ML14324A808

Issuance of Amendments - Leak Detection System Setpoint and Allowable Value Changes

ML14324A808
Person / Time
Site:	Limerick
Issue date:	12/29/2014
From:	Richard Ennis Plant Licensing Branch 1
To:	Pacilio M Exelon Nuclear
Ennis R
References
TAC MF3198, TAC MF3199
Download: ML14324A808 (25)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Decmeber 29, 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: LEAK DETECTION SYSTEM SETPOINT AND ALLOWABLE VALUE CHANGES (TAC NOS. MF3198 AND MF3199)

Dear Mr. Pacilio:

The Commission has issued the enclosed Amendment Nos. 213 and 174 to Renewed 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 December 6, 2013, as supplemented by letter dated September 19, 2014.

The amendments revise TS setpoints and allowable values for certain area temperature instrumentation associated with the leak detection system.

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

Sincerely, Richard B. Ennis, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-352 and 50-353

Enclosures:

1. Amendment No. 213 to Renewed NPF-39

2. Amendment No. 174 to Renewed NPF-85

3. Safety Evaluation cc w/encls: Distribution via Listserv

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. 213 Renewed 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 December 6, 2013, as supplemented by letter dated September 19, 2014, 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.

Enclosure 1

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 Renewed 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. 213, are hereby incorporated into this renewed 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 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 Renewed Facility Operating License Date of Issuance: December 29, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 213 RENEWED FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 Replace the following page of the Renewed 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 Insert Page 3 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 Insert 3/4 3-19 3/4 3-19 3/4 3-20 3/4 3-20

(2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and to use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (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 renewed 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 renewed license shall be completed as specified. Attachment 1 is hereby incorporated into this renewed license.

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix 8, as revised through Amendment No. 213, are hereby incorporated into this renewed license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-39 Amendment No. 213

TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

3. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. RWCS ~ Flow - High  :.,:; 54.9 gpm  :.,:; 65.2 gpm

b. RWCS Area Temperature - High  :.,:; 155°F or :.,:; 120°F**  :.,:; 160°F or :.,:; 125°F**

c. RWCS Area Ventilation

~Temperature - High  :.,:; 52°F or :.,:; 32°F**  :.,:; 60°F or :.,:; 40°F**

d. SLCS Initiation N.A. N.A.

e. Reactor Vessel Water Level -

Low, Low, - Level 2 2 -38 inches

• 2 -45 inches

f. Manual Initiation N.A. N.A.

4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION

a. HPCI Steam Line~ Pressure - High  :.,:; 974" H20  :.,:; 984" H20

b. HPCI Steam Supply Pressure - Low 2 100 psig 2 90 psig

c. HPCI Turbine Exhaust Diaphragm Pressure - High  :.,:; 10 psig  :.,:; 20 psig

d. HPCI Equipment Room Temperature - High 180°F 2 177°F, :.,:; 191°F

e. HPCI Equipment Room

~Temperature - High  :.,:; 104°F  :.,:; 108.5°F

f. HPCI Pipe Routing Area Temperature - High 180°F 2 177°F, :.,:; 191°F

g. Manual Initiation N.A. N.A.

h. HPCI Steam Line~ Pressure - Timer 3 :.,:; T:.,:; 12.5 seconds 2.5:.,:; T:.,:; 13 seconds LIMERICK - UNIT 1 3/4 3-19 Amendment No. JJ,&9,+GB,+&+,~.213

TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION

a. RCIC Steam Line 11 Pressure - High ~ 373" H20 ~ 381" H20

b. RCIC Steam Supply Pressure - Low ~ 64.5 psig ~ 56.5 psig

c. RCIC Turbine Exhaust Diaphragm Pressure - High ~ 10.0 psig ~ 20.0 psig

d. RCIC Equipment Room Temperature - High 180°F ~ 161°F, ~ 191°F

e. RCIC Equipment Room 11 Temperature - High ~ 109oF ~ 113. 5°F

f. RCIC Pipe Routing Area Temperature - High 180oF ~ 16l°F, ~ 191°F

g. Manual Initiation N.A. N.A.

h. RCIC Steam Line 11 Pressure Timer 3 ~ t ~ 12.5 seconds 2.5 ~ t ~ 13 seconds LIMERICK - UNIT 1 3/4 3-20 Amendment No. JJ, ~.-+/--%, 213

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. 174 Renewed 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 December 6, 2013, as supplemented by letter dated September 19, 2014, 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.

