ML18278A030

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Issuance of Amendments 232, 218 to Renewed Facility Operating Licenses Request to Revise the Technical Specifications Surveillance Requirement to Revise the Lower Setpoint Tolerance for Safety/Relief Valves
ML18278A030
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 12/19/2018
From: Bhalchandra Vaidya
Plant Licensing Branch III
To: Bryan Hanson
Exelon Generation Co
Vaidya B
References
EPID L-2018-LLA-0052
Download: ML18278A030 (20)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 19, 2018 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President an'd Chief Nuclear Officer Exelon Nuclear LaSalle County Station 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LASALLE COUNTY STATION, UNITS 1 AND 2- ISSUANCE OF AMENDMENTS TO RENEWED FACILITY OPERATING LICENSES RE:

LICENSE AMENDMENT REQUEST TO REVISE THE TECHNICAL SPECIFICATIONS SURVEILLANCE REQUIREMENT 3.4.4.1 AND THE LOWER SETPOINT TOLERANCES FOR SAFETY/RELIEF VALVES (EPID L-2018-LLA-0052)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (NRC or Commission) has issued the enclosed Amendment No. 232 to Renewed Facility Operating License No. NPF-11 and Amendment No. 218 to Renewed Facility Operating License No. NPF..:18 for the LaSalle County Station (LSCS), Units 1 and 2, respectively. The amendments are in response to your application dated February 27, 2018.

The amendments revised LSCS Technical Specification 3.4.4, "Safety/Relief Valves (S/RVs)."

Specifically, the amendments change the lower setpoint tolerances from -3 percent to -5 percent in Surveillance Requirement 3.4.4.1.

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

Sincerely, Bhalchandra K. Vaidya, Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-37 4

Enclosures:

1. Amendment No. 232 to NPF-11
2. Amendment No. 218 to NPF-18
3. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-373 LASALLE COUNTY STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 232 Renewed License No. NPF-11

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by the Exelon Generation Company, LLC (the licensee), dated February 27, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 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 set forth in 10 CFR Chapter I; 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 the Renewed Facility Operating License No. NPF-11 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised Through Amendment No. 232, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 45 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 0cJ)2* ~-,

David J. Wrona, hief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: December 1 9 , 2 O1 8

Attachment:

Revised License and Technical Specification Pages

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-37 4 LASALLE COUNTY STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 218 Renewed License No. NPF-18

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by the Exelon Generation Company, LLC (the licensee), dated February 27, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 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 set forth in 10 CFR Chapter I; 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 the Renewed Facility Operating License No. NPF-18 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 218, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 45 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION C)J} 2* of-----

David J. Wrona, hief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: December 1 9, 2 O1 8

Attachment:

Revised License and Technical Specification Pages

ATTACHMENT TO LICENSE AMENDMENT NOS. 232 AND 218 RENEWED FACILITY OPERATING LICENSE NOS. NPF-11 AND NPF-18 LASALLE COUNTY STATION, UNITS 1 AND 2 DOCKET NOS. 50-373 AND 50-37 4 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT License License NPF-11, Page 3 NPF-11, Page 3 NPF-18, Page 3 NPF-18, Page 3 TSs TSs TS 3.4.4-2 TS 3.4.4-2

Renewed License No. NPF-11 (3) Exelon Generation Company, LLC, 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; Am. 146 (4) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 01/12/01 30, 40, and 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 Am. 202 (5) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 07/21/11 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2, and such Class Band Class Clow-level radioactive waste as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, and Clinton Power Station, Unit 1.

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 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:

Am. 198 (1) Maximum Power Level 09/16/10 The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3546 megawatts thermal).

Am. 232 (2) Technical Specifications and Environmental Protection Plan 12/19/18 The Technical Specifications contained in Appendix A, as revised through Amendment No. 232, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Am. 194 (3) DELETED 08/28/09 Am. 194 (4) DELETED 08/28/09 Am. 194 (5) DELETED 08/28/09 Amendment No. 232

Renewed License No. NPF-18 (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and 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, and 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 Am. 189 (5) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 07/21/11 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2, and such Class Band Class Clow-level radioactive waste as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, and Clinton Power Station, Unit 1.

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 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:

Am. 185 (1) Maximum Power Level 09/16/10 The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3546 megawatts thermal). Items in Attachment 1 shall be completed as specified. Attachment 1 is hereby incorporated into this license.

