ML073461114

License Amendment Request - Proposed Administrative Changes to Technical Specification

ML073461114
Person / Time
Site:	Limerick
Issue date:	12/12/2007
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML073461114 (44)

Text

Exclon Nuclear

\\n~~u~+!.exeloniorp.corn zoo Exelon LVay Kennett Square, PA 19348 Nuclear 10 CFR 50.90 December 12,2007 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 n o c k e t - 5 0 - 3 5 2 50-w

SUBJECT:

License Amendment Request Proposed Administrative Changes to Technical Specifications Pursuant to 10 CFR 50.90, "Application for amendment of license or construction permit,"

Exelon Generation Company, LLC (Exelon), proposes changes to the Technical Specifications (TS), Appendix A of Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively.

The proposed administrative changes provide an editorial cleanup of the TS. The proposed changes involve: (I) correcting the index, (2) removing cycle specific requirements or notes that have since expired and are no longer applicable, (3) deleting references to previously deleted requirements, (4) changing references to the location of previously relocated information, and (5) editorial corrections. Evaluation of the proposed changes is provided in this attachment.

Markups of the proposed TS changes are provided in Attachment 2.

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

This amendment request contains no regulatory commitments.

Exelon requests approval of the proposed amendment by December 12,2008. Upon NRC approval, the amendment shall be implemented within 60 days of issuance.

These proposed changes have been reviewed by the Plant Operations Review Committee and approved in accordance with Nuclear Safety Review Board procedures.

License Amendment Request Administrative Changes Docket Nos. 50-352 and 50-353 December 12,2007 Page 2 We are notifying the State of Pennsylvania of this application for changes to the Technical Specifications 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 61 0-765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 12th day of December, 2007.

Respectfully, Pamela 6. &wan Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments:

1. Evaluation of Proposed Changes

2. Markup of Proposed Technical Specifications Pages cc:

Regional Administrator - NRC Region I W/ attachments NRC Senior Resident Inspector - Limerick Generating Station I t NRC Project Manager, NRR - Limerick Generating Station It Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection If

ATTACHMENT 1 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 EVALUATION OF PROPOSED CHANGES

Subject:

Proposed Administrative Changes to Technical Specifications

2.0 PROPOSED CHANGE

S

7.0 REFERENCES

License Amendment Request Administrative Changes Page 1 of 9 Docket Nos. 50-352 and 50-353 Evaluation of Proposed Changes

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, Application for amendment of license or construction permit, Exelon Generation Company, LLC (Exelon), proposes changes to the Technical Specifications (TS), Appendix A of Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively.

The proposed administrative changes provide an editorial cleanup of the TS. The proposed changes involve: (1) correcting the index, (2) removing cycle-specific requirements or notes that have since expired and are no longer applicable, (3) deleting references to previously deleted requirements, (4) changing references to the location of previously relocated information, and (5) editorial corrections. Evaluation of the proposed changes is provided in this attachment. Markups of the proposed TS changes are provided in Attachment 2.

2.0 PROPOSED CHANGE

S The changes requested by this amendment application are described below.

1. Index Corrections a) LGS, Units 1 and 2 - Revise Index page ii to reflect the previously approved addition of TS Definition 1.35, "RECENTLY IRRADIATED FUEL" to TS page 1-6 and TS Definition 1.37a, "RESTRICTED AREA" to TS page 1-7, and the appropriate renumbering of the remaining TS Definitions.

b) LGS, Units 1 and 2 - Change the word "Semiannual" to "Annual" in the title of the Radioactive Effluent Release Report on TS Index page xxvii to reflect the appropriate title for the report as specified under Section 6.9.1, "Routine Reports,"

on TS page 6-17.

2. Cycle-Specific Requirements or Notes that are No Longer Applicable a) LGS, Unit 1 - Delete Limiting Condition for Operation (LCO) 3.1.3.6, Action c. and associated Surveillance Requirement (SR) 4.1.3.6.d relative to repositioning uncoupled control rod 50-27. Also, delete the words "except as in 3.1.3.6.c or" from LCO 3.1.3.6, Action a.2. These requirements were limited to LGS, Unit 1, Cycle 7 only. LGS, Unit 1 is currently operating in Cycle 12. Therefore, these requirements are no longer applicable.

b) LGS, Unit 2 - Delete the double asterisk and associated footnote from the 145oF drywell average air temperature limit specified in TS LCO 3.6.1.7 and associated TS Action. This footnote was limited to LGS, Unit 2, Cycle 9 only. LGS, Unit 2 is currently operating in Cycle 10. Therefore, this footnote is no longer applicable.

License Amendment Request Administrative Changes Page 2 of 9 Docket Nos. 50-352 and 50-353 Evaluation of Proposed Changes

3. References to Previously Deleted Requirements a) LGS, Units 1 and 2 - Delete the reference to TS SR 4.8.1.1.3 from TS SR 4.8.1.2 on TS page 3/4 8-9 since TS SR 4.8.1.1.3 was previously deleted and is no longer in TS.

4. References to the Location of Previously Relocated Information a) LGS, Units 1 and 2 - Replace the word "UFSAR" with the word "TRM" in the relocation information provided of TS pages 3/4 3-68, 3/4 3-110, 3/4 6-49, 3/4 6-51, 3/4 6-51a, and 3/4 8-21.

5. Editorial Changes a) LGS, Unit 2 - Capitalize the first letter "r" in the word "radioactive" in the title of the Radioactive Effluent Controls Program referenced under Item No. (1) in TS Definition 1.24 (OFFSITE DOSE CALCULATION MANUAL) on TS page 1-4.

b) LGS, Unit 2 - Replace misspelled word "RELOCATD" with the word "RELOCATED" on TS page 3/4 3-110.

c) LGS, Units 1 and 2 - Replace capitalized word "INOPERABLE" with lower case word "inoperable" in TS LCO 3.4.3.1, Actions B, C, D, and E on TS page 3/4 4-8.

d) LGS, Unit 2 - Replace the word "or" with the word "of" in TS SR 4.4.3.1.b on TS page 3/4 4-8a.

e) LGS, Units 1 and 2 - Replace the word "patch" with the word "path" in TS LCO 3.5.1.a.2 on TS page 3/4 5-1.

f) LGS, Units 1 and 2 - Delete the "s" at the end of the word "valves" and replace the word "the" with the word "each" in TS LCO 3.6.3 on TS page 3/4 6-17.

g) LGS, Unit 1 - Change the paragraph indent for TS SR 4.7.1.2.b.2 to line up with the paragraph indent for TS SR 4.7.1.2.b.1 on TS page 3/4 7-4.

h) LGS, Unit 2 - Delete the duplicate word "both" from the first sentence in TS LCO 3.7.2, Action b.2 on TS page 3/4 7-6.

3.0 BACKGROUND

As a result of various changes to TS that have been requested and approved over time, a number of minor inconsistencies or editorial errors that have no safety impact have been introduced into the LGS Unit 1 and Unit 2 TS. The purpose of this amendment request is to correct those inconsistencies or errors.

License Amendment Request Administrative Changes Page 3 of 9 Docket Nos. 50-352 and 50-353 Evaluation of Proposed Changes

4.0 TECHNICAL ANALYSIS

The Technical justification for each of the changes proposed in Section 2.0 of this request is provided below.

