Text

Issuance of Amendments Elimination of Certain Technical Specifications Reporting Requirements
	ML13214A092
Person / Time
Site:	Limerick
Issue date:	03/26/2014
From:	Richard Ennis; Plant Licensing Branch 1
To:	Pacilio M; Exelon Nuclear
	Ennis R
References
	TAC MF1399, TAC MF1400
	Download: ML13214A092 (32)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 26, 2014 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LIMERICK GENERATING STATION, UNITS 1 AND 2-ISSUANCE OF AMENDMENTS RE: ELIMINATION OF CERTAIN TECHNICAL SPECIFICATION REPORTING REQUIREMENTS (TAC NOS. MF1399 AND MF1400)

Dear Mr. Pacilio:

The Commission has issued the enclosed Amendment Nos. 211 and 172 to Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station, Units 1 and 2, respectively.

These amendments consist of changes to the Technical Specifications (TSs) and Facility Operating Licenses in response to your application dated April 9, 2013.

The amendments eliminate certain TS reporting requirements.

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket Nos. 50-352 and 50-353

Enclosures:

1. Amendment No. 211 to License No. NPF-39

2. Amendment No. 172 to License No. NPF-85

3. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, Richard B. Ennis, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-352 LIMERICK GENERATING STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 211 License No. NPF-39

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Exelon Generation Company; LLC (the licensee), dated April 9, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-39 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 211, are hereby incorporated into this license.

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

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: March 26, 2014 FOR THE NUCLEAR REGULATORY COMMISSION Meena K. Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 211 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 Replace the following page of the Facility Operating License with the revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

Remove Page 3 Insert Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove 3/4 5-3 3/4 6-8 3/4 7-9 3/4 7-18 6-15 6-16 Insert 3/4 5-3 3/4 6-8 3/4 7-9 3/4 7-18 6-15 6-16 (3)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Pursuant to the Act and 10 CFR Parts 30, 40, 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility, and to receive and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and fuel channels from the Shoreham Nuclear Power Station.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.0. below) and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level Exelon Generation Company is authorized to operate the facility at reactor core power levels not in excess of 3515 megawatts thermal

( 1 00% rated power) in accordance with the conditions specified herein and in Attachment 1 to this license. The items identified in Attachment 1 to this license shall be completed as specified. Attachment 1 is hereby incorporated into this license.

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 211, are hereby incorporated into this license.

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

Amendment No. 211

EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

c.

(Continued)

For the HPCI system:

1.

With the HPCI system inoperable, provided the CSS, the LPCI system, the ADS and the RCIC system are OPERABLE, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to~ 200 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With the HPCI system inoperable, and one CSS subsystem, and/or LPCI subsystem inoperable, and provided at least one CSS subsystem, three LPCI subsystems, and ADS are operable, restore the HPCI to OPERABLE within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be in HOT SHUTDOWN in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and in COLD SHUTDOWN in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

Specification 3.0.4.b is not applicable to HPCI.

d.

For the ADS:

1.

With one of the above required ADS valves inoperable, provided the HPCI system, the CSS and the LPCI system are OPERABLE, restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to~ 100 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to~ 100 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e.

With a CSS and/or LPCI header 6P instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or determine the ECCS header 6P locally at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; otherwise, declare the associated CSS and/or LPCI, as applicable, inoperable.

f.

DELETED LIMERICK UNIT 1 3/4 5-3 Amendment No. JJ, 94, +69, 211

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.5 The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.5.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.1.5.1 The structural integrity of the exposed accessible interior and exterior surfaces of the primary containment, including the liner plate, shall be determined by a visual inspection of those surfaces.

This inspection shall be performed in accordance with the Primary Containment Leakage Rate Testing Program.

4.6.1.5.2 DELETED LIMERICK - UNIT 1 3/4 6-8 Amendment No. t+&, 211

PLANT SYSTEMS 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig.

ACTION:

a.

