ML18345A110

Redacted Version, Issuance of Amendment Nos. 263 and 266 for Risk-Informed Approach to Resolve Construction Truss Design Code Nonconformances

ML18345A110
Person / Time
Site:	Point Beach
Issue date:	03/26/2019
From:	Mahesh Chawla Plant Licensing Branch III
To:	Nazar M Point Beach
Haskell R
References
EPID L-2017-LLA-0209
Download: ML18345A110 (78)

Text

OFFIOIA:L USE ONLY SECURln&a.RELATEB INFORM:ATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0.C. 20555-0001 March 26, 2019 Mr. Mano Nazar President and Chief Nuclear Officer Nuclear Division NextEra Energy Point Beach, LLC Mail Stop: EX/JB 700 Universe Boulevard Juno Beach, FL 33408

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS FOR RISK-INFORMED APPROACH TO RESOLVE CONSTRUCTION TRUSS DESIGN CODE NONCONFORMANCES (EPID L-2017-LLA-0209)

Dear Mr. Nazar:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 263 and 266 to Renewed Facility Operating License Nos. DPR-24 and DPR-27 for the Point Beach Nuclear Plant (Point Beach), Units 1 and 2, respectively. The amendments revise the Point Beach, Units 1 and 2 operating licenses by adding a license condition to resolve construction truss design code nonconformances in response to your application dated March 31, 2017, as supplemented by letters dated April 12, 2018 (two),

-May 29, 2018, August 30, 2018, and March 13, 2019.

These amendments approve a risk-informed (RI) approach to resolve legacy design code nonconformances associated with construction trusses in the containment buildings of Point Beach, Units 1 and 2, using an RI approach, following the guidance in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, issued May 2011.

The NRC has determined that the related safety evaluation (SE) provided in Enclosure 3 contains security-related information pursuant to Title 10 of the Code of Federal Regulations, Section 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, the NRC has also prepared a non-security-related version of the SE, which is provided in Enclosure 4.

Enclosure 3 to this letter contains security-related information. When separated from Enclosure 3, this letter is DECONTROLLED.

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OFFICIAL USE ONLY SECURITV-RELATED INFORMATION M. Nazar The Notice of Issuance will be included in the Commission's biweekly Federal Register Notice.

Sincerely, Mahesh L. Chawla, Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosures:

1. Amendment No. 263 to DPR-24

2. Amendment No. 266 to DPR-27

3. Safety Evaluation (Security-Related)

4. Safety Evaluation (non-Security-Related) cc without Enclosure 3: ListServ OFFICIAL USE ONLY - SECURITV-RELA'fEf> INfOfitMA'flON

ENCLOSURE 1 AMENDMENT NO. 263 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-266

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 263 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 March 31, 2017, as supplemented by letters dated April 12, 2018 (two), May 29, 2018, August 30, 2018, and March 13, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and Paragraph 4.1 of Renewed Facility Operating License No. DPR-24 is hereby amended to read as follows:

I. Containment Building Construction Truss NextEra Energy Point Beach shall complete implementation items 1, 2, 3, 5, and 6 included in Attachment 3 of licensee letter NRC 2019-0007, dated March 13, 2019, by the completion date specified in the Attachment.

3. This license amendment is effective as of the date of issuance and shall be implemented in accordance with the requirements specified in Paragraph 4.1 of Renewed Facility Operating License No. DPR-24.

FOR THE NUCLEAR REGULATORY COMMISSION

{)J9 ¥ ___-

David J. Wrona, Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License Date of issuance: March 26, 201 9

ATTACHMENT TO LICENSE AMENDMENT NO. 263 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 POINT BEACH NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-266 Replace the following pages of Renewed Facility Operating License No. DPR-24 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 DPR-24 REMOVE INSERT G. Secondary Water Chemistry Monitoring Program NextEra Energy Point Beach shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation. This program shall include:

1. Identification of a sampling schedule for the critical parameters and control points for these parameters;

2. Identification of the procedures used to quantify parameters that are critical to control points; 3 Identification of process sampling points;

4. Procedure for the recording and management of data;

5. Procedures defining corrective actions for off control point chemistry condition; and

6. A procedure for identifying the authority responsible for the interpretation of the data, and the sequence and timing of administrative events required to initiate corrective action.

H. The licensee is authorized to repair Unit 1 steam generators by replacement of major components. Repairs shall be conducted in accordance with the licensee's commitments identified in the Commission approved Point Beach Nuclear Plant Unit No. 1 Steam Generator Repair Report, dated August 9, 1982 and revised March 1, 1983 and additional commitments identified in the staff's related safety evaluation.

I. Containment Building Construction Truss NextEra Energy Point Beach shall complete implementation items 1, 2, 3, 5, and 6 included in Attachment 3 of licensee letter NRC 2019-0007, dated March 13, 2019, by the completion date specified in the Attachment.

J. Deleted K. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule Withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.

L. Mitigation Strategy Strategies shall be developed and maintained for addressing large fires and explosions that include the following key areas:

Renewed License No. DPR-24 Amendment No. 263

1. Fire fighting response strategy with the following elements:

a. Pre-defined coordinated fire response strategy and guidance

b. Assessment of mutual aid fire fighting assets

c. Designated staging areas for equipment and materials

d. Command and control

e. Training of response personnel

2. Operations to mitigate fuel damage considering the following:

a. Protection and use of personnel assets

b. Communications

c. Minimizing fire spread

d. Procedures for implementing integrated fire response strategy

e. Identification of readily-available pre-staged equipment

f. Training on integrated fire response strategy

g. Spent fuel pool mitigation measures

3. Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders M. Additional Conditions The additional conditions contained in Appendix C, as revised through Amendment No. 241, are hereby incorporated into this license. NextEra Energy Point Beach shall operate the facility in accordance with the additional conditions.

5. The issuance of this renewed operating license is without prejudice to subsequent licensing action which may be taken by the Commission with regard to the ongoing rulemaking hearing on the Interim Acceptance Criteria for Emergency Core Cooling Systems (Docket No. RM 50-1 ).

6. This renewed operating license is effective as of the date of issuance, and shall expire at midnight on October 5, 2030.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By R. W. Borchardt, Deputy Director Office of Nuclear Reactor Regulation Attachments:

1. Appendix A - Technical Specifications

2. Appendix B - Environmental Technical Specifications

3. Appendix C - Additional Conditions Date of Issuance: December 22, 2005 Renewed License No. DPR-24 Amendment No. 263

ENCLOSURE 2 AMENDMENT NO. 266 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-301

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 266 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 March 31, 2017, as supplemented by letters dated April 12, 2018 (two), May 29, 2018, August 30, 2018, and March 13, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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

H. Containment Building Construction Truss NextEra Energy Point Beach shall complete implementation items 3, 4, 5, and 6 included in Attachment 3 of licensee letter NRC 2019-0007, dated March 13, 2019, by the completion date specified in the Attachment.

3. This license amendment is effective as of the date of issuance and shall be implemented in accordance with the requirements specified in Paragraph 4.H of Renewed Facility Operating License No. DPR-27.

FOR THE NUCLEAR REGULATORY COMMISSION Q~na.2ef q/__-

Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License Date of issuance: March 2 6, 2 O1 9

ATTACHMENT TO LICENSE AMENDMENT NO. 266 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 POINT BEACH NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-301 Replace the following page of Renewed Facility Operating License No. DPR-27 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Renewed Facility Operating License DPR-27 REMOVE INSERT H. Containment Building Construction Truss NextEra Energy Point Beach shall complete implementation items 3, 4, 5, and 6 included in Attachment 3 of licensee letter NRC 2019-0007, dated March 13, 2019, by the completion date specified in the Attachment.

I. Deleted J. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.

K. Mitigation Strategy Strategies shall be developed and maintained for addressing large fires and explosions that include the following key areas:

1. Fire fighting response strategy with the following elements :

a. Pre-defined coordinated fire response strategy and guidance

b. Assessment of mutual aid fire fighting assets

c. Designated staging areas for equipment and materials

d. Command and control

e. Training of response personnel

2. Operations to mitigate fuel damage considering the following:

a. Protection and use of personnel assets

b. Communications

c. Minimizing fire spread

d. Procedures for implementing integrated fire response strategy

e. Identification of readily-available pre-staged equipment

f. Training on integrated fire response strategy

g. Spent fuel pool mitigation measures

3. Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders Renewed License No. DPR-27 Amendment No. 266

ENCLOSURE 4 (NON-SECURITY-RELATED)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 263 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AND AMENDMENT NO. 266 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301 Security-related information pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations has been redacted from this document.

Redacted information is identified by blank space enclosed within double brackets (( )).

OFFICIAL USE ONLY SECURITY RELATED INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 263 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AND AMENDMENT NO. 266 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301

1.0 INTRODUCTION

By letter dated March 31, 2017 (Reference 1), as supplemented by letter NRC 2018-0014, dated April 12, 2018 (Reference 2; hereafter referred to as NRC 2018-0014), letter NRC 2018-0001, dated April 12, 2018 (References 3and 4), letter NRC 2018-0030, dated May 29, 2018 (Reference 5), letter NRC 2018-0041, dated August 30, 2018 (Reference 6), and letter NRC 2019-0007, dated March 13, 2019 (Reference 7), NextEra Energy Point Beach, LLC (NextEra, the licensee) submitted License Amendment Request (LAR) 278, "Risk-Informed Approach to Resolve Construction Truss Design Code Nonconformances," to the U.S. Nuclear Regulatory Commission (NRC) for review and approval. The LAR sought a risk-informed (RI) resolution for legacy design code nonconformances associated with construction trusses 1 in the containment buildings of Point Beach Nuclear Plant (Point Beach or PBNP), Units 1 and 2. The enclosure to letter NRC 2018-0001, dated April 12, 2018 (Reference 4), contains sensitive unclassified non-safeguards information and has been withheld from public disclosure pursuant to Section 2.390, "Public inspections, exemptions, requests for withholding," of Title 10 of the Code of Federal Regulations (10 CFR).

The supplemental letters dated April 12, 2018 (two), May 29, 2018, August 30, 2018, and March 13, 2019, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on June 19, 2017 (82 FR 27890).

1 The terms construction dome trusses, construction trusses, and trusses are used interchangeably in this safety evaluation (SE) and refer to the same structure.

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2.0 REGULATORY EVALUATION

2.1 Description of System The construction trusses in each unit were originally installed to provide support for the containment dome liner and initial dome concrete pour during original construction. After the initial concrete pour cured, the trusses were lowered a few inches away from the containment liner, no longer providing structural support to the dome, and remained in place. The licensee then used the truss as an attachment point for containment spray piping, containment recirculation cooling system (VNCC) ductwork, post-accident containment ventilation piping, and miscellaneous lights and associated conduits. An initial analysis of seismic adequacy was performed by the construction vendor.

The licensee discovered a discrepancy between the as-built configuration of the trusses and the design drawing used for the analysis during walkdowns as well as reviews of plant photos.

Specifically, the lower diagonal bracing framework of the trusses, and the bottom lower diagonal bracing location on the trusses, were found to be different from the design drawing. The licensee performed subsequent refinements of the analysis for the trusses, which resulted in identifying nonconformances to the design code of record for postulated seismic loads.

Follow-on inspection of the trusses identified additional nonconformances regarding the available clearance between the trusses and the containment liner at a limited number of locations in each unit.

The identified nonconformances are for the postulated design-basis seismic loading event and postulated design-basis accident thermal loading. The nonconformances for seismic loading were for the Units 1 and 2 construction trusses, the attached containment spray ring header piping and supports, and the containment liner. The nonconformances for thermal loading were for the Units 1 and 2 construction trusses as well as the containment liner and wall. The nonconformances were being tracked in the site's Corrective Action Program. The NRC staff inspection related to the containment truss is documented in the PBNP, Units 1 and 2 NRC Integrated Inspection Report 05000266/2014004; 05000301/2014004; and 07200005/2014001, dated October 30, 2014 (Reference 8).

