ML052550369

License Amendment, Relocate the Operability and Surveillance Requirements for the Safety/Relief Valve Position Instrumentation

ML052550369
Person / Time
Site:	Limerick
Issue date:	09/27/2005
From:	Travis Tate NRC/NRR/DLPM/LPD1
To:	Crane C Exelon Generation Co, Exelon Nuclear
Tate T, NRR/DLPM, 415-8474
References
TAC MC3454, TAC MC3455
Download: ML052550369 (16)

Text

September 27, 2005 Mr. Christopher M. Crane President and Chief Nuclear Officer Exelon Nuclear Exelon Generation Company, LLC 200 Exelon Way, KSA 3-E Kennett Square, PA 19348

SUBJECT:

LIMERICK GENERATING STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: RELOCATION OF OPERABILITY AND SURVEILLANCE REQUIREMENTS FOR THE SAFETY/RELIEF VALVE POSITION INSTRUMENTATION (TAC NOS. MC3454 AND MC3455)

Dear Mr. Crane:

The Commission has issued the enclosed Amendment No. 179 to Facility Operating License No. NPF-39 and Amendment No. 141 to Facility Operating License No. NPF-85 for the Limerick Generating Station, Units 1 and 2 (LGS). These amendments consist of changes to the technical specifications (TSs) in response to your application dated June 1, 2004.

These amendments relocate the operability and surveillance requirements (SRs) for the reactor coolant system safety/relief valve position instrumentation from the TSs to the LGS Technical Requirements Manual (TRM). Specifically, the amendments relocate LGS TS 3.4.2.c; SR 4.4.2.1; Table 3.3.7.5-1, Item 10; and Table 4.3.7.5-1, Item 10, from the TSs to the TRM.

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

Sincerely,

/RA/

Travis L. Tate, Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-352 and 50-353

Enclosures:

1. Amendment No. 179 to License No. NPF-39

2. Amendment No. 141 to License No. NPF-85

3. Safety Evaluation cc w/encls: See next page

Limerick Generating Station, Unit Nos. 1 and 2 cc:

Site Vice President Associate General Counsel Limerick Generating Station Exelon Generation Company, LLC Exelon Generation Company, LLC 4300 Winfield Road P.O. Box 2300 Warrenville, IL 60555 Sanatoga, PA 19464 Correspondence Control Desk Plant Manager Exelon Generation Company, LLC Limerick Generating Station P.O. Box 160 Exelon Generation Company, LLC Kennett Square, PA 19348 P.O. Box 2300 Sanatoga, PA 19464 Regional Administrator, Region I U.S. Nuclear Regulatory Commission Regulatory Assurance Manager - Limerick 475 Allendale Road Exelon Generation Company, LLC King of Prussia, PA 19406 P.O. Box 2300 Sanatoga, PA 19464 Senior Resident Inspector U.S. Nuclear Regulatory Commission Senior Vice President - Nuclear Services Limerick Generating Station Exelon Generation Company, LLC P.O. Box 596 4300 Winfield Road Pottstown, PA 19464 Warrenville, IL 60555 Library Vice President - Operations, Mid-Atlantic U.S. Nuclear Regulatory Commission Exelon Generation Company, LLC Region I 200 Exelon Way, KSA 3-N 475 Allendale Road Kennett Square, PA 19348 King of Prussia, PA 19406 Vice President Director, Bureau of Radiation Protection Licensing and Regulatory Affairs Pennsylvania Dept. of Environmental Exelon Generation Company, LLC Protection 4300 Winfield Road Rachel Carson State Office Building Warrenville, IL 60555 P.O. Box 8469 Harrisburg, PA 17105-8469 Director Licensing and Regulatory Affairs Chairman Exelon Generation Company, LLC Board of Supervisors of Limerick Township 200 Exelon Way, KSA 3-E 646 West Ridge Pike Kennett Square, PA 19348 Linfield, PA 19468 Manager Licensing Dr. Judith Johnsrud Limerick Generating Station National Energy Committee Exelon Generation Company, LLC Sierra Club 200 Exelon Way, KSA 3-E 433 Orlando Avenue Kennett Square, PA 19348 State College, PA 16803