Enclosure 2

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 Renewed 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. 174, are hereby incorporated into this renewed 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 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 Renewed Facility Operating License Date of Issuance: December 29, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 174 RENEWED FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 Replace the following page of the Renewed 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 Insert Page 3 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 Insert 3/4 3-19 3/4 3-19 3/4 3-20 3/4 3-20

(2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and to use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (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 renewed 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 ( 100 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. 174, are hereby incorporated into this renewed license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-85 Amendment No. 174

TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

3. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. RWCS ~Flow- High .::: 54.9 gpm .::: 65.2 gpm

b. RWCS Area Temperature - High .::: 155°F or~ 120°F** .::: 160°F or ~ 125°F**

c. RWCS Area Ventilation

~Temperature - High .::: 52°F or ~ 32°F** .::: 60°F or ~ 40°F**

d. SLCS Initiation N.A. N.A.

e. Reactor Vessel Water Level -

Low, Low, - Level 2 2: -38 inches

• 2: -45 inches

f. Manual Ini ti ati on N.A. N. A.

4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION

a. HPCI Steam Line~ Pressure - High _:: 974" H20 _:: 984" H20

b. HPCI Steam Supply Pressure - Low 2: 100 psig 2: 90 psig

c. HPCI Turbine Exhaust Diaphragm Pressure - High .::: 10 psig .::: 20 psig

d. HPCI Equipment Room Temperature - High 180°F 2: 177°F, ~ 191°F

e. HPCI Equipment Room

~Temperature - High _:: 104°F .::: 108.5°F f 0 HPCI Pipe Routing Area Temperature - High 180oF 2: 177°F, _:: 191oF g 0 Manual Initiation N.A. N.A.

h. HPCI Steam Line ~ Pressure - Timer 3.:::'.::: 12.5 seconds 2.5.::: '.::: 13 seconds LIMERICK - UNIT 2 3/4 3-19 Amendment No. 49,~.~.+&4, 174

TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION

a. RCIC Steam Line L\.

Pressure - High S 373" H20 S 381" H20

b. RCIC Steam Supply Pressure - Low ~ 64.5 psig ~ 56.5 psig

c. RCIC Turbine Exhaust Diaphragm Pressure - High s 10.0 psig s 20.0 psig

d. RCIC Equipment Room Temperature - High 180°F ~ 161°F, S 191oF

e. RCIC Equipment Room L\. Temperature - High S 109oF S 113.5oF

f. RCIC Pipe Routing Area Temperature - High 180oF ~ 161°F, ~ 191oF

g. Manual Initiation N.A. N.A.

h. RCIC Steam Line L\. Pressure Timer 3 s' s 12.5 seconds 2.5 s 's 13 seconds LIMERICK - UNIT 2 3/4 3-20 Amendment No.%, +/--, 174

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 213 AND 174 TO RENEWED 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 December 6, 2013, as supplemented by letter dated September 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML13343A024 and ML14262A127, respectively), Exelon Generation Company, LLC (Exelon, the licensee), requested changes to the technical specifications (TSs) for Limerick Generating Station (LGS), Units 1 and 2. The proposed amendment would revise TS setpoints and allowable values for certain area temperature instrumentation associated with the leak detection system (LOS). The purpose of the LOS is to detect and provide the signals necessary to isolate leakage from the reactor coolant pressure boundary (RCPB) before pre-determined limits are exceeded. The affected TS instrumentation monitor ambient temperature in the reactor water cleanup system (RWCS) area, the high pressure coolant injection (HPCI) equipment room and pipe routing area, and the reactor core isolation cooling (RCIC) equipment room and pipe routing area. The temperature setpoints, for the LOS instrumentation described above, are established to provide system isolations in the event of a postulated 25 gallon per minute (gpm) steam leak.

The proposed amendment would also change the leakage design basis from 25 gpm to 35 gpm for the turbine enclosure main steam line (MSL) tunnel temperature isolation setpoint (the setpoint of this instrumentation is not being changed).

