4.3.2-1.

The approach for this revision is consistent with the Improved Standard Technical Specifications (STS) described in NUREG-1433, Standard Technical Specifications, General Electric Plants (BWW4), and changes previously approved by the NRC for other licensees, including Seabrook and River Bend stations.

License Amendment Request Docket Nos. 50-352, 353 December 21,2005 Page 2 to this letter describes the proposed changes and provides justification for the changes. Attachments 2 and 3 to this letter provide the marked-up Technical Specifications and marked-up Technical Specifications Bases pages, respectively. Attachments 4 and 5 to this letter provide the typed Technical Specifications and Technical Specifications Bases pages, respectively.

Exelon has concluded that the proposed changes present no significant hazards consideration under the standards set forth in 10CFR 50.92.

Exelon requests approval of the proposed amendment by March 1,2006 to support the upcoming Unit 1 1R11 Refueling outage. Upon NRC approval, the amendment shall be I

implemented within 30 days of issuance.

These proposed changes have been reviewed by the Plant Operations Review Committee. We are notifying the State of Pennsylvania of this application for changes to the TS 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 Doug Walker at 610-765-5726.

I declare under penalty of perjury that the foregoing is true and correct.

Respectfully, Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1 Evaluation of the Proposed Change 2 Marked-Up Technical Specifications Pages 3 Marked-Up Technical Specifications Bases Pages 4 Typed Technical Specifications Pages 5 Typed Technical Specifications Bases Pages cc: Regional Administrator - NRC Region I w/attachments NRC Senior Resident Inspector - Limerick Generating Station I, NRC Project Manager, NRR - Limerick Generating Station I6 Director, Bureau of Radiation Protection - Pennsylvania Department r:

of Environmental Protection

Attachment 1 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Relocation of the Reactor Coolant System Pressure Isolation Valve Table from Technical Specifications to the Technical Requirements Manual EVALUATION OF PROPOSED CHANGE

License Amendment Request Docket Nos. 50-352,353 ATTACHMENT 1 EVALUATION OF PROPOSED CHANGE 1.O DESCRIPTION In accordance with 10 CFR 50.90, Application for amendment of license or construction permit, Exelon Generation Company, LLC (i.e., Exelon) requests changes to Technical Specifications (TS), Appendix A, of Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively.

The amendment revises the Appendix A, TS and the associated Bases relating to reactor coolant system leakage. Specifically, the amendment deletes Table 3.4.3.2-1, Reactor Coolant System Pressure Isolation Valves from the LGS Units 1 and 2 TS. In addition, the references to Table 3.4.3.2-1 are being removed from TS Limiting Condition for Operation 3.4.3.2.e and TS ACTION 3.4.3.2.d, as well as from Surveillance Requirement (SR) 4.4.3.2.2 and 4.4.3.2.3.

The information contained in Table 3.4.3.2-1 would be relocated to the Technical Requirements Manual (TRM). Relocating the Table from the TS will eliminate the burden of processing License Amendments when changes are made to the Pressure Isolation Valve (PIV) Table and will facilitate the more effective utilization of NRC and Exelon resources.

The proposed change is consistent with Generic Letter 91-08, Removal of Component Lists from Technical Specifications, which provides guidance to remove component lists from the Technical Specifications. This request meets all conditions outlined in the Generic Letter.

Additionally, the revision does not alter the requirements for pressure isolation valve and alarm instrumentation operability currently in the Technical Specifications. The LCO and Surveillance Requirements will be retained in the revised Technical Specifications and the proposed change will not affect the meaning, application, and function of the current Technical Specification requirements for the PlVs in Table 3.4.3.2-1.

The proposed change is consistent with the improved Standard Technical Specifications (STS) described in NUREG-1433, Standard Technical Specifications, General Electric Plants (BWW4), and changes previously approved by the NRC for other licensees including River Bend and Seabrook stations.

2.0 PROPOSED CHANGE

LGS has separate TS for Unit 1 and Unit 2; however, the proposed change is identical for both units.

TS Section 3/4.4.3.2, Reactor Coolant System, Operational Leakage contains the Limiting Conditions for Operation (LCO), Actions, and Surveillance Requirements applicable to Reactor Coolant System (RCS) operational leakage and includes RCS PIV leakage requirements. Table 3.4.3.2-1, Reactor Coolant System Pressure Isolation Valves contains a list of PIVs, as well as Alarm Setpoints and Alarm Allowable Values. The following changes are proposed to support relocation of Table 3.4.3.2-1 to the Limerick TRM:

License Amendment Request Docket Nos. 50-352, 353 Attachment 1 Evaluation of Proposed Change Page 2 of 8

1. TS Index, page xi, will be revised to delete reference to Table 3.4.3.2-1

2. TS LCO 3.4.3.2.e will be revised to delete reference to Table 3.4.3.2-1.

3. TS Action 3.4.3.2.d will be revised to delete reference to Table 3.4.3.2-1.

4. TS SR 4.4.3.2.2 will be revised to delete reference to Table 3.4.3.2-1.

5. TS SR 4.4.3.2.3 will be revised to delete reference to Table 3.4.3.2-1.

6. TS Table 3.4.3.2-1 will be deleted and will be relocated in its entirety to the Limerick TRM.

7. TS Bases 3/4.4.3.2 will be revised to replace the reference to Table 3.4.3.2-1 with a reference to the TRM.

The current RCS PIV leakage specifications themselves, i.e., LCO 3.4.3.2.e, TS Action 3.4.3.2.d, and Surveillance Requirements 4.4.3.2.2 and 4.4.3.2.3, will remain unchanged except for deletion of the references to Table 3.4.3.2-1.

3.0 BACKGROUND

The function of RCS PlVs is to separate the high pressure RCS from an attached low pressure system. This protects the RCS pressure boundary described in 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A. PlVs are described in NUREG-1433 as any two normally closed valves in series within the reactor coolant pressure boundary (RCPB).

The RCS PIV LCO allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety. The PIV leakage limit applies to each individual valve. Leakage through these valves is not included in any allowable LEAKAGE specified in LCO 3.4.3.2.

Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit provides indication that the PlVs between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components.

In May 1991, Generic Letter 91-08, Removal of Component Lists from Technical Specifications, was issued to provide guidance to remove component lists from the Technical Specifications. The guidance stipulates that the TS requirements are stated in general terms that describe the types of components to which the requirements apply, and that the removal of component lists do not alter existing TS requirements or those components to which they apply.

In addition, the removed lists must be included in a plant procedure that is subject to the change control provisions for plant procedures in the Administrative Controls section of TS.

License Amendment Request Docket Nos. 50-352,353 Attachment 1 Evaluation of Proposed Change Page 3 of 8 Generic Letter 91-08 provides guidance for preparing a request for a license amendment to remove component lists from technical specifications (TS). The nuclear industry and the U.S.

Nuclear Regulatory Commission (NRC) identified this line-item TS improvement during investigations of TS problems.