1. Index Corrections By letter dated August 23, 2006 (Reference 1), the NRC issued Amendment No. 185 to Facility Operating License No. NPF-39 and Amendment No. 146 to Facility Operating License No. NPF-85 for LGS, Units 1 and 2, respectively. These amendments revised LGS TS to support application of an alternate source term methodology. These amendments approved the addition of Definition No. 1.35 (RECENTLY IRRADIATED FUEL) to TS Section 1.0, "DEFINITIONS," on TS page 1-

6. As a result of this addition, pre-existing Definition Nos. 1.35 (REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY), 1.36 (REPORTABLE EVENT), 1.37 (ROD DENSITY), 1.38 (SHUTDOWN MARGIN), and 1.39 (SITE BOUNDARY) were renumbered to 1.36 through 1.40, respectively.

By letter dated June 29, 2007 (Reference 2), the NRC issued Amendment No. 187 to Facility Operating License No. NPF-39 and Amendment No. 148 to Facility Operating License No. NPF-85 for LGS, Units 1 and 2, respectively. These amendments revised LGS TS to incorporate revised requirements from Title 10 of the Code of Federal Regulations (CFR), Part 20. These amendments added Definition No. 1.37a (RESTRICTED AREA), to TS Section 1.0 on TS page 1-7.

By letter dated August 10, 1994 (Reference 3), the NRC issued Amendment No. 73 to Facility Operating License No. NPF-39 and Amendment No. 35 to Facility Operating License No. NPF-85 for LGS, Units 1 and 2, respectively. These amendments revised various LGS TS sections to change the frequency for submitting the Semiannual Radioactive Effluent Release Report to the NRC from semiannually to annually. These changes involved changing the title of TS Section 6.9.1.8 on TS page 6-17 from "Semiannual Radioactive Effluent Release Report" to "Annual Radioactive Effluent Release Report."

All of these amendments were approved as requested by Exelon; however, the Exelon requests did not include a revised TS Index to reflect the associated changes to TS Sections 1.0 or 6.9.1. The proposed changes described in this amendment request revise the TS Index to appropriately reflect the changes to these sections of TS that were previously approved by the amendments referenced above. The proposed changes are administrative in nature and do not involve any physical changes to structures, systems, or components (SSCs) in the plant, or the way SSCs are operated or controlled.

License Amendment Request Administrative Changes Page 4 of 9 Docket Nos. 50-352 and 50-353 Evaluation of Proposed Changes

2. Cycle-Specific Requirements or Notes that are No Longer Applicable By letter dated January 16, 1998 (Reference 4), the NRC issued Amendment No. 124 to Facility Operating License No. NPF-39 for LGS, Unit 1. This amendment revised TS LCO 3.1.3.6 and associated TS SR 4.1.3.6 to allow operation of the unit with control rod 50-27 uncoupled from its drive for the remainder of operating Cycle 7.

This amendment specified the conditions under which control rod 50-27 could be manipulated and modified the surveillance requirements to verify control rod position using neutron instrumentation. LGS, Unit 1 completed operating Cycle 7 in April, 1998, when the unit was shutdown for refueling outage 1R07, during which time the control rod drive was replaced and control rod 50-27 was re-coupled. LGS, Unit 1 is currently operating in Cycle 12. As a result, these requirements are no longer applicable. Therefore, the proposed change deletes LCO 3.1.3.6.c and SR 4.1.3.6.d from the LGS, Unit 1 TS, and deletes the words "except as in 3.1.3.6.c or" from LGS, Unit 1 LCO 3.1.3.6, Action a.2.

By letter dated July 7, 2006 (Reference 5), the NRC issued Amendment No. 145 to Facility Operating License No. NPF-85 for LGS, Unit 2. This amendment approved a one-time change to TS LCO 3.6.1.7 by adding a footnote (**) to the TS limit for drywell average air temperature of 145 degrees Fahrenheit (oF) to allow continued operation of LGS, Unit 2, with drywell average air temperature no greater than 148oF for the remainder of operating Cycle 9, or until the next shutdown of sufficient duration to allow for unit cooler fan repairs, whichever came first. LGS, Unit 2 completed operating Cycle 9 in March, 2007, when the unit was shutdown for refueling outage 2R09, during which the unit cooler fans were repaired. LGS, Unit 2 is currently in operating Cycle 10. As a result, this footnote is no longer applicable.

Therefore, the proposed change deletes this footnote from LGS, Unit 2, LCO 3.6.1.7.

The proposed changes are administrative in nature and do not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled.

3. References to Previously Deleted Requirements By letter dated November 6, 2007 (Reference 6), the NRC issued Amendment No.

189 to Facility Operating License No. NPF-39 and Amendment No. 150 to Facility Operating License No. NPF-85 for LGS, Units 1 and 2, respectively. These amendments modified emergency diesel generator (EDG) testing requirements to improve EDG reliability by reducing potential equipment degradation due to excessive testing. These amendments approved the deletion of SR 4.8.1.1.3 which required a special report to be submitted to the NRC within 30 days of all EDG failures.

TS Section 3.8.1 specifies the LCO and SR requirements for both A.C. Sources -

Operating (LCO 3.8.1.1) and A.C. Sources - Shutdown (LCO 3.8.1.2). SR 4.8.1.1.3 was previously specified under A.C. Sources - Operating. However, the operability of A.C. Sources - Shutdown is demonstrated by performing SR 4.8.1.2, which references the surveillance requirements under A.C. Sources - Operating, including

License Amendment Request Administrative Changes Page 5 of 9 Docket Nos. 50-352 and 50-353 Evaluation of Proposed Changes SR 4.8.1.1.3. SR 4.8.1.1.3 was an administrative requirement for reporting to the NRC and did not, in and of itself, help to demonstrate the operability of the A.C.

Sources. Deleting this reporting requirement did not impact the safe operation of the plant and was approved by the NRC in the amendment referenced above. Therefore, the proposed change deletes the reference to previously deleted SR 4.8.1.1.3 from SR 4.8.1.2, and makes an associated editorial change to SR 4.8.1.2 regarding the reference to the remaining SRs 4.8.1.1.1 and 4.8.1.1.2.

The proposed change is administrative in nature and does not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled.

4. References to the Location of Previously Relocated Information The information provided in TS Sections 3/4.3.7.2, 3/4.3.8, 3.6.5.2.1, 3.6.5.2.2 and 3/4.8.4.1 was previously relocated to licensee documents that are controlled under the requirements of 10 CFR 50.59 based on NRC approved amendments.

Specifically, by letter dated August 29, 1994 (Reference 7), the NRC issued Amendment No. 75 to Facility Operating License No. NPF-39 and Amendment No. 36 to Facility Operating License No. NPF-85 for LGS, Units 1 and 2, respectively. These amendments relocated the seismic monitoring instrumentation requirements from TS Section 3/4.3.7.2 to the Updated Final Safety Analysis Report (UFSAR).

By letter dated August 24, 1995 (Reference 8), the NRC issued Amendment No. 100 to Facility Operating License No. NPF-39 and Amendment No. 64 to Facility Operating License No. NPF-85 for LGS, Units 1 and 2, respectively. These amendments relocated the turbine overspeed protection system requirements from TS Section 3/4.3.8 to the UFSAR.

By letter dated November 20, 1995 (Reference 9), the NRC issued Amendment No.

105 to Facility Operating License No. NPF-39 and Amendment No. 69 to Facility Operating License No. NPF-85 for LGS, Units 1 and 2, respectively. These amendments relocated the Reactor Enclosure and Refueling Area Secondary Containment Automatic Isolation Valves from TS Sections 3.6.5.2.1 and 3.6.5.2.2, respectively, to the UFSAR.