Mr. Christopher M. Crane President and Chief Nuclear Officer Exelon Nuclear Exelon Generation Company, LLC 200 Exelon Way, KSA 3-E Kennett Square, PA 19348

SUBJECT:

LIMERICK GENERATING STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: RELOCATION OF OPERABILITY AND SURVEILLANCE REQUIREMENTS FOR THE SAFETY/RELIEF VALVE POSITION INSTRUMENTATION (TAC NOS. MC3454 AND MC3455)

Dear Mr. Crane:

The Commission has issued the enclosed Amendment No. 179 to Facility Operating License No. NPF-39 and Amendment No. 141 to Facility Operating License No. NPF-85 for the Limerick Generating Station, Units 1 and 2 (LGS). These amendments consist of changes to the technical specifications (TSs) in response to your application dated June 1, 2004.

These amendments relocate the operability and surveillance requirements (SRs) for the reactor coolant system safety/relief valve position instrumentation from the TSs to the LGS Technical Requirements Manual (TRM). Specifically, the amendments relocate LGS TS 3.4.2.c; SR 4.4.2.1; Table 3.3.7.5-1, Item 10; and Table 4.3.7.5-1, Item 10, from the TSs to the TRM.

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

Sincerely,

/RA/

Travis L. Tate, Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-352 and 50-353

Enclosures:

1. Amendment No. 179 to License No. NPF-39

2. Amendment No. 141 to License No. NPF-85

3. Safety Evaluation cc w/encls: See next page DISTRIBUTION:

PUBLIC PDI-2 R/F DRoberts TTate MOBrien MShanbaky, RI OGC ACRS GHill (4) TBoyce CSchulten DLPM DPR Accession Number: ML052550369 *SE input date. No major changes made.

OFFICE PDI-2/PM PDI-2/LA IROB/SC* OGC PDI-2/SC NAME TTate MOBrien TBoyce JBonanno DRoberts DATE 9/21/05 9/21/05 6/9/05 9/21/05 9/27/05 Official Record Copy

EXELON GENERATION COMPANY, LLC DOCKET NO. 50-352 LIMERICK GENERATING STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 179 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 June 1, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions 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:

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 179, are hereby incorporated in the 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

/RA/

Darrell J. Roberts, Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 27, 2005

ATTACHMENT TO LICENSE AMENDMENT NO. 179 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 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 Insert 3/4 3-85 3/4 3-85 3/4 3-87 3/4 3-87 3/4 4-7 3/4 4-7

EXELON GENERATION COMPANY, LLC DOCKET NO. 50-353 LIMERICK GENERATING STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 141 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 June 1, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions 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:

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 141, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Darrell J. Roberts, Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 27, 2005

ATTACHMENT TO LICENSE AMENDMENT NO. 141 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 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 Insert 3/4 3-85 3/4 3-85 3/4 3-87 3/4 3-87 3/4 4-7 3/4 4-7

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 179 AND 141 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 June 1, 2004 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML041610374), Exelon Generation Company, LLC (the licensee) requested changes to the technical specifications (TSs) for Limerick Generating Station, Units 1 and 2 (LGS). The proposed changes would revise the LGS TSs operability and surveillance requirements (SRs) for the reactor coolant system (RCS) safety/relief valve (SRV) position instrumentation. Specifically the proposed changes would revise:

• TS Table 3.3.7.5-1, Accident Monitoring Instrumentation

• TS Table 4.3.7.5-1, Accident Monitoring Instrumentation Surveillance Requirements

• TS 3.4.2, Reactor Coolant System, Safety/Relief Valve, Action c

• TS 4.4.2, Reactor Coolant System, Safety/Relief Valve, SR 4.4.2.1.