With the RCIC system inoperable, operation may continue provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

DELETED

c.

Specification 3.0.4.b is not applicable to RCIC.

SURVEILLANCE REQUIREMENTS 4.7.3 The RCIC system shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.

2.

Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

3.

Verifying that the pump flow controller is in the correct position.

b.

In accordance with the Surveillance Frequency Control Program by verifying that the RCIC pump develops a flow of greater than or equal to 600 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1040 + 13, - 120 psig.*

The prov1s1ons of Specification 4.0.4 are not applicable, provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam pressure to less than 150 psig within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

LIMERICK - UNIT 1 3/4 7-9 Amendment No. ~. ~. ~. +&e, 211

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b.

Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months.

Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

c.

Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.

4.7.5.3 DELETED LIMERICK - UNIT 1 3/4 7-18 Amendment No. 211

ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT

6. 9. 1. 1 Deleted

6. 9. 1. 2 Deleted 6.9.1.3 Deleted ANNUAL REPORTS*

6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year unless otherwise noted.

6.9.1.5 Reports required on an annual basis shall include:

a.

Deleted

A single submittal may be made for a multiple unit station.

LIMERICK - UNIT 1 6-15 Amendment No. ~.++S, 211

ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued)

b.

(Deleted)

c.

(Deleted)

d.

(Deleted)

MONTHLY OPERATING REPORTS*

6.9.1.6 Deleted ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*

6.9.1.7 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year.

The report shall include summaries, interpreta-tions, analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period.

The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

A single submittal may be made for a multiple unit station.

LIMERICK - UNIT 1 6-16 Amendment No. ~.4G,4S,~,+&9.

-+/-+§., 211

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-353 LIMERICK GENERATING STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 172 License No. NPF-85

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Exelon Generation Company, LLC (the licensee), dated April 9, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-85 is hereby amended to read as follows:

2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 172, are hereby incorporated into this license.

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

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

'-Y'vv~

Meena K. Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: March 26, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 172 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 Replace the following page of the Facility Operating License with the revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

Remove Page 3 Insert Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove 3/4 5-3 3/4 6-8 3/4 7-9 3/4 7-18 6-15 6-16 Insert 3/4 5-3 3/4 6-8 3/4 7-9 3/4 7-18 6-15 6-16 (4)

Pursuant to the Act and 10 CFR Parts 30, 40, 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility, and to receive and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and fuel channels from the Shoreham Nuclear Power Station.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.0. below) and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level Exelon Generation Company is authorized to operate the facility at reactor core power levels of 3515 megawatts thermal (1 00 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 172, are hereby incorporated into this license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

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

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

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

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

Amendment No. 172

EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

c.

(Continued)

For the HPCI system:

1.

With the HPCI system inoperable, provided the CSS, the LPCI system, the ADS and the RCIC system are OPERABLE, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to~ 200 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With the HPCI system inoperable, and one CSS subsystem, and/or LPCI subsystem inoperable, and provided at least one CSS subsystem, three LPCI subsystems, and ADS are operable, restore the HPCI to OPERABLE within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be in HOT SHUTDOWN in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and in COLD SHUTDOWN in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

Specification 3.0.4.b is not applicable to HPCI.

d.

For the ADS:

1.

With one of the above required ADS valves inoperable, provided the HPCI system, the CSS and the LPCI system are OPERABLE, restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to~ 100 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to ~ 100 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e.

With a CSS and/or LPCI header 6P instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or determine the ECCS header 6P locally at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; otherwise, declare the associated CSS and/or LPCI, as applicable, inoperable.

f.

DELETED LIMERICK - UNIT 2 3/4 5-3 Amendment No. &S, ~. 172

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.5 The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.5.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.1.5.1 The structural integrity of the exposed accessible interior and exterior surfaces of the primary containment, including the liner plate, shall be determined by a visual inspection of those surfaces.

This inspection shall be performed in accordance with the Primary Containment Leakage Rate Testing Program.