The licensee's March 31, 2017, submittal proposed a resolution of the nonconformances using an RI approach, following the guidance in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, issued May 2011 (Reference 9), for the above-described legacy design code nonconformances associated with construction trusses of both units. 2 The licensee performed engineering evaluations and risk assessments to support the submittal. The engineering evaluations used evaluation methods and acceptance criteria that differed from those in the licensing basis {LB). The revised LB subsequent to the completion of the 2 RG 1.174, Revision 3 (Reference 10), was issued in January 2018 during the NRC staffs review of this application. The licensee stated that RG 1.174, Revision 2, was used as the guidance for its application. The NRC staff notes that RG 1.174, Revision 3, expands the guidance on the topic of defense-in-depth and discussion of uncertainties as compared to RG 1.174, Revision 2. Those changes add clarity and do not change the underlying guidance as compared to Revision 2. RG 1.174, Revision 3, also adds discussion on the use of the guide for new reactors which is irrelevant to this application. Therefore, the NRC staff notes that the consideration of the changes in RG 1.174, Revision 3, neither impacts this application nor changes the NRC staffs review or conclusions documented here.

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OFFICIAL USE ONLY SECURITY RELATED INFORMATION implementation items in the license condition proposed by the licensee will include evaluation methods and acceptance criteria used in the engineering evaluations supporting this application.

2.2 Description of Licensee's Proposed Change The licensee requested NRC acceptance of the proposed modified configuration of the Unit 1 construction truss and associated equipment, and the current configuration of the Unit 2 construction truss and associated equipment, based on the RI approach described in the submittal and the supplements.

In support of the request, the licensee also proposed the following implementation items in Attachment 3 of licensee letter NRC 2019-0007, dated March 13, 2019:

Item 1 Upon approval of LAR 278 (Amd. No. 263), a modification will be implemented to the Unit 1 construction truss to improve clearance between the truss and the containment liner. The modification includes trimming the top chord at the first panel point of six trusses, specifically trusses 1, 2, 3, 7, 8, and 15, as identified in NextEra Energy Point Beach, LLC letter NRC 2018-0014, "Construction Truss License Amendment Request 278, Response to Request for Additional Information," dated April 12, 2018 (ML181028164).

The spatial clearance modification requires approval of LAR 278 (Amd. No. 263) prior to implementation as the small amount of material removal modifies the sectional properties of the construction truss, which impacts the available seismic margin.

Item 2 Unit 1 containment spray pipe support Sl-301 R-1-H202 will be modified to achieve additional seismic capacity. The modification will increase the size of the support's U-bolt diameter.

Item 3 NextEra will implement new seismic operating limits applicable to Unit 1 and Unit 2 to maintain stresses in the construction trusses within elastic stress limits.

The applicable bounding limits are:

Horizontal: 0.05g Peak Ground Acceleration Vertical: 0.04g Peak Ground Acceleration The existing seismic monitors will be used to detect the new operating seismic limits.

Site procedures will be revised to initiate actions to commence a controlled dual Unit backdown to hot shutdown upon reaching the specified peak ground OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION acceleration limits. The procedure will specify that the Units may be reduced in power individually in series, rather than concurrently, to ensure station staff have the proper resources and focus on each Unit for maximum safety and operational excellence during this off-normal operation. The procedure will also initiate inspection and/ or evaluation actions for the construction trusses, equipment supported by the trusses, and the containment liners.

Item 4 NextEra will implement a new modification to reroute the control cables for Unit 2 pressurizer power operated relief valve (PORV) 2RC-431 C and associated block valve 2RC-515 to ensure they are protected from a postulated falling object.

Item 5 NextEra will implement new thermal operating limits applicable to Unit 1 and Unit 2 construction trusses and attached components for any thermal excursion or occurrence resulting in exceeding elastic stress limits. The proposed new operational containment temperature limits are:

Unit 1: 227° F containment atmospheric temperature Unit 2: 236° F containment atmospheric temperature The station implementation of these new limits will include taking the affected Unit offline and performing an inspection and/ or evaluation, as necessary, to confirm that the construction truss' future stability during a postulated design basis accident was not compromised by the thermal occurrence.

Detection of the thermal occurrence and initiation of actions to address the new limiting values will be addressed in site procedures. The site procedures will be revised to initiate inspection and/or evaluation actions prior to Unit startup for the affected Unit's construction truss, equipment supported by the truss (as necessary), and. the containment liner. No new instrumentation is required to support this change. Existing station instrumentation will be utilized for containment temperature identification.

Item 6 NextEra will revise the Point Beach Units 1 and 2 UFSAR to include the evaluation methods and acceptance criteria identified in the Table in response to RAl-1.a, in NextEra letter NRC 2018-0014, dated April 12, 2018. NextEra will revise the Point Beach Units 1 and 2 UFSAR to reclassify the construction truss as stated in its response to RAl-17 in NextEra letter NRC 2018-0014. of the LAR included discussion of a modification for Units 1 and 2 to install a 24-hour backup pneumatic supply to the pressurizer power operated relief valves (PORVs) and route the supply tubing and electrical cables for the PORVs such that they are protected from a postulated falling object. Those modifications were part of the implementation of the licensee's approved use of the National Fire Protection Association (NFPA) Standard 805 as the LB for its OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION fire protection program. The NRC staff approved the licensee's transition to NFPA 805 by letter dated September 8, 2016 (Reference 11 ). In response to RAl-11.e in the letter NRC 2018-0014, and in the letter dated August 30, 2018, the licensee stated that those modifications were complete in both units.

The licensee performed engineering evaluations and risk assessments to support the request.

The engineering evaluations were performed to determine the behavior of the trusses as well as attached or adjacent structures, systems, and components (SSCs) under seismic and thermal loading conditions. The engineering evaluations were used (1) to determine the structural integrity at design basis seismic and thermal loading conditions, (2) to identify the need for the proposed modifications listed above, and (3) to determine the limiting truss members3 to determine the fragilities which would be used in the risk assessment. The engineering evaluations used evaluation methods and acceptance criteria that differed from those in the LB.

The evaluation methods and acceptance criteria used for the engineering evaluations in support of the request were listed in the response to RAl-1 in Enclosure 1 to letter dated April 12, 2018 (Reference 3). The licensee, in the enclosure, also proposed the evaluation methods and acceptance criteria used to support the request as part of the revised LB.

Further, in response to RAl-17, the licensee proposed classifying the trusses as 'Seismic Class Ill' structures.

2.3 Description of Regulatory Requirements Technical Specification (TS) 3.6.1, "Containment," addresses operability and surveillance requirements for containment structures. The containment liner is part of the containment structure. TS 3.6.6, "Containment Spray and Cooling Systems," addresses operability and surveillance requirements for the containment spray and containment cooling systems.

The principal criteria and safety objectives for the design of PBNP are defined in the plant's Updated Final Safety Analysis Report (UFSAR; Reference 12). The PBNP principal design criteria in Section 1.3 of the UFSAR pre-date the General Design Criteria (GDC) published in 1971 that are in Appendix A to 10 CFR Part 50. The PBNP principal criteria from the UFSAR relevant to this application are as follow:

• GDC 2, "Performance Standards" (Section 1.3 of the UFSAR), defines the performance standard for systems and components to withstand, without undue risk to the health and safety of the public, natural phenomena.

• GDC 10, "Reactor Containment" (Section 1.3 of the UFSAR), defines the principal criteria and safety objectives for the design of the containment.

• GDC 41, "Engineered Safety Features Performance Capability" (Section 1.3 of the UFSAR), defines the performance capability for engineered safety features such as the containment heat removal system.

3 Truss members or truss components refer to individual structural constituents which collectively form the truss.

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• GDC 49, "Reactor Containment Design Basis" (Section 1.3 of the UFSAR), defines the principal criteria and safety objectives for the design of the reactor containment structure, including openings and penetrations, and any necessary containment heat removal systems.

• GDC 52, "Containment Heat Removal System" (Section 1.3 of the UFSAR), defines the performance capability for active containment heat removal systems under accident conditions.

Section 1.4 of the plant's UFSAR states that the plant operational and support activities are conducted under NextEra Energy Quality Assurance Topical Report (QATR), FPL-1 and that the QATR satisfies the requirements of Appendix B of 10 CFR Part 50. The regulation under 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," requires, in part, that the design control measures include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled.

Design changes, including field changes, are to be subjected to design control measures commensurate with those applied to the original design.

PBNP Seismic Category I structures, other than containment, were designed in accordance with American Concrete Institute (ACI) 318-63, "Building Code Requirements for Reinforced Concrete," while the containment structures were designed in accordance with American Institute of Steel Construction (AISC) Steel Construction Manual, 5th Edition, April 1963.

2.4 Description of Regulatory Guidance Regulatory Guide (RG) 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

May 2011 (Reference 9), provides guidance for using risk information in support of voluntary (i.e., licensee-initiated) LB changes. The regulatory guide provides general guidance concerning one approach that the NRC has determined to be acceptable for analyzing issues associated with proposed changes to a plant's LB and for assessing the impact of such proposed changes on the risk associated with plant design and operation.

RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009 (Reference 13),

describes an acceptable approach for determining whether the probabilistic risk assessment (PRA), in total or the parts that are used to support an application, is acceptable for use in regulatory decisionmaking for light-water reactors.

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," Standard Review Plan (SRP) Section 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load," Revision 3, September 2012 (Reference 14), provides guidance to the NRC staff for evaluating the acceptability of a licensee's PRA results when used to request RI changes to the LB.

NUREG-0800, SRP Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance," Revision 0, June 2007 OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION (Reference 15), provides guidance to the NRC staff for evaluating risk information used by a licensee to support permanent RI changes to the LB.

NUREG-0800, SRP Section 3.8.4, "Other Seismic Category I Structures," Revision 4, September 2013 (Reference 16), provides guidance to the NRC staff on the review of areas relating to all Seismic Category I structures and other safety-related structures that may not be classified as Seismic Category I.

2.5 Description of Licensee's Proposed Changes to Operating Licenses The licensee proposed to amend the Point Beach Units 1 and 2 operating licenses by adding the following license conditions related to this application as Paragraph 4.1 and 4.H, respectively:

4.1 Containment Building Construction Truss NextEra Energy Point Beach shall complete implementation items 1, 2, 3, 5, and 6 included in Attachment 3 of licensee letter NRC 2019-0007, dated March 13, 2019, by the completion date specified in the Attachment.

4.H Containment Building Construction Truss NextEra Energy Point Beach shall complete implementation items 3, 4, 5, and 6 included in Attachment 3 of licensee letter NRC 2019-0007, dated March 13, 2019, by the completion date specified in the Attachment.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the licensee's proposed changes in accordance with SRP Section 19.2.

The scope of the NRC staff's review included determining whether the submittal meets the guidance in RG 1.174, Revision 2, and RG 1.200, Revision 2. RG 1.174, Revision 2, provides the framework for the staff to evaluate licensee-initiated LB change requests that use traditional engineering analyses as well as the RI approach. Further, the RG describes an acceptable method for the licensee and staff to use in assessing the nature and impact of LB changes when the licensee chooses to support the changes with risk information. For RI decisionmaking, RG 1.174, Revision 2, states that LB changes are expected to meet a set of key principles. Further, RG 1.174, Revision 2, states that risk analysis techniques can be, and are encouraged to be, used to help ensure and show that those principles are met.

The principles of RI decisionmaking as listed in RG 1.174, Revision 2, are:

• Principle 1: The proposed LB change meets the current regulations unless it is explicitly related to a requested exemption (i.e., a specific exemption under 10 CFR 50.12).

• Principle 2: The proposed LB change is consistent with the defense-in-depth philosophy.

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• Principle 3: The proposed LB change maintains sufficient safety margins.

• Principle 4: When proposed LB changes result in an increase in risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement (Reference 17).

• Principle 5: The impact of the proposed LB change should be monitored using performance measurement strategies.

The licensee addressed the five principles of RI decisionmaking as part of the submittal.

Further, the licensee used core damage frequency (CDF) and large early release frequency (LERF), which are the metrics in RG 1.174, Revision 2, to quantify the incremental risk from the proposed change and compared the same against the acceptance guidelines in RG 1.174, Revision 2. Therefore, RG 1.174, Revision 2, is directly applicable to the submittal and the NRC staff's review.

SRP Section 19.2, provides general guidance to the NRC staff on how the information from a PRA can be combined with other pertinent information in the process of making a regulatory decision. SRP Section 19.2 states that for cases where it is determined that an application could justify a review that is less than the full scope described in SRP Section 19.2, reviewers should choose the relevant and applicable parts of SRP Section 19.2, for guidance. According to SRP Section 19.2, the necessary sophistication of the review of the PRA, its supporting analyses, and its results depend on the contribution the risk assessment provides to the integrated decisionmaking.