By letter dated June 22, 1995 (Reference 10), the NRC issued Amendment No. 93 to Facility Operating License No. NPF-39 and Amendment No. 57 to Facility Operating License No. NPF-85 for LGS, Units 1 and 2, respectively. These amendments relocated the primary containment conductor protection device requirements from TS Section 3/4.4.8.1 to the UFSAR.

TS pages 3/4 3-68, 3/4 3-110, 3/4 6-49, 3/4 6-51, 3/4 6-51a, and 3/4 8-21 each indicate that the information from their respective TS sections was relocated to the UFSAR. Subsequent to implementation of the above license amendments, LGS implemented a Technical Requirements Manual (TRM), which is incorporated into the

License Amendment Request Administrative Changes Page 6 of 9 Docket Nos. 50-352 and 50-353 Evaluation of Proposed Changes LGS UFSAR by reference and is controlled under the requirements of 10 CFR 50.59.

The TRM contains the technical requirements that are relocated from TS as approved by the NRC via license amendments. Once the TRM was established, the technical requirements that had previously been removed from TS and relocated to the UFSAR via the license amendments referenced above were transferred from the UFSAR to the TRM under the requirements of 10 CFR 50.59. Although the reference to the UFSAR on these TS pages is technically correct, since the TRM is incorporated into the UFSAR by reference, the proposed change replaces the word "UFSAR" with the word "TRM" to more explicitly state the current location of the technical requirements relocated from these TS pages. The proposed change is administrative in nature and does not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled.

5. Editorial Changes The changes described under Item 5 of Section 2.0 strictly involve editorial changes to correct typographical errors, grammar, and paragraph indentation, and delete duplicate wording. These proposed changes are non-substantive changes that have no impact on safe operation of the plant in that they do not involve any physical changes to SSCs in the plant, or the way SSCs are operated or controlled.

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 amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1.

Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. The proposed changes are administrative in nature and do not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not impact the initiators or assumptions of analyzed events, nor do they impact mitigation of accidents or transient events. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

License Amendment Request Administrative Changes Page 7 of 9 Docket Nos. 50-352 and 50-353 Evaluation of Proposed Changes

2.

Do the proposed changes 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 alter plant configuration, require that new plant equipment be installed, alter assumptions made about accidents previously evaluated, or impact the function of plant SSCs or the manner in which SSCs are operated, maintained, modified, tested, or inspected. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Do the proposed changes involve a significant reduction in a margin of safety?

Response: No. The proposed changes are administrative in nature and do not involve any physical changes to plant SSCs or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not involve a change to any safety limits, limiting safety system settings, limiting conditions of operation, or design parameters for any SSC. The proposed changes do not impact any safety analysis assumptions and do not involve a change in initial conditions, system response times, or other parameters affecting an accident analysis. 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.

5.2 Applicable Regulatory Requirements/Criteria In Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, the Nuclear Regulatory Commission (NRC) established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plants TSs.

The proposed changes are administrative in nature and do not involve any physical changes to plant SSCs or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not involve a change to any safety limits, limiting safety system settings, limiting control settings, limiting conditions of operation, surveillance requirements, design features, or administrative controls required by 10 CFR 50.36.

License Amendment Request Administrative Changes Page 8 of 9 Docket Nos. 50-352 and 50-353 Evaluation of Proposed Changes 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 10 CFR 20, 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.

7.0 REFERENCES

1. Letter dated August 23, 2006, from R. Guzman, USNRC to C. Crane, Exelon Nuclear, "Limerick Generating Station, Units 1 and 2 - Issuance of Amendments RE:

Application of Alternate Source Term Methodology (TAC Nos. MC2295 and MC2296)."

2. Letter dated June 29, 2007, from P. Bamford, USNRC to C. Crane, AmerGen Energy Company, "Limerick Generating Station, Units 1 and 2 - Issuance of Amendment RE:

Incorporation of Revised 10 CFR Part 20 Requirements into the Technical Specifications (TAC Nos. MD4946 and MD4947)."

3. Letter dated August 10, 1994, from F. Rinaldi, USNRC to G. Hunger, Philadelphia Electric Company, "Frequency for Submitting Semiannual Radioactive Effluent Release Report, Limerick Generating Station, Units 1 and 2 (TAC Nos. M89175 and M89176)."

4. Letter dated January 16, 1998, from B. Buckley, USNRC to G. Hunger, PECO Energy Company, "Revision to Technical Specifications Regarding Control Rod 50-27, Limerick Generating Station, Unit 1 (TAC Nos. M99854 and M99855)."

5. Letter dated July 7, 2006, from R. Guzman, USNRC to C. Crane, Exelon Nuclear, "

Limerick Generating Station, Unit 2 - Issuance of Amendment RE: One-Time Change to the Drywell Average Air Temperature Limit (TAC No. MD2315)."

License Amendment Request Administrative Changes Page 9 of 9 Docket Nos. 50-352 and 50-353 Evaluation of Proposed Changes

6. Letter dated November 6, 2007, from P. Bamford, USNRC to C. Crane, Exelon Generation Company, LLC, "Limerick Generating Station, Units 1 and 2 - Issuance of Amendment RE: Changes to Technical Specification Emergency Diesel Generator Testing Requirements (TAC Nos. MD3710 and MD3711)."

7. Letter dated August 29, 1994, from F. Rinaldi, USNRC to G. Hunger, Philadelphia Electric Company, "Relocation of Seismic Monitoring Equipment, Limerick Generating Station, Units 1 and 2 (TAC Nos. M88608 and M88655)."

8. Letter dated August 24, 1995, from F. Rinaldi, USNRC to G. Hunger, PECO Energy Company, "Relocation of Turbine Overspeed Protection System Requirements, Limerick Generating Station, Units 1 and 2 (TAC Nos. M90375 and M90376)."

9. Letter dated November 20, 1995, from F. Rinaldi, USNRC to G. Hunger, PECO Energy Company, "Limerick Generating Station, Units 1 and 2 (TAC Nos. M93690 and M93691)."

10. Letter dated June 22, 1995, from F. Rinaldi, USNRC to G. Hunger, PECO Energy Company, "Relocation of Primary Containment Conductor Protection Device Requirements, Limerick Generating Station, Units 1 and 2 (TAC Nos. M90375, M90376, M90512, and M90513)."

ATTACHMENT 2 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Proposed Administrative Changes to Technical Specifications Markup of Proposed Technical Specifications Pages Unit 1 TS Pages ii xxvii 3/4 1-11 3/4 1-12 3/4 3-68 3/4 3-110 3/4 4-8 3/4 5-1 3/4 6-17 3/4 6-49 3/4 6-51 3/4 6-51a 3/4 7-4 3/4 8-9 3/4 8-21 Unit 2 TS Pages ii xxvii 1-4 3/4 3-68 3/4 3-110 3/4 4-8 3/4 4-8a 3/4 5-1 3/4 6-10 3/4 6-17 3/4 6-49 3/4 6-51 3/4 6-51a 3/4 7-6 3/4 8-9 3/4 8-21

JNDEX 1.22b MAPFAC( P) - (POWER DEPENDENT MAPLHGR MULTIPLIER)...........

1-4 SECTION DEFINITIONS (Continued)

PAGE i

1.20a 1-4 1.21 (DELETED)...................................................

1-4 LOW (PBWER) TRIP SETPOINT (LTSP)........................

