ML24303A340

Issuance of Amendment Nos. 275 and 277 Regarding Revision to Technical Specification 3.6.5, Containment Air Temperature

ML24303A340
Person / Time
Site:	Point Beach
Issue date:	11/21/2024
From:	Scott Wall Plant Licensing Branch III
To:	Coffey B Point Beach
Wall, S P
References
EPID L-2024-LLA-0088
Download: ML24303A340 (1)

Text

November 21, 2024 Robert Coffey Executive Vice President, Nuclear Division and Chief Nuclear Officer Florida Power & Light Company Mail Stop: EX/JB 700 Universe Blvd Juno Beach, FL 33408

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 275 AND 277 REGARDING REVISION TO TECHNICAL SPECIFICATION 3.6.5, CONTAINMENT AIR TEMPERATURE (EPID L-2024-LLA-0088)

Dear Robert Coffey:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 275 and 277 to Renewed Facility Operating License Nos. DPR-24 and DPR-27, respectively, for the Point Beach Nuclear Plant, Units 1 and 2, in response to your application dated June 26, 2024, as supplemented by letter dated August 9, 2024.

The amendments revise Technical Specification (TS) 3.6.5, Containment Air Temperature, by specifying a single containment average air temperature limit and deleting current TS limiting condition for operation (LCO) 3.6.5a, 3.6.5b, and 3.6.5c.

A copy of the related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Scott P. Wall, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosures:

1. Amendment No. 275 to DPR-24

2. Amendment No. 277 to DPR-27

3. Safety Evaluation cc: Listserv

NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 275 License No. DPR-24

1.

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

A.

The application for amendment by NextEra Energy Point Beach, LLC (the licensee), dated June 26, 2024, as supplemented by letter dated August 9, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment, and paragraph 4.B of Renewed Facility Operating License No. DPR-24 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 275, are hereby incorporated in the renewed operating license. NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert Kuntz, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: November 21, 2024 ROBERT KUNTZ Digitally signed by ROBERT KUNTZ Date: 2024.11.21 11:12:43 -05'00'

NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 277 License No. DPR-27

1.

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

A.

The application for amendment by NextEra Energy Point Beach, LLC (the licensee), dated June 26, 2024, as supplemented by letter dated August 9, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment, and paragraph 4.B of Renewed Facility Operating License No. DPR-27 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 277, are hereby incorporated in the renewed operating license. NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert Kuntz, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: November 21, 2024 ROBERT KUNTZ Digitally signed by ROBERT KUNTZ Date: 2024.11.21 11:13:21 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 275 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AND LICENSE AMENDMENT NO. 277 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 DOCKET NOS. 50-266 AND 50-301 Replace the following pages of Renewed Facility Operating License Nos. DPR-24 and DPR-27 and 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.

Renewed Facility Operating License No. DPR-24 REMOVE INSERT Renewed Facility Operating License No. DPR-27 REMOVE INSERT Appendix A, Technical Specifications REMOVE INSERT 3.6.5-1 3.6.5-1 Renewed License No. DPR-24 Amendment No. 275 D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach 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 E.

Pursuant to the Act and 10 CFR Parts 30 and 70, NextEra Energy Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.

4.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; 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 below:

A.

Maximum Power Levels NextEra Energy Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1800 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 275, are hereby incorporated in the renewed operating license.

NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.

C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensees application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.

Renewed License No. DPR-27 Amendment No. 277 C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach to receive, possess and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed source for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach 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 E.

Pursuant to the Act and 10 CFR Parts 30 and 70, NextEra Energy Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.

4.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; 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 below:

A.

Maximum Power Levels NextEra Energy Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1800 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 277, are hereby incorporated in the renewed operating license.

NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.

C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensees application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.

Containment Air Temperature 3.6.5 Point Beach 3.6.5-1 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be 120°F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Containment average air temperature not within limit.

A.1 Restore containment average air temperature to within limit.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B.

Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is within limit.

In accordance with the Surveillance Frequency Control Program Unit 1 - Amendment No. 275 Unit 2 - Amendment No. 277

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 275 AND 277 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27, RESPECTIVELY NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301

1.0 INTRODUCTION

By application to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated June 26, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24178A265), as supplemented by letter dated August 9, 2024 (ML24222A401), NextEra Energy Point Beach, LLC (the licensee) submitted a license amendment request (LAR) for the Point Beach Nuclear Plant, Units 1 and 2 (Point Beach).

The proposed amendments consist of changes to the Point Beach technical specifications (TSs) that would revise TS 3.6.5, Containment Air Temperature, by specifying a single containment average air temperature limit and deleting current TS limiting condition for operation (LCO) 3.6.5a, 3.6.5b, and 3.6.5c.

The supplemental letter of August 9, 2024, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 6, 2024 (89 FR 63992).