4.6.1.5.2 DELETED LIMERICK - UNIT 2 3/4 6-8 Amendment No. S+/-, 172

PLANT SYSTEMS 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig.

ACTION:

a.

With the RCIC system inoperable, operation may continue provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

DELETED

c.

Specification 3.0.4.b is not applicable to RCIC.

SURVEILLANCE REQUIREMENTS 4.7.3 The RCIC system shall be demonstrated OPERABLE:

a.

In accordance with the Surveillance Frequency Control Program by:

1.

Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.

2.

Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

3.

Verifying that the pump flow controller is in the correct position.

b.

In accordance with the Surveillance Frequency Control Program by verifying that the RCIC pump develops a flow of greater than or equal to 600 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1040 + 13, - 120 psig.*

The prov1s1ons of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressure to less than 150 psig within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

LIMERICK - UNIT 2 3/4 7-9 Amendment No.~.~.+++. 172

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b.

Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months.

Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

c.

StartuR sources and fission detectors - Each sealed startup source*

and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.

4.7.5.3 DELETED

Except the Cf-252 startup sources which shall be tested within 6 months prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.

LIMERICK - UNIT 2 3/4 7-18 Amendment No. 172

ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the fallowing reports shall be submitted to the Regi anal Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT 6.9.1.1 Deleted 6.9.1.2 Deleted 6.9.1.3 Deleted ANNUAL REPORTS*

6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year unless otherwise noted.

6.9.1.5 Reports required on an annual basis shall include:

a.

Deleted

A single submittal may be made for a multiple unit station.

LIMERICK - UNIT 2 6-15 Amendment No. +QG,+&J, 172

ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued)

b.

(Deleted)

c.

(Deleted)

d.

(Deleted)

MONTHLY OPERATING REPORTS*

6.9.1.6 Deleted ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*

6.9.1.7 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year.

The initial report shall be submitted prior to May 1 of the year following initial criticality.

The report shall include summaries, interpretations, analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period.

The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

A single submittal may be made for a multiple unit station.

LIMERICK - UNIT 2 6-16 Amendment No.~.~.~.~. 172

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 211 AND 172 TO FACILITY OPERATING LICENSE NOS. NPF-39 AND NPF-85 EXELON GENERATION COMPANY, LLC LIMERICK GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353

1.0 INTRODUCTION

By application dated April 9, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13100A020), Exelon Generation Company, LLC (Exelon, the licensee), requested changes to the Technical Specifications (TSs) for Limerick Generating Station, Units 1 and 2 (Limerick). The proposed amendments would eliminate certain TS reporting requirements.

2.0 REGULATORY EVALUATION

The U.S. Nuclear Regulatory Commission's (NRC's or the Commission's) regulatory requirements related to the content of the TSs are set forth in Title 10 of the Code of Federal Regulations ( 10 CFR) Section 50. 36, "Technical specifications." This regulation requires that the TSs include items in the following five specific categories: ( 1) safety limits, limiting safety system settings, and limiting control settings; (2) Limiting Conditions for Operation (LCOs);

(3) Surveillance Requirements (SRs); (4) design features; and (5) administrative controls.

The regulation does not specify the particular requirements to be included in a plant's TSs.

As discussed in 10 CFR 50.36(c)(2), LCOs are the lowest functional capability or performance level of equipment required for safe operation of the facility. When LCOs are not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCOs can be met.

As discussed in 10 CFR 50.36(c)(3), SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

As discussed in 10 CFR 50.36(c)(5), administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

Appendix J to 10 CFR Part 50, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," requires that the results of the containment integrated leak rate testing, including the visual inspection, be documented and readily available on site for inspection by the NRC.

The requirements for making prompt telephone notifications and submitting written reports to the NRC are discussed in detail in 10 CFR 50.72, "Immediate notification requirements for operating nuclear power reactors," and in 10 CFR 50.73, "Licensee event report system."