The licensee's amendment request indicated that the proposed changes are being made in order to establish adequate margins such that normal variations in the maximum operating temperatures for the affected plant areas do not result in system isolation.

The supplement dated September 19, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register on April 1, 2014 (79 FR 18333).

Enclosure 3

2.0 REGULATORY EVALUATION

The NRC staff review took into consideration the regulatory requirements and guidance documents discussed below.

The NRC's regulatory requirements related to the content of the TSs are set forth in Title 10 of the Code of Federal Regulations (1 0 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 of 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 LCO 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.

Appendix A to 10 CFR Part 50, General Design Criterion (GDC) 13, "Instrumentation and control," requires, in part, that instrumentation be provided to monitor variables and systems, and that controls are provided to maintain these variables and systems within prescribed operating ranges.

Appendix A to 10 CFR Part 50, GDC 20, "Protection system functions," requires, in part, that the protection system be designed to initiate automatically the operation of appropriate systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences.

Appendix A to 10 CFR Part 50, GDC 30, "Quality of reactor coolant pressure boundary,"

requires, in part, that means for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

As discussed in 10 CFR 50.55a(h), "Protection and safety systems," for nuclear power plants with construction permits issued after January 1, 1971, but before May 13, 1999 (such as LGS, Units 1 and 2), protection systems must meet the requirements stated in either Institute of Electrical and Electronics Engineers (IEEE) Standard 279, "Criteria for Protection Systems for Nuclear Power Generating Stations," or in IEEE Standard 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations," and the correction sheet dated January 30, 1995. Section 6.8, "Setpoints," of IEEE 603-1991 requires that the allowance for uncertainties between the process analytical limit and the device setpoint be determined using a documented methodology. The IEEE standard makes reference to Instrument Society of America (ISA) (now referred to as the International Society of Automation Standard S67.04-1987.

The NRC Regulatory Guide (RG) 1.1 05, Revision 3, "Setpoints for Safety-Related Instrumentation," issued December 1999 (ADAMS Accession No. ML993560062), describes a

method acceptable to the staff of the NRC for complying with the NRC's regulations for ensuring that setpoints for safety-related instrumentation are initially within and remain within the TS limits.

The NRC Standard Review Plan Section 10.3, Main Steam Supply System, Revision 4, dated March 2007 (ADAMS Accession No. ML070380206), provides guidance for NRC staff review of the main steam supply system. This guidance includes reviewing the system with respect to the capability to detect and control system leakage and to isolate portions of the system in case of excessive leakage or component malfunctions.

3.0 TECHNICAL EVALUATION

3.1 Proposed Changes The proposed amendment would revise the TSs as described below in safety evaluation (SE)

Section 3.1.1. The amendment would also change the design bases for LGS, Units 1 and 2, as described in SE Section 3.1.2.

3.1 .1 TS Changes The proposed amendment would revise the setpoints and allowable values for LGS, Units 1 and 2, for the following instrumentation in TS Table 3.3.2-2, "Isolation Actuation Instrumentation Setpoints:"

Trip Description Current Proposed Current Proposed Function (Instrument Loop) TS TS TS TS Setpoint Setpoint Allowable Allowable Value Value RWCS Area 3.b Temperature - High  : :; 132 °F  :::;; 120 °F  : :; 137 °F  : :; 125 °F (Heat Exchanger Rooms)

(TE-044-1 N016A)

HPCI Equipment Room ~ 218 °F, ~ 177 °F, 4.d Temperature - High

= 225 °F = 180 °F  : :; 247 OF  : :; 191 °F (TE-055-1 N030B)

HPCI Pipe Routing Area ~ 165 °F, ~ 177 °F, 4.f Temperature - High

= 175 °F = 180 °F  : :; 200 °F  : :; 191 °F (TE-055-1 N025B)

RCIC Equipment Room ~198°F, ~161 °F, 5.d Temperature - High

= 205 °F = 180 °F  : :; 227 °F  : :; 191 °F (TE-049-1 N023A)

RCIC Pipe Routing Area ~ 165 °F, ~ 161 °F, 5.f Temperature - High

= 175 °F = 180 °F  : :; 200 °F  : :; 191 °F TE-049-1 N025A)

3.1.2 Design Basis Changes The proposed amendment would also change the design basis for LGS, Units 1 and 2, to approve the use of a 35-gpm leak as the basis, during winter operating conditions, for the MSL isolation setpoint associated with the Turbine Enclosure- Main Steam Line Tunnel Temperature

- High trip function {TS Table 3.3.2-2, trip function 1.g). No TS changes are proposed associated with this design basis change.