The removal of component lists from TS permits administrative control of changes to these lists without processing a license amendment. Any change to component lists contained in plant procedures is subject to the requirements specified in the Administrative Controls section of the TS on changes to plant procedures. Therefore, the change control provisions of the TS provide an adequate means to control changes to these component lists, when they have been incorporated into plant procedures, without including them in TS.

An Enclosure to the Generic Letter provided additional guidance for changing individual TS sections. At the time of issuance in 1991, the Enclosure to GL 91-08 specifically addressed the issue of PlVs stating:

Guidance on removing from the TS the list of reactor coolant system pressure isolation valves is pending the NRC staffs resolution of generic concerns with existing lists for these valves. In the interim, licensees should not submit proposals to remove this list from the TS.

The NRC has since resolved the Generic Safety Issue referenced in the GL Enclosure. On July 1, 1993, NUREG 1463, Regulatory Analysis for the Resolution of Generic Safety Issue 105:

Interfacing System Loss-of-Coolant Accident in Light-Water Reactors was issued. The NUREG addressed the outstanding Generic Safety Issue (GSI) 105 regarding Interfacing Systems Loss-of-Coolant Accident (ISLOCA) and PIVs. Additionally, the NRC has since approved NUREG-1433, Standard Technical Specifications General Electric Plants, BWW4, which does not include PIV Tables. In addition, the NRC has approved specific LARSfor relocation of PIV Tables from TS.

4.0 TECHNICAL ANALYSIS

Exelon proposes to delete Table 3.4.3.2-1, Reactor Coolant System Pressure Isolation Valves, and any reference thereto from TS 3.4.3.2, and to relocate the PIV Table to the Limerick TRM.

The TRM is an Exelon controlled document that has been developed to contain requirements relocated from the TS. Revisions to the TRM are reviewed pursuant to 10 CFR 50.59, and summaries of changes are provided to the NRC in the periodic 10 CFR 50.59 report.

Relocating the table from the TS will eliminate the burden of processing license amendments when changes are made to the PIV Table and will facilitate the more effective utilization of NRC and Exelon resources.

On May 6, 1991, the Commission issued Generic Letter 91-08 (GL 91-08) relating to the issue of removing component lists from the TS. GL 91-08 stated in part:

This guidance includes the incorporation of lists into plant procedures that are subject to the change control provisions for plant procedures In the Administrative Controls section of the TS.

License Amendment Request Docket Nos. 50-352, 353 Attachment 1 Evaluation of Proposed Change Page 4 of 8 The removal of component lists from TS permits administrative control of changes to these lists without processing a license amendment, as is required to update TS component lists.

Any change to component lists contained in plant procedures is subject to the requirements specified in the Administrative Controls section of the TS on changes to plant procedures.

Therefore, the change control provisions of the TS provide an adequate means to control changes to these component lists, when they have been incorporated into plant procedures, without including them in TS.

Specific issues identified in Enclosure 1 to GL 91-08 to be addressed with a request to remove component lists from the TS include:

1. Each TS should include an appropriate description of the scope of the components to which the TS requirements apply. Components that are defined by regulatory requirements or guidance need not be clarified further. However, the Bases section of the TS should reference the applicable requirements or guidance.

2. If the removal of a component list results in the loss of notes that modify or provide an exception to the TS requirements, the specification should be revised to incorporate that modification or exception. The modification or exception should be stated in terms that identify any group of components by function rather than by plant identification number, if practical.

3. Licensees should confirm that the lists of components removed from the TS are located in appropriately controlled plant procedures. The list of components may be included in the next update of the FSAR. The Bases section of individual specifications also may reference the plant procedures or other documents that identify each component list.

With regard to item (1) above, PlVs are described in NUREG-1433 as any two normally closed valves in series within the reactor coolant pressure boundary. The TS requirements for LCO, Actions, and SR relating to PlVs remain applicable. Therefore, deletion of reference to Table 3.4.3.2-1 from LCO 3.4.3.2.e does not affect the scope of components to which the TS requirements apply. Per the proposed changes in Attachment 3, the TS Bases now describe PIVs, which is consistent with the NUREG-1433 Standard Technical Specifications, General Electric Plants, BWW4 Bases. The same argument applies to removal of the reference to Table 3.4.3.2-1 from TS Action 3.4.3.2.d, as well as surveillance requirements 4.4.3.2.2 and 4.4.3.2.3.

With regard to item (2) above, there are no notes, exceptions, or modifications listed directly in the PIV Table 3.4.3.2-1.

With regard to item (3), Exelon will confirm, prior to implementation of this LAR that the list of PlVs is located in the TRM, which is an appropriately controlled plant procedure.

Generic Letter 91-08 provided the guidance for changing individual TS sections. The guidance written in the Generic Letter was written prior to the resolution of GSI 105, which discusses Interfacing Systems Loss of Coolant Accidents. The enclosure to GL 91-08 specifically addresses the issue of PlVs and this GSI stating:

License Amendment Request Docket Nos. 50-352,353 Attachment 1 Evaluation of Proposed Change Page 5 of 8 Guidance on removing from the TS the list of reactor coolant system pressure isolation valves is pending the NRC staffs resolution of generic concerns with existing lists for these valves. In the interim, licensees should not submit proposals to remove this list from the TS.

Explicit guidance on removal of lists of PlVs from the TS has not been issued by the NRC.

However, in July 1993, the NRC issued NUREG-1463, Regulatory Analysis for the Resolution of Generic Safety Issue 105: Interfacing System Loss-of-CoolantAccident (LOCA) in Light-Water Reactors.

NUREG-1463 concluded the most viable course of action to resolve Generic Issue 105 is licensee participation in individual plant examinations (IPEs). Limerick Individual Plant Examination (IPE) was completed in response to GL 88-20, which was accepted by the NRC (ref. NRC Review of LGS IPE Submittal, Letter dated December 19,1994, from NRC to G. A.

Hunger).

Additionally, NUREG-1443, Standard Technical Specifications, General Electric Plants, BWW4, I does not contain a list of PIVs. The list of PlVs is not included in STS section 3.4.5, RCS Pressure Isolation Valve (PIV) Leakage, and STS Basis section 3.4.5 indicates that PlVs are listed in the FSAR.

Exelon concludes that the proposed change to TS 3.4.3.2 is administrative in that it merely relocates the PIV Table from the TS to the TRM and maintains the requirements for PIV testing and the acceptance criteria for the testing in the Limiting Condition for Operation 3.4.3.2.e.

Equipment test methods, frequencies, and acceptance criteria are not affected by this proposed change.

Exelon determined that the relocation of Table 3.4.3.2-1 does not eliminate the requirements for the licensee to ensure that the RCS pressure isolation valves are capable of performing their safety function. Although Table 3.4.3.2-1 is relocated from the TSs to the TRM, the information being relocated will be controlled and further revisions to the TRM Table will be subject to evaluation pursuant to 10 CFR 50.59.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Exelon has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No.