1.1

System Description

In Section 2.1 of the LAR, NextEra provided the following system description:

The Point Beach Unit 1 and 2 containments are horizontally and vertically pre-stressed, post-tensioned concrete cylinders positioned on reinforced concrete slabs and covered by pre-stressed, post-tensioned shallow concrete domes. Their structural members have sufficient capacity to accept a combination of normal operating loads, functional loads due to a loss of coolant accident (LOCA), and the loadings imposed by the safe shutdown earthquake (SSE) without exceeding specified stress limits. The design pressure and temperature of the containments are in excess of the peak pressure and temperature occurring as the result of the complete blowdown of the reactor coolant through any rupture of the reactor coolant system up to and including the hypothetical severance of a reactor coolant pipe. The pressure and temperature loadings obtained by analyzing various LOCAs, when combined with operating loads and maximum wind or seismic forces, do not exceed the load carrying capacity of the structures.

The containment air temperature monitoring system provides indication in the control room so operators can verify that the average containment temperature is below the initial conditions assumed in the design basis accident (DBA) analyses. The maximum allowable average containment air temperature is 120°F (degree Fahrenheit), which represents the initial condition in the safety analyses. In the current TS, there are three different allowable limits depending on the availability of instrumentation and these limits incorporate instrument uncertainty directly into the individual TS values. The licensee states this allows for operational flexibility when one or more containment channels are unavailable while preserving the initial condition assumptions in the DBA analyses.

1.2 Description of Changes The current LCO for TS 3.6.5 requires the containment average air temperature to be below a certain average air temperature threshold based on the available number of instrument channels.

The current LCO statement for TS 3.6.5 reads:

a. 116.3°F based on three averaged temperature channels,

b. 115.7°F based on two averaged temperature channels, or

c. 112.5°F based on a single temperature channel The proposed change for the LCO statement reads:

Containment average air temperature shall be 120°F.

This proposed change modifies LCO 3.6.5 by specifying a single containment average air temperature limit and by deleting current TS LCO 3.6.5a, 3.6.5b, and 3.6.5c.

1.3 Reason for the Proposed Change In section 2.3 of the LAR, NextEra provided the following reason:

The proposed change aligns TS 3.6.5, Containment Air Temperature, with NUREG-1431, Standard Technical Specifications, Westinghouse Plants

[ML21259A155], which presents a single value for the containment average air temperature and relocates to licensee control the procedural details associated with an accounting for instrument uncertainty based on the available number of measurement channels.

2.0 REGULATORY EVALUATION

2.1 Applicable Regulatory Requirements Under Title 10 of the Code of Federal Regulations (10 CFR) 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a) (regarding, among other things, consideration of the operating procedures, the facility and equipment, the use of the facility, and other technical specifications, or the proposals) and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commission's regulations.

The regulation at 10 CFR 50.36(c)(2) requires that TSs include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

With regards to the requirements of 10 CFR 50.36(c)(2), in the NRC Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132), the Commission states:

If a licensee elects to apply these criteria, the requirements of the removed specifications will be relocated to the FSAR [Final Safety Analysis Report] or other licensee-controlled documents. Licensees are to operate their facilities in conformance with the descriptions of their facilities and procedures in their FSAR. Changes to the facility or to procedures described in the FSAR are to be made in accordance with 10 CFR 50.59.

Although the licensee makes reference to TS Bases requirements, the NRC staff notes that, in accordance with 10 CFR 50.36(a)(1), TS Bases are not TS requirements. Further, the NRC staff notes that the methodology for calculating instrument uncertainty contained in the licensees proposed change to the TS Bases will continue to be maintained in accordance with 10 CFR 50.59.

The regulation at 10 CFR 50.36(c)(3) requires that TSs include surveillance requirements, which 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 LCO will be met.

The regulations in 10 CFR Part 50, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components (SSCs) important to safety. Point Beach was licensed prior to the GDC of 10 CFR Part 50, Appendix A. The Point Beach Final Safety Analysis Report (FSAR), section 1.3 (ML24116A054), lists plant-specific GDCs to which Point Beach was licensed, which are similar in content to the 1967 proposed GDCs published for public comment. The applicable Point Beach GDCs for this submittal include:

Point Beach GDC 10 (Criterion 16 in 10 CFR Part 50, Appendix A), Reactor Containment, states that the containment structure shall be designed (a) to sustain, without undue risk to the health and safety of the public, the initial effects of gross equipment failures, such as a large reactor coolant pipe break, without loss of required integrity, and (b) together with other engineered safety features as may be necessary, to retain for as long as the situation requires, the functional capability of the containment to the extent necessary to avoid undue risk to the health and safety of the public.

Point Beach GDC 12 (Criterion 13 in 10 CFR Part 50, Appendix A), Instrumentation and Control, states that instrumentation and controls shall be provided as required to monitor and maintain within prescribed operating ranges essential reactor facility operating variables.