3.2 Instrumentation Review The NRC staff reviewed the proposed amendment from an instrumentation perspective as described in SE Sections 3.2.1, 3.2.2, and 3.2.3 below.

3.2.1 Methodology As shown in the table in SE Section 3.1.1, the licensee proposes revisions to the TS instrument setpoints and allowable values in Table 3.3.2-2, associated with the LOS.

This license amendment request (LAR) affects the LOS instrument setpoints of the circuits that monitor the ambient temperature for the RWCS Area, HPCI/RCIC equipment rooms, and HPCI/RCIC pipe routing areas. The associated setpoints provide system isolations on high ambient temperature. The LOS does not provide any automatic trip function for protection against a violation of a reactor core Safety Limit (SL) or a reactor coolant system pressure SL, during an anticipated operational occurrence, a normal operational transient, or steady state operation.

The LGS setpoint methodology, which is currently contained in Exelon Procedure CC-MA-1 03-2001, is based on the NRC-approved General Electric (GE) Topical Report NEDC-31336P-A, 1 "General Electric Instrument Setpoint Methodology," dated September 1996. The NRC staff previously found the LGS setpoint methodology acceptable as discussed in an NRC letter dated February 16, 1995, "Revised Maximum Authorized Thermal Power Limit, Limerick Generating Station, Unit No. 2" (ADAMS Accession No. ML011560773).

Since the regulatory basis and technical basis for the GE setpoint methodology (i.e., NEDC-31336P-A) and LGS setpoint methodology was previously evaluated and approved for use by the NRC staff, this SE does not re-examine the acceptability of those documents. The acceptability of the subject LAR was evaluated against the criteria set forth in the approved GE setpoint methodology.

In the application dated December 6, 2013, the licensee explained that LGS Calculation -1001, "Compartment Temperature Transients for Steam and Water Leaks," contains the detailed design analysis that supports setting the appropriate analytical limits. The analytical limits derived from this calculation were then used as inputs to the LGS setpoint methodology to calculate the TS allowable values and trip setpoints. The licensee also indicated that Calculation -1001 was completed using the CONCOIL-FLUD (CFLUD) program ("CONCOIL-FLUD (CFLUD), Version 1.0, 11 Thermofluid Dynamics for a System of Interconnected 1

GE topical report NEDC-31336P-A is a non-public, proprietary version of the GE setpoint methodology (ADAMS Accession No. ML072950103). A public version of the methodology is available as GE topical report NED0-31336-A (ADAMS Accession No. ML073450560).

Compartments"). The CFLUD program was used for the computation performed to support a 2011 license amendment for LGS, which was issued by the NRC on May 11, 2011 (ADAMS Accession No. ML111101429).

The outputs of the CFLUD computations were summarized in Attachment 3 to the application dated December 6, 2013, which also contained graphs showing the temperature response for the Turbine Enclosure-Main Steam Line Tunnel room, RWCS Area, HPCI/RCIC equipment rooms and HPCI/RCIC pipe routing areas room as a function of time following small leaks and 25 gpm steam leak events. The room and piping area temperature analytical limits for isolation setpoints are based on the maximum room temperature resulting from a 25 gpm equivalent leak for summer and winter conditions. The minimum of these two values is the recommended analytical limit for isolation setpoint calculations. For the lower temperature cases (which occur during the winter), the reactor building is assumed to start at 65 degrees Fahrenheit (°F) based on the minimum design temperature for the reactor building. For the higher temperature cases (which occur during the summer), the reactor building is started at 98.5 °F, and the outside atmosphere is held constant at 95 oF and 47 percent(%) relative humidity.

In establishing adequate margins, the recommended analytical limit was compared to the maximum expected temperature for normal balance of plant systems operations or post loss-of-coolant accident (LOCA)/station blackout (SBO) temperature for HPCI and RCIC equipment rooms. The recommended minimum margin between the expected maximum operating temperature and the analytical limit for the isolation setpoint is 20 oF to ensure that normal variations in the maximum operating temperature do not result in spurious system isolation.