Am. 218 (2) Technical Specifications and Environmental Protection Plan 12/19/18 The Technical Specifications contained in Appendix A, as revised through Amendment No. 218, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 218

S/RVs 3.4.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 - - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - -

Less than or equal to two required S/RVs may be changed to a lower setpoint group.

Verify the safety function lift setpoints In accordance of the required S/RVs are as follows: with the INSERVICE Number of Setpoint TES TI NG PROGRAM S/RVs (psiq)

+ 36.1 2 1205 - 60.2

+ 35.8 3 1195 - 59.7

+ 35.5 2 1185 - 59.2

+ 35.2 4 1175

- 58.7

+ 34.5 2 1150 - 57.5 Following testing, lift settings shall be within+/- 1%.

LaSalle 1 and 2 3.4.4-2 Amendment No. 232/218

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 232 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-11 AND AMENDMENT NO. 218 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-18 EXELON GENERATION COMPANY, LLC LASALLE COUNTY STATION, UNITS 1 AND 2 DOCKET NOS. 50-373 AND 50-374

1.0 INTRODUCTION

By application dated February 27, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18058A257), Exelon Generation Company, LLC (EGC or the licensee), requested amendments to Renewed Facility Operating License Nos. NPF-11 and NPF-18 for the LaSalle County Station (LSCS), Units 1 and 2, respectively. The proposed amendments would revise LSCS Technical Specification (TS) 3.4.4, "Safety/Relief Valves (S/RVs)."

Specifically, EGC proposes a new safety function lift setpoint lower tolerance for the S/RVs as delineated in Surveillance Requirement (SR) 3.4.4.1. The proposed revision would change the lower setpoint tolerances from -3 percent to -5 percent.

2.0 REGULATORY EVALUATION

The NRC staff considered the following regulatory requirements and guidance to review the license amendment request {LAR):

(A) General Design Criteria (GDCs)

The NRC staff identified the following GDC as being applicable to this LAR:

  • GDC 10, "Reactor design," requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

exceeded during any condition of normal operation, including anticipated operational occurrences.

  • GDC 16, "Containment design," requires that the reactor containment and associated systems be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
  • GDC 35, "Emergency Core Cooling," requires a system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that: ( 1) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (2) clad metal-water reaction is limited to negligible amounts.
  • GDC 50, "Containment design basis," requires that the reactor containment structure, including access openings, penetrations, and containment heat removal system be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident (LOCA). This margin shall reflect consideration of: ( 1) effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and as required by Title 10 of the Code of Federal Regulations (10 CFR), Section 50.44, energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) limited experience and experimental data available for defining accident phenomena and containment responses, and (3) conservatism of the calculational model and input parameters.

(B) Technical Specification Requirements The regulation at 10 CFR 50.36(a)(1) states that "Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications."

In 10 CFR 50.36, "Technical specifications," the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36(c), TSs are required to include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCO); (3) SRs; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plant's TSs.

Per 10 CFR 50.36(b), TSs must be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto.

Per 10 CFR 50.36(c)(2), LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

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 LCOs will be met.

(C) Other Regulations and Guidance The regulation, 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems [ECCS]

for light-water nuclear power reactors," in part, establishes standards for the calculation of ECCS accident performance and acceptance criteria for that calculated performance.

Anticipated transients without scram (AlWS) is defined in 10 CFR 50.62(b) as an anticipated operational occurrence (AOO) followed by the failure of the reactor trip portion of the protection system specified in GDC-20. In 10 CFR 50.62, "Requirements for reduction of risk from anticipated transients without scram (AlWS) events for light-water-cooled nuclear power plants," states, in part:

  • Each boiling-water reactor (BWR) must have a standby liquid control (SLC) system with the capability of injecting into the reactor vessel a borated water solution with reactivity control at least equivalent to the control obtained by injecting 86 gallons per minute of a 13 weight-percent sodium pentaborate decahydrate solution at the natural boron-10 isotope abundance into a 251-inch inside diameter reactor pressure vessel for a given core design.
  • Each BWR must have equipment to trip the reactor coolant recirculation pumps automatically under conditions indicative of an AlWS. This equipment must be designed to perform its function in a reliable manner.

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Revision 2, March 2007, provides guidance to U.S. Nuclear Regulatory Commission (NRC) staff in performing safety reviews of construction permit (CP) or operating license (OL) applications (including requests for amendments) under 10 CFR Part 50.