2.2 Applicable Regulatory Guidance NRC Regulatory Guides (RGs) provide one way to ensure that the codified regulations continue to be met. The NRC staff considered the following guidance, along with industry guidance endorsed by the NRC, during its review of the proposed changes:

RG 1.105, Setpoints for Safety-Related Instrumentation, Revision 4, February 2021 (ML20330A329), describes an approach to meeting regulatory requirements to ensure that setpoints for safety-related instrumentation are established to protect nuclear power plant safety and analytical limits, and the maintenance of instrument channels implementing these setpoints ensures they are functioning as required, consistent with the plant TSs.

RG 1.105 endorses American National Standards Institute (ANSI)/International Society of Automation (ISA) Standard 67.04.01-2018, Setpoints for Nuclear Safety-Related Instrumentation. The NRC staff used this guidance to establish the adequacy of the licensees setpoint calculation methodologies.

RG 1.97, Revision 3, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, May 1983 (ML003740282), defines the design and qualification criteria for instrumentation Categories 1 through 3 in Table 1. Types A, B, C, D, and E are listed in Table 2 (BWR

[boiling water reactor] Variables) and Table 3 (PWR [pressurized water reactor]

Variables) of this guidance.

NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [light-water reactor] Edition (SRP):

o Chapter 7, Instrumentation and Controls, Branch Technical Position 7-12, Guidance on Establishing and Maintaining Instrument Setpoints, Revision 6, August 2016 (ML16019A200) provides guidelines for the NRC staffs review of the process an applicant or licensee uses to establish and maintain instrument setpoints.

o Chapter 16, Section 16.0, Technical Specifications, March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared standard technical specifications (STSs) for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with the applicable STSs, as modified by NRC-approved travelers. The applicable STSs for Point Beach are:

NRC NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Volume 1, Specifications, and Volume 2, Bases, Revision 5, September 2021 (ML21259A155 and ML21259A159, respectively).

NUREG/CR-3659, Mathematical Model for Assessing the Uncertainties of Instrumentation Measurements for Power and Flow of PWR Reactors, February 1985, Pacific Northwest Laboratory (ML081550335 (non-public)).

3.0 TECHNICAL EVALUATION

3.1 Evaluation of Temperature Measurement Instrument Uncertainty In the LAR, the licensee indicates that the uncertainty calculations were performed in accordance with Point Beach Design and Installation Guidelines Manual DG-101, Instrument Setpoint Methodology Revision 8 (Attachment 1 of the August 9, 2024, supplement). The methodologies described in DG-101, Revision 8, are based on the industry guidance of ANSI/ISA Standard ISA S67.04.01-2000, Part I, Setpoints for Nuclear Safety-Related Instrumentation, and ISA RP67.04.02, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation, Part II.

The NRC staff used ISA 67.04.01-2018, as endorsed by RG 1.105, Revision 4, to evaluate the licensees methodology of the instrument uncertainty calculation.

The NRC staff reviewed the following references in the licensees LAR related to the methodology to calculate the instrument uncertainty for the containment average air temperature limits:

DG-101, Instrument Setpoint Methodology Revision 8 (Attachment 1 of the August 9, 2024, supplement). This document describes the design guide to establish a consistent approach for the analysis of instrument loop uncertainties and their effect on the safety-related setpoint and process indication.

Point Beach LAR No. 262, Revision to Technical Specification Operating Limits to Include Measurement Uncertainty, dated March 23, 2011 (ML110830009), includes Calculation 2006-0035, Revision 1, RWST, Temperature, Containment Average Air Temperature and Spray Additive Tank Level Uncertainty/Setpoint Calculation, that used the results of the total loop uncertainties (errors) for the Containment Average Air Temperature limits from Calculation 2005-0028 Revision 0, Containment Air Temperature Indication Loop Uncertainty. Calculation 2006-0035, Revision 1 converts these uncertainties in percent (%) of Span to degree Fahrenheit (ºF) (in Section 8.1) and calculates the average uncertainties for two and three air temperature indicators (in Section 8.1.1 and 8.1.2).

The previously issued license amendment Point Beach Nuclear Plant, Units 1 and 2 -

Issuance of Amendment to Revise Technical Specification Operating Limits to Include Measurement Uncertainty (TAC Nos. ME5906 and ME5907), dated January 30, 2012 (ML113540147). This amendment authorized the current TS LCO 3.6.5a, 3.6.5b, and 3.6.5c, requested in the March 23, 2011, LAR.

Calculation 2005-0028, Revision 0, Containment Air Temperature Indication Loop Uncertainty (Attachment 2 of the August 9, 2024, supplement), describes the methodology to calculate the uncertainty in Section 7.0, Methodology, and the Total Loop Error calculation in Section 8.3.1, Total Loop Error - Control Room Indicator Leg.

Beggs, W.J., Statistics for Nuclear Engineers and Scientist, Part 1: Basic Statistical Inference, Department of Energy (DOE) Research and Development Report No. WAPD-TM-1292, February 1981. Table III, Normal Distribution provides the value of normal deviate (z) which corresponds to a probability value (i.e., probability 95 percent value and 75 percent value).