The proposed relocation of Technical Specification Table 3.4.3.2-1 does not alter the requirements for pressure isolation valve operability or surveillance currently in the Technical Specifications. The proposed change to remove the pressure isolation valve

License Amendment Request Docket Nos. 50-352,353 Attachment 1 Evaluation of Proposed Change Page 6 of 8 table from TS and relocate the information to an administratively controlled document, and to revise the wording in TS to reflect this change, will have no impact on any safety related structures, systems or components. The probability of occurrence of a previously evaluated accident is not increased because this change does not introduce any new potential accident initiating conditions. The consequences of accidents previously evaluated in the UFSAR are not affected because the ability of the PlVs to limit leakage through these valves in amounts that do not compromise safety is not affected. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes are administrative in nature and do not result in physical alterations or changes in the method by which any safety related system performs its intended function@). The proposed changes do not impact any safety analysis assumptions. The proposed changes do not create any new accident initiators or involve an activity that could be an initiator of an accident of a different type.

All PlVs and alarm instrumentation will continue to be tested to the same rigorous requirements as defined in the Technical Specification Surveillance Requirements. The proposed revision does not make changes in any method of testing or how any safety related system performs its safety functions. Therefore, the possibility of an accident of a different type than any previously evaluated in the UFSAR is not created.

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

Response: No.

The administrative change to relocate Technical Specification Table 3.4.3.2-1 to the Technical Requirements Manual does not alter the basic regulatory requirement for Reactor Coolant System pressure isolation and will not affect the isolation capability for credible accident scenarios. Future revisions to the Technical Requirements Manual Table will be subject to evaluation pursuant to 10CFR50.59.

Additionally, the proposed relocation does not alter the requirements for pressure isolation valve and alarm instrumentation operability currently in the Technical Specifications. The LCO and Surveillance Requirements will be retained in the revised Technical Specifications. The proposed change will not affect the meaning, application, and function of the current Technical Specification requirements for the valves in Table 3.4.3.2-1. Therefore, the proposed changes do not result in a significant reduction in the margin of safety.

License Amendment Request Docket Nos. 50-352,353 Attachment 1 Evaluation of Proposed Change Page 7 of 8 Based on the above, Exelon concludes that the proposed changes present 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.

5.2 Applicable Regulatory Requirements General Design Criterion (GDC) 30, Quality of Reactor Coolant Pressure Boundary, of Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants, requires that means be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage. As previously noted, the proposed change involves relocating Technical Specification Table 3.4.3.2-1 to the Technical Requirements Manual. The proposed change does not involve a change in the design or operation of the RCS leakage detection system which will continue to meet the requirements of GDC 30.

The RCS operational leakage limits and PIV leakage limits protect the Reactor Coolant Pressure Boundary (RCPB) described in 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A. The current RCS operational leakage LCOs, Applicability, Actions, and Surveillance Requirements for RCPB leakage, unidentified leakage and total leakage will remain unaltered in TS Section 3/4.4.3.2. The current LCO, Applicability, Actions, and Surveillance Requirements for RCS PIV Leakage will also remain unaltered in TS Section 3/4.4.3.2.

The administrative change to relocate Technical Specification Table 3.4.3.2-1 to the Technical Requirements Manual was generically approved by the NRC in NUREG-1433, Standard Technical Specifications, General Electric Plants (BW W4), which is consistent with the NRC Final Policy Statement and the revised 10 CFR 50.36 rule. The proposed change is consistent with TS changes previously approved by the NRC for other stations including Seabrook and River Bend stations.

Generic Letter 91-08, Removal of Component Lists from Technical Specifications, provides guidance to remove component lists from the Technical Specifications.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review 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 10CFR20, 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 the individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical

License Amendment Request Docket Nos. 50-352,353 Attachment 1 Evaluation of Proposed Change Page 8 of 8 exclusion set forth in 10 CFR 51.22(~)(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.

7.0 REFERENCES

1. 10 CFR 50.36, Technical Specifications

2. NUREG-1433, Standard Technical Specifications General Electric Plants, B W W 4 Rev. 3.

3. Generic Letter 91-08, Removal of Component Lists from Technical Specifications

4. NUREG-1463, Regulatory Analysis for the Resolution of Generic Safety Issue 105:

Interfacing System Loss-of-Coolant Accident (LOCA) in Light-Water Reactors

5. General Design Criterion (GDC) 30, Quality of Reactor Coolant Pressure Boundary,

6. 10 CFR 50.2, Definitions

7. 10 CFR 50.55aI Codes and standards

8. General Design Criterion (GDC) of 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants

9. NRC Safety Evaluation Related to Amendment No. 44 (Reactor Coolant System Pressure Isolation Valves - Delete Table 3.4-l), Seabrook Station, dated November 28, 1995

10. NRC Safety Evaluation Related to Amendment No. 76, River Bend Station, dated March 8, 1995

Attachment 2 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Relocation of the Reactor Coolant System Pressure Isolation Valve Table from Technical Specifications to the Technical Requirements Manual Marked-up Technical Specification Pages Unit 1 & 2 TS Paaes xi 3i4 4-9 314 4-10 314 4-1 1

INDEX SECTION eaGE PFACTOR COO1 ANT SYSTFY (Continued)

Figure 3.4.1.1-1 Deleted .. ........... . ...... ........... 3/4 4-3 I J e t Pumps ...................................................... 3/4 4-4 R e c i r c u l a t i o n Pumps ........................................... 3/4 4-5 I d l e R e c i r c u l a t i o n Loop Startup ............................... 3/4 4-6 3/4.4.2 SAFETYiRELIEF VALVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . 3 / 4 4-7 3f4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems ..................................... 3/4 4-8 Operational Leakage.. ..............., .............. .. ........ 3/4 4-9 Table 3.4.3.2-1 3/4.4.4 (Deleted) The information from pages 3/4 4-12 through 314 4-14 has been i n t e n t i o n a l l y omitted.

Refer t o note on page 3/4 4-12 ................................ 314 4-12 3/4.4.5 SPECIFIC ACTIVITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 4-15 Table 4.4.5-1 Primary Coolant Specific A c t i v i t y Sample and Analysis Program .............. 314 4-17 3/4.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System ........................................ 314 4-18 Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel Metal Temperature Vs. Reactor Vessel Pressure ....................... 314 4-20 Table 4.4.6.1.3-1 Deleted .............................. 3/4 4-21 Reactor Steam Dome ............................................ 314 4-22 3/4.4.7 M A I N STEAM LINE ISOLATION VALVES .,..............,.............3/ 4 4-23 3f4.4.8 STRUCTURAL INTEGRITY ................. ......................... 314 4-24 LIMERICK - UNIT 1 xi Amendment No. W, 444, 177

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE V

3.4.3.2 R e a c t o r c o o l a n t system leakage s h a l l be l i m i t e d to:

a. No PRESSURE BOUNDARY LEAKAGE.

b. 5 gpm UNIDENTIFIED LEAKAGE.

c. 30 gpm t o t a l leakage.

d. 25 gpm t o t a l leakage averaged over any 24-hour period.

e. 1 gpm leakage a t a r e a c t o r coolant system press r e a c t o r c o o l a n t system pressure i s o l a t i o n valve

@.** I

f. 2 gpm increase i n UNIDENTIFIED LEAKAGE over a 24-hour period.