The regulation in 10 CFR 50.55a(f) requires that systems and components of boiling water reactors must meet the requirements for preservice and inservice testing requirements of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code and ASME Operation and Maintenance (OM) Code, in this case the main steam (MS) SRVs.

Proposed alternatives to requirements imposed by 10 CFR 50.55a{b) through (h} may be used when authorized by the Director of the Office of Nuclear Reactor Regulation under 10 CFR 50.55a(z).

3.0 TECHNICAL EVALUATION

3.1 System Description The LSCS reactor pressure relief system consists of S/RVs located on the main steam lines between the reactor vessel and the first isolation valve within the drywall. These valves protect against overpressure of the RCS. These valves were manufactured by Crosby Valve and Gauge Company to ASME Code, Section Ill, 1971 Edition up to and including Summer 1972 Addenda. They comply with ASME Code, Section Ill, Paragraph NB-7640, as safety valves with auxiliary actuating devices.

The Crosby Dual Function Safety/Relief Valve, Model HB-65-BP, is designed to perform either as a safety valve or as a relief valve. The safety mode of operation is independent and separate from the relief mode of operation.

The S/RVs provide three main protection functions:

a. Overpressure relief operation - The valves open automatically to limit a pressure rise.
b. Overpressure safety operation - The valves function as safety valves and open (self-actuated operation if not already automatically opened for relief operation) to prevent nuclear system over pressurization.
c. Depressurization operation - The automatic depressurization system (ADS) valves open automatically as part of the ECCS for events involving small breaks in the nuclear system process barrier (i.e., RCS).

In the safety mode (or spring mode of operation), the direct action of the steam pressure in the main steam lines will act against a spring loaded disk that will pop open when the valve inlet pressure exceeds the spring force. In the relief mode (or power actuated mode of operation), a pneumatic piston/cylinder and mechanical linkage assembly are used to open the valve by overcoming the spring force, even with the valve inlet pressure equal to O pounds per square inch gauge (psig). The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure reaches the spring lift set pressures. In the relief mode, valves may be opened manually or automatically at the selected preset pressure.

The overpressure protection system must accommodate the most severe pressure transient.

The licensee has determined that the most severe transient is the closure of all main steam isolation valves (MSIVs) followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position). For the purpose of the analyses in its safety analysis report (SAR), Section 5.2.2, 12 of the S/RVs are assumed to operate in the safety mode. The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110 percent of vessel design pressure ( 11 O percent x 1250 psig = 1375 psig ).

The S/RV safety setpoints are established to ensure that the ASME Section Ill Code limit on peak reactor pressure is satisfied. The ASME Code specifies that the lowest safety valve is to be set at or below vessel design pressure (1250 psig) and the highest safety valve be set so the total accumulated pressure does not exceed 110 percent of the design pressure for over pressurization conditions. The transient evaluations involving the safety mode are based on these setpoints but also include the additional uncertainties of+/- 3 percent of the nominal setpoint to account for potential setpoint drift to provide an added degree of conservatism.

TS SR 3.4.4.1 requires that the licensee verify that the S/RVs will open at the pressures assumed in the safety analysis. The demonstration of the S/RV safety function lift settings must be performed during shutdown since this is a bench test and in accordance with the licensee's inservice testing program, as required by 10 CFR 50.55a(f). The lift setting pressure must correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The current S/RV setpoint is+/- 3 percent for OPERABILITY; however, the LSCS test procedures require that the valves are reset to +/- 1 percent during the surveillance to allow for drift.

3.2 Proposed TS changes The LSCS, Units 1 and 2, TS SR 3.4.4.1 currently reads as follows:

SURVEILLANCE FREQUENCY

N()TE------------------------

SR 3.4.4.1 Less than or equal to two required S/RVs may be changed to a lower setpoint group.

Verify the safety function lift setpoints In accordance of the required S/RVs are as follows: with the INSERVICE Number of Setpoint TESTING

$/RVs (psig) PROGRAM 2 1205 +/- 36.1 3 1195 +/- 35.8 2 1185 +/- 35.5 4 1175 +/- 35.2 2 1150 +/- 34.5 Following testing, lift settings shall be within+/- 1%.

The upper and lower tolerances listed after each nominal setpoint represent +3 percent and -3 percent of each nominal setpoint. The licensee proposed changing the lower tolerance for each listed nominal setpoint to -5 percent of each nominal setpoint.