3.1.1 Overview of DG-101 Calculations DG-101 states that the purpose of this setpoint methodology design guide is to establish a consistent approach for the analysis of instrument loop uncertainties and their effect on safety-related setpoints and process indications. The Point Beach setpoint methodology provides a methodology that is consistent with the requirements within ISA S67.04, Part I, and ISA RP67.04, Part II. The NRC staff reviewed DG-101 within the scope of the methodology of the total loop uncertainty calculations for the Containment Average Air Temperature limits to confirm that this method is consistent with the NRC-approved methodology in ISA 67.04.01-2018.

An overview of relevant sections of DG-101 is summarized below:

Section 3.1, General All instruments in Point Beach can be included in at least one of the following categories:

A: Reactor Protection System (RPS)/Engineered Safety Feature Actuation System (ESFAS) TS Setpoint Instrument Loop and Post Accident Monitoring (PAM)

Type A in RG 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident B: Other TS Instrument Loops C: Other than PAM Type A Instrument Loops D: Other Safety-related Instrument Loops E: Non-safety-related Instrument Loops.

The licensee notes that the categories listed above pertain to the instrument loop classifications as described in DG-101 only. These categories are not interpreted Types A, B, C, D, E variables in RG 1.97 (Category A in this list includes the PAM, Type A, in RG 1.97).

Section 3.2, Graded Approach This section states, in part, The setpoint methodology will utilize a grade approach in which the probability/confidence level of random uncertainty may be varied based on the safety significance of the setpoint or indication. In addition, this section states that the highest safety significance, defined in the TSs as Type A in accordance with RG 1.97, will combine uncertainties that have 95 percent probability at a 95 percent confidence level (95/95). Non-Type A (i.e., Types B, C, D, E) instrument loop may use more relaxed, 75 percent probability at a 75 percent confidence level (75/75).

Section 3.3.1, Description of Methodology for Combining Uncertainties This section establishes a consistent approach for the analysis of instrument loop uncertainties and their effect on safety-related setpoints and process indications. The determination of loop indication errors is required to provide plant operators with the necessary information to make timely and correct manual actions in response to plant transients and abnormal operating conditions.

In this document, the square-root-sum-of-square (SRSS) and arithmetic are techniques for combining errors effects, including effects characterized as independent, dependent, random, or non-random. The random independent errors are combined by SRSS. The random dependent errors are combined by algebraic, and non-random elements of the errors are combined according to their dependency by the following equation:

TLE = +/- [A2 + B2 + (C + D)2]1/2 +/- lXl + Y - Z (Equation 1)

(Same as the Total Loop Uncertainty (TLU) in ISA 67.04.01-2018)

Where:

A and B: Random and independent elements of uncertainty (are combined by SRSS)

C and D: Random and dependent elements of uncertainty (are combined algebraically according to their dependency and combined with other independent random terms by SRSS)

X, Y, Z:

Non-random elements are combined algebraically according to their sign with the result of the SRSS computation.

X:

Non-random variable with unknown sign Y:

Non-random positive bias Z:

Non-random negative bias As per the description above, this method for combining uncertainties is consistent with section 4.5, Combination of uncertainties, section 4.5.3, Formulas and Methodology Discussion, and section 4.5.3, Formulas and Methodology Discussion, of ISA 67.04.01-2018.

Section 3.3.3.13, Subsection of Section 3.3, Safety-related Setpoints This section indicates that the methodology will utilize:

o 95 percent probability at a 95 percent confidence criteria level (95/95) for instrument Category A (as described in Section 3.1 of DG-101: Type A in accordance with RG 1.97) o 75 percent probability a 75 percent confidence criteria level (75/75) for instrument Categories B & C (as described in Section 3.1 of DG-101: Other TS instrument loops and other than PAM Type A in accordance with RG 1.97)

The document is based on the assumption of sufficiently large samples of the probability value or confidence level that meets the two-standard deviation (2 Sigma) criteria. If a specified lower probability/confidence level (e.g., 1 Sigma), then data must be treated as a non-random variable.

This subsection described the methodology and the following equations to convert a 95/95 value to a 75/75 value.

µ = (Z/2

• sigma) / N1/2 Where:

µ = mean (standard error - equal to zero unless a bias is specified)

Z/2 = normal deviate Z for a one-sided distribution Sigma = standard deviation N1/2 = square root of sample size For a 95 percent value, each size (Z/2) will be 5% / 2 = 2.5% = 0.025 (Z0.025)

For a 75 percent value, each size (Z/2) will be 25% / 2 = 12.5% = 0.125 (Z0.125)

For this conversion, the document used data in Tabe III, Normal Distribution, in DOE Research and Development Report No. WAPD-TM-1292, to determine the error value for a 95/95 value (Error95/95) and for 75/75 (Error75/75).