APPLICABII ITY: OPERATIONAL CONDITIONS 1, 2, and 3.

a. With any PRESSURE BOUMDARY LEAKAGE, be i n a t l e a s t HOT SHUTDOWN w i t h i n 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i n COLD SHUTDOWN w i t h i n the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With any r e a c t o r c o o l a n t system leakage g r e a t e r than t h e l i m i t s i n b, c and/or d above, reduce the leakage r a t e t o w i t h i n t h e l i m i t s w i t h i n 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> o r be i n a t l e a s t HOT SHUTDOWN w i t h i n t h e n e x t 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i n COLD SHUTDOWN w i t h i n t h e f o l l o w i n g 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C- . With any r e a c t o r cool a n t system pressure i s o l a t i o n V a l ve leakage g r e a t e r than t h e above l i m i t , i s o l a t e t h e h i g h pressure p o r t i o n o f the a f f e c t e d system f r o m t h e low pressure p o r t i o n w i t h i n 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use o f a t l e a s t one other c l o s e d manual, deactivated automatic, o r check* v a l v e s , o r be i n a t l e a s t HOT SHUTDOWN w i t h i n t h e next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i n COLD SHUTDOWN w i t h i n the f o l l o w i n g 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d. With one ressure i n t e r f a c e v a l v e leakage pressure moni t o r s inoperable, r e s t o r e t h e inoperable monitor( t h i n 7 days o r v e r i f y t h e pressure t o be l e s s than t h e a l a r m s e t p o i n t a t l e a s t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; r e s t o r e t h e inoperable m o n i t o r ( s 1 t o OPERABLE s t a t u s w i t h i n 30 days o r be i n a t l e a s t HOT SHUTDOWN w i t h i n t h e next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i n COLD SHUTDOWN w i t h i n t h e f o l l owing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

e. With any r e a c t o r c o o l a n t system leakage g r e a t e r than t h e l i m i t i n f above, i d e n t i f y t h e source o f leakage w i t h i n 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> o r be i n a t l e a s t HOT SHUTDOWN w i t h i n t h e n e x t 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i n COLD SHUTDOWN w i t h i n t h e f o l l o w i n g 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

• Which have been v e r i f i e d n o t t o exceed t h e allowable leakage l i m i t a t the l a s t r e f u e l i n g outage o r a f t e r t h e l a s t time the valve was disturbed, whichever i s more recent.

• • Pressure i s o l a t i o n v a l v e leakage i s n o t included i n any o t h e r allowable operational leakage s p e c i f i e d i n Section 3.4.3.2. I LIMERICK - UNIT 1 3/4 4 - 9 Amendment No. 28, 4-9, 172

REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated t o be w i t h i n each of the above limits by:

a. Monitoring the primary containment atmospheric gaseous radioactivity a t l e a s t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (not a means of quantifying leakage),

b. Monitoring the drywell floor drain sump and drywell equipment drain tank f l o w r a t e a t l e a s t once per eight (8) hours, C. Monitoring the drywell u n i t coolers condensate flow r a t e a t l e a s t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

d. Monitoring t h e primary containment pressure a t l e a s t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (not a means of quantifying leakage),

e. Monitoring the reactor vessel head flange leak detection system a t l e a s t once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and

f. Monitoring the primary containment temperature a t l e a s t once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (not a means o f quantifying leakage).

reactor cool ant system pressure is01 ation valve shall be demonstrated OPERABLE by leak t e s t i n g p 0.5 and verifying the leakage o f each valve t o b specified limit:

a. A t l e a s t once per 24 months, and I

b. Prior t o returning the valve t o service following maintenance, repair o r replacement work on the valve which could a f f e c t its leakage r a t e .

The provisions of Specification 4.0.4 a r e not applicable for entry i n t o OPERATIONAL CONDITION 3.

4.4.3.2.3 The highllow pressure interface valve leakage pressure monitors shall RABLE w i t h alarm s e t p o i n t s set less than the by performance o f a:

a. CHANNEL FUNCTIONAL TST a t l e a s t once per 31 days, an

b. CHANNEL CALIBRATION a t l e a s t once per 24 months.

LIMERICK - UNIT 1 314 4-10 Amendment No. 33, 49, 71 JUL 2 8 1994

P U TA8LE 3.4.3.2-1 REACTOR COOLANT SYSTEH PRESSURE ISOUTION VALVES (Psf4) VALUE (pstg) -

SERVICE

'A' LPCI Injection

'8' LPCI Injection

'C' LPCI Injection

'0' LPCI Injection

+4 3 -

SECTION eaGE 4 - s (Continued)

Figure 3.4.1.1-1 Deleted ............................... 314 4 - 3 1 J e t Pumps ..................................................... 3/4 4-4 R e c i r c u l a t i o n Pumps ........................................... 3/4 4-5 I d l e R e c i r c u l a t i o n Loop Startup ............................... 3/4 4-6 3/4.4.2 SAFETY/RELIEF VALVES .......................................... 3/4 4-7 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems ..................................... 3/4 4-8 Leakage ...........................................

Operational Table 3.4.3.2-1 &

/4 4-9 3/4 4-11 3/4.4.4 (Deleted) The information from pages 3/4 4-12 through 3/4 4-14 has been i n t e n t i o n a l l y omitted .

Refer t o note on page 3/4 4-12 ............................... 3/4 4-12 3/4.4.5 SPECIFIC ACTIVITY ............................................. 3/4 4-15 Table 4.4.5-1 Primary Coolant Speciffc A c t i v i t y Sample and Analysis Program .............. 3/4 4-17 314.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System ........................................ 3/4 4-18 Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel Metal Temperature Vs Reactor Vessel Pressure ....................... 3/4 4-20 Table 4.4.6.1.3-1 Deleted .............................. 3/4 4-21 Reactor Steam Dome ............................................ 3/4 4-22 314.4.7 M A I N STEAM LINE ISOLATION VALVES .............................. 3/4 4-23 3/4.4.8 STRUCTURAL INTEGRITY. ......................................... 314 4-24 LIMERICK . UNIT 2 xi Arnendmcrnt. Nn . U L l llii 120

REACTOR COOLANT SYSTEM QPERATIONAL LEAKAGE 3.4.3.2 Reactor coolant system leakage shall be limited to:

a. N o PRESSURE BOUNDARY LEAKAGE.

b. 5 gpm U N I D E N T I F I E D LEAKAGE.

c. 30 gpm t o t a l leakage.

d. 25 gpm t o t a l leakage averaged over any 24-hour period.

e. 1 gprn leakage a t a reactor coolant system press reactor coolant system pressure isolation valve

f. 2 gpm increase i n U N I D E N T I F I E D LEAKAGE over a 24-hour period.