These two regulations cover a broad spectrum of events, including emergency system actuation.

3.0 TECHNICAL EVALUATION

3.1 LCO 3.5.1, Action f-Emergency Core Cooling System (ECCS) Actuation Limerick LCO 3.5.1, Action f, currently states that:

In the event an ECCS system is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0. 70.

As discussed in the application dated April 9, 2013, the licensee proposed to delete the above TS reporting requirement based on the requirements in 10 CFR 50.73(a)(2)(iv). Specifically, that regulation provides the requirements for a licensee to submit a Licensee Event Report (LER) in the event of an ECCS actuation. In accordance with 10 CFR 50.73(a), the LER must be submitted within 60 days.

Although 10 CFR 50.73 requires the licensee to submit an LER in the event of an ECCS actuation, 10 CFR 50.73 does not cover all the requirements currently specified in LCO 3.5.1, Action f. Specifically, LCO 3.5.1, Action f, requires that the licensee provide the "total accumulated actuation cycles to date" in the Special Report. The current value of the usage factor is also to be provided in the Special Report "whenever its value exceeds 0.70." The purpose of these requirements is to provide assurance of the integrity of the components in the reactor coolant pressure boundary (RCPB) for the design life of the plant due to thermal cycles (i.e., injection of water into the reactor vessel by the ECCS results in a thermal cycle). As discussed in Limerick Updated Final Safety Analysis Report (UFSAR) Section 5.2.2.4.2, the specified operating transients for components within the RCPB are given in UFSAR Table 5.9-2, "RCPB Operating Thermal Cycles." Similar information is contained in UFSAR Table 3.9-2, "Plant Events." In addition, TS 5.6, requires that the thermal and pressure transient limits for the reactor vessel be maintained within the limits of TS Table 5. 6.1-1, "Component Cyclic or Transient Limits." Limerick surveillance test procedures ST-1-107-640-1 (Unit 1) and ST-1-107-640-2 (Unit 2), both titled, "Reactor Vessel Thermal Transient Monitoring," record and tabulate transient cycle events to ensure that the limits specified in UFSAR Tables 3.9-2 and 5.2-9 and TS Table 5.6.1-1 are not exceeded. Corrective action is taken if the cumulative cycles exceed 80 percent of the UFSAR or TS limits.

The NRC staff finds that the reporting requirements in 10 CFR 50.73, the limits in UFSAR Tables 3.9-2 and 5.9-2, the limits in TS Table 5.6.1-1, and the monitoring provided by the Limerick surveillance procedures noted above, provide reasonable assurance that ECCS actuations will be appropriately tracked consistent with assuring that the integrity of the components in the RCPB will be maintained within design limits. Therefore, the NRC staff concludes that deletion of LCO 3.5.1, Action f, is acceptable.

3.2 SR 4.6.1.5.2-Primary Containment Structural Integrity Limerick SR 4.6.1.5.2 currently states the following:

Reports Any abnormal degradation of the primary containment structure detected during the above required inspections shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days. This report shall include a description of the condition of the liner and concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken.

As discussed in the application dated April 9, 2013, the licensee proposed to delete the above TS reporting requirement based on the requirements in 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."

The "above required inspections" cited in SR 4.6.1.5.2 refers to the inspections required by SR 4.6.1.5.1, which states:

The structural integrity of the exposed accessible interior and exterior surfaces of the primary containment, including the liner plate, shall be determined by a visual inspection of those surfaces. This inspection shall be performed in accordance with the Primary Containment Leakage Rate Testing Program.

The Primary Containment Leakage Rate Testing Program is discussed in Limerick UFSAR Section 6.2.6. The UFSAR states, in part, that the program complies with 10 CFR Part 50, Appendix J, Option B, "Performance-Based Requirements," to the greatest extent practicable.