The NRC staff's review to determine if the information presented in the submittal and supplements satisfies the principles of RI decisionmaking is presented in Sections 3.2 through 3.6 of this SE.

3.1 Engineering Evaluations Section 111.2 of SRP Section 19.2 states that in order to make findings regarding the acceptability of a proposed license amendment, the reviewers should use an integrated assessment of traditional engineering evaluations and probabilistic information. On the topic of engineering evaluations, RG 1.174, Revision 2, states that:

• The scope, level of detail, and technical adequacy of the engineering analyses conducted to justify any proposed LB change should be appropriate for the nature and scope of the proposed change.

• The licensee should appropriately consider uncertainty in the analysis and interpretation of findings.

• The licensee should use judgment on the complexity and difficulty of implementing the proposed LB change in deciding upon appropriate engineering analyses to support regulatory decisionmaking. Thus, the licensee should consider the appropriateness of qualitative and quantitative analyses, as well as OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION analyses using traditional engineering approaches and those techniques associated with the use of PRA findings.

The licensee performed engineering evaluations and risk assessments for seismic and thermal loading conditions to support the licensee's request. The NRC staff's review of the engineering evaluations is provided hereunder. Section 3.1.1 discusses the NRC staff's review of the approach and inputs used for the seismic evaluations and Section 3.1.2 does the same for the thermal evaluations. Sections 3.1.3 through 3.1. 7 discuss the NRC staff's review on the conclusions drawn by the license from the seismic and thermal engineering evaluations. The NRC staff's review of the risk assessments is discussed in Section 3.5 of this SE.

The NRC staff's findings related to the engineering evaluations, including the proposed approaches and the proposed acceptance criteria, are limited to this application and are not to be interpreted as being generic.

3.1.1 Approach and Inputs for Seismic Evaluations to Support RI Resolution The licensee provided details of the seismic evaluations performed to support this application in of the LAR. The NRC staff's review of the seismic evaluations was informed by regulatory audits (References 18 and 19) during which the NRC staff had access to the documents containing additional details of the evaluations.

The seismic evaluations determined the structural integrity of the trusses as well as the attached or adjacent SSCs under seismic loading conditions, and the seismic fragility of the trusses in the as-proposed configurations. Sections 5 and 6 of Enclosure 5 to the LAR provided details on the technical approach used by the licensee for various aspects of the seismic evaluation supporting this application.

The licensee used the site-specific ground motion response spectrum (GMRS) based on the site specific reevaluated hazard and the in-structure seismic demand based on soil structure interaction analysis (Reference 20). The NRC staff has previously reviewed the licensee's seismic hazard reevaluation and concluded that "the licensee conducted the seismic hazard reevaluation using present day methodologies and regulatory guidance, appropriately characterized the site given the information available, and met the intent of the guidance for determining the reevaluated seismic hazard" (Reference 21 ). The NRC staff verified that the mean seismic hazard curve as well as the different percentiles used to support this application were obtained from the seismic hazard reevaluation performed by the licensee in response to the Near-Term Task Force (NTTF} Recommendation 2.1. Because the same seismic hazard curves were used and a separate seismic hazard analysis is not necessary for this application, the NRC staff finds that a separate review of the seismic hazard curves used for this application is not warranted and the seismic hazard curves used by the licensee to support this application are acceptable.

The licensee performed a soil-structure interaction analysis for use in analyzing the trusses, containment spray lines, and attached components. The licensee stated that development of the ground motion time histories that match the GMRS, for use in the soil-structure interaction analysis, was performed according to American Society of Civil Engineers (ASCE)/Structural Engineering Institute (SEI) Standard 43-05, "Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities," with the limitations identified in NUREG/CR-6926, OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION "Evaluation of the Seismic Design Criteria in ASCE/SEI Standard 43-05 for Application to Nuclear Power Plants," March 2007 (Reference 22), and that the approach was consistent with SRP Section 3.7.1, "Seismic Design Parameters," December 2014 (Reference 23). The licensee explained that the in-structure response spectra was developed consistent with the guidance in SRP Section 3.7.2, "Seismic System Analysis," September 2013 (Reference 24),

and noted that the approach was chosen because NUREG/CR-6926 stated that some provisions of ASCE Standard 4-98, "Seismic Analysis of Safety-Related Nuclear Structures," do not agree with current NRC regulatory documents and staff positions. Section 6.2 of to the LAR provides additional details on the licensee's soil-structure interaction approach used for the seismic evaluations. The NRC staff reviewed the licensee's soil-structure interaction analysis approach and finds it to be acceptable for this application because ( 1) the approach was technically justified, and (2) it was performed consistent with the staff's guidance on performance of soil-structure interaction and the development of the in-structure response spectra.

The NRC staff's review noted that in Section 2 of Enclosure 5 of the LAR the licensee used 4 percent damping factor for the containment sprays and indicated that the damping factors used in the design of welded steel framed and bolted steel-framed structures were 2 percent and 5 percent, respectively. However, in. Section 5.3 of the same enclosure, the licensee stated that the damping value used in the engineering evaluations supporting this application was 7 percent for bolted steel with bearing connections and that the value was from ASCE/SEI Standard 43-05, "Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities," with the limitations identified in NUREG/CR 6926 and was consistent with SRP Sections 3.7.1 and 3.7.2. In RAl-20.a, the NRC staff requested the licensee to justify the use of 7 percent damping. In response to RAl-20.a, the licensee discussed the geometry and the connections of the truss structure. The licensee explained that the transfer of load between each of the truss members, and between the truss and the containment building, was through bolted connections. While each of the individual 18 truss members is a welded planar truss assembly, the transfer of load between the 18 truss members is through a bolted brace system.

Furthermore, the licensee stated that the 7 percent damping was used in accordance with Table 3-2 of ASCE/SEI 43-05 for Response Level 2, which was applicable to most of the highly stressed truss components. The licensee further explained that for a selected number of locations of the truss, a 10 percent damping could have been used in accordance with Table 3-2 of ASCE/SEI 43-05 instead of the 7 percent value that was selected. The NRC staff reviewed the licensee's use of the damping factor for the trusses and finds that the transfer of load between each of the truss members as well as between the truss and the containment building is through bolted connections and using the selected damping for those connections is appropriate for this application. The NRC staff also confirmed that Table 3-2 of ASCE/SEI 43-05 permits the use of 7 percent damping for bolted connections. The NRC staff also notes that a higher damping can be used in accordance with the ASCE/SEI 43-05 in some locations of the truss. Therefore, based on its review, the NRC staff finds the use of structural damping of 7 percent for truss structures per ASCE/SEI 43-05, Table 3-1, Response Level 2 and structural damping of 4 percent for containment spray piping attached to the truss structures per Table 3 of RG 1.61, Revision 1, "Damping Values for Seismic Design of Nuclear Power Plants," March 2007 (Reference 25), to be acceptable for the seismic evaluations supporting this application.

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OFFICIAL USE ONLY SECURITY RELATED INFORMATION Development of Seismic Fragility of the Trusses The licensee discussed the development of seismic fragility for the trusses in Section 6.4.2.1 of to the LAR. The high confidence of low probability of failure (HCLPF) capacity was determined by the licensee and used to compute the median fragility of the truss which was then used in the risk assessments supporting this application. The HCLPF for a SSC is the seismic acceleration at which there is 95 percent confidence that the failure probability of the SSC is 5 percent or less. Details of the Conservative Deterministic Failure Margin (CDFM) calculation used to determine the HCLPF capacity were provided in Section 5. 7 of Enclosure 5 of the LAR.

The licensee applied capacity adjustment factors, such as those for load redistribution and inelastic energy absorption, to the calculations of the equivalent peak ground acceleration (PGA), which the licensee termed "PGAc" in the submittal. The licensee further concluded that the PGA calculated from the equivalent static analysis was higher than that from the elastic analysis and used the higher PGA to derive the seismic fragility of the trusses used in the risk assessments supporting this application. The NRC staff requested the licensee to provide additional information about the capacity adjustment factors used by the licensee and to justify the use of the higher PGA. In response to RAl-16, the licensee clarified that two different sets of capacity adjustment factors were used to determine the HCLPF capacity, depending on the type of analysis. The licensee explained that HCLPF based on the elastic analysis was calculated considering the PGA at which the limiting truss member demand was equal to the capacity, as determined in accordance with the proposed acceptance criteria for stress. The licensee further explained that the HCLPF based on the equivalent static analysis used the proposed alternate strain-based acceptance criteria for the top and bottom chord members of the truss and that the analysis accounted for load redistribution and inelastic energy absorption.

The licensee stated that due to the approach used for determination of the HCLPF for the trusses the capacity factors used for the equivalent static analysis did not include any increases for load redistribution. Further, the capacity factors used for the equivalent static analysis included a reduced increase, compared to the elastic analysis, for inelastic energy absorption to reflect a limited number of members that have stress exceeding the proposed acceptance criteria. The licensee stated that the maximum strain determined in the equivalent static analysis would be below the proposed alternate strain-based acceptance criteria even if it was increased using the capacity factor for inelastic energy absorption in the analysis. Based on its review of the licensee's submittal and supplements, the NRC staff finds that the licensee's use of the HCLPF based on the equivalent static analysis, which has a higher PGA, is appropriate for this application because it provides a more realistic structural evaluation of the truss accounting for load redistribution and inelastic energy absorption. The NRC staff also finds that the licensee appropriately applied the capacity factor for the HCLPF calculation based on the equivalent static analysis because the capacity factor properly accounts for the assumption of inelastic behavior at a limited number of truss members.

The CDFM calculation performed by the licensee used the PGA corresponding to the mean seismic hazard curve. The CDFM approach, as described in Electric Power Research Institute (EPRI) report NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, August 1991 (Reference 26), indicated that the 84 percent non-exceedance hazard curve should be used. Therefore, in RAl-8.a, the NRC staff requested justification for the licensee's use of the PGA from the mean hazard curve for its HCLPF calculation supporting this application. In letter NRC 2018-0014, the licensee stated that OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION EPRI NP-6041-SL discussed different alternatives to the selection of the seismic margins earthquake. The licensee stated that the second alternative in that report discussed the use of the uniform hazard spectrum, which was a response spectrum anchored to a given PGA and a uniform annual frequency of exceedance, as the seismic margins earthquake. The licensee indicated that according to the report, the spectral shape corresponding to either the 84 percent non-exceedance or the mean uniform hazard spectrum could be used, depending on the objectives of the review. The licensee explained that the site-specific GMRS used in the seismic evaluation of the trusses followed the second alternative in the EPRI report and that the second alternative provided results in terms of annual exceedance frequency which was compatible with the RI approach in the submittal. The licensee further explained that the approach used to calculate the HCLPF for this application anchored the HCLPF to the PGA irrespective of the shape of the ground response spectra. Based on its review of the licensee's approach, the submittal, and the supplements, the NRC staff finds the licensee's approach to calculate the HCLPF to be acceptable for this application because it is consistent with the state-of-practice and with the RI approach that utilizes annual exceedance frequencies.

The licensee used the hybrid method discussed in EPRI report 1025287, "Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic" (Reference 27), to compute the median fragility of the truss based on the calculated HCLPF. The median fragility was then used in the risk assessments supporting this application. The licensee used the composite, randomness, and uncertainty variability values (also known as the lognormal standard deviations) from EPRI 1025287 to characterize the uncertainty in the fragility computation. The NRC staff reviewed the licensee's approach for determining the fragility of the trusses based on the hybrid method and finds that the use of the hybrid method to be appropriate for this application because the method provides sufficient realism in the fragility computation to support this application. In addition, based on its review, the NRC staff finds the chosen variability values to be acceptable for this application because the values are expected to be representative of the trusses which are passive structures mounted at a high elevation.

In summary, based on its review of the seismic evaluations from the submittal as well as the supplements, the NRC staff finds the approach and inputs for the seismic evaluations supporting this application to be acceptable for this application because ( 1) it uses re-evaluated site-specific hazard information which is based on present day methodologies, (2) it is technically justified and consistent with staff guidance on such analysis, and (3) it provides sufficiently realistic predictions of the truss and attached components during a seismic event.