1.24 OFFSITE DOSE CALCULATION MANUAL...........................

1-4 1.25 OPERABLE - OPERABILITY....................................

1-4 1.26 1-5 1.27 PHYSICS TESTS...............................................

1-5 OPERATIONAL CONDITION - CONDITION..........................

1.28 PRESSURE BOUNDARY LEAKAGE..................................

1-5 1.29 PRIMARY CONTAINMENT INTEGRITY..............................

1-5 1.30 PROCESS CONTROL PROGRAM.

1-5 1-6 1

1.31 PURGE - PURGING............................................

1.32 RATED THERMAL POWER........................................

1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY.........

1-6 ME....................

INTEGRITY............

• @ - " c C C ) c / C ~

4 0

/

~

I Y

C

~

+$+;&7 0

S c w 3 > SHUTDOWN MARG I N............................................

1-7 cxa SITE BOUNDARY.............................................

1 1.41 SOURCE CHECK...............................................

1-7 LIMERICK - UNIT 1 Amendment No. $3, $8, 52, ii FEB 1 o 1994 66

Deleted Deleted....................................................

6-9 Deleted.................................................... 6-9 De'leted.................................................... 6-10 Deleted.................................................... 6-10 Deleted....................................................

6-10 Deleted....................................................

6-10 Deleted.................................................... 6-10 Deleted....................................................

6-12 6.5.3 Deleted........................................~...........6-12 5.6 RFPOR TION TABLE EVENT AC

............................................ 6-12a 6.7 SAFFTY I I M I T VIOLATIOt(.............................................

6-12a 6.8 PROCEDURFS AND PRO~RAMS............................................ 6-13 6.9 RFpORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS............................................

6-15 Startup Report.............................................6-15 Annual Reports.............................................

6-15 Monthly Operating Reports..................................

6-16 Annual Radiological Environmental Operating Report......... 6-16 e n n u a l Radioactive E f f l u e n t Release Report............. 6-17 CORE OPERATING LIMITS' REPORTS..............................

6-18a 6.9.2 SPECIAL REPORTS............................................

6-18a 5-10 DELETED

.... 6-19 6.11 RADIATION PROTFC TION P R O M...................................... 6-20 6.13 HIGH RADIATION ARF$...............................................

6-20 I

I I

I I

I I

I I

I LIMERICK - UNIT 1 xxvi i Amendment No. 32, 42, 96, 176

REACTIVITY CONTROL SYSTEMS CONTROL ROD D R I V E COUPLING LIMITING CO NDITION FOR OPER AT I ON 3.1.3.6 A l l c o n t r o l rods shall be coupled t o t h e i r d r i v e mechanisms.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 5".

ACTION:

a.

I n OPERATIONAL CONDITIONS 1 and 2 w i t h one c o n t r o l rod n o t coupled t o i t s associated d r i v e mechanism, w i t h i n 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

1.

I f permitted by t h e RWM, i n s e r t t h e c o n t r o l rod d r i v e mechanism t o accompl i sh recoupl i ng and v e r i f y recoupl i n g by withdrawing t h e c o n t r o l rod, and:

a )

Observing any i n d i c a t e d response o f t h e nuclear instrumenta-b )

Demonstrating t h a t t h e c o n t r o l rod w i l l n o t go t o t h e o v e r -

t i o n, and t r a v e l p o s i t i o n.

d a c Otherwise, 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 />.

2.

I f recoupling i s n o t accomplis n o t permitted by t h e RWM, then I

p e r m i t t e d by t h e RWM, declare t h e c o n t r o l rod and disarm t h e associated d i r e c t i o n a l c o n t r o l valves** e i t h e r :

a )

E l e c t r i c a l l y, o r b)

Otherwise, 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 />.

b.

I n OPERATIONAL CONDITION 5" w i t h a withdrawn c o n t r o l r o d n o t coupled H y d r a u l i c a l l y by c l o s i n g t h e d r i v e water and exhaust water i s o l a t i o n valves.

t o i t s associated d r i v e mechanism, w i t h i n 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> e i t h e r :

1.

I n s e r t t h e c o n t r o l rod t o accomplish recoupling and v e r i f y recoup-l i n g by withdrawing t h e c o n t r o l r o d and demonstrating t h a t t h e c o n t r o l rod w i l l n o t go t o t h e o v e r t r a v e l p o s i t i o n, o r d i sarm t h e associ ated d i r e c t i onal c o n t r o l valves** e i t h e r :

a )

E l e c t r i c a l l y, o r b )

2.

I f recoupling i s n o t accomplished, i n s e r t t h e c o n t r o l r o d and H y d r a u l i c a l l y by c l o s i n g t h e d r i v e water and exhaust water e withdrawn when 1 owi ng c o n d i t i o n s

1) 2 )

No o t h e r uncoupled c o n t r o l rod i s withdrawn.

The uncoupled c o n t r o l rod may n o t be withdrawn past notch p o s i t i o n 46, and A t l e a s t each withdrawn c o n t r o l rod.

Not a p p l i c a b l e t o c o n t r o l rods removed per S p e c i f i c a t i o n 3.9.10.1 o r 3.9.10.2.

t e s t i n g associated w i t h r e s t o r i n g t h e c o n t r o l rod t o OPERABLE s t a t u s.

• • May be rearmed i n t e r m i t t e n t l y, under a d m i n i s t r a t i v e c o n t r o l, t o p e r m i t LIMERICK - UNIT 1 3/4 1-11 I

Amendment No. -44, 224, 169

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIWEN TS 4.1.3.6 Each affected control rod shall be demonstrated t o be coupled t o i t s drive mechanisnr by observlng any indicated response o f the nuclear instrumen-

~

tation while withdrawing the control rod to the fully withdrawn posltton and then verjfying that the control rod drfve does not go to the overtravel posi ti on z Prior t o reactor criticality after completing CORE ALTERATIONS t h a t could have affected the control rod drive coupling integrity, Anytime the control rod i s withdrawn t o the "Full out" position in subsequent operati on, Following malntenance on or modification to the control rod or control rod drive system which could have affected the control rod

a.

b.

c.

t e

g r

1 0

g the uncoupled control rod per Specificati-3.1.3.6,d the uncoupled rod's position shall be verffqed to have followed the control rod drive by neutron instrumentatjon (LPRM or TIP).

drive out t o i t s final position, then the rod shall be completely inserted and the control rdd directional valves disarmed as stated I f the control blade can not be verified t o have followed the 1IMERICK - UNIT 1 314 1-12 JAM161998 krrnhnath. l24 I

Section 3.3.7.2 (Deleted)

FROM THIS TECHNICAL CTION HAS BEEN RELOCATED TECHNICAL SPECIFICATIONS HROUGH 3/4 3-72 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.

LIMERICK - UNIT 1 314 3-68

gG 2 y 1994 Amendment No. 11, 4 6, 75

Section 3/4.3.8 (Deleted)

THE INFORMATION FRON THIS TECHNICAL SPEC RELOCATED I F I CAT IONS TO SECT1

.I*-

BEEN TEw HAS BEEN INTENTIONALLY OMITTED.

TECH ICAC SPEC I F ICATI ONS 3f4 3-111 LIMERICK - UNIT 1 SE? 2.1 Amendment No. 100 314 3-110

REACTOR C C ; x A r ; i SYSTEM 3 / 4. 4. 3 REPLTOR COOLANT SYSTEM LEAKAGE LEAKAGE DFTECTION S YSTEMS 3.4.3.1 The following reactor coclant leakage detection systems shall be OPFRAHLE :

a.