TS Table 3.3.7.5-1, Item 10; TS Table 4.3.7.5-1, Item 10; TS 3.4.2.c; and SR 4.4.2.1 would be deleted from the LGS TSs and relocated to the LGS Technical Requirements Manual (TRM).

Additionally, the associated footnotes to SR 4.4.2.1 would also be deleted and relocated to the TRM.

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36 establishes the regulatory requirements for TS content. The requirements emphasize preventing accidents and mitigating accident consequences. Applicants are expected to incorporate into their TSs those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity (see Statement of Consideration, Technical Specifications for Facility Licenses; Safety Analysis Reports, of December 17, 1968 (33 FR 18610)). Section 50.36 of 10 CFR requires that TSs include items in the following specific categories: (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation (LCOs);

(3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. However, the rule does not specify the format and content for TS categories.

IMPROVED STANDARD TECHNICAL SPECIFICATIONS PROGRAM Interim Policy Statement on TS Improvements The Nuclear Regulatory Commission (NRC or Commission) and industry representatives sought to develop guidelines for improving nuclear power plant TS content and quality. On February 6, 1987, the Commission issued an Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors (52 FR 3788). In September 1992, the Commission issued NUREG-1433, Standard Technical Specifications, General Electric Plants, [Boiling Water Reactor] BWR/4 (STS), which were developed using the guidance and criteria contained in the Commissions Interim Policy Statement. The STSs are a model for developing improved TSs (ITSs) for General Electric plants. The Interim Policy Statement criteria ensure that ITS would consistently reflect system configurations and operating characteristics for the BWR/4 design. In addition, the generic Bases statements provide the basis for each of the STS requirements.

Final Policy Statement on TS Improvements On July 22, 1993, the Commission issued its Final Policy Statement indicating that satisfying the guidance in the policy statement also satisfies Section 182a of the Atomic Energy Act and 10 CFR 50.36 (58 FR 39132). The Final Policy Statement described the improved STS safety benefits and encouraged licensees to use the improved STS as the basis for plant-specific TSs amendments and for complete conversions to the improved STS. Further, the Final Policy Statement gave guidance for evaluating the required scope of the ITS and defined the guidance criteria for determining which of the LCOs and associated surveillances should remain in the ITS. Using this approach, licensees should keep existing LCO requirements that fall within or satisfy any of the Final Policy Statement criteria in the TSs. Those LCO requirements that do not fall within or satisfy these criteria may be relocated to licensee-controlled documents.

The Commission codified the four criteria in 10 CFR 50.36 (60 FR 36953, July 19, 1995).

Final Policy Statement Criteria The Final Policy Statement criteria are as follows:

Criterion 1 Installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary (RCPB).

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to fission product barrier integrity.

Criterion 3 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to fission product barrier integrity.

Criterion 4 A structure, system, or component which operating experience or

probabilistic safety assessment has shown to be significant to public health and safety.

3.0 TECHNICAL EVALUATION

In Section 3.0 of the application, the licensee stated that the LGS design licensing basis for overpressure protection for the RCPB is provided by 14 dual-function SRVs. Further, the licensee stated that each SRV is instrumented with acoustic sensors to provide position indication and alarm in the control room. [.....] Two sensors per valve are mounted on piping downstream of the SRVs. When an SRV is actuated, the sensor produces a signal proportional to the flow noise through the valve. The two sensors and their connecting cables constitute active and passive channels. The signals from both channels are routed to a pre-amplifier outside containment where the active signal is amplified and routed to the position indication cabinet in the control room. The passive channel signal is available for use at the pre-amplifier if a component of the active channel should fail.

The licensee stated in addition to this instrumentation specified in the TS, an SRV tailpipe temperature monitoring system provides a backup for the dual channel acoustic position instrumentation. Furthermore, a redundant safety-grade suppression pool temperature monitoring system provides another diverse means of SRV position indication. The SRV position indication instrumentation addressed in TSs 3.4.2, 4.4.2.1, 3.3.7.5, and 4.3.7.5 performs only alarm and open/close position indication functions, and does not provide input to a control or trip function.