The UFSAR states that, in accordance with the program, a general inspection of the accessible interior and exterior surfaces of the primary containment structure and components is performed to uncover any evidence of structural deterioration that could affect either the containment structural integrity or leak-tightness. If there is evidence of structural deterioration, corrective action is taken. The structural deterioration and corrective action are reported in accordance with 10 CFR Part 50, Appendix J. Appendix J requires that the results of the containment leak rate testing, including the visual inspection, be documented and available on site for inspection by the NRC.

As discussed above in safety evaluation (SE) Section 2.0, the requirements for SRs is provided in 10 CFR 50.36(c)(3). SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met. The LCO associated with SRs 4. 6.1. 5.1 and 4.6.1.5.2, LCO 3.6.1.5, requires that the structural integrity of the primary containment be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.5.

The NRC staff finds that the inspections required by SR 4.6.1.5.1 provide reasonable assurance that the structural integrity of the primary containment will be maintained as required by LCO 3.6.1.5. As such SR 4.6.1.5.1, in itself, is sufficient to meet the requirements in 10 CFR 50.36(c)(3). Therefore, the NRC staff concludes that it is acceptable to delete SR 4.6.1.5.2.

The deletion of SR 4.6.1.5.2 does not change the existing reporting requirements associated with the Primary Containment Leakage Rate Testing Program, as described in the UFSAR, and as required by Appendix J of 10 CFR Part 50.

3.3 LCO 3.7.3, Action b-Reactor Core Isolation Cooling (RCIC) System Actuation Limerick LCO, Action b, currently states that:

In the event the RCIC system is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

As discussed in the application dated April 9, 2013, the licensee proposed to delete the above TS reporting requirement based on the requirements in 10 CFR 50.73(a)(2)(iv). Specifically, that regulation provides the requirements for a licensee to submit an LER in the event of an RCIC actuation. In accordance with 10 CFR 50.73(a), the LER must be submitted within 60 days.

Although 10 CFR 50.73 requires the licensee to submit an LER in the event of an RCIC actuation, 10 CFR 50.73 does not cover all the requirements currently specified in LCO 3.7.3, Action b. Specifically, LCO 3.7.3, Action b, requires that the licensee provide the "total accumulated actuation cycles to date" in the Special Report. The purpose of this requirement is to provide assurance of the integrity of the components in the RCPB for the design life of the plant due to thermal cycles (i.e., injection of water into the reactor vessel by the RCIC system results in a thermal cycle). As discussed in Limerick UFSAR Section 5.2.2.4.2, the specified operating transients for components within the RCPB are given in UFSAR Table 5.9-2, "RCPB Operating Thermal Cycles." Similar information is contained in UFSAR Table 3.9-2, "Plant Events." In addition, TS 5.6, requires that the thermal and pressure transient limits for the reactor vessel be maintained within the limits of TS Table 5.6.1-1, "Component Cyclic or Transient Limits." Limerick surveillance test procedures ST-1-107-640-1 (Unit 1) and ST-1-107-640-2 (Unit 2), both titled, "Reactor Vessel Thermal Transient Monitoring," record and tabulate transient cycle events to ensure that the limits specified in UFSAR Tables 3.9-2 and 5.2-9 and TS Table 5.6.1-1 are not exceeded. Corrective action is taken if the cumulative cycles exceed 80 percent of the UFSAR or TS limits.

The NRC staff finds that the reporting requirements in 10 CFR 50. 73, the limits in UFSAR Tables 3.9-2 and 5.9-2, the limits in TS Table 5.6.1-1, and the monitoring provided by the Limerick surveillance procedures noted above, provide reasonable assurance that RCIC system actuations will be appropriately tracked consistent with assuring that the integrity of the components in the RCPB will be maintained within design limits. Therefore, the NRC staff concludes that deletion of LCO 3.73, Action b, is acceptable.

3.4 SR 4.7.5.3-Sealed Source Contamination Limerick SR 4.7.5.3 currently states the following:

Reports -A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable contamination.

As discussed in the application dated April 9, 2013, the licensee proposed to delete the above TS reporting requirement.