The recommended analytical limits identified in Calculation -1001 were used as the design basis input for calculating the subject TS instrument setpoint and allowable value changes.

3.2.2 Evaluation of TS Changes The NRC staff reviewed Attachments 1 and 4 to the application dated December 6, 2013, to determine whether the setpoint basic formulas, bounding conditions, and assumptions are consistent with the LGS methodology for determining new setpoint values. In addition, the staff evaluated the licensee's description of its design inputs to determine whether the information included as input to the determination of total loop uncertainty were reasonable. to the application describes basic formulas and assumptions used in the instrument loop uncertainty calculations. In Attachment 4, the licensee provided applicable excerpts of LGS instrument loop uncertainty calculations for the following channels affected by the proposed revision to the TSs: TE-044-1N016A, TE-055-1N0308, TE-055-1N0258, TE-049-1N023A, and TE-049-1 N025A.

The NRC staff reviewed the licensee's setpoint methodology and associated uncertainty calculation. The LGS setpoint methodology is a statistically based methodology consisting of calculating the error terms, allowable value, and nominal trip setpoint (NTSP), and then establishing the actual setpoint.

The approach used to determine the total channel uncertainty was to combine all terms considered random using the square root of the sum of the squares (SRSS) methodology and then algebraically add any bias terms identified. The loop uncertainty calculation identified the

data needed for uncertainty errors associated with miscellaneous and random bias effects, device accuracy, and maintenance and test equipment. The error terms were normalized to 2 sigma. In addition, the loop uncertainty calculation identified the data associated with device accuracy temperature effect, device humidity effect, device seismic effect, and leave-alone-zone tolerance limits.

The allowable value was calculated using an equation representative of a process variable that increases-to-trip. Based on the GE setpoint methodology adopted in the LGS methodology, the allowable value is equal to the analytical limit minus the SRSS of the errors associated with device performance accuracy, calibration accuracy (test equipment uncertainty), primary element (if applicable) uncertainty and process measurement uncertainty, minus any applicable bias terms.

The nominal trip setpoint was calculated using an equation representative of a process variable that increases-to-trip. Based on the GE setpoint methodology adopted in the LGS methodology, the nominal setpoint value is equal to the analytical limit minus the SRSS of the errors associated with device performance accuracy, calibration accuracy (test equipment uncertainty),

primary element (if applicable) uncertainty, process measurement uncertainty, device drift, minus any applicable bias terms. After calculating the NTSP, the licensee calculates the actual trip setpoint {ATSP), which is the NTSP minus any additional margin assigned by the calculation preparer. Attachment 1 to the application dated December 6, 2013, explains how the additional margin is estimated. The ATSP represents the value proposed by the licensee for the TS setpoint.

The GE setpoint methodology (NEDC-31336P-A) discusses the results of a GE evaluation of field data on performance of Rosemount transmitters and trip units in relation to the design assumptions for drift. The accuracy and drift methodology described in Section 2.0 of the GE setpoint methodology is based on the use of Rosemount (1151, 1152-T0280, 1153 Series B, and 1154) or Gould (3018, 3200, and 3218) transmitters with Rosemount (510) trip units.

The NRC SE that approved the GE setpoint methodology indicated that licensees are to discuss their plant-specific use of instrumentation that differs from the instrumentation described in Section 2.0 of the GE setpoint methodology. This staff SE states, in part, that:

Where instruments are used that are different from those presented in Section 2 of NEDC-31336, the licensee must demonstrate that instrument performance can be quantified either through vendor data or plant specific surveillance test data.

The licensee must confirm that the observed measurements of instrument performance are bounded by the design allowances used in the plant specific analysis, for the chosen calibration interval, in accordance with the criteria stated in NEDC-31336.

The NRC staff's review of the LAR noted that the licensee uses instruments that are different from the instrumentation described in the NRC-approved GE setpoint methodology.

Specifically, the temperature elements and temperature indicating switches in the instrument loops associated with the LAR are of different manufacturers and model numbers than those contained in Section 2.0 of NEDC-31336P-A.

Additionally, the NRC staff noted that the specific line break detection applications that are the subject of the LAR are not covered among the 25 specific instrument setpoint descriptions discussed in Section 3.0 of NEDC-31336P-A.