A P P L I C A B I I I TY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be i n a t l e a s t HOT SHUTDOWN w i t h i n 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i n COLD SHUTDOWN w i t h i n the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. W i t h any reactor coolant system leakage greater t h a n the l i m i t s i n b , c and/or d above, reduce the leakage r a t e t o w i t h i n the limits w i t h i n 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be i n a t l e a s t HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i n COLD SHUTDOWN w i t h i n the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. W i t h any reactor coolant system pressure isolation valve leakage greater t h a n the above l i m i t , i s o l a t e the high pressure portion of the affected system f r o m the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of a t l e a s t one other closed manual, deactivated automatic, or check* valves, or be in a t l e a s t HOT SHUTDOWN w i t h i n the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i n COLD SHUTDOWN w i t h i n the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d. W i t h one ressure interface valve leakage pressure monitors inoperable, restore the i noperabl e monitor(s1 t o OPERABLE status w i t h i n 7 days or verify the pressure t o be less t h a n the alarm setpoint a t ? e a s t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; restore the inoperable monitor(s1 to OPERABLE status w i t h i n 30 days or be i n a t l e a s t HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i n COLD SHUTDOWN w i t h i n the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

e. With any reactor coolant system leakage greater t h a n the limit i n f above, identify the source of leakage w i t h i n 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be i n a t l e a s t HOT SHUTDOWN within the next 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and in COLD SHUTDOWN w i t h i n the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• Which have been verified not t o exceed the allowable leakage limit a t the l a s t refueling outage or a f t e r the l a s t time the valve was disturbed, whichever i s more recent.

• • Pressure i s o l a t i o n valve leakage i s n o t included i n any other allowable operational leakage specified i n Section 3.4.3.2.

LIMERICK - U N I T 2 3/4 4-9 Amendment No. &,%,134

REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS -

4.4.3.2.1 The reactor coolant system leakage shall be demonstrated t o be w i t h i n each of the above limits by:

a. Monitoring the primary conta nment atmospheric gaseous radioactivity a t least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (not a means of quant i fyi ng 1ea kage) ,

b. Monitoring the drywell floor drain sump and drywell equipment drain tank f l o w r a t e a t l e a s t once per eight (8) hours,

c. -Monitoring t h e drywel.1 u n i t -coolers -condensate f l o w -rate at least Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

d. Monitoring the primary containment pressure a t l e a s t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (not a means of quantifying leakage),

e. Monitoring the reactor vessel head flange leak detection system a t l e a s t once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />? and

f. Monitoring the primary containment temperature a t l e a s t once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (not a means of quantifying leakage).

reactor cool ant system pressure is01a t ion Val ve shall be demonstrated OPERABLE by leak testing pursuan 0.5 and verifying the leakage o f each valve t o be w i t h i n the

a. A t l e a s t once per 24 months, and I

b. Prior t o returning the valve t o service following maintenance, repair o r replacement work on the valve which could a f f e c t its leakage rate.

The provisions o f Specification 4.0.4 are not applicable f o r entry into OPERATIONAL CONDITION 3.

4.4.3.2.3 The high/low pressure interface valve leakage pressure monitors shall RABLE w i t h alarm setpoints set less t h a n t h by performance o f a:

a. CHANNEL FUNCTIONAL TEST a t least once per 31 days, a

b. CHANNEL CALIBRATION a t l e a s t once per 24 months. 1 LIMERICK - UNIT 2 3/4 4-10 Amendment No. 12, 34 2 8 1994

TABLE 3.4.3.2-1 REACTOR COOLANT SYSTUI PRESSURE ISOUTION VALVES

'A' LPCI Injection

'B' LPCI Injection

'C' LPCI Injection w

cn I

Attachment 3 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Relocation of the Reactor Coolant System Pressure Isolation Valve Table from Technical Specifications to the Technical Requirements Manual Marked-up Technical Specification Bases Pages Unit 1 & 2 TS Paaes B 314 4-3e

REACTOR COOLANT SY STEV BASES ___ __ ~_ __

____________ _ ._ ~-

3/4.4.3.2 OPERATIONAL LEAKAGE (Continued)

The A C T I O N requirements f o r pressure i s o l a t i o n valves are conjunction with the system s p e c i f i c a t i o n s f o r which P I V s a d in a n d w i t h primary contai nrnent i sol a t i o n valve requi rements t o ensure t h a operation i s appropriately l i mi t e d .

The Survei 11 ance Requi rements f o r the RCS pressure i sol a t i o n valves provide added assurance o f valve i n t e g r i t y thereby reducing the probability o f gross valve f a i l u r e a n d consequent intersystem LOCA. Leakage from the RCS pressure i s o l a t i o n valves i s n o t included i n any o t h e r allowable operational leakage specified i n Section 3.4.3.2.

3/4.4.4 (Deleted) I N F O R M A T I O N FROM THIS S E C T I O N RELOCATED TO T H E TRM LIMERICK - UNIT 1 B 3/4 4-3e 4-72, 174 Amendment No. 44.0,

REACTOR COOLANT S Y S T E M BASES

- - __ ~ I

~ - _ _ __ -__

-~I __ _

3 / 4.4.3 .2 OPE RAT I ON A t LEAKAGE ( Con t in ued 1 1 The ACTION r e q u i r e m e n t s f o r p r e s s u r e i s o l a t i o n v a l v e s c o n j u n c t i o n w i t h t h e system s p e c i f i c a t i o n s f o r which P I V s a r and w i t h p r i m a r y c o n t a i n m e n t i s o l a t i o n v a l v e r e q u i r e m e n t s t o e n s u r e t h a operation i s appropriately l i m i t e d .

The S u r v e i 11 ance R e q u i r e m e n t s f o r t h e RCS p r e s s u r e i s o l a t i o n v a l v e s p r o v i d e added a s s u r a n c e o f v a l v e i n t e g r i t y t h e r e b y r e d u c i n g t h e p r o b a b i 1it y o f g r o s s v a l v e f a i 1 u r e and consequent i n t e r s y s t e m LOCA. Leakage f r o m t h e RCS p r e s s u r e i s o l a t i o n v a l v e s i s n o t i n c l u d e d i n any o t h e r a l l o w a b l e o p e r a t i o n a l l e a k a g e s p e c i f i e d i n S e c t i o n 3 . 4 . 3 . 2 .

3/4.4.4 (Deleted). INFORMATION FROM T H I S S E C T I O N RELOCATED TO THE TRM LIMERICK - UNIT 2 B 3/4 4-3e Amendment No. 44.3,434, 136

Attachment 4 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Relocation of the Reactor Coolant System Pressure Isolation Valve Table from Technical Specifications to the Technical Requirements Manual Typed Technical Specification Pages Unit 1 & 2 TS Paaes xi 314 4-9 314 4-10 3/44-1 1

INDEX I O N AND SECT I O N PAGE REACTOR COOLANT SYSTEM ( C o n t i n u e d )

Figure 3.4.1.1-1 Deleted .............................. 3/4 4-3 J e t Pumps ................................................... 3/4 4-4 R e c i r c u l a t i o n Pumps ......................................... 3/4 4 - 5 I d l e R e c i r c u l a t i o n Loop S t a r t u p ............................. 3/4 4-6 3/4.4.2 SAFETY/RELIEF VALVES ........................................ 3/4 4-7 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage D e t e c t i o n Systems ................................... 3/4 4 - 8 O p e r a t i o n a l Leakage ......................................... 3/4 4 - 9 Table 3.4.3.2-1 Deleted .............................. 3/4 4-11 I 3/4.4.4 ( D e l e t e d ) The i n f o r m a t i o n f r o m pages 3 / 4 4 - 1 2 t h r o u g h 3 / 4 4 - 1 4 has been i n t e n t i o n a l l y o m i t t e d .