The NRC staffs findings related to the methodology for the seismic evaluations are limited to this application and are not to be interpreted as being generic.

3.1.2 Approach and Inputs for Thermal Evaluations to Support RI Resolution The licensee provided details of the thermal evaluations performed to support this application in of the LAR. The NRC staffs review of the thermal evaluations was informed by regulatory audits (References 18 and 19) during which the NRC staff had access to the documents containing additional details of the evaluations.

The thermal evaluations determined the structural integrity of the trusses as well as the attached or adjacent SSCs under thermal loading conditions, and the thermal fragility of the trusses in the as-proposed configurations. The construction trusses were evaluated for thermal loads due to OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION differential thermal displacements between the trusses and the containment spray lines and between the trusses and the containment liner. The main steam line break (MSLB) inside containment was chosen as the limiting event for the thermal evaluation. Sections 4, 5.8, 6.4.1, and 6.5.1 of Enclosure 5 to the LAR provided details on the technical approach used by the licensee for the thermal evaluations supporting this application.

According to the response to RAl-19.a, the MSLB case chosen for the thermal evaluations was the same as that referenced in Chapter 14 of the Point Beach UFSAR (Reference 28). The NRC staff's review finds that the MSLB represents the limiting thermal loading challenge to the truss for this application because of the magnitude as well as the rate of temperature rise during that event compared to the other design-basis accidents as well as thermal initiators considered in the risk assessments supporting this application.

Development of Thermal Fragility of the Trusses According to Section 5.8 of Enclosure 5 to the LAR, the licensee used engineering judgment, supported by thermal evaluations, to assign a probability of failure of the truss at various temperatures (i.e., thermal fragility curve). Section 6.4.1 of the same enclosure provided details of the thermal fragility curves for the truss in Units 1 and 2. However, the licensee used a common thermal fragility curve for both units in the risk assessments supporting this application as stated in Section 2.2.1 of Enclosure 4 to the LAR and Section 2.3.1 of Enclosure 2 to the supplement to letter NRC 2018-0014.

The NRC staff requested details of the development and assignment of the thermal fragility curve as well as justification for use of a common curve for both units. In response to RAl-5.a, the licensee provided details on the rationale and supporting evaluations for assigning probability of failure of the truss at the various containment temperatures. The licensee stated that due to the lack of information about the failure probability of a structure whose components are subjected to stresses that result from thermal loading, guidance from literature on seismic loading was used to assign probability of failure of a structure subjected to thermal loads based on the component stress level. The licensee performed engineering analysis of the Unit 1 and Unit 2 trusses in the as-found configurations under thermal loading. The containment temperature at which contact between the truss and the containment steel liner initially occurred was assigned a very low failure probability. The licensee stated that the low failure probability was because of the lack of thermal loading from the restraint imposed by the containment wall, the negligible localized loads in the bottom chord of the truss due to differential expansion between the construction truss and the containment spray piping, and the truss carrying only its self-weight and the weight of attached components.

The licensee considered the condition when a fully plastic hinge would first develop to be analogous to the seismic capacity corresponding to 1 percent failure probability. To support such an analogy, the licensee cited EPRI Report 1025287 which stated that the 1 percent probability of failure could be determined using the CDFM approach in EPRI Report NP-6041-SL. The licensee explained that EPRI Report NP-6041-SL stated that plastic design methods may be used to determine the capacity without amplification to the applied loads. Therefore, the licensee assigned a failure probability of 0.01 when the containment reaches the temperature at which the first fully plastic hinge was developed. The licensee also explained that engineering analysis demonstrated that only one of the truss connections reached a fully plastic condition and the truss was demonstrated to maintain structural integrity OFFICIAL USE ONLY SECURITY RELMED INFORMMION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION for this condition. At the containment design temperature (286 °F), the engineering analysis for the Unit 1 truss showed plastic hinges forming at multiple locations and therefore, the licensee assigned a failure probability of 0.1 at that temperature for the Unit 1 truss. The licensee stated that the assignment was comparable to the methodology in ASCE/SEI 43-05, which defined 10 percent probability of unacceptable performance as the acceptance criteria for a ground motion equal to 150 percent of the design-basis earthquake ground motion.

For the Unit 2 truss, the engineering analysis for the Unit 2 truss did not show any plastic hinges at the containment design temperature. Therefore, the licensee assigned a failure probability of 0.001 at that temperature for the Unit 2 truss due to the margin before the development of full plastic hinges, the liner contact force being less than the liner capacity, and overall loading not challenging the integrity of the truss structure. The licensee assigned a failure probability of 0.99 when the containment reaches the temperature at which truss "instability or unacceptable force on the containment steel liner" would occur because that temperature represented either near collapse of the truss or exceedance of the maximum permissible strain of the containment steel liner.

In response to RAl-5.b, the licensee, citing relevant engineering evaluations, stated that the number of locations and the contact force with the liner for the Unit 1 truss with the modifications proposed in the application were less than the corresponding parameters for the Unit 2 truss at the containment design temperature. Based on those results, the licensee considered the thermal fragility curve for the Unit 2 truss to bound the modified Unit 1 truss and used a single thermal fragility curve for the risk assessments for thermal initiators supporting this application.

In response to RAl-5.c, the licensee performed a sensitivity study to quantitatively demonstrate the impact of changing the failure probability as a function of temperature for the trusses on the results of the risk assessments for thermal initiators. The licensee increased the failure probabilities by an order of magnitude for the sensitivity. The results of the sensitivity study showed that the CDF for the risk assessments for thermal initiators increased by an order of magnitude but remained below the acceptance guidelines in RG 1.174, Revision 2, both individually and when combined with the results for the seismic initiator. The NRC staff notes that due to the linear change in the CDF for risk assessments for thermal initiators with the change in the thermal fragility curve as demonstrated by the sensitivity, the failure probability would have to increase substantially from that developed by the licensee and would correspond to the collapse of the truss at the containment design temperature. The NRC staff considers such a scenario to be unrealistic.

In summary, based on its review of the thermal evaluations from the submittal as well as supplements and informed by regulatory audits, the NRC staff finds the approach and inputs for the thermal evaluations supporting this application to be acceptable for this application because (1) it uses the limiting thermal loading on the trusses, (2) it is technically justified, and (3) it provides sufficiently realistic predictions of the truss and attached components during a thermal loading event. Further, the NRC staff's review finds that the licensee's approach to developing the thermal fragility curve for use in the risk assessments is acceptable for this application because ( 1) it is based on the thermal evaluations performed to support this application and the NRC staff has found the approach and inputs for those evaluations to be acceptable for this application, (2) it is based on the component stresses that result from thermal loading thereby relating it to the structural stability of the trusses under thermal loading, and (3) the sensitivity OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION study performed by the licensee demonstrated that consideration of the uncertainty in the thermal fragility determination would not change the NRC staff's decision.

The NRC staff also finds the use of a single thermal fragility curve for the risk assessments for thermal initiators to be acceptable for this application because the stresses from thermal loading in the Unit 2 truss will bound those in the Unit 1 truss subsequent to the completion of the implementation items in the proposed license conditions.

The NRC staff's findings related to the methodology for the thermal evaluations, including the development of the thermal fragility curve, are limited to this application and are not to be interpreted as being generic.

3.1.3 Proposed Acceptance Criteria The licensee proposed alternate acceptance criteria for the trusses based on the engineering evaluations. The licensee proposed using the American National Standard Institute (ANSl)/American Institute of Steel Construction (AISC) N690-1994, "Specification for the Design, Fabrication, and Erection of Steel Safety Related Structures for Nuclear Facilities,"

including Supplement 2(R2004 ), as the code for the evaluation of the trusses. Further, the evaluations demonstrated that the proposed configuration of the trusses did not meet the allowable stresses of AISC N690-1994 (R2004) at the top and/or bottom chords of the trusses in combined axial and flexure or flexure only. Therefore, the licensee proposed and used a strain based acceptance criteria for the top and/or bottom chords of the trusses in combined axial and flexure or flexure only.

According to Section 5.4 of Enclosure 5 to the LAR, the proposed acceptance criteria for the trusses are based on ASCE/SEI 43-05 for Limit State D, "Essentially Elastic," within the limitations of NUREG/CR-6926. ASCE/SEI 43-05 identifies a set of seismic analysis parameters and design code options associated with this limit state. Based on Section 4.2.4 of ASCE/SEI 43-05, the licensee used ANSI/AISC N690-1994 (R2004), as the code for the evaluation of the trusses. The licensee stated that the primary limitation in NUREG/CR-6926 was addressed by the application of Limit State D in combination with the use of the GMRS.

The NRC staff reviewed the selection and application of the proposed code for the trusses. The NRG staff's review notes that Section C.1.3.8.4.5 of RG 1.206, "Combined License Applications for Nuclear Power Plants," June 2007 (Reference 29), endorsed the use of ANSI/AISC N690-1994 (R2004) as the acceptance criteria for seismic Category I structures.

Further, for design and analysis procedures, SRP Sections 3.8.4.1.4 and 3.8.4.11.5 state that ANSI/AISC N690-1994 (R2004) is applicable for steel structures. Based on its review of the licensee's application of the proposed code for the evaluation of the trusses and applicable guidance, the NRG staff finds that the licensee's proposed use of ANSI/AISC N690-1994 (R2004) with ASCE/SEI 43-05 for Limit State Dis acceptable for this application because ( 1) the licensee has used the code that is endorsed as the acceptance criteria for seismic Category I structures as well as steel structures, (2) the engineering evaluations document meeting the acceptance criteria for majority of the truss members and to provide reasonable confidence about the structural stability of the trusses, and (3) the licensee has applied the code consistent with staff guidance.

In Sections 5.4 and 6.5.1.2 of Enclosure 5 of the LAR, the licensee indicated that the truss top and bottom chord members would exceed the ANSI/AISC N690-1994 (R2004) allowable stress.

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OFFICIAL USE ONLY SECURITY RELATED INFORMATION The licensee provided a discussion in Section 5.5 of Enclosure 5 of the LAR the basis for the use a strain-based acceptance criteria of 1.5 percent as an alternative for the truss top and bottom chord members. SRP Section 3.8.4.111.5 states that the NRC staff evaluates the justification provided to show that structural integrity will not be affected if the licensee proposes to exceed some of the limits of ANSI/AISC N690-1994 (R2004). In RAl-20.b, the NRC staff requested that the licensee justify the selection of the proposed alternative strain-based acceptance criterion and explain the impact on the safety margins from the use of that criterion.

In response to RAl-20.b, the licensee stated that an allowable strain of approximately 1.5 percent accounted for the significant ductility present in steel components subject to varied loading conditions and was computed based on the ductility factor specified in the ANSI/AISC N690-1994 (R2004) for the truss. The licensee explained that ANSI/AISC N690-1994 (R2004) specified a reduced allowable ductility factor for open cross-section flexural members. The NRC staff's review noted that the licensee's proposed strain-based acceptance criteria was determined by calculating the allowable strain based on the reduced ductility factor relative to the strain at yield. Engineering evaluations supporting this application, which include the proposed configuration of the trusses, demonstrated that the maximum calculated strain limit, determined in the seismic evaluation of the Unit 2 truss, was below the proposed 1.5 percent strain based acceptance criterion. The licensee explained that the 1.5 percent strain acceptance criterion was slightly larger than the strain value at which strain hardening begins for ASTM (formerly known as American Society for Testing and Materials) A36 steel using minimum material properties. Based on its review of the submittal as well as the supplements and informed by the regulatory audits, the NRC staff finds that using an allowable strain limit of 1.5 percent as an alternative acceptance criterion for the top and bottom chord truss members is acceptable for this application because ( 1) the ductility factor of flexural members used in the determination of the acceptance criteria is representative of the truss top and bottom chord members, and (2) engineering evaluations, which the NRC staff found to be acceptable for this application as noted in Sections 3.1.1 and 3.1.2, demonstrate that the structural integrity of the trusses at design basis loading conditions would not be affected subsequent to the completion of the implementation items in the proposed license conditions.

3.1.4 Determination of Proposed Modifications As noted in Section 2.21 of this SE, the licensee proposed modifications in support of this application. The modifications include modification to the Unit 1 truss to improve clearance between the construction truss and the containment liner at six specified locations and modification to one Unit 1 containment spray pipe support to achieve additional seismic capacity.