As discussed above in SE Section 2.0, the requirements for SRs is provided in 10 CFR 50.36(c)(3). SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met. The LCO associated with SR 4.7.5.3, LCO 3.7.5, requires that:

Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcurie of removable contamination.

The Action statement for LCO 3.7.5 requires, in part, that:

a. With a sealed source having removable contamination in excess of the above limit, withdraw the sealed source from use and either:

1. Decontaminate and repair the sealed source, or

2. Dispose of the sealed source in accordance with Commission Regulations.

SRs 4.7.5.1 and 4.7.5.2 provide the test requirements and test frequencies, respectively to ensure the LCO requirements are met. With respect to test requirements, SR 4. 7.5.1 states that each sealed source shall be tested for leakage and/or contamination and that the test method has a sensitivity of at least 0.005 microcuries per test sample. With respect to test frequencies, SR 4.7.5.2, provides specific test frequencies for sources in use, stored sources not in use, and startup sources and fission detectors.

The NRC staff finds that SRs 4.7.5.1 and 4.7.5.2 are sufficient to provide reasonable assurance that LCO 3.7.5 will be met. The NRC further finds that the reporting requirements in SR 4.7.5.3 are not needed to meet the requirements in 10 CFR 50.36(c)(3). Therefore, the NRC staff concludes that it is acceptable to delete SR 4.7.5.3.

3.5 TS6.9.1.1, 6.9.1.2, and6.9.1.3-StartupReport Limerick TSs 6.9.1.1, 6.9.1.2, and 6.9.1.3 currently state the following:

6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.

6. 9.1. 2 The startup report shall address each of the tests identified in Subsection 14.2.12 of the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

6. 9.1. 3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the startup report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.

As discussed in the application dated April 9, 2013, the licensee proposed to delete the above three TS reporting requirements pertaining to the Startup Report.

As discussed above in SE Section 2.0, with regard to the content of the administrative controls section of the TSs, 10 CFR 50.36(c)(5) states, that "administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner." In general, TS Section 6.0, "Administrative Controls," contains those requirements not covered by other TS sections, but which are necessary to assure operation of the facility in a safe manner.

Limerick UFSAR Section 1.8 states that the Limerick TS reporting requirements are based on NUREG-0123, "Standard Technical Specifications for General Electric Boiling Reactors,"

Revision 2, dated August 1979. NUREG-0123, Revision 2, TSs 6.9.1.1, 6.9.1.2, and 6.9.1.3, contain startup report requirements nearly identical to Limerick TSs 6. 9.1.1, 6. 9.1.2, and 6. 9.1. 3.

In August 1975, the NRC published for comment, Revision 4 of Regulatory Guide (RG) 1.16, "Reporting of Operating Information-Appendix A, Technical Specifications." As discussed in the "Regulatory Position" section of the RG, in addition to the applicable reporting requirements in 10 CFR, the information discussed in the RG was intended to provide an acceptable basis to the NRC staff for meeting the reporting requirements of the TSs. Regulatory Position C.1.a provided reporting requirements associated with the Startup Report. The Startup Report information in the RG is nearly identical to that in to Limerick TSs 6.9.1.1, 6.9.1.2, and 6.9.1.3.

In a Federal Register notice dated August 11, 2009 (74 FR 40244), the NRC withdrew RG 1.16 because it was no longer needed. Specifically, the notice stated that reporting requirements are contained in 10 CFR Part 50, as well as other parts of 10 CFR Chapter I. In addition, the Federal Register notice stated that guidance on the content and frequency of required reports is provided in Chapter 5, "Administrative Controls," of the Standard Technical Specifications (STS) contained in NUREG-1430, NUREG-1431, NUREG-1432, NUREG-1433, and NUREG-1434.

The current STS for General Electric BWR/4 plants (such as Limerick) is NUREG-1433, Revision 4, dated April2012. Section 5.6 of NUREG-1433 contains the necessary reporting requirements. This section does not contain any requirements for a Startup Report.