The proposed revised SR 3.4.4.1 for each Unit would read:

SURVEILLANCE FREQUENCY SR 3.4.4.1 -----------------------------------N()TE-------------------------------- In accordance with Less than or equal to two required S/RVs may be the INSERVICE changed to a lower setpoint group. TESTING PR()GRAM Verify the safety function lift setpoints of the required S/RVs are as follows:

Number of Setpoint S/RVs (Q§ig}

+ 36.1 2 1205

- 60.2

+ 35.8 3 1195

- 59.7

+ 35.5 2 1185

- 59.2

+ 35.2 4 1175

- 58.7

+ 34.5 2 1150

- 57.5 Following testing, lift settings shall be within +/- 1%.

3.3 Evaluation of Licensee's Proposed S/RV Setpoint Tolerances

()n March 8, 1993, the NRC issued a safety evaluation (SE) for the General Electric Nuclear Energy Licensing Topical Report, NEDC-31753P, submitted by the Boiling Water Reactor

()wners Group (BWR()G). In the SE, the NRC stated that a generic change of setpoint tolerance to +/- 3 percent is acceptable provided that it is evaluated in the plant-specific safety analysis. The required analysis was completed for LSCS and the change was approved by the NRC in Amendment No 108 transmitted by a letter dated January 3, 1996 (ADAMS Accession No. ML021130181 ). The current operability of the S/RVs is based on the TS SR acceptance criteria with a setpoint tolerance of +/- 3 percent. If any S/RV exceeds the tolerance, an issue report for each S/RV that exceeds the tolerance is entered into the LSCS corrective action program to evaluate the test failure. In addition, test failures outside of+/- 3 percent would result

in testing additional valves to comply with the ASME OM Code requirements.

The licensee provided an evaluation of the proposed TS changes in Section 3.2 of the LAR.

The licensee evaluated the effect of the change on analyses related to: thermal limits, ASME Code overpressure, LOCA, high pressure system performance, ATWS, and containment systems, structures, or components' (SSCs) ability to manage energy release during S/RV actuation. The license evaluated the effect of the change on plant operating margins in Section 3.3 of the LAR. The licensee provided surveillance test history for the valves in of the LAR which was summarized in LAR Section 3.4.

ASME OM Code Requirements In LAR Section 4.1.2, the licensee states:

The ASME OM Code Mandatory Appendix I, lnservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants, Guiding Principles, paragraph 1-1310(e ), Acceptance Criteria, allows the owner to establish setpoint acceptance criteria for relief valves tested under the 1ST Program; therefore, no relief will be required with regard to the setpoint tolerance change from -3%

to -5%.

Additionally, in LAR Section 1.0, the licensee states that "the proposed change relaxes an unnecessarily restrictive SR." The licensee thereby, indicates that a request pursuant to 10 CFR 50.55a(z) for NRC authorization of an alternative to ASME OM Code requirements imposed by 10 CFR 50.55a is not required in order to change the S/RV lower lift setpoint tolerance in the TSs from -3 percent to -5 percent.

The S/RVs are Class 1, Category C, valves. ASME OM Code, Mandatory Appendix I, "lnservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants,"

paragraph 1-3300, requires the periodic testing frequency of S/RVs at BWRs to be determined in accordance with the paragraphs in 1-1300, "Guiding Principles." Paragraph 1-1320, "Test Frequencies, Class 1 Pressure Relief Valves," states, in part:

(c) Requirements for testing additional valves shall be tested in accordance with the following requirements:

(1) For each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either the +/- tolerance limit of the owner

[licensee]-established set-pressure acceptance criteria of 1-131 O(e) or

+/- 3 percent of valve nameplate set-pressure, two additional valves shall be tested from the same valve group.

(2) If the as-found set-pressure of any of the additional valves tested in accordance with 1-1320 (c )( 1) exceeds the criteria noted therein, then all remaining valves of that same valve group shall be tested.

The licensee indicated that the S/RVs will continue to be tested in accordance with these ASME OM Code requirements.