In RAl-4.a and RAl-4.b, the NRC staff requested that the licensee clarify the modifications that were proposed to the trusses in the submittal and provide technical justification for the selection of the proposed modifications. In letter NRC 2018-0014, the licensee stated that the modifications proposed for the Unit 1 truss structure consisted of trimming the top chord of the truss at the first panel point of six truss locations. The locations at which the modifications would occur were provided by the licensee. The licensee's response to RAl-19.a clarified that the trimming consisted of removal of a portion of the flange and web followed by welding of a new plate to replace the removed flange at five locations and removing a portion of each end of the outer face of the flange adjacent to the containment liner or the remaining at the sixth OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION location. Additional details of the trimming were provided in the licensee's response to RAl-19.a. The licensee stated that no modifications were proposed for the Unit 2 truss structure.

The NRC staff's review of the submittal and the supplements noted that the proposed modifications to the Unit 1 truss and containment spray pipe support were based on seismic and thermal engineering evaluations as well as risk assessments supporting this application. The engineering evaluations performed for Units 1 and 2 included the modifications as well as the acceptance criteria proposed in the submittal. In addition, the engineering evaluations also considered the attached or adjacent SSCs. The licensee determined the locations that required modification through an iterative process such that the proposed configuration of the trusses and adjacent or attached SSCs met the corresponding. proposed acceptance criteria. The licensee's engineering evaluations performed for Units 1 and 2, which included the proposed modifications and the proposed acceptance criteria, demonstrated that:

• The trusses in both units maintained structural integrity for applied loading from design-basis accidents (thermal loading) and seismic loading.

• The truss members and connections for the truss in both units maintained stress values within limits of the proposed acceptance criteria. Those truss members that exceeded the stress limits remained within the proposed permissible strain limit acceptance criteria.

• The truss to containment liner contact load remained below the proposed allowable limit for both units.

• The attached VNCC ductwork for both units remained within original design-basis allowable limits and remained fully capable of performing its intended safety functions.

• The containment spray piping, for applied thermal loading, remained within original design-basis allowable limits and remained fully capable of performing its intended safety functions for both units.

• Other supported equipment such as lighting, conduit, and post-accident containment ventilation piping, which have no specified safety functions, maintained their structural integrity for both units.

The NRC staff's review of the containment liner evaluation and the corresponding proposed allowable limit is discussed in Section 3.1.6 of this SE. The NRC staff's review of the evaluation for the containment spray piping and VNCC ductwork is provided in Section 3.1. 7 of this SE.

The licensee explained that its engineering evaluations also demonstrated that the containment spray piping that was attached to the modified Unit 1 truss remained within original design-basis allowable limits under applied design-basis seismic loading and remained fully capable of performing its intended safety functions except for one containment spray pipe support. That specific containment spray pipe support required modification to meet design code of record allowable values as proposed by the licensee.

The licensee stated that the risk assessments supporting the application demonstrated that minimal risk reduction would be gained by performing additional modifications beyond those OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION proposed in the submittal. The licensee also stated that the risk assessment and its comparison against the acceptance guidelines in RG 1.174, Revision 2, ensured that the health and safety of the public was maintained while minimizing potential safety risks that could arise from rigging and handling of large structural members at high elevations, and the potential for dropped objects. The NRC staffs review of the risk assessments supporting this application is discussed in Section 3.5 of this SE.

Based on its review of the submittal as well as the supplements, and the NRC staff's findings in Sections 3.1.1, 3.1.2, 3.1.6, and 3.1. 7 of this SE, the NRC staff finds the proposed modifications to the Unit 1 truss and one Unit 1 containment spray pipe support to be acceptable for this application because (1) they were determined based on the seismic and thermal evaluations performed to support this application and the NRC staff has found the approach and inputs for those evaluations to be acceptable for this application as discussed in Sections 3.1.1 and 3.1.2, respectively, of this SE, (2) the proposed acceptance criteria for the trusses and adjacent or attached SSCs will be met subsequent to the completion of the implementation items in the proposed license conditions, (3) the structural integrity of the trusses as well as adjacent or attached SSCs, including the containment spray piping, at design basis thermal and seismic loading conditions would not be affected subsequent to the completion of the implementation items in the proposed license conditions, and (4) the risk assessments, reviewed by the NRC staff in Section 3.5 of this SE, demonstrate that the acceptance guidelines in RG 1.174, Revision 2, are met with the proposed modifications.

According to Enclosure 2, Section 1.2 of Enclosure 4 of the LAR, no modifications were proposed for the Unit 2 truss. However, Section 6.4.2.2 of Enclosure 5 stated that "[t]rimming the first panel point at ... 11 locations for Unit 2 ... " and "[t]he Unit 2 truss was analyzed ... "

Therefore, in RAl-4.c, the NRC staff requested the licensee to confirm that the configuration of the Unit 2 truss that was analyzed remained consistent with the configuration of the Unit 2 truss proposed in the submittal. In response to RAl-4.c in letter NRC 2018-0014, the licensee clarified that the evaluations performed for the Unit 2 truss included two configurations:

(1) unmodified and (2) with all modifications necessary to meet ANSI/AISC N690-1994 (R2004) limits without exceptions. The licensee explained that the unmodified Unit 2 truss was the configuration proposed in the submittal. The licensee confirmed that the configuration of the Unit 2 truss with all modifications was only utilized in the determination of the change in risk for this application.

Based on its review, the NRC staff finds that the Unit 2 truss configurations evaluated by the licensee are consistent with the proposed configuration and the approach used to determine the change in risk for the risk assessments supporting this application.

3.1.5 Determination of Proposed Seismic and Thermal Operating Limits Section 3.2 of Enclosure 1 of the LAR provided the proposed new seismic operating limits and Section 2.2.2 of the same enclosure provided the maximum ground accelerations in the horizontal and vertical directions for the operational basis earthquake (QBE). The proposed new seismic operating limit for the vertical direction exceeded the ground acceleration for the OBE. Additional information on the basis and justification for the selection of the new seismic operating limits and its relation to the QBE was requested in RAl-2. In letter NRC 2018-0014, dated April 12, 2018, in response to RAl-2, the licensee stated that the proposed new seismic operating limits represented the peak ground accelerations that, based on the seismic OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION evaluations, resulted in trusses as well as attached or adjacent equipment remaining within the corresponding proposed acceptance criteria. The licensee further stated that the calculated seismic accelerations which were used for the new seismic operating limits did not replace the site QBE and safe shutdown earthquake (SSE) design-basis accelerations. The licensee stated that existing site procedures would be revised to initiate actions to commence a dual-unit backdown to hot shutdown upon reaching the new seismic operating limits to ensure that both units are positioned into a condition that would permit inspection and/or evaluation of the trusses, equipment supported by the trusses, and the containment liner, as necessary. The licensee further stated the implementation of the new seismic operating limits would be consistent with current practice and therefore, would result in the new operating limit in the vertical direction being the same as the QBE. In Section 3.2 of Enclosure 1 to the LAR, the licensee stated that existing seismic monitors would be used to detect the new seismic operating limits. The NRC staff reviewed the licensee's approach to determine the proposed new seismic operating limits and its implementation and finds that the licensee's approach to determining the proposed new seismic operating limits and its implementation is acceptable for this application because ( 1) the proposed new seismic operating limits maintain the truss components and attached or adjacent SSCs within the proposed acceptance criteria and essentially elastic limits during seismic loading, which does not result in permanent strain deformation, (2) the proposed new seismic operating limits are based on the seismic evaluations which include the proposed modifications as well as acceptance criteria and were found to be acceptable for this application as discussed in Sections 3.1.1 and 3.1.3 of this SE, (3) exceedance of the proposed new seismic operating limits will result in inspections to determine structural integrity of the truss equipment supported by the trusses, and the containment liner, as necessary, and (4) the new seismic operating limit in the vertical direction, subsequent to its implementation in accordance with the proposed license conditions, will be the same as the site QBE.

The implementation items in the proposed license conditions included the implementation of a proposed new thermal operating limit for each unit. In Section 3.2 of Enclosure 1 of the LAR, the licensee provided the proposed new thermal operating limit for each unit and stated that those limits addressed continued operation after the limiting thermal event, and ensured that the affected SSCs remain capable of withstanding a subsequent design-basis thermal event. The licensee explained that the proposed new thermal operating limits are based on the thermal evaluations supporting this application. The licensee further stated that existing site procedures would be revised to take the affected unit offline after the thermal event and initiate inspection and/or evaluation for the affected unit's truss, equipment supported by the truss (as necessary),

and the containment liner prior to the unit startup. According to the licensee, existing station instrumentation will be utilized for containment temperature identification. The licensee explained that existing containment temperature indicators provided remote control room indication and were capable of identifying exceedance of the new thermal operating limits. The NRC staff reviewed the licensee's approach to determine the new thermal operating limits and its implementation, in accordance with the proposed license conditions, in the submittal and the supplements. Based on the review, the NRC staff finds that the licensee's approach to determining the new thermal operating limits and its implementation, in accordance with the proposed license conditions, is acceptable for this application because ( 1) the proposed new thermal operating limits maintained the truss components and supported attached or adjacent SSCs within the corresponding proposed acceptance criteria and essentially elastic limits during thermal loading, which does not result in permanent strain deformation, (2) the proposed new thermal operating limits are based on the thermal evaluations which include the proposed OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION modifications as well as acceptance criteria and were found to be acceptable for this application as discussed in Sections 3.1.2 and 3.1.3 of this SE, (3) the proposed new thermal operating limits are set below the containment design temperature providing margin for the implementing subsequent actions, and (4) exceedance of the proposed new thermal operating limits will result in inspections and/or evaluation for the affected unit's truss, equipment supported by the truss (as necessary), and the containment liner prior to the unit startup.

3.1.6 Evaluation of the Impact on the Containment Liner The licensee provided details of the evaluations performed for the containment liner in Section 6.3 of Enclosure 5 to the LAR. The evaluations used the following acceptance criteria:

• Allowable contact load on the containment liner near the truss ,structure under seismic or design-basis accident (i.e., thermal) loads as the minimum of the load that develops a maximum primary stress intensity of 0.9 times the ultimate strength or two-thirds of the maximum sustainable load.

• Liner integrity for applied cyclic loading assessed by comparing the accumulation in strains and the change in strains between cycles in combination with the fatigue curve from Section Ill of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV) Code, 1983.

• Localized exceedance of the permissible concrete strain limit of 0.003 in/in per ACI 318-63 at locations where the truss can contact the containment liner using concrete compressive strength from test data.

The evaluations for the containment liner included consideration of the concrete behind the liner. To assess liner integrity under repeated seismic loading from a truss contact point, the evaluation included repeated loading and unloading cycles which were determined from the shapes of the ground motion time histories used for the soil-structure interaction analysis.

Based on the evaluations, the licensee concluded that adequate structural capability of the liner was maintained subsequent to the completion of the implementation items in the proposed license conditions even though the trusses may come in contact with the containment liner as a result of a seismic or thermal event. The licensee stated that the acceptance criteria for the containment liner were based on the requirement to maintain leak-tight integrity of the liner and that the allowable stress contact load was determined based on ASME B&PV Code.

The concrete directly behind the contact point was expected to have strains that exceeded the permissible limit based on ACI 318-63 due to the highly localized stresses in the relatively small contact area. The licensee provided clarification on the proposed criteria for the localized strain exceedance in the concrete behind the liner in Attachment 1 of letter NRC 2018-0041, dated August 30, 2018, and stated that the evaluation performed for the containment concrete and liner modeled the concrete as integral with the containment liner such that the contact surface was a composite structure. The licensee provided details of the evaluation and determination of the portion of local wall area that exceeded the permissible strain limit based on ACI 318-63.

The licensee stated that the area exceeding the permissible strain was less than 1 percent of the local area which reflected the licensee's intent for the criteria of demonstrating that the containment structure shell strength was not "significantly reduced."