The value of Error95/95 corresponds to a probability (Z0.025) is 1.96 and The value of Error75/75 corresponds to a probability (Z0.125) is 1.15 The ratio of the two error values:

Error75/75 / Error95/95 = (1.15 / 1.96)

Error75/75 = (1.15 / 1.96)

• Error95/95 (Equation 2)

The NRC staff verified that this conversion is consistent with the information in Table III of DOE Research and Development Report No. WAPD-TM-1292.

Section 3.3.4, Element of Device Uncertainty This section defined the error:

Error = Indicated value - ideal value As per the description above, this definition is consistent with the definition of Error in Section 3.9 of ISA 67.04.01-2018.

3.1.2 Overview of Calculation 2005-0028, Revision 0 The NRC staff reviewed Calculation 2005-0028, Revision 0, which is based on the methodology of DG-101, to verify whether these calculations used a methodology that is consistent with an NRC-approved methodology to calculate the instrument uncertainties for TS 3.6.5 containment average air temperature limits.

Calculation 2005-0028 declares in the following sections that:

Section 7.1, Uncertainty Determination The total loop uncertainties are calculated in accordance with DG-101. This methodology uses the SRSS method to combine random and independent errors, and algebraic addition of non-random or bias error.

Section 7.1.A, Treatment of 95/95 and 75/75 Values The instrumentation uncertainties for air temperature indicators are evaluated as Category B, Other Technical Specifications Loops (as defined by section 3.1 of DG-101). Therefore, the total loop uncertainties (errors) will be expressed as a 75/75 value. This section states that the error 95/95 was converted to error 75/75 in accordance with section 3.3.3.13 of the DG-101.

Section 7.1.2, Total Loop Error Equation Summary (TLE)

The TLE is determined in accordance with the requirements of DG-101. This methodology uses the SRSS method to combine the applicable random and independent errors, and algebraic addition of non-random or bias errors.

Section 8.1, Device Uncertainties Analysis This section introduces all applicable uncertainties for the devices that comprise the Point Beach Containment Air Temperature Indication Instrumentation Loop. This section states, in part, that: From Section 3.3.4.3 of DG-101, the drift values statistically derived from as-found/as-left instrument calibration data normally include the error effects under normal conditions of drift, accuracy, power supply, plant vibration, calibration temperature, normal radiation, normal humidity, M&TE [measuring & test equipment]

used for calibration, and instrument readability. If it is determined that the calibration conditions are indicative of the normal operating conditions, the environmental effects need not be included separately. All device uncertainty terms are considered random and independent unless otherwise noted.

The device uncertainties of the Point Beach Containment Air Temperature Indication Instrumentation Loop are determined in section 8.1.1, Process Interface Uncertainties, and the results are in section 8.2 of this calculation.

Section 8.2, Device Uncertainty Summary The NRC staff summarized the device uncertainties applicable to the control room indicators in Section 8.2 as shown below:

Section 8.2.1 - Sensor Uncertainties Parameter Abbreviation Uncertainty (+/- %Span)

Sensor Primary Element Accuracy Spea 0.629 Sensor Lead Wire Effect Slwe 0.2086 (bias)1 Sensor Insulation Resistance Effect Sire 0.000 Section 8.2.2. - RV [Resistance - Voltage] Converter Uncertainties Parameter Abbreviation Uncertainty (+/- %Span)

RV Converter Accuracy RVa 0.500 RV Converter Drift RVd 0.500 RV Converter M&TE RVm 0.075 RV Converter Setting Tolerance RVv 0.500 RV Converter Power Supply Effect RVp 0.250 RV Converter Temperature Effect RVt 0.222 RV Converter Humidity Effect RVh 0

RV Converter Radiation Effect RVr 0

RV Converter Seismic Effect RVs 0

Section 8.2.3 - VI [Voltage - Current] Converter Uncertainties (Random and independent elements of uncertainty)

Parameter Abbreviation Uncertainty (+/- %Span)

VI Converter Accuracy VIcra 0.500 VI Converter Drift VIcrd 0.500 VI Converter M&TE VIcrm 0.092 VI Converter Setting Tolerance VIcrv 0.500 VI Converter Power Supply Effect VIcrp 0.500 VI Converter Temperature Effect VIcrt 0.200 VI Converter Humidity Effect VIcrh 0

VI Converter Radiation Effect VIcrr 0

VI Converter Seismic Effect VIcrs 0

Section 8.2.4, Control Room Indicator Uncertainties 1 Per Annex G of ISA-RP67.04.02-2000, "Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation," lead wire resistance changes from environmental temperature variations should be considered for the two-and three-wire Resistance Temperature Detectors (RTDs). This effect is a bias. Table 18.4 of NUREG/CR-5560, Aging of Nuclear Plant Resistance Temperature Detectors (ML062540186), does not specify a direction for this error. Therefore, this bias error will be treated as a double sided (+/-) bias and applied to the Total Loop Uncertainty accordingly.