The NRC staff reviewed the licensee's proposed amendment as it relates to the ASME OM Code requirements imposed by 10 CFR 50.55a(f). ASME OM Code, Mandatory Appendix I,

Paragraph l-1310{e), "Guiding Principles," states: "the Owner, based upon system and valve design basics or technical specification, shall establish and document acceptance criteria for tests required by this Appendix." ASME OM Code section l-1310(e) allows the owner [licensee]

to set the acceptance criteria for the S/RV periodic tests based on the limits in the Technical Specifications, and 1-1320(c )( 1) allows the owner to establish acceptance criteria beyond

+/- .3 percent of nameplate set-pressure. Therefore, the staff concludes that a licensee request that the NRC authorize an alternative pursuant to 10 CFR 50.55a(z) is not necessary for the licensee to change the S/RV lower setpoint tolerances from -3 percent to -5 percent.

3.4 Evaluation on System Functions 3.4.1. ASME Code Overpressure Protection Overpressure protection for the RCPB during power operation is provided by S/RVs and the reactor protection system. GDC-15 requires that the RCS and associated auxiliary, control, and protection systems be designed with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal operation, including AOOs.

For the LSCS current operation, the limiting transient for ASME Code overpressure consideration is the postulated MSIV closure. Analysis of the ASME Code overpressure event has been based on the S/RV upper tolerance settings (i.e., +3 percent setting). Therefore, the NRC staff finds that the proposed change in lower S/RV tolerance from -3 percent to -5 percent is acceptable because the proposed change has no impact on the ASME Code overpressure analysis and the analysis is in compliance with GDC 15.

3.4.2 ATWS Overpressure Analysis An ATWS is defined in 10 CFR 50.62(b) as an AOO followed by the failure of the reactor portion of the protection system specified in GDC-20. Requirements related to ATWS events are specified in 10 CFR 50.62. The NRC staff review of ATWS events includes confirming that the peak reactor vessel bottom pressure will be less than the ASME Code Service Level C limit of 1500 psig (NUREG-0800, Section 15.2.8, SRP acceptance criterion 2) during an ATWS overpressure event as protected by the S/RVs and systems required in accordance with 10 CFR 50.62 (e.g., alternate rod injection system, SLC system).

The methodology used for the LSCS ATWS overpressure analysis is similar to the ASME Code overpressure event (MSIV closure) discussed above, except that a complete failure of scram is assumed. In any pressurization case, an earlier opening of the S/RV due to lower setpoint would produce improved results and gain safety margin. Analysis of the ATWS peak pressure and suppression pool temperature are conservatively based on the S/RV upper tolerance settings (i.e., +3 percent setting). The NRC staff finds that the proposed change in the lower S/RV tolerance from -3 percent to -5 percent is acceptable because the proposed change has no impact on the ATWS analysis and the ATWS analysis meets the acceptance criterion for the ATWS-related requirement in 10 CFR 50.62(b).

3.4.3 Thermal Limits Assessment The GDC 10 provides requirements related to core design and protection systems in order to assure that acceptable fuel damage limits are not exceeded.

The safety limit minimum critical power ratio (SLMCPR) ensures that 99.9 percent of the fuel rods are protected from boiling transition during normal operation or AOO. The operating limit minimum critical power ratio ensures that the SLMCPR will not be exceeded as a result of an AOO. The maximum average planar linear heat generation rate (MAPLHGR) and linear heat generation rate {LHGR) limits ensure compliance with limits, such as peak cladding temperature (PCT) of 2200 °F, oxidation limit, etc., established in 10 CFR 50.46. The MAPLHGR limit is determined by analyzing the limiting LOCA for the given plant (see the next section for details).

The LHGR limit is determined by the fuel rod thermal-mechanical design analysis.

The licensee determined that all AOO thermal limits analyses credit the relief mode operation of the S/RV. The S/RV safety function pressure setpoints expanded to the proposed -5 percent, are still greater than the S/RV relief function pressure setpoints assumed in the SAR analyses, Section 15; thus, the S/RV relief function pressure setpoints at which the S/RVs lift is assumed in the SAR analyses are not affected. The NRC staff finds that the proposed change in lower S/RV tolerance from -3 percent to -5 percent is acceptable because the change is consistent with the AOO safety analyses and AOO safety analyses still meet the GDC 10 that the specified acceptable fuel design limits will not be exceeded during normal and AOO conditions.

3.4.4 ECCS-LOCA Performance Assessment The NRC staff review related to ECCS and LOCA performance is based on: (1) 10 CFR 50.46, which establishes standards for the calculation of ECCS performance, such as PCT of 2200 °F, oxidation limit, etc.; and (2) GDC 35, which requires that a system to provide abundant emergency core cooling shall be provided. It also requires that the ECCS function be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that: (a) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (b) clad metal-water reaction is limited to negligible amounts.