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION The NRC staff reviewed the licensee's approach for the evaluation of the impact on the containment liner and the concrete behind the liner from the submittal as well as the supplement and informed by regulatory audits. Based on the review the NRC staff finds that the evaluation of the impact on the containment liner and the concrete behind the liner, including the approach and inputs, are acceptable for this application because (1) the inputs are consistent with those used for the seismic and thermal evaluations for the trusses, (2) the evaluation accounts for the relevant implementation items in the proposed license conditions, and (3) the evaluation includes the impact of repeated seismic loading on the liner and concrete behind it. The NRC staff's review also finds the proposed evaluation criteria for the affected areas of containment liner as well as the concrete behind the liner to be acceptable for this application because (1) they are consistent with the provisions in the ASME B&PV Code and ACI 318-63, (2) they continue to maintain the leak tightness as well as structural integrity of the containment liner and the concrete behind it near the truss structure, and (3) the exceedance of the permissible concrete strain limit of ACI 318-63 is highly localized and is a very small fraction of the local area. The NRC staff notes that other criteria for the containment liner as well as the concrete behind the liner not included in the evaluations supporting this application remain unchanged from the LB.

3.1. 7 Evaluation of the Impact on Containment Spray Piping and VNCC Ductwork The licensee provided information on the evaluations performed for the containment spray piping and VNCC ductwork in Section 6.5.3 of Enclosure 5 to the LAR. The NRC staff reviewed the information and requested additional details regarding the evaluation. In letter NRC 2018-0014, dated April 12, 2018, the licensee stated that the containment spray piping ring headers that were supported from the trusses in Units 1 and 2, including a portion of upstream piping, were evaluated for seismic loading. The evaluation was based on the seismic response spectra that was used to evaluate the trusses. The licensee stated that the evaluation demonstrated that all associated pipe supports remained within the code of record stress limits excluding one support. As noted previously, modification to the identified support to achieve additional seismic capacity was included as one of the implementation items in the proposed license conditions. The licensee further stated that, based on its evaluations, design margin existed against code of record allowable values for piping and pipe support stresses from the applied seismic accelerations, which were developed from the site specific GMRS. According to the licensee, additional margin existed when the stresses were compared against higher operability limits for piping and pipe supports allowed by site design guidelines which were consistent with Section Ill, Appendix F of the ASME B&PV Code. The licensee stated that accounting for the permissible operability limits, the containment spray piping was considered to have a median capacity exceeding that calculated for the trusses. The licensee's response also included a discussion on the evaluation of the VNCC ductwork in both units for applied seismic loading. Based on the results of those evaluations, the licensee stated that the VNCC ductwork loading had significant margin to allowable design limits and concluded that the VNCC ductwork was considered to have a median capacity exceeding that calculated for the truss structures.

The NRC staff's review of the evaluation for the containment spray and VNCC ductwork finds that evaluation to be acceptable for this application because ( 1) it is performed using inputs that are consistent with those used for the seismic evaluation of the trusses, and (2) the approach is technically justified. Based on the review as well as the NRC staffs findings on the seismic evaluation of the trusses as well as the containment spray piping and VNCC ductwork, the NRC OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION staff finds reasonable confidence exists for higher expected median capacity of the containment spray piping, pipe supports, and the VNCC ductwork, subsequent to the completion of the implementation items in the proposed license conditions, as compared to the containment trusses in each unit because ( 1) margin exists to allowable design limits using the site-specific GMRS, and (2) additional margin is expected to be present when operability limits for piping and pipe supports from Section Ill, Appendix F of the ASME B&PV Code are considered.

3.2 Conformance with Current Regulations - Principle 1 Regulatory Position 1 in RG 1.174, Revision 2, states that the licensee should affirm that the proposed LB change meets the current regulations unless the proposed change is explicitly related to a proposed exemption (i.e., a specific exemption under 10 CFR 50.12). The licensee proposed acceptance of the configuration of the construction trusses, including the attached containment spray piping and VNCC ductwork, and the containment liner adjacent to the trusses, with the modifications in Attachment 3 of letter NRC 2018-0041, dated August 30, 2018, using an RI approach. The licensee did not seek an exemption under 10 CFR 50.12 as part of this submittal.

The licensee stated that no changes were proposed to the TS governing the operability and surveillance requirements for the containment structures including the containment liner as well as the containment spray and containment cooling systems. The licensee stated that no relief from the relevant licensed site-specific GDC requirements, as listed in Section 2.2.2 of to the LAR was requested. The licensee stated that the results of the engineering evaluations performed to support this application demonstrated that the construction trusses and components supported by the trusses continued to perform their designated design functions during a design-basis seismic or thermal event. The NRC staff's review of and findings on the engineering evaluations is discussed in Section 3.1 of this SE. The NRC staff finds that the proposed change continues to meet the current regulations because ( 1) relief from relevant site-specific GDC requirements is not sought, and (2) there is reasonable assurance that the construction trusses and components supported by the trusses will continue to perform their designated design functions. Further, the NRC staff finds that exemptions from regulations have neither been proposed nor been granted for this application.

3.3 Defense-in-Depth - Principle 2 Section 2.1.1 of RG 1.174, Revision 2, provides guidance on the evaluation that the proposed change is consistent with the defense-in-depth philosophy and states that consistency with the defense-in-depth philosophy is maintained if the following seven considerations are adequately addressed:

• A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation.

• Over-reliance on programmatic activities as compensatory measures associated with the change in the LB is avoided.

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

• System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g.,

no risk outliers). *

• Defenses against potential common-cause failures are preserved, and the potential for the introduction of new common-cause failure mechanisms is assessed.

• Independence of barriers is not degraded.

• Defenses against human errors are preserved.

• The intent of the plant's design criteria is maintained.

In Section 3.1.3 of Enclosure 1 to the LAR, the licensee provided an assessment of the impact of the proposed changes on the defense-in-depth philosophy and addressed each of the above items.

3.3.1 Reasonable Balance between Core Damage, Prevention of Containment Failure, and Consequence Mitigation is Preserved The NRC staff's review of and findings on the engineering evaluations are discussed in Section 3.1 of this SE. The NRC staff found the licensee's approach, inputs, and acceptance criteria used for the seismic and thermal engineering evaluations of the trusses and attached or adjacent SSCs to be acceptable for this application. Those engineering evaluations demonstrated that ( 1) the construction trusses and components supported by the trusses retained their structural stability during a design-basis seismic or thermal event without failure, (2) the construction trusses and components supported by the trusses continued to perform their design functions during a design-basis seismic or thermal event, and (3) the impact on the containment liner would not result in a breach of the liner material.

Based on its review, as discussed in Section 3.1. 7 of this SE, the NRC staff found reasonable confidence for higher expected median capacity of the containment spray piping, pipe supports, and the VNCC ductwork, after the completion of the implementation items in the proposed license conditions, as compared to the containment trusses in each unit.

The NRC staff's review notes that the seismic evaluations show that the containment trusses have a median capacity under a seismic event of approximately 0.5 g which is four times the SSE for each unit. Because the median capacity of the trusses exceeds the SSE and there is reasonable confidence that the median capacity of the containment spray piping, pipe supports, and the VNCC ductwork, subsequent to the completion of the implementation items in the proposed license conditions, would be higher than the containment trusses, the NRC staff finds that there is reasonable confidence that sufficient defense-in-depth exists beyond the site's seismic design basis.

The NRC staff, based on its review of the risk assessments supporting this application, discussed in Section 3.6 of this SE, notes that the VNCC ductwork supported by the construction truss is never credited as functional and the containment spray piping is postulated OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION to fail along with the postulated construction truss failure due to overstress in the risk assessments.

Based on the NRC staff's review of the submittal, the supplements, and the review as well as findings on the licensee's engineering evaluations, the NRC staff concludes that a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved for this application and none of the design-basis functions of the three fission product barriers are impacted by the proposed changes.

3.3.2 Over-Reliance on Programmatic Activities is Avoided The licensee has not proposed any compensatory measures to support the proposed changes.

As discussed in Section 2.1 of this SE, the licensee proposed modifications to support this application.

As noted in Section 3.1.5 of this SE, the implementation items in the proposed license conditions included the implementation of proposed new seismic operating limits applicable to both units. In Section 3.1.3 of Enclosure 1 of the LAR, the licensee also stated that the proposed new seismic and thermal operating limits were established to ensure the construction trusses and equipment supported by the trusses, as well as the containment liner, are inspected and/or evaluated if an event occurs that results in exceeding the proposed new operating limits and that those proposed new limits do not impose administrative actions for frequently occurring events. Further, in Section 3.1.3 of Enclosure 1 of the LAR, the licensee stated that changes to existing design functions or accident mitigation strategies were not proposed. The NRC staff's review of and findings on the proposed new seismic and thermal operating limits is discussed in Section 3.1.5 of this SE. Based on its review, the NRC staff notes that the proposed new seismic and thermal operating limits as well as the related procedure changes represent permanent changes and do not constitute a recurring or ongoing compensatory measure.

Therefore, the NRC staff finds that over-reliance on human actions or compensatory measures due to the proposed changes is avoided because the proposed changes do not substitute programmatic activities for design features.

3.3.3 System Redundancy, Independence, and Diversity is Preserved System redundancy, independence, and diversity result in high availability and reliability of safety functions and also help ensure that system functions are not reliant on any single feature of the design. The licensee has not proposed changes to the design, configuration, or mitigation functions of any plant systems. Further, the licensee's risk assessment supporting this application included the impact of the postulated truss failure on different systems. The NRC staff notes that the application does not change existing accident mitigation.

In response to RAl-11.e and the letter dated August 30, 2018, the licensee stated that proposed modifications to install a 24-hour backup nitrogen supply to the PORVs were completed in both units. Those modifications were part of the implementation of the licensee's approved use of the National Fire Protection Association (NFPA) Standard 805 as the LB for its fire protection program. The NRC staff approved the licensee's transition to NFPA 805 by letter dated September 8, 2016 (Reference 11 ). The licensee stated that the nitrogen supply system and the instrument air system, which is the primary supply source to the PORVs, did not depend on each other or on common equipment including power supply. Furthermore, the licensee OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION explained that the nitrogen supply system met seismic qualification requirements and was installed in accordance with station design standards while the instrument air system was not seismically qualified. The NRC staff notes that the nitrogen supply system enhances redundancy and independence for operation of the PORVs and consequently, the ability to perform feed and bleed (F&B) cooling.

The NRC staff's review finds that system redundancy, independence, and diversity are preserved because the proposed changes do not result in a significant increase in the expected frequency of challenges to SSCs or the consequences of failure of system functions.

3.3.4 Defenses Against Potential Common-Cause Failures is Preserved A seismic event has the potential to be a common-cause initiator for an overstress condition of the trusses in both units. However, the likelihood of the occurrence of such an event is independent of the proposed changes. As mentioned above, the licensee stated that based on the results of engineering evaluations, the trusses and supported equipment maintain their structural capability during a design-basis seismic event. Further, the proposed changes neither increased the existing common-cause failure potential nor introduced new common-cause failure mechanisms for the SSCs in a particular unit. The impact of the postulated failure of the construction truss in each unit, considering the relevant implementation items in the proposed license conditions, due to a seismic event was included in the licensee's risk assessments supporting this application. The licensee also indicated that a thermal event would be unit-specific and would not be a common-cause failure mechanism for the postulated failure of the truss of the other unit. In RAl-3, the NRC staff requested additional information about thermal events that can occur as a consequence of seismic events. The NRC staff's review of the licensee's response to RAl-3 is discussed in Section 3.5.1 of this SE. Based on its review, the NRC staff finds that the licensee has adequately assessed the potential for the introduction of new common-cause failure mechanisms because the proposed changes do not (1) introduce a new common-cause failure mechanism or event, (2) increase the probability of a cause or event that could cause simultaneous component failures, and (3) degrade defenses against potential common-cause failures. Furthermore, the licensee's risk assessment considers the impact of the common-cause initiator.

3.3.5 Independence of Barriers is Not Degraded The licensee stated that the three fission product barriers for each unit would remain intact as a result of the postulated seismic or thermal events. The licensee further stated that although the trusses may come in contact with the containment liner as a result of a seismic or thermal event, the engineering evaluations supported the determination that the liner would maintain adequate structural stability during a design-basis seismic or thermal event. Section 6.3 of Enclosure 5 to the LAR contains details on the engineering analysis performed by the licensee for the containment liner. The licensee has proposed an acceptance criteria for the liner contact load and localized containment concrete strain exceedance in the response to RAl-1. The NRC staff's review and findings on the evaluation for the liner and concrete behind the liner as well as the proposed criteria are discussed in Section 3.1.6 of this SE.