In a request for additional information (RAI) dated July 29, 2014 (ADAMS Accession No. ML 1421 OA576), the NRC staff requested the licensee to demonstrate that the instrument accuracy and drift values for the temperature elements and temperature indicating switches are bounded by the design allowances used in the plant-specific analysis for the chosen calibration intervals, and are consistent with the criteria stated in NEDC-31336P-A.

In the response to the RAI dated September 19, 2014 (ADAMS Accession No. ML14262A127),

the licensee identified that the temperature elements were manufactured by Pyco, model number 102-9039, and that the temperature switches were Nuclear Measurement Analysis and Control (NUMAC), model number 304A3714 (manufactured by GE). In this response, the licensee also indicated that the surveillance tests (STs) for the NUMAC temperature switches, listed below, are performed every 24 months.

• ST 025-404-1 (Unit 1, Division 1) - (This is the ST identified in the Instrument Loop Uncertainty Calculation Instrument Data table for TE-044-1 N016A, and TE-049-1 N025A).

• ST-2-025-405-1 (Unit 1, Division 2)- (This is the ST identified in the Instrument Loop Uncertainty Calculation Instrument Data table for TE-049-1 N023A, TE-055-1 N025B, and TE-055-1 N030B).

The licensee stated in its RAI response that the instrumentation performance trending, for the past five operating cycles (1 0 years), revealed that there were no indications of unsatisfactory surveillance testing results for the two-year (24 months) calibration interval. Instrument channel surveillance test results verified that the observed channel performance (which, the NRC staff notes includes the effects of accuracy and drift) was within 1 percent and bounded by the value in the plant specific analysis. The staff notes that the temperature elements are safety-related and environmentally qualified. Additionally, it is noted that the qualification testing identified that the component remained within the limit of error of+/- 1 oc, or+/- 0.75 percent per the specification, which is the device accuracy used in the setpoint calculations.

Based on its review of the information provided by the licensee, the NRC staff concludes that the proposed TS changes will serve to maintain acceptable margins. Also, based on its evaluation, the staff concludes that the proposed TS changes are consistent with the approved setpoint methodology.

Based on the above evaluation, the NRC staff concludes that the changes to the TSs are acceptable from an instrumentation perspective.

3.2.2 Evaluation of Design Basis Changes The NRC staff reviewed the licensee's request to change the design basis from a 25 gpm to a 35 gpm equivalent steam leak during winter operations for the turbine enclosure MSL tunnel.

The staff reviewed the application dated December 6, 2013, and the supplement dated September 19, 2014.

During the licensee's preparation of the revision to Calculation -1 001, low margins were noted for the temperature TS isolation setpoint in the turbine enclosure MSL tunnel for a 25 gpm leak.

Currently, with a 25 gpm leak at winter conditions, only a 2.5 oF margin exists. The licensee considered reducing the temperature TS isolation setpoint, in TS Table 3.3.2-2, trip function 1.g, from 165 oF to 155 oF to restore LOS operational margin during winter conditions. In 2013, the licensee noted that several high temperature readings of 153 oF existed during normal operations during the summer, for several of the temperature elements supplying isolation readings. The licensee determined that decreasing the setpoint to 155 oF to restore LOS operational margin during winter conditions would not be prudent, since it would significantly increase the probability of an inadvertent MSL isolation and turbine trip due to normal operational transients, such as loss of turbine enclosure heating, ventilation and air conditioning.

The licensee determined that station operational margin should be maintained by changing the design basis for the isolation temperature setpoint during winter operations, while maintaining the current TS isolation setpoint at 165 °F, rather than decreasing the setpoint to 155°F.

The NRC staff reviewed Attachment 3, to the supplement dated September 19, 2014, "Temperature Response Curve for Turbine Enclosure, Main Steam Tunnel with a 35-gpm Leak in Winter Conditions," which shows the temperature response as a function of time following a 25 gpm steam leak, and a 35 gpm steam leak. At a 25 gpm steam leak, the temperature is 167.3 °F, and at a 35 gpm steam leak, the temperature is 184.9 °F. The Turbine Enclosure MSL Tunnel Temperature - High TS setpoint is not impacted by the design basis change; the TS isolation setpoint remains unchanged.