The LSCS ECCS-LOCA analysis credits the automatic depressurization system (ADS) function of the S/RVs during LOCA as the ADS function and is not based on pressure setpoints. During a large break LOCA, the reactor vessel is automatically depressurized due to the nature of the event and the S/RV safety function is not actuated. The NRC staff finds that the proposed change in the lower S/RV tolerance from -3 percent to -5 percent is acceptable because the proposed change has no impact on the ECCS-LOCA analyses and the analyses meet 10 CFR 50.46 and GDC 35 requirements.

3.4.5 High Pressure System Performance Assessment The high pressure systems include the high pressure core spray (HPCS), reactor core isolation cooling (RCIC) and SLC systems. The NRC staff review for all these high pressure systems are based on: (1) GDC 35, which requires these systems to provide abundant emergency core cooling, and (2) 10 CFR 50.62(c)(4), which requires that the SLC system be capable of reliably injecting a borated water solution into the reactor pressure valve at a boron concentration, boron enrichment, and flow rate that provides a set level of reactivity control.

For LSCS, the most significant effect of changing the S/RV setpoint tolerance on the HPCS, RCIC, and SLC systems operations is the maximum reactor pressure at which they are required to deliver flow to the reactor. The limiting safety analyses on these systems are conservatively based on the S/RV upper tolerance settings (i.e., +3 percent setting). The NRC staff finds that the proposed change in lower S/RV tolerance from -3 percent to -5 percent is acceptable because lower tolerance setpoint is not used in the assessment and thus the proposed change

has no impact on the analyses for high pressure systems. Therefore, the systems still meet the above high pressure systems requirements in GDC 35 and 10 CFR 50.62(c)(4).

3.4.6 Containment Analysis The NRC staff review related to containment and S/RV is based on the requirements of GDCs 16 and 50 described in Section 2.0 of this SE.

For LSCS, the containment-related analyses such as peak suppression pool temperature, peak suppression pool pressure, discharge line dynamic load, and submerged structure loads are all most limited by high mass/energy flow rates from the vessel into the containment structure. The flow rate through the S/RVs is directly proportional to the pressure differential across the valves.

Since the proposed change on S/RV setpoint tolerance is only at the minimum value and does not affect the limiting analyses that were performed at the maximum value of the S/RV pressure setpoint, there is no impact on the containment SSCs' ability to manage energy release during S/RV actuation.

The NRC staff finds that the proposed change in the lower S/RV setpoint tolerance from -3 percent to -5 percent is acceptable for the containment related analyses because the proposed change does not impact the containment related analyses and, thus, the analyses still meet the containment requirements in GDC 16 and GDC 50.

3.4. 7 S/RV Safety Mode Performance and Impact of Proposed Change of Lower Setpoint Tolerance The licensee evaluated S/RV safety mode performance using surveillance test historical data covering a period from refueling outage 02 in 1988 to refueling outage 16 in 2017 for Unit 2, and refueling outage 02 in 1988 to refueling outage 16 in 2016 for Unit 1. There were 14 valve tests out of 133 test results that failed outside the -3 percent tolerance range with 3 of these valves testing outside the -5 percent tolerance range. The data includes tests for valves that span up to 6 years between bench tests. The NRC staff reviewed and accepted the licensee's conclusion that an excessive S/RV drift over 6 years is not to be expected because the test data has demonstrated that the valve performance is consistent.

The proposed change in lower setpoint tolerance from -3 percent to -5 percent would tend to decrease the valve simmer margin (difference between normal plant operating pressure and the valve setpoint) and that will increase the probability of valve leakage. The licensee uses the Crosby Valve and Gage Company Procedure 1-11069, "Instruction Manual for Crosby Style 6xRx1 O HB-65-BP Safety Relief Valve for Main Steam Service," Revision 1, which discusses valve performance in Section 3. Step 3.1.1.4 in Section 3 states the following:

Setpoint repeatability for on line service should be within a tolerance of +1, -3% of nameplate set pressure. However, the low limit of the setpoint tolerance may be extended to 1067 psig if all other valve functional requirements are met.