Parameter Abbreviation Uncertainty (+/- %Span)

Indicator Accuracy Ia 0

Indicator Drift Id 1.525 Indicator M&TE Im 0

Indicator Setting Tolerance Iv 2.000 Indicator Power Supply Effect Ip 0

Indicator Temperature Effect It 0

Indicator Humidity Effect Ih 0

Indicator Radiation Effect Ir 0

Indicator Seismic Effect Is 0

Indicator Readability Irea 0

Section 8.3.1, Total Loop Error - Control Room Indicator Leg (TLECRI)

The TLECRI calculation uses the TLE equation given in section 3.3.1 of DG-101, and as Equation 1 in section 3.1.1 of this SE, and substitutes the uncertainties applicable to the control room indicators that are listed in sections 8.2.1 through 8.2.4, as shown above.

This specified TLECRI is considered 2 sigma (95/95) and the result from this calculation is:

TLECRI = +/- 2.939 %Span +/- 0.2086 %Span (bias)

(95/95)

Converting from 95/95 to 75/75, of the random and independent errors, using the equation given in Section 3.3.3.13 of DG-101 and as Equation 2 in section 3.1.1 of this SE, as shown below.

The TLECRI (95/95) is converted to TLECRI (75/75) by the equation below:

TLECRI = +/- 2.939 * (1.15 / 1.96) = 1.724 %Span (75/75)

Converting from % of Span to process units and rounding to procedure precision:

The range of the control room indicator is 50oF - 350oF for a span of 300oF (as stated in Procedures P.1 through P.4 in Attachment A of this calculation).

TLECRI = +/- 1.724 %Span * (300ºF/100%Span) = +/- 5.172 ºF (Random)

TLECRI = +/- 0.2086 %Span * (300ºF/100%Span) = +/- 0.626 ºF (Bias)

TLECRI = (+/- 5.172 +/- 0.626) ºF TLECRI = +/- 5.80ºF 3.1.3 Overview of Calculation 2006-0035, Revision 1 The NRC staff reviewed Calculation 2006-0035, Revision 1, which is based on the methodology of DG-101, to verify whether these calculations used a methodology that is consistent with an NRC-approved methodology to calculate the instrument uncertainties for TS 3.6.5 containment average air temperature limits.

Calculation 2006-0035, Revision 1, declares in the following sections that:

Section 8.1 - Average Uncertainty for Three Indicators This section states that the determination of the individual containment temperature indication channel uncertainty for temperature loop (TLECRI) on both units was in Calculation 2005-0028 (the summary of this calculation and its results are as shown in Section 3.1.2 above).

The licensee calculated the uncertainty in average temperature (Tave) that obtained from n similar independent indicators by using equations in NUREG/CR-3659 (found in the section titled Development of the Uncertainty Method).

The NRC staff reviewed NUREG/CR-3659 and recognized that the equations to calculate the uncertainty associated with any one similar independent temperature sensors on page 14, that are as shown below:

Tave = (T1 + T2 + T3 + Tn) / n The uncertainty of Tave (UTave) will be:

UTave = UTi / n1/2 (Equation 3)

Where UTi is the uncertainty associated with any one of the similar temperature sensors.

From Equation 3 above, the overall uncertainty of Tave was obtained by dividing the random portion of the individual uncertainty by the square root of the number of indicators used to obtain the average.

For the uncertainty of an individual temperature indication, this section states, The uncertainty for individual indicator is found by correcting for the readability of the individual control board indicators 1 (2) TI-3292, -3293, and -3295 that perform TS surveillance SR 3.6.5.1, Verify containment average air temperature is within limit.

In Attachment A to DG-101, the table of Unit 1, Containment Temperature Indicator 1TI-3292/3293, shows indicator readability is 1/2 of a minor scale division of 5°F (2.5ºF).

Therefore, the uncertainty is conservatively rounded up to the nearest readable value, which is 7.5ºF.

Sections 8.1.1, Average Uncertainty for Three Indicators, and Section 8.1.2, Average Uncertainty for Two Indicators The random portion of the average uncertainty UTave of three indicators and two indicators are calculated by using Equation 3 above and combined with the bias term.

The results are as shown below:

Number of Indicators Random Portion of UTave (% of Span)

Combined with the bias term and converting from % of Span to ºF Three Indicators 1.724% / 31/2 = 0.995%

300 ºF * (0.00995 + 0.002086) = 3.7 ºF Two Indicators 1.724% / 21/2 = 1.219%

300 ºF * (0.01219 + 0.002086) = 4.3 ºF One Indicator Per Equation 3 discussion above 7.5ºF Based on the results in the table above, when identical indication loops are averaged, the average uncertainties of two indicators and of three indicators are different, and they are less than the individual loop uncertainty. Therefore, the number of available channels used to determine the containment average air temperature impacts the magnitude of measurement uncertainty.