Limerick TS 6.9.1.3 provides the time frames for submittal of the reports required by TSs 6.9.1.1 and 6.9.1.2 (e.g., 90 days following completion of the startup test program, 90 days following resumption or commencement of commercial power operation, or 9 months following initial criticality). Given these timeframes, report completion and submittal are clearly not necessary to assure safe operation of the facility for the timeframe between completion of the testing and submittal of the report. In addition, there is no requirement for NRC approval of the information provided in the report after it is submitted. In essence, these reports merely provide the NRC a mechanism to review the appropriateness of the licensee activities after-the-fact.

The NRC staff has determined that the proposed elimination of the Startup Report requirements would not eliminate the need to perform the necessary testing. Appropriate testing and documentation of startup testing will continue to be performed in accordance with the requirements of 10 CFR Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," Criterion XI, "Test Control." Specifically, Criterion XI requires that testing be performed to demonstrate that systems, structures and components perform satisfactorily in service and that test results be documented. In addition, as discussed in Limerick UFSAR Section 14.2.6, "Test Records," test records are retained for the life of the plant. Based on the testing, documentation and retention requirements stated above, the NRC staff may review the test results on site as needed.

On the basis that the Startup Report requirements in TSs 6.9.1.1, 6.9.1.2, and 6.9.1.3, are not necessary to assure operation of the facility in a safe manner, the NRC staff finds that deletion of these TS requirements is consistent with 10 CFR 50.36(c)(5). Therefore, the NRC staff concludes that deletion of TSs 6.9.1.1, 6.9.1.2, and 6.9.1.3 is acceptable.

3.6 TS 6.9.1.5.c-Unique Reports Limerick TS 6.9.1.5.c currently states that:

Any other unit unique reports required on an annual basis.

As discussed in the application dated April 9, 2013, the licensee proposed to delete the above TS reporting requirement.

Limerick TS 6.9.1.5.c is based on the NUREG-0123, Revision 2, TS 6.9.1.5.b which states

"(Any other unit unique reports required on an annual basis.)" As discussed in Regulatory Position C.3, "Unique Reporting Requirements," in RG 1.16, Revision 4, the intent of the unique reports TS is for the licensee to submit reports necessary due to any unique plant design features or other factors.

As discussed above in SE Section 3.5, in a Federal Register notice dated August 11, 2009 (74 FR 40244}, the NRC withdrew RG 1.16 because it was no longer needed. In addition, the Federal Register notice stated that guidance on the content and frequency of required reports is provided in Chapter 5, "Administrative Controls," of the Standard Technical Specifications (STS) contained in NUREG-1430, NUREG-1431, NUREG-1432, NUREG-1433, and NUREG-1434.

The current STS for General Electric BWR/4 plants (such as Limerick) is NUREG-1433, Revision 4, dated April2012. Section 5.6 of NUREG-1433 contains the necessary reporting requirements. This section does not contain any requirements for Unique Reports.

As discussed above in SE Section 2.0, with regard to the content of the administrative controls section of the TSs, 10 CFR 50.36(c)(5) states that "administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner." TS 6.9.1.5.c currently contains no specific reports that need to be submitted. Furthermore, for Limerick, reports that are required to be submitted on an annual basis are already contained in other portions of TS 6.9.1 (e.g., TS 6.9.1.7, "Annual Radiological Environmental Operating Report," and TS 6.9.1.8, "Annual Radioactive Effluent Release Report") or in 10 CFR Part 50 (e.g.,

10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors"). On the basis that the Unique Reports requirements in TS 6.9.1.5.c is not necessary to assure operation of the facility in a safe manner, the NRC staff finds that deletion of this TS is consistent with 10 CFR 50.36(c)(5). Therefore, the NRC staff concludes that deletion of TS 6.9.1.5.c is acceptable.