The primary containment atmosphere gaseous radioactivity monitoring system,

b.

The drywell f l o o r drain sump f l o w monitoring system,

c.

The drywell u n i t coolers condensate f l o w r a t e monitoring system, and d.

The primary containment pressure and temperature monitoring system.

A P P L I C A B I L I T Y :

OPERATIONAL CONDITIONS 1, 2, and 3.*

• - The primary containment gaseous radioactivity monitor i s not required t o be operable u n t i l Operational Condition 2.

ACTIONS:

A.

B.

C.

0.

I With the primary containment atmosphere gaseous radioactivity moni toring system inoperable, analyze grab samples o f primary containment atmosphere 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 /> AND restore primary containment atmosphere gaseous radioactivity monitoring system t o OPERABLE status within 30 days.

With the drywell f l o o r drain sump flow monitoring drywell f l o o r drain sump f l o w monitoring system t o AN0 increase monitoring frequency o f drywell u n i t cooler condensate f l o w rate (SR 4.4.3.2.1.c) t o once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

A-case With the drywell u n i t coolers condensate flow rate AND the primary containment atmosphere gaseous radioactivity OPERABLE, perform a channel check o f the primary containment atmosphere gaseous radioactivity monitoring system (SR 4.4.3.1.a) once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

mary containment pressure and temperature moni t o r i n g system restore the primary containment pressure and temperature monitoring ERABLE status within 30 days.

NOTE:

A l l other Tech Spec L i m i t i n q ce Rea.ui rement s associated with the primarv co ntainrnent pressure/tgmper ature moni torincr S Y S tern s t i l l app l v.,Affected Tech Spec Sec tions include:

3/4. 3.7.5.

4.4.3.2.1.

3/4.6.1.6.

a nd 3/4.6.1.7, mary containment atmosphere gaseous radioactivity monitoring system NO the drywell unit coolers condensate flow r a t e monitoring system restore the primary containment atmosphere gaseous radioactivity stem t o OPERABLE status within 30 days OR restore the drywell u n i t coolers condensate flow rate monitoring system t o OPERABLE status within 30 days.

With the primary containment atmosphere gaseous radioactivity monitoring system inoperable, analyze grab samples o f primary Containment atmosphere 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 />.

w c LIMERICK - UEIIT 1 3/4 4-8 Amendment U, 444, 169

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooling systems shall be OPERABLE with:

a.

b.

C.

d.

APPLICABILITY :

The core spray system (CSS) consisting o f two subsystems with each subsystem comprised of:

2.

An OPERABLE o f taking suction from the the water through the spray sparger t o the reactor vessel.

The low pressure coolant injection (LPCI) system o f the residual heat removal system consisting o f four subsystems with each subsystem comprised of:

1.

One OPERABLE LPCI pump, and

2.

An OPERABLE flow path capable o f taking suction from the suppression chamber and transferring the water t o the reactor vessel.

The high pressure coolant injection (HPCI) system consisting of:

1.

One OPERABLE HPCI pump, and

2.

An OPERABLE flow path capable o f taking suction from the suppression chamber and transferring the water t o the reactor vessel.

The automatic depressurization system (ADS) with a t least f i v e OPERABLE ADS valves.

OPERATIONAL CONDITION 1, 2" ** #, and 3* ** ##.

• The HPCI system i s not required t o be OPERABLE when reactor steam dome pressure i s less than o r equal t o 200 psig.

• • The ADS i s not required t o be OPERABLE when the reactor steam dome pressure i s less t h a t or equal t o 100 psig.

1. See Special Test Exception 3.10.6.

1. 1. Two LPCI subsystems o f the RHR system may be inoperable i n that they are aligned i n the shutdown cooling mode when reactor vessel pressure i s less than the RHR Shutdown cooling permissive setpoint.

CONTAINMENT SYSTEMS 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES 3.6.3 Each primary ainment i s o l a t i o n v rumentation l i n e excess flow check v a l a l l be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIQNS 1, 2, and 3.

ACTION :

a.

With one or more o f the primary containment i s o l a t i o n valves inoperable,**

maintain a t least one i s o l a t i o n v a l v e OPERABLE i n each a f f e c t e d penetration t h a t i s open and 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 /> either:

1.

Restore the inoperable valve(s1 t o OPERABLE status, o r

2.

I s o l a t e each affected penetration by use o f a t l e a s t one de-activated automatic valve secured i n the i s o l a t e d position,* or

3.

I s o l a t e each affected penetration by use o f a t l e a s t one closed manual valve o r b l i n d flange."

Otherwise, be i n 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 t h e following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

With one o r more o f the instrumentation l i n e excess f l o w check valves inoperable, operation may continue and the provisions o f Specification 3.0.3 are not applicable provided t h a t 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 /> e i t h e r :

1.

The inoperable valve i s returned t o OPERABLE status, or

2.

The instrument l i n e i s isolated and the associated instrument i

s decl ared i noperabl e.

Otherwise, be i n a t 7east 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 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.

With one o r more scram discharge volume vent o r d r a i n valves inoperable, perform the applicable actions specified i n Specification 3.1.3.1.

• I s o l a t i o n valves closed t o s a t i s f y these requirements may be reopened on an i n t e r m i t t e n t basis under administrative control.

• • Except f o r the scram discharge volume vent and drain valves.

LIMERICK - UNIT 1 314 6-17 Amendment No. 29, 446, W, 169

LIHERICK - UNIT 1 J

J

't DEC 2 0 1995 314 6-49 Amendment No. 23,105

LIMERICK - UNIT 1 THE INFORMATION FROM THIS TECHNICAL SPEC FICATIONS SECTION HAS BEEN RELOCATED TO THE&

c 7 R m DEC 2 0 1995 Amendment No. 23,105 314 6-51

LIMERICK - UNIT 1 THE INFORMATION FROM THIS TECHNICAL SECTION HAS BEEN RELOCATED TO THE 314 6-51a DEC 2 0 l9¶5 Amendment No. 23,105

PLANT SYSTEMS M l T I N G W I T I O N FOR OPFRATION (Continued)

ACTION:

(Continued)

4.

With three ESW pump/diesel generator pairs** inoperable, r e s t o r e a t l e a s t one inoperable ESW pump/diesel generator pair** t o OPERABLE status w i t h i n 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, o r be i n 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 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 />.

5.

With f o u r ESW pump/diesel generator pairs** inoperable, r e s t o r e a t l e a s t one inoperable ESW pumpidiesel generator pair** t o OPERABLE status w i t h i n 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-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 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 />.

b.

In OPERATIONAL CONDITION 4 o r 5:

1.

With only one emergency service water pump and i t s associated flowpath OPERABLE, restore a t l e a s t two pumps w i t h a t l e a s t one f l o w path t o OPERABLE status w i t h i n 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> o r declare t h e associated safety r e l a t e d equipment inoperable and take t h e ACTION required by Specifications 3.5.2 and 3.8.1.2.

C.

I n OPERATIONAL CONDITION *

1.

With only one emergency service water pump and i t s associated f l o w path OPERABLE, r e s t o r e a t l e a s t two pumps w i t h a t l e a s t one f l o w path t o OPERABLE status w i t h i n 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> o r v e r i f y adequate cooling remains a v a i l a b l e f o r t h e d i e s e l generators required t o be OPERABLE o r declare t h e associated diesel generator(s1 inoperable and take t h e ACTION required by S p e c i f i c a t i o n 3.8.1.2.

are not applicable.