As described above, the licensee's engineering judgment involved the evaluation of the applicable conditions and pertinent information as related to plant safety. The NRC staff evaluated the licensee's description of its engineering judgment employed, and agrees with the licensee's decision to maintain the current isolation setpoint.

Based on the above evaluation, the NRC staff concludes that the change to the design basis, for the isolation temperature setpoint during winter operations, is acceptable from an instrumentation perspective.

3.2.3 Instrumentation Conclusion The NRC staff has reviewed the licensee's justification for the proposed changes to the setpoints and allowable values for trip functions 3.b, 4.d, 4.f, 5.d, and 5.f, located in TS Table 3.3.2-2, and found reasonable assurance that the proposed TS changes continue to meet the requirements of GDCs 13, 20, and 30 of Appendix A to 10 CFR Part 50, 10 CFR 50.55a(h},

and 10 CFR 50.36. The staff also concludes that the licensee setpoint calculations are consistent with the guidance in RG 1.1 05, Revision 3. The NRC staff further concludes that the design inputs and methodology employed for evaluating total instrument channel uncertainty are appropriate and consistent with the currently approved methodologies for LGS.

3.3 Containment and Ventilation Review The NRC staff reviewed the proposed amendment from a containment and ventilation perspective as described in SE Sections 3.3.1 and 3.3.2 below.

3.3.1 Evaluation of TS Changes The setpoints, for the trip functions listed in the table in SE Section 3.1.1, provide for system isolations in the event of a postulated 25 gpm steam leak to prevent the excessive loss of reactor coolant and the release of significant amounts of radioactive steam and water from the nuclear reactor coolant boundary. The LOS, as described in Section 7.6.1.3.3 of the LGS Updated Final Safety Analysis Report (UFSAR), provides initiation signals to isolation systems, as described in UFSAR Section 7.3.1.1.2.4, but does not provide any automatic trip function for protection against a violation of a reactor core SL, or a reactor coolant system pressure SL, during an anticipated operational occurrence, a normal operational transient, or steady state operation.

The NRC staff requested additional information regarding the 25 gpm steam leak. In the RAI response dated September 19, 2014, the licensee stated that the 25 gpm is the liquid water equivalent flow rate with part or all being flashed in the room as steam. In the same RAI response, the licensee provided the thermal-hydraulic compartment conditions for the leakage that were verified to the NRC staff to represent a high energy line leak.

The LOS system includes temperature indicating switches (TISs) TIS-025-101 A, B, C, 0 and TIS-025-201 A, B, C, D. The TISs receive data from temperature elements located in the areas being monitored and determine if the temperature has reached its alarm or isolation setpoints.

If reached, an automatic isolation/alarm signal is initiated and an annunciator is activated in the control room. The TIS setpoints are selected high enough to avoid spurious isolations, yet low enough to provide timely detection of a small leak and isolation at 25 gpm for the subject areas.

The decision for the licensee to propose the setpoint changes was a result of revising a leak detection Calculation -1001, "Compartment Temperature Transients for Steam and Water Leaks." The licensee found that the current high temperature setpoints (i.e., before this proposed change) in the RWCS area, HPCI equipment room and RCIC equipment room were set too high during winter conditions to detect a 25 gpm steam leak. In other words, the rooms subjected to the low temperature surrounding rooms or outside during winter may not be heated up sufficiently by the 25 gpm steam leak to reach the current detection setpoints. Also, an inconsistency was noted between the HPCI/RCIC pipe routing area isolation setpoints and the HPCIIRCIC equipment room isolation setpoints. The licensee proposed to change the associated setpoints to the same setpoint for both the pipe routing areas and the respective equipment rooms.

The NRC staff evaluated the methodology used in the revised Calculation -1001. The calculation contains the detailed design analysis to provide data to set the appropriate analytical limits for the LOS. The staff finds that the design inputs were reasonably chosen. A representative case for the Main Steam Tunnel (Turbine Building) and Condenser Area was provided for review. The trend of room temperature response as a function of time following small leaks has been reviewed by the staff and found to be reasonable.