Section 8.1.3, Technical Specification, states that chapter 14 of the Point Beach FSAR assumes that containment average air temperature is at or below 120°F to support the plant accident analyses. Therefore, with the maximum allowable containment average air temperature limit of 120°F, the three average air temperature channels should be:

Number of Air Temperature Channels (Indicators)

The average air Temperature TS Containment Average Air Temperature requirements Three Channels 120 ºF - 3.7 ºF = 116.3 ºF 116.3 ºF Two Channels 120 ºF - 4.3 ºF = 116.3 ºF 115.7 ºF Single Channel 120 ºF - 7.5 ºF = 112.5 ºF 112.5 ºF Based on the description and discussion above, the NRC staff recognizes that the methodology to calculate the containment temperature indication channel uncertainty for temperature loop (TLECRI) in Calculations 2005-0028 and 2006-0035 are based on the Instrument Setpoint Methodology (DG-101), which is consistent with the setpoint methodology in ISA 67.04.01-2018.

3.1.3 Conclusion Regarding Temperature Measurement Instrument Uncertainty Methodology Based on the discussion and description above, the NRC staff notes the following related to the methodology of Calculations 2005-0028 and 2006-0035:

Calculations 2005-0028 and 2006-0035 are based on the methodology of DG-101 to calculate the instrument total loop uncertainties of the Containment Average Air Temperature limits. The NRC staff confirms that the methodology in DG-101 is consistent with Section 4.5, Combination of uncertainties, Section 4.5.3, Formulas and Methodology Discussion, and Section 3.9, Error, of ISA 67.04.01-2018. In addition, this methodology is based on sufficiently large samples of the probability value or confidence levels that meets the two-standard deviation (2 Sigma) and 1 Sigma criteria.

Calculations 2005-0028 and 2006-0035 are based on NUREG/CR-3659 to calculate the uncertainty in average temperature that obtained from n similar independent indicators.

Calculations 2005-0028 and 2006-0035 demonstrate that (1) the number of available channels used to determine the containment average air temperature impacts the magnitude of measurement uncertainty and (2) by adding in the conservative direction the calculated uncertainty for a given number of available channels to the maximum allowable containment average air temperature limit of 120°F, the containment average air temperature assumed as an initial condition in plant safety analyses is preserved.

The NRC finds that the methodology used in Calculations 2005-0028 and 2006-0035 is consistent with the NRC-approved methodology in ISA 67.04.01-2018. In addition, these calculations conform to the NRCs regulatory guidance found in NUREG/CR-3659, for ensuring that the setpoints for safety-related instrumentation are initially within and remain within the TS limits. Therefore, the licensees methodology of the TLECRI and UTAVE calculations is an acceptable method.

Based on the above evaluations, the NRC staff finds that the licensee's proposed changes to TS 3.6.5 values incorporate measurement uncertainty to ensure that the operating parameters are maintained within the limits assumed as initial conditions in the accident analysis, consistent with Point Beach GDC 12, and the TS operating limits continue to meet Criterion 2 of 10 CFR 50.36(c)(2)(ii).

3.2 Evaluation of Change to TS 3.6.5 In the LAR, the licensee states, in part:

TS 3.6.5 was modified to impose voluntarily limits on the containment average air temperature measurement which account for the magnitude of instrument uncertainty based on the available number of measurement channels. NextEra has since found the added elements of TS 3.6.5 to be overly restrictive and unnecessary for TS inclusion.

The licensee proposes to change TS LCO 3.6.5 containment average air temperature limit to 120°F. As discussed in Point Beach Updated Final Safety Analysis Report (UFSAR) Section 14, Safety Analysis (ML24116A065), the initial containment average air temperature assumed in the accident analysis is 120°F. The results of the analyses demonstrate that the calculated transient containment air temperature is acceptable for the DBA LOCA and steam line break (SLB). As discussed in NUREG-1431, Bases for LCO 3.6.5, the containment average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a LOCA or SLB. Further, the NUREG-1431 limit for LCO 3.6.5 is 120°F.

The NRC staff reviewed the proposed change provided in LAR Attachment 1, Point Beach Technical Specifications Page Markups, and described in section 1.2 of this SE. Based on the above technical evaluation, the NRC staff finds the proposed change in the LCO for TS 3.6.5 from its current presentation to its analytical limit of 120°F is consistent with the analyses and evaluation included in the Point Beach UFSAR and the guidance contained in the Westinghouse STS NUREG-1431, and is therefore acceptable. The NRC staff finds that the requirements of 10 CFR 50.36(c)(2) will continue to be met because the LCO for TS 3.6.5 will continue to specify the lowest functional capability or performance levels of equipment required for safe operation of the facility.