R e f e r t o n o t e on page 3 / 4 4 - 1 2 . . . . . . . . . . . . . . . . . . ..... 3/4 4-12 3/4.4.5 S P E C I F I C ACTIVITY ........................................... 3/4 4-15 Table 4.4.5-1 Primary Coolant S p e c i f i c A c t i v i t y Sample and A n a l y s i s Program ............. 3 / 4 4 - 1 7 3/4.4.6 PRESSURE/TEMPERATURE LIMITS R e a c t o r C o o l a n t System ...................................... 3/4 4-18 Figure 3.4.6.1-1 Minimum R e a c t o r P r e s s u r e Vessel M e t a l Temperature Vs . R e a c t o r Vessel P r e s s u r e ...................... 3/4 4-20 Table 4.4.6.1.3-1 Deleted ............................. 3/4 4-21 R e a c t o r Steam Dome .......................................... 3/4 4-22 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES ............................ 3/4 4-23 3/4.4.8 STRUCTURAL INTEGRITY ........................................ 3 / 4 4-24 LIMERICK . UNIT 1 xi Amendment No . 4-64. 4-74. 4-7.4,

REACTOR COOLANT SYSTEM OPE RAT ION A L LEAKAGE L IM I T IN G C O N D I T I O N FOR_ OP E RAT ION ~- - - --

3.4.3.2 R e a c t o r c o o l a n t s y s t e m l e a k a g e s h a l l be l i m i t e d t o :

a. No PRESSURE BOUNDARY LEAKAGE.

b. 5 gpm UNIDENTIFIED LEAKAGE.

c. 30 gpm t o t a l l e a k a g e .

d. 25 gpm t o t a l l e a k a g e averaged o v e r any 2 4 - h o u r p e r i o d .

e. 1 gpm l e a k a g e a t a r e a c t o r c o o l a n t system p r e s s u r e o f 950 210 p s i g f r o m any r e a c t o r c o o l a n t s y s t e m p r e s s u r e isol a t i on v a l v e . *

• I

f. 2 gpm i n c r e a s e i n UNIDENTIFIED LEAKAGE o v e r a 24-hour period.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. W i t h any PRESSURE BOUNDARY LEAKAGE, be i n a t l e a s t HOT SHUTDOWN w i t h i n 12 h o u r s and i n COLD SHUTDOWN w i t h i n t h e n e x t 24 h o u r s .

b. W i t h any r e a c t o r c o o l a n t system leakage g r e a t e r t h a n t h e l i m i t s i n b, c and/or d above, reduce t h e leakage r a t e t o w i t h i n t h e l i m i t s w i t h i n 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> o r be i n a t l e a s t HOT SHUTDOWN w i t h i n t h e n e x t 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and i n COLD SHUTDOWN w i t h i n t h e f o l l o w i n g 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. W i t h any r e a c t o r c o o l a n t system p r e s s u r e i s o l a t i o n v a l v e l e a k a g e g r e a t e r t h a n t h e above l i m i t , i s o l a t e t h e h i g h p r e s s u r e p o r t i o n of t h e a f f e c t e d system f r o m t h e l o w p r e s s u r e p o r t i o n w i t h i n 4 h o u r s b y use o f a t l e a s t one o t h e r c l o s e d manual, d e a c t i v a t e d a u t o m a t i c , o r check* v a l v e s , o r be i n a t l e a s t HOT SHUTDOWN w i t h i n t h e n e x t 12 h o u r s and i n COLD SHUTDOWN w i t h i n t h e f o l 1 owing 24 h o u r s .

d. W i t h one o r more o f t h e h i g h / l o w p r e s s u r e i n t e r f a c e v a l v e l e a k a g e p r e s s u r e m o n i t o r s i n o p e r a b l e , r e s t o r e t h e i n o p e r a b l e moni t o r ( s ) t o OPERABLE s t a t u s I w i t h i n 7 days o r v e r i f y t h e p r e s s u r e t o be l e s s t h a n t h e a l a r m s e t p o i n t a t l e a s t once p e r 12 h o u r s ; r e s t o r e t h e i n o p e r a b l e m o n i t o r ( s 1 t o OPERABLE s t a t u s w i t h i n 30 days o r be i n a t l e a s t HOT SHUTDOWN w i t h i n t h e n e x t 12 h o u r s and i n COLD SHUTDOWN w i t h i n t h e f o l l o w i n g 24 h o u r s .

e. W i t h any r e a c t o r c o o l a n t system l e a k a g e g r e a t e r t h a n t h e l i m i t i n f above, i d e n t i f y t h e s o u r c e of l e a k a g e w i t h i n 4 h o u r s o r be i n a t l e a s t HOT SHUTDOWN w i t h i n t h e n e x t 12 h o u r s and i n COLD SHUTDOWN w i t h i n t h e f o l l o w i n g 24 h o u r s .

• Which have been v e r i f i e d n o t t o exceed t h e a l l o w a b l e l e a k a g e l i m i t a t t h e l a s t r e f u e l i n g o u t a g e o r a f t e r t h e l a s t t i m e t h e v a l v e was d i s t u r b e d , w h i c h e v e r i s more recent.

• • P r e s s u r e i s o l a t i o n v a l v e l e a k a g e i s n o t i n c l u d e d i n any o t h e r a l l o w a b l e o p e r a t i o n a l l e a k a g e s p e c i f i e d i n S e c t i o n 3.4.3.2.

LIMERICK - UNIT 1 3/4 4-9 Amendment No. 8, 49,&?2,

REACTOR COOLANT SYSTEM

&QUI REMENTs - ~~ -~ - __ ~

4.4.3.2.1 The r e a c t o r c o o l a n t system l e a k a g e s h a l l be d e m o n s t r a t e d t o be w i t h i n each o f t h e above l i m i t s b y :

a. M o n i t o r i n g t h e p r i m a r y c o n t a nment a t m o s p h e r i c gaseous r a d i o a c t i v t y a t l e a s t once p e r 12 h o u r s ( n o t a means o f q u a n t i f y i n g l e a k a g e ) ,

b. M o n i t o r i n g t h e d r y w e l l f l o o r d r a i n sump and d r y w e l l equipment d r a n t a n k f l o w r a t e a t l e a s t once p e r e i g h t (8) h o u r s ,

c. M o n i t o r i n g t h e d r y w e l l u n i t c o o l e r s c o n d e n s a t e f l o w r a t e a t l e a s t once p e r 12 h o u r s ,

d. M o n i t o r i n g t h e p r i m a r y c o n t a i n m e n t p r e s s u r e a t l e a s t once p e r 12 h o u r s

( n o t a means o f q u a n t i f y i n g l e a k a g e ) ,

e. M o n i t o r i n g t h e r e a c t o r v e s s e l head f l a n g e l e a k d e t e c t i o n system a t l e a s t once p e r 24 h o u r s , and

f. M o n i t o r i n g t h e p r i m a r y c o n t a i n m e n t t e m p e r a t u r e a t l e a s t once p e r 24 h o u r s ( n o t a means o f q u a n t i f y i n g l e a k a g e ) .