Calculation -1001 calculated the maximum room temperature resulting from a 25 gpm equivalent leak for summer and winter conditions. The room and piping area temperature analytical limits for isolation setpoints are then based on the room temperature resulting from a 25 gpm equivalent leak for summer and winter conditions. The minimum of these two values is the recommended analytical limit for isolation setpoint calculations. For the lower temperature cases (winter), the reactor building is assumed to start at 65 oF based on the minimum design temperature for the reactor building. For the higher temperature cases (summer), the reactor building is started at 98.5 oF while the outside atmosphere is held constant at 95 oF and 47%

relative humidity. The maximum room temperatures as obtained from these temperature responses for each room were compared to the maximum expected temperature for normal balance of plant systems operations or post LOCA/SBO temperature for the HPCI and RCIC equipment rooms. The comparison indicates that margin exists so that if the currently existing analytical limits are lowered by the same amount (hereafter, the lowered analytical limits will be referred as the recommended analytical limits), the normal variations in the maximum operating temperature can still avoid any spurious system isolation. The recommended analytical limits were then used as the design basis input for calculating the subject TS instrument setpoint and allowable value changes in accordance with the setpoint calculation methodology described above in SE Section 3.2.1. This methodology was used to determine changes to the allowable value, actual trip setpoint, nominal trip setpoint, and setting tolerance for the subject TS instrument loops.

The NRC staff's review of the proposed TS changes found that the environmental qualification of required equipment in the subject rooms and pipe routing areas is not affected by the proposed changes to the high temperature isolation trip setpoints. This is because the mass and energy release used to determine the proposed setpoint changes (i.e., 25 gpm of water release from relatively high energy line breaks or leakage), is bounded by the mass and energy release used in the accident analysis described in LGS UFSAR Section 15.6.4, "Steam System Piping Break Outside Primary Containment." Furthermore, as shown in the table in SE Section 3.1.1, the proposed TS setpoint changes for three of the trip functions (3.b, 4.d, and 5.d) will lead to an earlier LOS detection and isolation (i.e., proposed setpoints are lower than current setpoints). Although the setpoint for two of the trip functions (4.f and 5.f) will both increase by 5 oF (i.e., 175 oF to 180 °F), the time to initiate isolation is estimated by the NRC staff to be approximately 3 minutes {based on review of graph 7.11 in Attachment 3 of the application dated December 6, 2013). This is considered acceptable in terms of the extra 3 minutes leakage of mass and energy, as compared to the bounding mass and energy release from a main steam line break (MSLB).

Based on the above evaluation, the NRC staff concludes that the proposed TS changes are bounded by the existing design basis with respect to mass and energy release.

3.3.2 Evaluation of Design Basis Changes As discussed above in SE Sections 3.1.2 and 3.2.2, the licensee has proposed to change the design basis from a 25 gpm to a 35 gpm equivalent steam leak, during winter operations, for the Turbine Enclosure MSL Tunnel Temperature - High trip function {TS Table 3.3.2-2, trip function 1.g). No TS changes are proposed associated with this design basis change.

As discussed on page 6 of Attachment 1 to the licensee's application dated December 6, 2013, the licensee has calculated that a 35 gpm leak represents a loss of reactor coolant of

4.66 pounds mass (Ibm) per second. The licensee further stated that any leak would be isolated within 15 minutes by the LOS. As such, the total integrated coolant loss for a 35 gpm leak would be: 4.66 Ibm/second x 15 minutes x 60 seconds= 4194 Ibm. As discussed in Section 15.6.4.4 of the LGS UFSAR, the total integrated coolant loss due to a design basis MSLB is 108,785 Ibm. As such, the leakage due to a 35 gpm leak is bounded by the MSLB analysis. Therefore, the NRC staff concludes that the change in design basis for the LOS is acceptable from a containment perspective.

3.4 Technical Evaluation Conclusion Based on the evaluation in SE Sections 3.2 and 3.3, the NRC staff concludes that the proposed amendment 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 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 (79 FR 18333). 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) 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: S. Peng D. Spaulding-Yeoman R. Ennis Date: December 29, 2014

• ML14324A808 OFFICE LPL 1-2/PM LPL 1-2/LA SCVB/BC EICB/BC OGC LPL 1-2/BC LPL 1-2/PM NAME REnnis ABaxter RDennig JThorp MSpencer MKhanna REnnis (MHenderson for)

DATE 11/20/14 11/26/14 12/3/14 12/4/14 12/8/14 12/9/14 12/29/14