3.3 Evaluation of Effect of TS 3.6.5 Surveillance Criteria For containment air temperature, the licensee proposes to use the analytical value of 120°F as the sole TS LCO limiting value without incorporating instrument uncertainty into the TS value itself. Currently, the TS 3.6.5 limits apply the calculated instrument uncertainty in the conservative direction to the containment average air temperature of 120°F assumed in plant safety analyses. These values are then rounded in the conservative direction to a value which allows the operators to accurately read the containment temperature instrument indicator in the control room. The proposed change explains the basis for measuring containment average air temperature that accounts for instrument uncertainty.

The NRC staff requested clarification on the relationship between the containment temperature analytical limit, the proposed additions to the TS Bases discussion on instrument uncertainty, and the TS surveillance criteria, specifically for TS SR 3.6.5.1, which verifies containment air temperature is within limit. In the August 9, 2024, supplement, the licensee stated:

The uncertainty calculations discussed in the TS Bases for SR 3.6.5.1 are implemented in Point Beach Operations Daily Logsheets, PBF-2034 and PBF-2035, for Units 1 and 2, respectively. These Logsheets direct plant operators to document the average containment air temperature every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, as required by the Point Beach Surveillance Frequency Control Program (SFCP),

using containment temperature indicators, T-3292, T-3293, and T-3295.

As stated above, the licensee proposes to control the instrument loop uncertainty using station operating procedures and surveillance acceptance criteria that reflects the instrument inaccuracies of the containment temperature measuring and testing equipment. This surveillance acceptance criteria will be less than the 120°F analytical limit by the amount calculated as described in the TS Bases. If the containment average temperature calculated from the surveillance test is greater than the acceptance criteria, operators will comply with the required actions of TS LCO 3.6.5, ACTION A to restore containment temperature within limits in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or, failing that, shutdown the reactor. Taking this action ensures compliance with the analytical limit of 120°F maximum initial containment temperature assumed in the safety analyses when accounting for instrument loop uncertainties.

This LAR also proposes to eliminate the current TS 3.6.5c authorization to determine the containment average air temperature employing a single temperature measurement instrument.

The NRC staff requested clarification on any criteria, justification, or limitations placed on instrument selection to verify that the arithmetic average containment air temperature would remain below those values used as initial conditions in accident analyses.

In the August 9, 2024, supplement, the licensee confirmed that the instruments are classified in accordance with RG 1.97, Revision 2, as:

Type D - which applies to variables providing information to indicate the operation of individual safety systems and other systems important to safety, and Category 2 - which applies to instrumentation designated for indicating system operating status.

The corresponding TS Bases for TS LCO 3.6.5 explains that the containment average air temperature is measured using the arithmetic average of all the available temperature instruments. A single temperature instrument would not be sufficient because an average cannot be obtained from a single measurement. The licensee also stated that the temperature indicators are repaired or replaced to the existing design specification in accordance with the maintenance rule provisions of 10 CFR 50.65 and any changes to the temperature indication design are subject to the regulatory change provisions of 10 CFR 50.59.

The NRC staff reviewed the licensees evaluation provided in section 3.0 of the LAR and the information in the August 9, 2024, supplement. Based on the above technical evaluation, the NRC staff finds the proposed change in the LCO for TS 3.6.5 from its current presentation to its analytical limit of 120°F is consistent with the analyses and evaluation included in the Point Beach UFSAR and the guidance contained in the Westinghouse STS, NUREG-1431, and is therefore acceptable. The NRC staff finds that the requirements of 10 CFR 50.36(c)(3), and Point Beach GDCs 10 and 12 will continue to be met.

3.4 Regulatory Conclusions Based on the above evaluations, the NRC staff finds that the licensees proposed changes to the LCO for TS 3.6.5 continue to ensure that the operating parameters are maintained within the limits assumed as initial conditions in the accident analysis, consistent with Point Beach GDCs 10 and 12, and the TS operating limits continue to meet 10 CFR 50.36(c)(2) and 10 CFR 50.36(c)(3). Therefore, the NRC staff concludes that the proposed changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, on October 17, 2024, the Wisconsin State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 or change surveillance requirements. 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 published a proposed finding in the Federal Register on August 6, 2024 (89 FR 63992) that the amendments involve no significant hazards consideration, and there has been no public comment on such finding.

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 Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: H. Vu, NRR M. Hamm, NRR C. Ashley, NRR B. Lee, NRR H. Wagage, NRR J. Ambrosini, NRR A. Sallman, NRR Date of Issuance: November 21, 2024

ML24303A340 NRR-058 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DEX/EICB/BC NAME SWall SRohrer FSacko DATE 10/29/2024 10/30/2024 10/30/2024 OFFICE NRR/DSS/STSB/BC NRR/DSS/SCPB/BC NRR/DSS/SNSB/BC NAME SMehta MValentin PSahd DATE 11/01/2024 10/31/2024 11/09/2024 OFFICE OGC - NLO NRR/DORL/LPL3/BC (A)

NRR/DORL/LPL3/PM NAME PLom RKuntz SWall DATE 11/19/2024 11/19/2024 11/21/2024