4.4.3.2.2 Each r e a c t o r c o o l a n t s y s t e m p r e s s u r e i s o l a t i o n v a l v e s h a l l be d e m o n s t r a t e d I OPERABLE by l e a k t e s t i n g p u r s u a n t t o S p e c i f i c a t i o n 4 . 0 . 5 and v e r i f y i n g t h e l e a k a g e o f each v a l v e t o be w i t h i n t h e s p e c i f i e d l i m i t :

a. A t l e a s t once p e r 24 months, and

b. P r i o r t o r e t u r n i n g t h e v a l v e t o s e r v i c e f o l l o w i n g maintenance, r e p a i r o r r e p l a c e m e n t work on t h e v a l v e w h i c h c o u l d a f f e c t i t s l e a k a g e r a t e .

The p r o v i s i o n s o f S p e c i f i c a t i o n 4 . 0 . 4 a r e n o t a p p l i c a b l e f o r e n t r y i n t o OPERATIONAL CONDITION 3.

4.4.3.2.3 The h i g h / l o w p r e s s u r e i n t e r f a c e v a l v e l e a k a g e p r e s s u r e m o n i t o r s s h a l l be d e m o n s t r a t e d OPERABLE w i t h a l a r m s e t p o i n t s s e t l e s s t h a n t h e s p e c i f i e d a l l o w a b l e v a l u e s by p e r f o r m a n c e o f a : l

a. CHANNEL FUNCTIONAL TEST a t l e a s t once p e r 3 1 days, and

b. CHANNEL CALIBRATION a t l e a s t once p e r 24 months.

LIMERICK - UNIT 1 3/4 4-10 Amendment No. 3 , 4-9, a,

TABLE 3 . 4 . 3 . 2 ~ 1 ( Del e t e d 1 THE INFORMATION FROM THIS TECHNICAL SPECIFICATION SECTION HAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM)

LIMERICK - UNIT 1 3/4 4-11 Amendment No. 14,

INDEX LIMITING CONDITIONS FOROPERATION AND SURVEILLANCE R-EQUIREMENTS- ....___-

SECTION PAGE REACTOR COOLANT SYSTEM ( C o n t i n u e d )

Figure 3.4.1.1-1 Deleted .............................. 3/4 4-3 J e t Pumps ................................................... 3/4 4-4 R e c i r c u l a t i o n Pumps ......................................... 3/4 4-5 I d l e R e c i r c u l a t i o n Loop S t a r t u p ............................. 3/4 4-6 3/4.4.2 SAFETY/RELIEF VALVES ........................................ 3/4 4-7 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage D e t e c t i o n Systems ................................... 3/4 4 - 8 O p e r a t i o n a l Leakage ......................................... 3/4 4-9 Table 3.4.3.2-1 Deleted ............................... 3/4 4-11 I 314.4.4 ( D e l e t e d ) The i n f o r m a t i o n f r o m pages 3 / 4 4 - 1 2 t h r o u g h 3 / 4 4 - 1 4 has been i n t e n t i o n a l l y o m i t t e d .

R e f e r t o n o t e on page 3 / 4 4 - 1 2 .............................. 3/4 4-12 3/4.4.5 SPECIFIC ACTIVITY ........................................... 3/4 4-15 Table 4.4.5-1 P r i m a r y Coolant S p e c i f i c A c t i v i t y Sample and A n a l y s i s Program ............. 3 / 4 4 - 1 7 3/4.4.6 PRESSURE/TEMPERATURE LIMITS R e a c t o r C o o l a n t System ...................................... 3/4 4-18 Figure 3.4.6.1-1 Minimum R e a c t o r P r e s s u r e Vessel M e t a l Temperature Vs . R e a c t o r Vessel P r e s s u r e ...................... 3/4 4-20 Table 4.4.6.1.3-1 Deleted ............................. 3/4 4-21 R e a c t o r Steam Dome .......................................... 3/4 4-22 3/4.4.7 M A I N STEAM LINE ISOLATION VALVES ............................ 3/4 4-23 3/4.4,8 STRUCTURAL INTEGRITY ........................................ 3/4 4-24 LIMERICK . UNIT 2 xi Amendment No . W. 4-36. 4-29.

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE L I_

_ M I_

T I- NG C ON D I T ION- F 0 R-0 P ERA T I0N - ~ - -_- - -- __

3.4.3.2 Reactor coolant system leakage s h a l l be l i m i t e d t o :

a. N o PRESSURE BOUNDARY LEAKAGE.

b. 5 gpm U N I D E N T I F I E D LEAKAGE.

c. 30 gpm t o t a l leakage.

d. 2 5 gpm t o t a l leakage averaged over a n y 24-hour period.

e. 1 gpm leakage a t a r e a c t o r coolant system pressure of 950 t10 p s i g from any r e a c t o r coolant system pressure i s o l a t i o n valve.**

f. 2 gpm increase in U N I D E N T I F I E D LEAKAGE over a 24-hour period.

APPLICABILITY: OPERATIONAL C O N D I T I O N S 1, 2 , a n d 3.

ACTION:

a. With a n y PRESSURE BOUNDARY LEAKAGE, be i n a t l e a s t HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> a n d in COLD SHUTDOWN within t h e next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With any r e a c t o r coolant system leakage g r e a t e r t h a n t h e l i m i t s i n b , c and/or d above, reduce the leakage r a t e t o within t h e l i m i t s within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be i n a t l e a s t HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> a n d i n COLD SHUTDOWN within t h e following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. W i t h a n y reactor coolant system pressure i s o l a t i o n valve leakage g r e a t e r t h a n the above l i m i t , i s o l a t e t h e h i g h pressure portion of the a f f e c t e d system from t h e low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of a t l e a s t one other closed manual, deactivated automatic, o r check* valves, or be i n a t l e a s t HOT SHUTDOWN within t h e next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> a n d in COLD SHUTDOWN within t h e following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d. W i t h one or more of the high/low pressure i n t e r f a c e valve leakage pressure monitors inoperable, r e s t o r e the inoperable monitor(s1 t o OPERABLE s t a t u s within 7 d a y s or v e r i f y t h e pressure t o be l e s s t h a n the alarm s e t p o i n t a t l e a s t once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; r e s t o r e t h e inoperable monitor(s) t o OPERABLE s t a t u s within 30 days or be i n a t l e a s t HOT SHUTDOWN within t h e next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> a n d in COLD SHUTDOWN w i t h i n t h e following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

e. W i t h any reactor coolant system leakage g r e a t e r t h a n t h e l i m i t i n f above, i d e n t i f y t h e source of leakage w i t h i n 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in a t l e a s t HOT SHUTDOWN w i t h i n t h e next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> a n d in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• Which have been v e r i f i e d n o t t o exceed t h e allowable leakage l i m i t a t t h e l a s t r e f u e l i n g outage or a f t e r t h e l a s t time the valve was d i s t u r b e d , whichever i s more recent.