The engineering evaluations performed by the licensee demonstrated that contact between the truss and liner does occur for the thermal events. In response to RAl-5.a and RAl-5.b, the licensee stated that at the design-basis temperature of 286 degrees Fahrenheit (°F) fully plastic OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION hinges are not formed and that the liner contact force is less than the liner capacity. The licensee further stated that based on the engineering evaluations, Unit 2 truss was bounding as compared to the Unit 1 truss subsequent to the completion of the implementation items in the proposed license condition for Unit 1. As discussed in Section 3.1.5 of this SE, the licensee proposed a modification to implement proposed new thermal operating limit for each unit. The detection of thermal occurrence that trigger the new operating limits would result in inspection and/or evaluation actions for the affected unit's construction truss, equipment supported by the truss (as necessary), and the containment liner. The NRC staff's review notes that the proposed limits are below the design-basis temperature and, therefore, capture any potential impact on the liner due to contact from the truss at beyond design-basis temperatures.

Based on the information presented by the licensee in Section 2.3 of Enclosure 2 of letter NRC 2018-0014, the NRC staffs review also notes the low frequency of occurrence for thermal events that result in the containment temperature reaching or exceeding the design-basis value (i.e., the product of the thermal-initiating events of interest and the probability of those events causing the containment temperature to reach or exceed the design-basis value) is small.

In Section 2.1.2 of Enclosure 2 of letter NRC 2018-0014, the licensee stated that the reinforced-concrete containment structure would not be perforated by falling truss debris because it was unlikely that the structure would be hit based on the trajectories of truss debris.

The licensee has also included the containment wall and liner as part of the containment truss target assessment submitted in response to RAl-15.a (Reference 4). The target assessment uses walkdown observations along with empirical approaches to semi-qualitatively assess the possibility of failure of components due to the postulated failure of the truss. Based on the evaluation for the containment wall performed in the target assessment, the licensee stated that

((

)), the containment wall would not be perforated. Based on the review and consideration of the information described above, the NRC staff concludes that the independence of fission-product barriers is not degraded because ( 1) the licensee has adequately considered the impact of the thermal loading of the truss as well as its postulated failure on the containment liner and wall, and (2) such consideration has provided reasonable assurance that the containment liner and wall will maintain their integrity. The NRC staff also notes that the licensee's existing programs for the inspection of the containment liner remain unaffected by the proposed changes.

3.3.6 Defenses Against Human Errors are Preserved In Section 2.2 of Enclosure 1 to the LAR, the licensee stated that human error was not a contributor to the postulated failures, and that the construction trusses and equipment supported by the trusses were passive components. The NRC staff's review determined that human intervention relative to the structural components or system operations was not required during design-basis accident response. Further, although the impact of the postulated failure of the construction trusses on operator actions is included in the "demonstrably conservative" risk analyses, the NRC staff finds that new operator actions specifically related to the proposed changes were not developed or included. The NRC staff notes that defenses against human error generally involve the use of procedures, training, and human engineering, none of which are impacted or altered by the proposed changes. The NRC staff notes that the change to procedures to incorporate proposed new seismic and thermal operating limits does not affect OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL IJSE ONLY SECIJRITY RELATED INFORMATION accident mitigation. Therefore, the NRC staff concludes that the proposed changes preserve defenses against human error and do not introduce new human error mechanisms.

3.3. 7 Intent of the Plant's Design Criteria is Maintained The licensee stated that engineering evaluations demonstrated that, following completion of the proposed modifications to the Unit 1 truss and containment spray pipe support, structural integrity would be maintained in both units in the event of a design-basis seismic or thermal event. Based on its review and findings on the engineering evaluations as discussed in Section 3.1 of this SE, the NRC staff concludes that, after the completion of the implementation items in the proposed license conditions, the trusses and the equipment supported by the trusses remain capable of performing their specified design functions during a design-basis seismic or thermal event. Further, as discussed in Section 3.1.6 of this SE, the NRC staff also concludes that the containment and the containment liner can perform their specified design functions. Therefore, the NRC staff finds that the proposed change maintains the intent of the plant's design criteria.

3.3.8 Conclusion on Defense-in-Depth In summary, based on its review and the discussion in this section, the NRC staff finds that the proposed changes with consideration of the relevant implementation items in the proposed license conditions does not significantly impact the seven considerations for defense-in-depth and defense-in-depth will continue to be maintained commensurate with the expected frequency and consequence of challenges to the system resulting from the proposed changes.

• 3.4 Safety Margins - Principle 3 Section 2.1.2 of RG 1.174, Revision 2, states that the engineering evaluations should assess whether the impact of the proposed change is consistent with the principle that sufficient safety margins are maintained. RG 1.174, Revision 2, further states that sufficient safety margins can be shown to exist if:

• Codes and standards or their alternatives approved for use by the NRC are met, and

• Safety analysis acceptance criteria in the licensing basis (e.g., final safety analysis report, supporting analyses) are met or proposed revisions provide sufficient margin to account for analysis and data uncertainty.

3.4.1 Meeting Codes and Standards Nonconformance to the code of record at select truss locations were due to exceedance of design allowable stresses and contact with the containment liner. Therefore, the licensee used alternate evaluation methods and acceptance criteria to verify structural stability and containment integrity. Based on the licensee's engineering evaluations, the containment spray piping and pipe supports were determined to be within their respective code limits, except for one support in Unit 1 which required modification. The licensee, in response to RAl-1, clarified that the proposed changes to the LB also sought the NRC staff's review and approval for the use of the alternate evaluation methods and acceptance criteria. The response to RAl-1 OFFICIAL IJSE ONLY SECIJRITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION included a list of the alternate evaluation methods and acceptance criteria proposed and used by the licensee. Enclosure 5 to the LAR provided additional details on the proposed evaluation methods and acceptance criteria. The licensee submitted Attachment 4 to letter dated August 30, 2018, which included the proposed changes to the UFSAR.

The NRC staff reviewed the alternate evaluation methods and acceptance criteria to determine their acceptability for the current application and to determine if sufficient safety margins will continue to be maintained.

Based on its review and the information provided by the licensee, as discussed in Section 3.1.1 of this SE, the NRC staff finds the use of 7 percent damping factor for the truss structure to be acceptable for this application because of the connections for load transfer in the truss and consistency with the ASCE/SEI 43-05 as well as staff guidance. Furthermore, the NRC staffs review finds that additional safety margin exits beyond the selected damping value because higher damping factor can be used in accordance with the ASCE/SEI 43-05 in some locations of the truss.

The NRC staff requested the licensee to discuss the impact of the alternate strain-based acceptance criteria on maintaining sufficient safety margins. In its response to RAl-20.b in letter NRC 2018-0014, the licensee explained that the 1.5 percent strain acceptance criterion was slightly larger than the strain value at which strain-hardening begins for ASTM (formerly known as American Society for Testing and Materials) A36 steel using minimum material properties.

The NRC staff notes that strain hardening would result in additional margin and acceptable performance of the trusses under loading conditions. The licensee explained that the maximum calculated strain limit in the engineering evaluations supporting this application, determined in the seismic analysis of the Unit 2 truss, was below the proposed 1.5 percent strain-based acceptance criterion. The NRC staff notes that the typical material strength of steel members exceeds the minimum design values assumed in the evaluation, which provides additional design margin. Based on its review, the NRC staff finds that the acceptance criteria of an allowable strain limit of 1.5 percent for the top and bottom chord truss members maintains sufficient safety margin subsequent to the completion of the implementation items in the proposed license conditions.

Section 2.1.1 of the LAR states that "the trusses were not included in the original final safety analysis report seismic classification tables. They were subsequently added to the UFSAR in 2013 as a Seismic Class I structure supporting Class I piping and ductwork." In RAl-17, the NRC staff requested the licensee to clarify whether the containment dome trusses in both units were qualified as Seismic Class I or to the original intended seismic criteria. In letter NRC 2018-0014, the licensee stated that the function of Seismic Class I systems supported from the truss was conservatively used as the basis for classifying the trusses as being required to meet Seismic Class I criteria in the 2013 revision of the UFSAR. The licensee clarified that qualification or assessments, such as material classifications and weld certifications, did not exist to demonstrate that the trusses met Seismic Class I criteria. In letter NRC 2018-0014, the licensee proposed classification of the trusses in the UFSAR as Seismic Class Ill structures while maintaining their functions of providing support to Seismic Class I systems and components as well as preventing interaction with Seismic Class I SSCs (so-called Seismic 11/1 interactions). Seismic Class Ill is a licensee designation, which, according to Section A.5.1 of the UFSAR, represents structures and components that are not directly related to reactor operation or containment. The guidance in RG 1.29, Revision 5, "Seismic Design Classification OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION for Nuclear Power Plants," July 2016 (Reference 30), only distinguishes between seismic Category I and non-seismic Category I SSCs with seismic Category I SSCs designated as those "SSCs that must be designed to remain functional if the SSE occurs." The NRC staff reviewed the licensee's proposed classification of the trusses as Seismic Class Ill structures to determine the appropriateness of the proposed change. The NRC staff's review notes that no changes were made to the construction and configuration of the trusses to increase their capacity or structural stability when they were classified as Seismic Class I in 2013. The NRC staff's review also notes that the seismic and thermal evaluations supporting this application used the limiting thermal loading scenario and the seismic loading based on the GMRS from the site-specific re-evaluated seismic hazard in addition to the normal operating loads. The NRC staff's review also notes that the engineering evaluations of the trusses used the acceptance criteria of ANSI/AISC N690-1994 (R2004) which is endorsed in Section C.1.3.8.4.5 of RG 1.206 (Reference 29) as the acceptance criteria for seismic Category I structures. Alternate acceptance criteria to the ANSI/AISC N690-1994 (R2004) stress limits were proposed for a limited part of the trusses and included in the engineering evaluations. The NRC staff found the approach, inputs, and acceptance criteria for the engineering evaluations to be acceptable for this application as discussed in Sections 3.1.1, 3.1.2, and 3.1.3 of this SE. Those evaluations demonstrated that the trusses maintained their structural stability under design basis seismic and thermal loading conditions. Based on the seismic evaluations, the HCLPF for the trusses, considering the relevant implementation items in the proposed license conditions, was calculated to be higher than the site SSE providing additional confidence about the structural stability of the trusses. Even under conservative assumptions, the risk assessments supporting this application also meet the applicable acceptance guidelines as discussed in Section 3.5 of this SE. The NRC staff finds the licensee's classification of the trusses as Seismic Class Ill structures to be acceptable on the following bases: ( 1) the engineering evaluations included limiting thermal loading scenario and the seismic loading based on the GMRS from the site-specific re-evaluated seismic hazard, which used present day methodologies, in addition to the normal operating loads, (2) the acceptance criteria used for the engineering evaluations of the trusses was based on stress limits from the code that is endorsed by the NRC for seismic Category I structures, (3) the truss functions of supporting Seismic Class I systems and preventing interaction with Seismic Class I systems are maintained, and (4) the risk assessments supporting this application demonstrate that, even under conservative assumptions, the risk of truss failure immediately following slight overstress meets the acceptance guidelines.

Based on the staff review of the licensee's submittal as well as supplements and the finding on the categorization of the trusses as Seismic Class Ill, the NRC staff concludes that the trusses will continue to maintain sufficient safety margin reasonably commensurate with Seismic Class I criteria.

3.4.2 Meeting Safety Analysis Acceptance Criteria In RAl-19, the NRC staff requested additional information to ensure that the safety analyses in the licensee's UFSAR, including the assumptions therein, are not impacted by the proposed change. Specifically, the NRC staff sought information on the impact of the proposed change on the containment and accident analyses documented in Chapter 14 of the licensee's UFSAR.