The provisions o f S p e c i f i c a t i o n 3.0.3 4.7.1.2 s h a l l be demonstrated OPERABLE:

A t l e a s t t h e above required emergency service water system loop(s)

a.

I n accordance w i t h t h e Surveillance Frequency Control Program by I

v e r i f y i n g t h a t each valve (manual, power-operated, o r automatic) t h a t i s n o t locked, sealed, o r otherwise secured i n p o s i t i o n, i s i n i t s c o r r e c t p o s i t i on.

b.

In accordance w i t h the Surveillance Frequency Control Program by I

v e r i f y i ng t h a t :

1.

Each automatic valve actuates t o i t s c o r r e c t p o s i t i o n on i t s appropriate ESW pump s t a r t s i g n a l.

2.

Each pump s t a r t s automatically when i t s associated diesel generator s t a r t s.

• When handling i r r a d i a t e d f u e l i n the secondary containment.

• • An ESW pump/diesel generator p a i r consists o f an ESW pump and i t s associated d i e s e l generator.

I f e i t h e r an ESW pump o r i t s associated d i e s e l generator becomes inoperable, then t h e ESW pump/diesel generator p a i r i s inoperable.

LIMERICK - UNIT 1 3/4 7 - 4 Amendment No. 2J, 48, 74, 186

ELECTRICAL muER SYnrWs LIMITING COM)ZTfON FOR OPERATION 3.8.1.2 As 0 mSnhme tha following A.C.

electrical pow? sources shill h OPERABE:

1.

One c h u f f bokrcHn the o f f s f k t r m S. i S S h network and the onsf-Class lE distribution system,. and' l o 20

3.

APPLICABILITY:

ACTIW:

I am W i t h less thur the above required A.C.

elcrctrfcal power sources OPERABLE, suspend W E ALTERATIOWS, W l f n g of irradiated furl i n the seconday containment, operations with (I potential for draining the mactOr vessel and cram oprrrtions ovw t)n spent fuel storr pool when fuel urWlhrr 813) stand tlmmtn.

In addltjon, ukn n OPERATIOlJAL ITI ION 5 with the water level less thur 22 fmt above the mutor pressure vessel f l a w, imdiately-initiate correct+ve action to -ston th r e q u i d powr sources to OPEMU status as soon as practical.

The pro~$sfons of SpdffUtiOn 3e0.3 am not 8pplhble.

r

b.

SURVEfLUNCE REao IRE?EMTS I

W h n handling fwrdirtad fwl i n the secondry totrtrimmnt.

LIERICK - UNIT 1

's-

Section 3/4.8.4.1 (Deleted)

SECTION PAGES 3/4 8-21 INTENTIONALLY OMITTED.

LIMERICK - UNIT 1 314 8-21 Amendment No. 32, 48, 31, 93 Jul 0 7 1995

INDEX SECTION DEFINITIONS (Continued)

PAGE 1.20a 1-4 LOW (POWER) TRIP SETPOINT (LTSP)............................

1-4 1.21 (DELETED)...................................................

1-4 1.22 MEMBER(S) OF THE PUBLIC....................................

1-4 1.22a MAPFAC( F).

(MAPLHGR FLOW FACTOR) 1-4 1.22b MAPFAC(P).

(POWER DEPENDENT MAPLHGR MULTIPLIER)...........

1-4 1.23 MINIMUM CRITICAL POWER RATIO (MCPR)........................

1-4 1.24 OFFSITE DOSE CALCULATION MANUAL............................

1-4 1.25 OPERABLE.

OPERABILITY.....................................

1-5 1.26 OPERATIONAL CONDITION.

CONDITION..........................

1.27 PHYSICS TESTS...............................................

1-5 1-5 1.28 PRESSURE BOUNDARY LEAKAGE..................................

1-5 1.29 PRIMARY CONTAINMENT INTEGRITY..............................

PROCESS CONTROL PROGWM....................................

1-5 1.30 PURGE.

PURGING 1-6 1.31 1.32 RATED THERMAL POWER........................................

1-6 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY..........

1.34 REACTOR PROTECTION SYSTEM RESPONSE T I EVENT...........................................

RGIN 1-7 SOURCE CHECK...............................................

1-7 1.41 LIMERICK.

UNIT 2 ii Amendment So. 11. 16. 48 JAN 3 7 1995

INDEX A D M I N I S T R A T I V F CONTROl S SECTION PAGE 6.5.2 Deleted Deleted.................................................... 6 - 9 Deleted....................................................

6 - 9 Deleted....................................................

6-10 Deleted....................................................

6-10 Deleted.................................................... 6-10 Deleted....................................................

6-10 Deleted.................................................... 6-10 Deleted....................................................

6-12 6.5.3 Deleted....................................................

6-12 6.6 REPORTABLE EVENT ACT10.............................................

6-12a 6. 7 N L I M I T V I O C A T I Q N.............................................

6-12a 6.8 PROC EDU RES AND PROGRAMS............................................

6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS............................................

6-15 Startup Report.............................................

6-15 Annual Reports.............................................

6-15 Monthly Operating Reports..................................

6-16 Annual Radiological Environmental Operating Report......... 6-16

@nudl Radioactive E f f l u e n t Release Report............. 6-17 CORE OPERATING LIMITS REPORTS..............................

6-18a 6.9.2 SPECIAL REPORTS............................................

6-18a 6.10 DELETED...........................................................

6-19 6.11 RADIATION PROTECTION PROGRAM......................................6-20 6.12 HIGH RADIATION AREA...............................................

6-20 I

I I

I I

I I

I I

I LIMERICK - UNIT 2 x x v i i Amendment No. 4, a, 138

DFFINITIWS l O G I C SYSTFM FUNCTIONAL TEST 1.20 A LOGIC SYSTEM FUNCTIONAL TEST shall be a t e s t o f a l l l o g i c components, i.e.,

a l l relays and contacts, a l l t r i p units, s o l i d state l o g i c elements, etc, o f a l o g i c c i r c u i t, from sensor through and including the actuated device, t o v e r i f y OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping o r t o t a l system steps such t h a t the e n t i r e l o g i c system i s tested.

!OW (POWER) TRIP SETPOINT (LTSP) 1.20a The low power t r i p setpoint associated with the Rod Block Monitor (RBM) rod block t r i p setting applicable between 30% and 65% reactor thermal power.

1.21 (Deleted MEMBFR(S1 OF THF PUBLIC 1.22 MEMBER OF THE PUBLIC means any individual except when t h a t individual i s receiving an occupational dose.

I MA 1 1.22a A core flow dependent m u l t i p l i c a t i o n f a c t o r used t o flow bias the standard Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) l i m i t.

MAPFAC(P1-(POWER DEPENDENT MAPIHGR MULTIPLIER) 1.22b A core power dependent mu1 t i p 1 i c a t i o n factor used t o power bias the standard Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) l i m i t.

MINIMUM CRITICAI POWER RATIO (MCPR) 1.23 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists i n the core ( f o r each class o f f u e l ).

Associated w i t h the minimum c r i t i c a l power r a t i o i s a core flow dependent (MCPR(F)) and core power dependent (MCPR( P ) ) m i nimurn c r i t i c a l power r a t i o.