ACCIDENT MONITORING INSTRUMENTATION SECTION 3.1 Evaluation for Relocating SRV Position Indication TSs The licensee proposes to relocate the Accident Monitoring Instrumentation TSs described below.

TS 3.3.7.5, Table 3.3.7.5-1, Function 10, Safety/Relief Valve Position Indicators This table specifies operability requirements for SRV position indication instrumentation. With less than the required minimum channels operable, the associated action statement (Action 80) requires the inoperable channel to be restored to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The required number of operable channels is one channel per valve (this is also the required minimum number of operable channels). The channels are required to be operable in Modes 1 and 2.

TS 3.3.7.5, Table 4.3.7.5-1, Function 10, Safety/Relief Valve Position Indicators This table specifies the SRs for SRV position indication instrumentation. A monthly channel check and refueling interval channel calibration are required.

Accident monitoring instrumentation is provided to monitor variables and systems over their anticipated ranges for accident conditions, as appropriate, to ensure adequate safety during and following accidents. These variables are used by the control room operating personnel to perform their role in the emergency plan in the evaluation, assessment, and monitoring of events, and execution of control room functions. American National Standards Institute/American Nuclear Society N4.5-1980, Criteria for Accident Monitoring Functions in

Light-Water-Cooled Reactors, delineates criteria for determining the variables to be monitored in the control room by the operator. Instrumentation designated as Types A through D aid the designer in selecting the accident monitoring instrumentation and applicable criteria.

Categories 1, 2 and 3 separate the Type criteria into groups for a graded approach to requirements depending on the importance to safety or the measurement of a specific variable.

In development of BWR/4 STSs, the NRC staff documented deterministic screening criteria for post-accident monitoring instrumentation in a letter dated May 7, 1988, from T.E. Murley (NRC) to R. F. Janecek (BWR Owners Group). In accordance with 10 CFR 50.36, the NRC staff requires all plant-specific Regulatory Guide (RG) 1.97, Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Type A instruments specified in the plant's Safety Evaluation Report (SER) on RG 1.97, and all RG 1.97 Category 1 instruments to be included in TSs. Type A instruments monitor primary information required to permit the control room operator to take specific manually-controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design-basis accident (DBA) events.

Category 1 instruments are designed for full qualification, redundancy, continuous realtime display, and onsite (standby) power. Accordingly, the staff position to include all plant-specific RG 1.97 Type A instruments specified in the plant's SER on RG 1.97, and all RG 1.97 Category 1 instruments to be included in TSs can be applied to LGS.

In Section 4.0 of the application, the licensee stated that the plant RG 1.97 SER designates primary system SRV position instrumentation as Type D, Category 2, instrumentation. Type D instruments provide information to indicate the operation of individual safety systems and other systems important to safety. Category 2 instruments are designed to less stringent qualifications that do not require seismic qualification, redundancy, or continuous display, and require only a high reliability power source (not necessarily standby power).

The staff reviewed the licensees technical analysis of the proposed changes to the accident monitoring instrumentation section of the TSs against the requirements in 10 CFR 50.36.

Based on the discussion above, the staff determined that the SRV position indication accident monitoring instrumentation does not meet the Type A or Category 1 instrumentation designations required to be maintained in the TSs in accordance with 10 CFR 50.36. In addition, the staff also determined that the proposed changes are consistent with the requirements in the STSs. Therefore, the staff finds that relocating the proposed SRV position indication accident monitoring instrumentation requirements from the LGS TSs to a licensee-controlled document is acceptable.

SRV SECTION 3.2 Evaluation for Relocating SRV Acoustic Monitors TSs The licensee proposes to relocate SRV acoustic monitors TSs described below.

TS 3.4.2.c This section specifies requirements for inoperable SRV acoustic monitors. With one or more safety/relief valve acoustic monitors inoperable, restore the inoperable acoustic monitors to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SR 4.4.2.1 This section specifies SRs for SRV acoustic monitors.