• • Pressure i s o l a t i o n valve leakage i s n o t included in any other allowable operational leakage s p e c i f i e d in Sect on 3.4.3.2.

LIMERICK - UNIT 2 3/4 4-9 Amendment N o . J.2, 2-34.,

REACTOR COOLANT SYSTEM 4.4.3.2.1 The r e a c t o r coolant system leakage s h a l l be demonstrated t o be within each of t h e above l i m i t s by:

a. Monitoring t h e primary containment atmospheric gaseous r a d i o a c t i v i t y a t l e a s t once per 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ( n o t a means of quantifying l e a k a g e ) ,

b. Monitoring t h e drywell f l o o r drain sump a n d drywell equipment drain t a n k flow r a t e a t l e a s t once per e i g h t ( 8 ) hours,

c. Monitoring the drywell u n i t coolers condensate flow r a t e a t l e a s t once per 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,

d. Monitoring t h e primary containment pressure a t l e a s t once per 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

( n o t a means of quantifying l e a k a g e ) ,

e. Monitoring the reactor vessel head flange leak d e t e c t i o n system a t l e a s t once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a n d

f. Monitoring t h e primary containment temperature a t l e a s t once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ( n o t a means of quantifying l e a k a g e ) .

4.4.3.2.2 Each r e a c t o r coolant system pressure i s o l a t i o n valve s h a l l be demonstrated I O P E R A B L E by leak t e s t i n g pursuant t o S p e c i f i c a t i o n 4.0.5 a n d v e r i f y i n g the leakage of each valve t o be within the s p e c i f i e d l i m i t :

a. A t l e a s t once per 24 months, a n d

b. Prior t o returning t h e valve t o s e r v i c e following maintenance, r e p a i r or replacement work on the valve which could a f f e c t i t s leakage r a t e .

The provisions o f S p e c i f i c a t i o n 4.0.4 a r e n o t applicable f o r e n t r y i n t o O P E R A T I O N A L CONDITION 3.

4.4.3.2.3 The high/low pressure i n t e r f a c e valve leakage pressure monitors s h a l l be demonstrated O P E R A B L E with a1 arm setpoi n t s s e t 1 e s s t h a n t h e s p e c i f i e d a1 1 owabl e values by performance of a :  !

a. C H A N N E L FUNCTIONAL TEST a t l e a s t once per 31 days, a n d

b. CHANNEL C A L I B R A T I O N a t l e a s t once per 24 months.

LIMERICK - UNIT 2 3/4 4-10 Amendment No, 4-2,34,

TABLE 3 . 4 . 3 . 2 - 1 (Deleted)

THE INFORMATION FROM THIS TECHNICAL SPECIFICATION SECTION HAS B E E N RELOCATED TO THE TECHNICL REQUIREMENTS MANUAL (TRM).

LIMERICK - UNIT 2 3/4 4-11 Amendment No.

Attachment 5 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Relocation of the Reactor Coolant System Pressure Isolation Valve Table from Technical Specifications to the Technical Requirements Manual Typed Technical Specification Bases Pages Unit 1 & 2 TS Paaes B 3/4 4-3e

REACTOR COOLANT SYSTEM BASES___,. .._=_________ ... __~_~ .. ______________

314.4.3.2 OPERATIONAL LEAKAGE ( C o n t i nued)

The f u n c t i o n o f R e a c t o r C o o l a n t System P r e s s u r e I s o l a t i o n V a l v e s ( P I V s ) i s t o s e p a r a t e t h e h i g h p r e s s u r e R e a c t o r C o o l a n t System f r o m an a t t a c h e d l o w p r e s s u r e system.

The ACTION r e q u i r e m e n t s f o r p r e s s u r e i s o l a t i o n v a l v e s a r e used i n c o n j u n c t i o n w i t h t h e s y s t e m s p e c i f i c a t i o n s f o r w h i c h PIVs a r e l i s t e d i n t h e T e c h n i c a l Requirements Manual and w i t h p r i m a r y c o n t a i n m e n t isol a t i on v a l v e r e q u i rements t o e n s u r e t h a t p l a n t operation i s appropriately limited.

The S u r v e i 11 ance Requi rements f o r t h e RCS p r e s s u r e is o l a t i o n v a l v e s p r o v i d e added assurance o f v a l v e i n t e g r i t y thereby r e d u c i n g t h e p r o b a b i l i t y o f gross v a l v e f a i l u r e and c o n s e q u e n t i n t e r s y s t e m LOCA. Leakage f r o m t h e RCS p r e s s u r e i s o l a t i o n v a l v e s i s n o t i n c l u d e d i n any o t h e r a l l o w a b l e o p e r a t i o n a l l e a k a g e s p e c i f i e d i n S e c t i o n 3 . 4 . 3 . 2 .

3/4.4.4 ( D e l e t e d ) INFORMATION FROM THIS SECTION RELOCATED TO THE TRM LIMERICK - UNIT 1 B 3/4 4-3e Amendment No. 4-443,&?2, 444,

REACTOR COOLANT SYSTEM 3/4.4.3.2 OPERATIONAL LEAKAGE ( C o n t i n u e d )

The f u n c t i o n o f R e a c t o r C o o l a n t System P r e s s u r e I s o l a t i o n V a l v e s ( P I V s ) i s t o s e p a r a t e t h e h i g h p r e s s u r e R e a c t o r C o o l a n t System f r o m an a t t a c h e d l o w p r e s s u r e system.

The ACTION r e q u i r e m e n t s f o r p r e s s u r e i s o l a t i o n v a l v e s a r e used i n c o n j u n c t i o n w i t h t h e s y s t e m s p e c i f i c a t i o n s f o r w h i c h PIVs a r e l i s t e d i n The T e c h n i c a l Requirements Manual and w i t h p r i m a r y c o n t a i n m e n t i sol a t i o n v a l v e r e q u i rements t o e n s u r e t h a t p l a n t operation i s appropriately limited.

The S u r v e i 11 ance Requi rements f o r t h e R C S p r e s s u r e isol a t i o n v a l v e s p r o v i d e added assurance o f v a l v e i n t e g r i t y t h e r e b y r e d u c i n g t h e p r o b a b i l i t y o f gross v a l v e f a i l u r e and consequent i n t e r s y s t e m LOCA. Leakage f r o m t h e RCS p r e s s u r e i s o l a t i o n v a l v e s i s n o t i n c l u d e d i n any o t h e r a l l o w a b l e o p e r a t i o n a l l e a k a g e s p e c i f i e d i n S e c t i o n 3.4.3.2.

3/4.4.4 (Deleted) INFORMATION FROM THIS SECTION RELOCATED TO THE TRM LIMERICK - UNIT 2 B 3/4 4-3e Amendment No. 443, 4-34,4.36,