OFFSITE DOSF CALCUIATION MANUAL 1.24 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used i n t h e calculation o f o f f s i t e doses r e s u l t i n g from radioactive gaseous and l i q u i d effluents, i n the calculation o f gaseous and l i q u i d e f f l u e n t m o r i n g alarm/trip setpoints, and i n the conduct o f the Radiological v i onmental Monitoring Program.

The ODCM shall also contain (1) @

i o a c t i ve E f f l uent Control s and Radio1 ogical Environmental Mon o

g Programs required by Section 6.8.4 and (2) descriptions o f the information t h a t should be included i n the Annual Radiological Environmental Operating and Annual Radioactive E f f l u e n t Release Reports required by Specifications 6.9.1.7 and 6.9.1.8.

OPERABLE - OPERABILITY 1.25 A system, subsystem, t r a i n, component o r device s h a l l be OPERABLE o r have OPERABILITY when i t i s capable o f performing i t s specified function(s) and when a l l necessary attendant instrumentation, controls, e l e c t r i c a l power, cooling o r seal water, l u b r i c a t i o n o r other a u x i l i a r y equipment t h a t are required for the system, subsystem, t r a i n, c c ~ ~ o n e n t,

o r device t o perform i t s function(s1 are also capable o f performing t r r related support function(s1.

LIMERICK - UNIT 2 Amendment No. 44, %,

148

Section 3.3.7.2 (Deleted)

FROM THIS TECHNICAL ION HAS BEEN RELOCATED CHNICAL SPECIFICATIONS PAGES 3/4 3-69 THROUGH 3/4 3-72 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.

LIMERICK - UNIT 2 314 3-68

Section 3/4.3.8 (Deleted)

THE INFORMATION FROn THIS TECHNICAL HAS BEEN INTENTIONALLY OUInED.

LIMERICK - UNIT 2 SIP 2 1 1995 311 3-110 Amendment No. 64

REACTOR COOL ANT SYSTEM 3/4.4,3 LFAKAGE DETFCTION SYSTFMS REACTOR COOL ANT SYSTEM LEAKAGE 3.4.3.1 be OPERABLE:

The fol'lowing reactor Coolant leakage detection systems shall

a.

The primary containment atmosphere gaseous radioactivity monitoring system,

b.

The drywell f l o o r drain sump flow monitoring system,

c.

The drywell unit coolers condensate f l o w rate monitoring system, and d.

The primary containment pressure and temperature monitoring system.

APPLICABILI TY:

OPERATIONAL CONDITIONS 1, 2, and 3."

• - The primary containment gaseous radioactivity monitor i s not required t o be operable u n t i l Operational Condition 2.

ACTIQNS:

I A.

With the primary containment atmosphere gaseous radioactivity monitoring system inoperable, analyze grab samples o f primary containment atmosphere 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 /> AND restore primary containment atmosphere gaseous r a d i o a c t i v i t y monitoring system t o OPERABLE status within 30 days.

B.

With the drywell f l o o r drain sump flow monitoring system drywell floor drain sump flow monitoring system t o AND increase monitoring frequency o f drywell u n i t cooler condensate flow r a t e (SR 4.4.3.2.1.c) t o once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C.

With the drywell u n i t coolers condensate flow rate monitoring s y s t e m W G 3 AND the primary containment atmosphere gaseous radioactivity monitoring system OPERABLE, perform a channel check o f the primary containment atmosphere gaseous radioactivity monitoring system (SR 4.4.3.1.a) once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

rimary containment pressure and temperature monitoring system restore the primary containment pressure and temperature monitoring ERABLE status within 30 days.

Note:

All other Tech SDec Limitinq r a t i o n and Survei 11 ance Reuui rernents assoc i a t e d with the Tech Spec S ections include:

3/4.3.7.5.

4.4.3.7.1.

3/4. 6.1.6. and 3/4.6.1. 7, g r i m a r v c ontainment pressure/temperature rnonitorinq S Y S tem s t i l l app IY.

Affected r i m a r y containment atmosphere gaseous r a d i o a c t i v i t y monitoring system the drywell u n i t coolers condensate flow r a t e monitoring system the primary containment atmosphere gaseous r a d i o a c t i v i t y OPERABLE status within 30 days OR restore the drywell u n i t coolers condensate f l o w rate monitoring system t o OPERABLE status within 30 days.

with the primary containment atmosphere gaseous r a d i o a c t i v i t y monitoring system inoperable, analyze grab samples o f primary containment atmosphere 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 />.

LIMERICK - UNIT 2 3/4 4-8 Amendment No. 34, 1-83, 132

REACTOR COOLANT SYSTEM ACTIONS (Continued)

,. With any o t h e r two o r more leak d e t e c t i o n systems inoperable o t h e r than ACTION E above OR w i t h r e q u i r e d Actions and associated Completion Time o f ACTIONS A, B,

C, D o r E n o t met, be i n 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 t h e n e x t 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS

-~

4.4.3.1 The r e a c t o r c o o l a n t system leakage d e t e c t i o n systems s h a l l be demonstrated operable by:

a.

Perform a CHANNEL CHECK o f t h e primary containment atmosphere gaseous r a d i o a c t i v i t y m o n i t o r i n g system i n accordance w i t h t h e Survei 11 ance Perform a CHANNEL FUNCTIONAL TEST Frequency C o n t r o l Program.

i n s t r u m e n t a t i o n i n accordance w i t h Program.

m o n i t o r i n g system.

qui r e d leakage d e t e c t i o n

b.

he S u r v e i l lance Frequency Control T h i s does n o t apply t o containment pressure and temperature

c.

Perform a CHANNEL CALIBRATION o f r e q u i r e d leakage d e t e c t i o n i n s t r u m e n t a t i o n n o t apply t o containment pressure and temperature m o n i t o r i n g system.

i n accordance w i t h t h e Survei 11 ance Frequency Control Program.

Thi s does I

I d.

Monitor p r i m a r y containment pressure &NJ primary containment temperature i n accordance w i t h t h e Survei 11 ance Frequency Control Program.

LIMERICK - UNIT 2 3/4 4-8a Amendment No. 1-Bc3,147

3 / 4 3 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cool iig systems shall be OPERABLE with:

a.

The core spray system (CSS) consisting of two subsystems with each subsystem comprised of:

suction from the the water through the spray sparger to the reactor vessel.

b.

The low pressure coolant injection (LPCI) system o f the residual heat removal system consisting o f four subsystems with each 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.

c.

The high pressure coolant injection (HPCI) system consisting of:

1.

One OPERABLE HPCI pump, and

2.

An OPERABLE flow path capable o f taking suction from the suppression chamber and transferring the water to the reactor vessel.

d.

The automatic depressurization system (ADS) with at least five OPERABLE ADS valves.

APPLICABILITY:

OPERATIONAL CONDITION 1, 2* ** #, and 3" ** X f.

• The HPCI system is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.

• • The ADS is not required to be OPERABLE when the reactor steam dome pressure is less that or equal to 100 psig.

%See Special Test Exception 3.10.6.

1. 1. Two LPCI subsystems o f the RHR system may be inoperable in that they are aligned in the shutdown cooling mode when reactor vessel pressure is less than the RHR Shutdown cooling permissive setpoint.