The acoustic monitor for each safety/relief valve shall be demonstrated operable with the setpoint verified to be 0.20 of the full open noise level## by performance of a

a. CHANNEL FUNCTIONAL TEST at least once per 92 days, and a

b. CHANNEL CALIBRATION at least once per 24 months**.

The associated footnotes to be deleted read: ** The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test, and ## Initial setting shall be in accordance with the manufacturers recommendation. Adjustment to the valve full open noise level shall be accomplished during the startup test program.

In the application, the licensee analyzed the safety basis for SRVs using the Final Policy Statement criterion of 10 CFR 50.36. The SRVs are a part of the primary success path in the Updated Final Safety Analysis Report (UFSAR) accident analysis in that they can actuate to mitigate a DBA and, therefore, meet Criterion 3. Accordingly, their operability is required by TS 3.4.2, Safety/Relief Valves. However, the SRV acoustic monitors do not detect or indicate a significant abnormal degradation of the RCPB considered by Criterion 1. SRV acoustic monitors are not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2. While the function of SRVs is part of the primary success path and the SRVs actuate to mitigate a DBA or transient, SRV acoustic monitors do not form a part of the primary success path since UFSAR accident analysis assumes the SRVs function as designed (i.e., the UFSAR analysis assumes no operator action based on SRV valve position for the SRVs to perform their primary success path function considered in Criterion 3). Furthermore, the licensee stated that the loss of this instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to public health and safety as considered in Criterion 4. Finally, failure of SRV acoustic monitors would not pose a significant challenge to the ability of the operating staff to respond to a DBA or transient, since the Emergency Operating Procedures (EOP) provide symptom-based instruction to the staff in mitigating an upset condition of the plant (i.e., level, pressure, and temperature provide EOP direction regardless of SRV status).

The staffs evaluation of the licensing design basis finds that relocating the SRV acoustic monitor TSs is consistent with the Commission's Final Policy Statement criteria in that Criterion 1 was intended to assure that TSs control those instruments specifically installed to detect RCS leakage which present a threat of significant compromise to the RCPB. Leakage detection systems inside the drywell are designed with the capability of detecting leakage less than the established leakage rate limits and providing appropriate alarm of excess leakage in the control room. These actions provide adequate response before a significant break in the RCPB can occur. Thus, identifying the source of RCS leakage does not include reliance on the

SRV acoustic monitors since this instrumentation is not used to identify leaks that present a threat of significant compromise to the RCPB.

The staff reviewed the licensees technical analysis of the proposed changes to the SRV acoustic monitor TS LCOs and SRs, including the associated footnotes, against the requirements in 10 CFR 50.36. Based on the discussion above, the staff determined that the SRV acoustic monitors do not satisfy any of the criteria in 10 CFR 50.36 for items required to be maintained in the TSs. In addition, the staff also determined that the proposed changes are consistent with the requirements in the STSs. Therefore, the staff finds that relocating the proposed SRV acoustic monitor instrumentation TS requirements and associated footnotes from the LGS TSs to a licensee-controlled document is acceptable.

3.3 Conclusion The proposed changes relocate operability and SRs for the RCS SRV position instrumentation from: (1) the Accident Monitoring Instrumentation section, TS 3.3.7.5 and SR 4.3.7.5, and (2) the SRV section, TS 3.4.2 and SR 4.4.2.1, including the associated footnotes, to the LGS TRM which is incorporated by reference into the LGS UFSAR. Based on the discussions above in Sections 3.1 and 3.2, the staff concludes that the proposed changes are acceptable.

Accordingly, future changes to the SRV position instrumentation and acoustic monitor operability and SRs will be performed under regulatory control pursuant to the requirements of 10 CFR 50.59.

4.0 STATE CONSULTATION

In accordance with the Commissions 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 change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (69 FR 62475). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). 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) such activities will be conducted in compliance with the Commissions 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 Contributor: C. Schulten Date: September 27, 2005