3.7 TS 6.9.1.5.d-Specific Activity Analysis Limerick TS 6.9.1.5.d currently requires that an annual report be submitted which provides the results of specific activity analysis in which the primary coolant exceeded the limit of TS 3.4.5.

The requirements in TS 6.9.1.5.d reads as follows:

The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5. The following information shall be included:

( 1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Cleanup system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) the time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

The above TS reporting requirement is based on the model TSs provided in Generic Letter (GL) 85-19, "Reporting Requirements on Primary Coolant Iodine Spikes," dated September 27, 1985 (ADAMS Accession No. ML031150725). The current STS for General Electric BWR/4 plants (such as Limerick) is NUREG-1433, Revision 4, dated April 2012. Section 5.6 of NUREG-1433 contains the necessary reporting requirements. This section does not contain any requirements related to specific activity analysis.

The licensee's application stated that specific activity analysis pertaining to primary coolant limits is reported to the NRC by means of the Performance Indicator (PI) Program, under the Reactor Oversight Process (ROP). As part of the ROP PI Program, Limerick currently provides monthly reactor coolant specific activity data on a quarterly basis to the NRC in accordance with Regulatory Issue Summary (RIS) 2000-08, Revision 1, "Voluntary Submission of Performance Indicator Data" (ADAMS Accession No. ML083290153), following the guidelines provided in Nuclear Energy Institute (NEI) 99-02, Revision 6, "Regulatory Assessment Performance Indicator Guideline," dated October 2009 (ADAMS Accession No. ML12167A098). The reactor coolant specific activity concentration is provided more frequently than that required by the current TS, regardless of whether or not the TS limit is exceeded.

As discussed in Section 2.3 of NEI 99-02, the purpose of the PI related to reactor coolant system specific activity is to monitor the integrity of the fuel cladding (one of the barriers to prevent release of fission products). In accordance with 10 CFR 50.72, licensees are required to provide notification to the NRC within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months regarding "(t]he condition of the nuclear power plant, including its principal safety barriers, being seriously degraded." The same condition needs to be reported to the NRC within 60 days via an LER in accordance with 10 CFR 50.73.

In addition to the above reporting requirements, TS 3/4.4.5, "Specific Activity," puts limitations on the specific activity of the reactor coolant. These limitations ensure that in the event of a releases of any radioactive material to the environment during a design-basis accident, radiation doses are maintained within the limits of 10 CFR Part 100. LCO 3.4. 5 requires that the specific activity of the primary coolant be limited to less than or equal to 0.2 microcurie per gram dose equivalent iodine 131. In the event that the specific activity of the primary coolant exceeds the limits of LCO 3.4.5, the LCO Actions require, in part, that the licensee perform sampling and analysis until the specific activity is restored to within its limit.

Based on the reporting requirements in 10 CFR 50.72 and 50.73, the specific activity reporting provided under the ROP PI Program, and the requirements in TS 3/4.4.5, the NRC staff finds that the reporting requirements in TS 6.9.1.5.d are not necessary to assure operation of the facility in a safe manner. Therefore, the NRC staff finds that deletion of this TS is consistent with 10 CFR 50.36(c)(5). On this basis, the NRC staff concludes that deletion of TS 6.9.1.5.d is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments relate to changes in recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(1 0). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: J. Lamb R. Ennis D~e: March 26, 2014

ML13214A092 OFFICE LPL 1-2/PM LPL 1-2/LAit LPL 1-2/LA STSB/BC NAME REnnis KBeckford ABaxter REIIiott DATE 3/24/14 2/25/14 3/5/14 3/13/14

v t e Letter Sequence Approval
TAC:MF1399, (Approved, Closed) TAC:MF1400, (Approved, Closed)
Initiation Acceptance... Administration Questions 29 January 2014 Answers Results Approval Other:
MONTHYEAR2014-01-29 [Table View]

ML13214A092

Text

SUBJECT:

Dear Mr. Pacilio:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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