LIMERICK - UNIT 2 314 5-1 Amendment No. %, 92

.EIoV 1 6 1998

CONTAINMENT SYSTEMS DRYWELL AVERAGE A I R TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.7 Drywell average a i r temperature s h a l l n o t exceed 1 4

5 e

d,/crle APPLICABILITY:

OPERATIONAL CONDITIONS I, 2, and 3.

abbh

- 0 ACTION With t h e d r y w e l l average a i r temperature greater than 145"@reduce the average a i r temperature t o w i t h i n t h e l i m i t w i t h i n 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-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 />.

SURVEILLANCE REQUIREMENTS 4.6.1.7 The drywell average a i r temperature s h a l l be the v o l u m e t r i c average of the temperatures a t t h e f o l l o w i n g l o c a t i o n s and s h a l l be determined t o be w i t h i n t h e l i m i t i n accordance w i t h t h e S u r v e i l l a n c e Frequency Control Program:

1 Approximate E l e v a t i o n a.

330 '

Number o f I n s t a l l e d Sensors*

3

b.

320 '

3 C.

260 '

3

d.

248' 6

A t l e a s t one reading from each e l e v a t i o n i s r e q u i r e d f o r a volumetric average a i r ternperatur u r r e n t o p e r a t i ng c y c l e n t d u r a t i o n t o a l l 0 cooler fan r e p a i r s, whichever comes f i r s t.

LIMERICK - UNIT 2 3/4 6-10 Amendment No. 424, 345, 147

CONTAINMENT SYSTEMS 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 Each primary containment i s o l a t i o n rumentation l i n e excess f l o w check valve shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACT I ON :

a.

With one o r more o f the primary containment i s o l a t i o n valves inoperable,**

maintain a t l e a s t one i s o l a t i o n valve OPERABLE i n each affected penetration that i s open and 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 /> either:

1.

Restore the inoperable valve(s) t o OPERABLE status, o r

2.

I s o l a t e each affected penetration by use o f a t l e a s t one de-activated automatic valve secured i n the i s o l a t e d position,* or

3.

I s o l a t e each affected penetration by use o f a t l e a s t one closed manual v a l v e or b l i n d flange.*

Otherwise, be i n 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 />.

b.

With one or more o f the instrumentation l i n e excess f l o w check valves inoperable, operation may continue and the provisions of Specification 3.0.3 are n o t applicable provided t h a t 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 /> e i t h e r :

1.

The inoperable valve i s returned t o OPERABLE status, o r

2.

The instrument l i n e i s isolated and the associated instrument i s declared inoperable.

Otherwise, be i n 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 />.

c.

With one or more scram discharge volume vent o r d r a i n valves inoperable, perform the applicable actions specified i n Specification 3.1.3.1.

• I s o l a t i o n valves closed t o s a t i s f y these requirements may be reopened on an intermi t t e n t basis under administrative control.

• • Except f o r the scram discharge volume vent and drain valves.

LIMERICK - UNIT 2 3/4 6-17 Amendment No. W, W, 132

LIMERICK - UNIT 2 THE INFORMATION FROM THIS TECHNICAL SPEC SECTION HAS BEEN RELOCATED TO THE 314 6-49 Amendment No. 69 DEC 2 0 895

THE INFORMATION FROM THIS TECHNICAL SP SECTION HAS BEEN RELOCATED TO THE LIMERICK - UNIT 2 3/4 6-51 Amendment No. 69 DEC 2 0 1995

LIMERICK - UNIT 2 THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO TH 314 6-51a Amendment No. 69 DEC 2 0 t995

PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FRESH A I R SUPPLY SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2 s h a l l be OPERABLE.

Two independent c o n t r o l room emergency f r e s h a i r supply system subsystems APPLICABILITY:

A l l OPERATIONAL CONDITIONS and when RECENTLY IRRADIATED FUEL i s being handled i n t h e secondary containment, o r d u r i n g operations w i t h a p o t e n t i a l f o r d r a i n i n g t h e reactor vessel.

ACTION :

a.

I n OPERATIONAL CONDITION 1, 2, o r 3:

1.

With t h e U n i t 1 d i e s e l generator f o r one c o n t r o l room emergency f r e s h a i r supply subsystem inoperable f o r more than 30 days, 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 owing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.

With one c o n t r o l room emergency f r e s h a i r supply subsystem inoperable, r e s t o r e the inoperable subsystem t o OPERABLE s t a t u s w i t h i n 7 days, o r be i n a t l e a s t HOT SHUTDOWN w i t h i n the 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 />.

3.

With one c o n t r o l room emergency f r e s h a i r supply subsystem inoperable and the o t h e r c o n t r o l room emergency f r e s h a i r supply subsystem w i t h an inoperable U n i t 1 diesel generator, r e s t o r e t h e inoperable subsystem t o OPERABLE s t a t u s or r e s t o r e t h e U n i t 1 d i e s e l generator t o OPERABLE s t a t u s w i t h i n 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 />.

4.

With t h e U n i t 1 d i e s e l generators f o r both c o n t r o l room emergency f r e s h a i r supply subsystems inoperable f o r more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, be i n 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 t h e f o l 1 owi ng 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

I n OPERATIONAL CONDITION 4, 5 o r when RECENTLY IRRADIATED FUEL i s being handled i n t h e secondary containment, o r d u r i n g operations w i t h a p o t e n t i a1 f o r d r a i n i n g t h e r e a c t o r vessel :

1.

2.

LIMERICK - UNIT 2 With one c o n t r o l room emergency f r e s h a i r supply subsystem inoperable, r e s t o r e t h e inoperable subsystem t o OPERABLE s t a t u s w i t h i n 7 days, o r i n i t i a t e and m a i n t a i n operation o f t h e OPERABLE subsystem i n t h e r a d i a t i o n i s o l a t i o n mode o f operation.

With b o t h 4 & & & t r o l room emergency f r e s h a i r supply subsystem secondary containment and operations w i t h a p o t e n t i a l f o r d r a i n i n g t h e r e a c t o r vessel.

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 3.0.3 a r e n o t appl i cab1 e.

dd&

i noperabl e, suspend hand1 i ng o f RECENTLY IRRADIATED FUEL i n t h e I

3/4 7-6 Amendment No. 432, 146

• '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 -ciI?cuft between 4he-offsite.transini-ssion network and the onsite Class lE distribution systm, and

b.

Two diesel generators each wfth:

1.

A day fuel tank containing a minimum o f 200 gallons of fuel.

2.

3.

A fuel transfer pump.

A fuel storage system containing a ainimm of 33,500 gallons o f fuel.

APPLICABILIfY: OPERATIONAL CONDITIONS 4, 5, and *.

ACTION:

a.

With less than the above required A.C.

electrical power sources OPERABLEs suspend CORE ALTERATIONS, handling o f irradiated fuel i n the secondary containment, operations with a potential for draining the reactor vessel and crane operations over the spent fuel storage pool when fuel assWlies are stored therein.

In addition, when i n 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 mot applicable.

SURVEILLANCE REQUIREHENTS 4.8.1.2 A t least the above required A.C. electrical Suwei 1 1 anct Requi rements qhen hand1 i ng i rrad fated fuel i n the secondary contai m n t.

LIMERICK = UNIT 2 314 8-9 M625-

Section 3/4 8.4.1 (Deleted) 7

/flH THE INFORMATION FROM TH AL SPECIFICATION SECTION HAS BEEN RELOCATED TO TH TECHNICAL SPECIFICATIONS INTENTIONALLY OMITTED.

PAGES 3/4 8-21 THROUGH F THIS SECTION HAVE BEEN LIMERICK - UNIT 2 314 8-21 Amendment No. 34, 57 Jut 0 7 1995