The proposed change would lower the minimum S/RV as-found set point acceptance criteria for the lowest setpoir.it from 1116 psig to 1093 psig which is still higher than the manufacturer's recommended low setpoint pressure limit of 1067 psig. The NRC staff, therefore, finds that the proposed change in the lower setpoint tolerance from -3 percent to -5 percent is acceptable because the change is consistent with the manufacturer's recommendation and provides sufficient simmering margin of more than 50 pounds per square inch so that the necessary

quality of reactor vessel and SRN is still maintained by the SR, as required by 10 CFR 50.36(c)(3).

3.4.8 Conclusion - Technical Evaluation The NRC staff reviewed the evaluations and justifications provided by the licensee and determined that lowering the S/RV tolerance range from -3 percent to -5 percent does not create an adverse impact on the AOO thermal limits because the current thermal limits bound the expanded S/RV range. The staff also determined that lowering the S/RV tolerance range from -3 percent to -5 percent does not have an adverse impact on: the ASME Code overpressure analysis result (as discussed in Section 3.4.1 of this SE), the AOO thermal limits analyses (as discussed in Section 3.4.3 of this SE), any type of LOCA (as discussed in Section 3.4.4 of this SE), high pressure system performance (as discussed in Section 3.4.5 of this SE),

the ATWS overpressure (as discussed inSection 3.4.2 of this SE) or suppression pool temperature analysis results, containment SSCs' ability to manage energy release during S/RV actuation (as discussed in Section 3.4.6 of this SE) or the margins during normal plant operation (as discussed in Section 3.4.7 of this SE). The staff determined that the licensee's evaluations justify the changes. The staff also found that TS SR 3.4.4.1, as modified by the proposed changes, will continue to meet 10 CFR 50.36 requirements because the SR will continue to provide assurance that the necessary quality of systems and components is maintained, that the facility operation will be within safety limits, and that LCOs will be met. Therefore, the staff concludes that the proposed changes are acceptable.

Based on the above evaluation, the NRC staff finds that because the licensee demonstrated through justifiable assumptions and analyses that the regulatory requirements GDCs 10, 15, 16, 35, 50, and 10 CFR 50.36, 50.46, and 50.62, will continue to be met, the relevant safety systems as potentially affected by the proposed change in lower S/RV tolerance from -3 percent to -5 percent shall maintain their design functions as required by the GDCs 10, 15, 16, 35, 50, and 10 CFR 50.36, 50.46 and 50.62. Therefore, the staff concludes that the proposed TS change is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendment on October 4, 2018. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR, Part 20, and changes SRs.

The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding (Federal Register (83 FR 15416) dated April 10, 2018). Accordingly, the amendment meets 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 amendment.

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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: M. Hamm, NRR/STSB A. Mink, NRR/EMIB S. Peng, NRR/SRXB Date of Issuance: December 19, 2018

B. Hanson

SUBJECT:

LASALLE COUNTY STATION, UNITS 1 AND 2- ISSUANCE OF AMENDMENTS TO RENEWED FACILITY OPERATING LICENSES RE:

LICENSE AMENDMENT REQUEST TO REVISE THE TECHNICAL SPECIFICATIONS SURVEILLANCE REQUIREMENT 3.4.4.1 AND THE LOWER SETPOINT TOLERANCES FOR SAFETY/RELIEF VALVES (EPID L-2018-LLA-0052) DATED DECEMBER 19, 2018 DISTRIBUTION:

PUBLIC PM File Copy RidsACRS_MailCTR Resource RidsNrrDorlLpl3 Resource RidsNrrDssStsb Resource RidsNrrLASRohrer Resource RidsNrrPMLaSalle Resource RidsRgn3MailCenter Resource RidsNrrDssSrxb Resource RidsNrrDeEmib Resource M. Hamm, NRR/STSB S. Peng, NRR/SRXB A. Mink, NRR/EMIB ADAMS Accession No.: ML18278A030 (**) no substantial change from SElnput memo OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/STSB/BC(**) NRR/DE/EMIB/BC NAME BVaidya SRohrer VCusumano SBailey DATE 10/10/18 10/10/18 10/5/18 10/11/18 OFFICE NRR/DSS/SRXB/BC(**) NRR/DE/EICB/BC OGC NLO/w comments NRR/DORL/LPL3/BC NAME JWhitman MWaters MYoung DWrona DATE 10/4/18 10/7/18 11/20/18 12/19/18 OFFICE NRR/DORL/LPL3/PM NAME BVaidya (KGreen for)

DATE 12/19/18 OFFICIAL RECORD COPY

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