In response to RAl-19.a-c, the licensee stated that Chapter 14 referenced the calculations for containment pressure and temperature responses to a loss-of-coolant accident (LOCA) and MSLB and that the containment heat sinks were an integral part of the calculations. The OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION licensee further stated that the calculations included the containment dome truss steel as a lumped heat sink. The licensee explained that the surface area and weight of the truss structures were calculated using the as-designed dimensions of the truss members for consideration as a heat sink and that some truss member lengths were conservatively calculated while some members were not included for simplicity of the calculation. The licensee stated that two discrepancies in the as-found configuration of the truss associated with the nonconformances were identified. The licensee explained that first discrepancy (i.e., the outermost bottom horizontal brace being framed into the bottom chord away from the truss support point) was not included in the original calculation of the total heat sink area (i.e.,

member was conservatively neglected) while the second discrepancy (i.e., the direction of the bottom chord bracing not matching the direction identified on the design drawing) was related to the direction of a truss member which would not impact the overall length and subsequently the corresponding surface area. The NRC staff reviewed the licensee's response and determination of the impact of the proposed change, including the proposed modifications to the trusses, on containment and accident analyses. The staff's review noted that the licensee determined the surface area of the portion of the truss that would be removed as part of the proposed modification to the Unit 1 truss under the conservative assumption of not crediting the surface area of the replacement material and demonstrated that the resulting reduction in surface area would have a negligible impact on the containment response calculations. The licensee further stated that the truss steel was considered to have a negligible impact in the source calculation for the containment free volume and therefore, was not explicitly included as a displacing volume in that calculation. The NRC staff also reviewed the licensee's basis for concluding that no new design-basis accidents were considered other than those currently postulated in the UFSAR. In its response to RAl-19.d, the licensee stated that the engineering evaluations performed to support the submittal demonstrated that the trusses maintained structural integrity and all associated equipment maintained the ability to perform its intended design functions. Therefore, the licensee concluded that no new accidents other than those currently postulated in its UFSAR were considered. The NRC staff's review of and findings on the engineering evaluations supporting this application are discussed in Section 3.1 of this SE.

The NRC staff's review of the risk assessments supporting this application, discussed in Section 3.6 of this SE, noted that the postulated collapse of the truss structures was included in the risk assessments.

Based on its review of the submittal as well as the supplements to the LAR, and supported by NRC staff findings on the engineering evaluations and risk assessments, the NRC staff finds that the safety analysis acceptance criteria in the LB for Units 1 and 2 will continue to be met because ( 1) as-found discrepancies did not impact the originally calculated surface area of the truss structures, (2) proposed modifications are expected to have a negligible, if any, impact on the containment response calculations, and (3) the proposed changes do not result in design-basis accidents other than those postulated in the UFSAR.

3.4.3 Additional Safety Margin Features The proposed changes are related to a legacy design condition originating from the station construction which affects passive structures. The licensee stated that there was no adverse impact to SSCs as a result of the changes and all SSCs remained fully capable of performing their designated design-basis accident mitigation functions with no change to the method of performing those functions and no need for human intervention.

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION In response to RAl-11.e, the licensee stated that the completed modifications for the 24-hour backup nitrogen supply to the PORVs in both units included the installation of nitrogen supply tubing and actuation control circuits in a configuration such that they were considered protected from a postulated falling object. In addition, the licensee identified that the original, as-built control cables for one bleed path in Unit 2 were not routed in a manner that would provide protection from a postulated falling object. The licensee proposed an implementation item in of licensee letter NRC 2019-0007, dated March 13, 2019, to reroute the control cables for the identified components to ensure they were protected from a postulated falling object. The installation of the supply tubing and actuation control circuits and the proposed modification to reroute the control cables for one bleed path in Unit 2 to ensure protection against a postulated falling object provide additional safety margin in the availability of the PORVs as a mitigation feature.

As stated in Section 2.2, the implementation items in the proposed license conditions included proposed new seismic and thermal operating limits as well as changes to existing site procedures when either of the new operating limits are reached. The NRC staff's review of the proposed new seismic and thermal operating limits is discussed in Section 3.1.5 of this SE. The NRC staff's review of the proposed new seismic and thermal operating limits found that the proposed new seismic and thermal operating limits maintained the truss components and supported attached or adjacent SSCs within the corresponding proposed acceptance criteria and essentially elastic limits during thermal loading, which does not result in permanent strain deformation; the proposed new thermal operating limits are set below the containment design temperature providing margin for the implementing subsequent actions; and exceedance of the proposed new seismic or thermal operating limits would result in inspections to ensure structural stability of the trusses. Based on its review of the proposed new seismic and thermal operating limits, the NRC staff finds that additional means will exist to ensure that sufficient safety margins continue to be maintained after the completion of the implementation items in the proposed license conditions.

3.5 Evaluation of Risk Increase - Principle 4 The licensee performed risk assessments to support the proposed changes to the LB. The risk assessments determined the change in risk associated with seismic and thermal events for the proposed configuration of the trusses. The licensee's original risk assessments were discussed in Sections 3.1.1 and 3.1.2 of Enclosure 1 as well as Enclosure 4 of the LAR. In response to RAls from the NRC staff, the licensee provided a revised risk assessment in Enclosure 2 of letter NRC 2018-0014. The licensee has performed and submitted the results from two assessments; the "bounding analysis" (hereafter referred to as BA) and the "demonstrably conservative analysis" (hereafter referred to as DCA). The licensee performed the BA and DCA for the original risk assessment as well as the revised one. A comparison of the key boundary conditions for each analysis is provided in the Table 1 below. Additional details on the approach OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION used for each analysis and the corresponding NRC staff review are provided in Section 3.5.2 of this SE.

Table 1: Key Boundary Conditions for Risk Assessments Supporting the Application Bounding Demonstrably Conservative Construction truss fails (i.e., Construction truss fails (i.e., collapses) if overstressed.

collapses) if overstressed.

Construction truss failure always Qualitative determination of the trajectory of leads to core damage. components/members from postulated falling truss member.

Semi-quantitative determination of the vulnerability of components to falling construction truss debris including consideration of the location and robustness of the barriers that would protect such components.

No credit for mitigating systems and Mitigating systems and operator actions credited for operator actions for seismic initiator; seismic initiator; F&B, along with associated operator Containment mitigation systems actions, as well as containment mitigation systems credited for thermal initiators. credited for thermal initiators.

The licensee performed engineering evaluations to support the risk assessments. The engineering evaluations were used to determine the strength capacity of the truss under seismic loading. The determination of the strength capacity under thermal loading was determined using expert elicitation and engineering judgement supported by the engineering evaluations. of the LAR provided additional details on the engineering evaluations for seismic and thermal loading. The resulting seismic and thermal median fragility values were used to calculate the corresponding conditional failure probabilities that were used in the risk assessments.

The licensee characterized the risk from the proposed change in terms of CDF and LERF, which are used as surrogates for latent and early fatality risks, respectively. As a result, the acceptance guidelines for the change in these risk metrics provided in RG 1.174, Revision 2, are directly applicable to the fourth principle of RI decisionmaking and the licensee used those to support the proposed change.

Section 2.3.2 of RG 1.174, Revision 2, states that "[t]he characterization of the problem should include establishing a cause-effect relationship to identify portions of the PRA affected by the issue being evaluated." Section 111.2.2.1 of SRP Section 19.2, states that such an approach includes identifying the specific PRA contributors for the application, assessing the portions of the model that should be modified for the application, and identifying supplemental analyses that could be used to support the application.

Section 2.3 of RG 1.174, Revision 2, indicates that the acceptability of a PRA analysis used to support a LAR is measured in terms of its appropriateness with respect to scope, level of detail, technical elements, and plant representation. These attributes are expected to be OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICI.O.b USE ONLY SECURITY RELATED INFORMATION commensurate with the application for which the PRA analysis is intended and the role the corresponding results play in the integrated decisionmaking process. The guidance in RG 1.174, Revision 2, and SRP Section 19.2, framed the NRC staff's review of the risk assessment performed by the licensee to support this application. The NRC staff's review was informed by regulatory audits (References 18 and 19) and a public meeting (Reference 31 ).

3.5.1 Scope of the Probabilistic Risk Assessment Analysis Initiating events are system perturbations to the steady-state of the plant that challenge plant control and safety systems whose failure could lead to core damage or radioactivity release.

According to Section 3.1.1 of Enclosure 1 and Section 1.3 of Enclosure 4 of the LAR, the initiating events considered were those that cause seismic and thermal loading events. The thermal loading events considered were postulated LOCAs including small, medium, and large break LOCAs (SLOCA, MLOCA, and LLOCA, respectively). These initiators were used in the risk assessments for both the BA and DCA. While the seismic and thermal loading events are directly related to the identified nonconformances, a risk assessment needs to consider all events and hazards that can credibly impact the structural integrity of the truss and the equipment supported from it.

RAl-3 requested the licensee to provide quantitative or qualitative technical justification for the exclusion of internal and external initiating events other than those considered in the risk assessments. In response to RAl-3, the licensee stated that only initiating events that could increase containment temperature above 250 °F were considered in the risk assessments based on the thermal fragility curve developed for this application. The licensee provided a listing of all initiators for internal events, internal fires, and external events that were included or excluded from the thermal analysis along with the corresponding bases. Based on the consideration of events performed by the licensee after the NRC staff's request, internal fire-induced SLOCA and MLOCA were found to can cause containment temperature to increase above 250 °F and were included as initiating events in the revised thermal analysis performed in of letter NRC 2018-0014.

In the listing provided in response to RAl-3, the licensee excluded seismically induced SLOCA, MLOCA, and LLOCA. The licensee stated that the basis for exclusion was the associated fragility, which would result in a low probability of occurrence. In case of SLOCA, the licensee, in response to RAl-8.c, stated that SLOCAs were considered in the revised PRA evaluation in of letter NRC 2018-0014. The licensee explained that the revised PRA evaluation considered SLOCA due to seismically induced failures. The licensee used the fragility associated with SLOCA from the NUREG/CR-4840, "Procedures for the External Event Core Damage Frequency Analyses for NUREG-1150," dated November 1990 (Reference 32). The licensee stated that the value from NUREG/CR-4840 was consistent with that used in the licensee's seismic PRA developed for its Individual Plant Examination for External Events (IPEEE) ("Point Beach Nuclear Plant Individual Plant Examination of External Events for Severe Accident Vulnerabilities: Summary Report"}, dated June 30, 1995, and in NRC's risk assessment of operational events (RASP) manual, Volume 2 (Reference 33). In addition, in response to RAl-3, the licensee explained that based on the fragility associated with SLOCA, the truss was more likely to fail due to a seismic event than a piping failure resulting in an SLOCA. In Section 9.2.1 of Enclosure 2 of letter NRC 2018-0014, the licensee provided the quantitative estimate of the initiating frequency for seismically induced MLOCA and SLOCA and demonstrated that the resulting cumulative frequency was appreciably lower than that used for OFFICI.O.b USE ONbY SECURITY REbMED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION the thermal analysis. The NRC staff notes that the addition of seismically induced LLOCA to the estimate would not cause the cumulative frequency to change noticeably. Although not an initiator, the licensee calculated the "frequency of successful F&B." Based on the low calculated frequency as compared to the initiating frequency for thermal initiators included in the analysis, the licensee excluded F&B as a thermal-initiating event for this application. The NRC staff notes that temperature increase in the containment due to successful F&B would occur after a prolonged period of time, in an intermittent manner, and its impact would be bounded by initiators included in the thermal analysis such as MSLB. The licensee's evaluation resulted in the thermal initiators being identified as LLOCA, MLOCA, SLOCA, fire-induced SLOCA, fire-induced MLOCA, feedwater line break inside containment (FLBIC), and steam line breaks inside containment (SLBIC). The NRC staff also reviewed the internal events initiators in the NRC's Standardized Plant Analysis Risk model for Point Beach. The NRC staff finds that the licensee has adequately considered internal initiators that may impact this application and identified the thermal initiators that are appropriate for this application.

In response to RAl-3, the licensee stated that external hazards were screened for the site and referred to Attachment I of Enclosure 2 of letter NRC 2018-0014, for details. The NRC staff notes that the screening of external hazards is provided in Attachment H and that the screening is based on the impact to the site rather than the impact on the current application. The NRC staff's review finds that the external hazards identified in Attachment H of Enclosure 2 of letter NRC 2018-0014, represent a reasonably comprehensive list of such hazards. Additionally, the NRC staff also finds that site-specific bases for screening of several external hazards are applicable to the current application and that the external hazards in Attachment H of of letter NRC 2018-0014, are not expected to impact this application. Therefore, the NRC staff concludes that licensee has adequately considered external hazards that can impact this application.

Section 2.3.1 of RG 1.174, Revision 2, states that a qualitative treatment of certain modes and initiating events may be sufficient when the licensee can demonstrate that those risk contributions would not affect the decision; that is, they do not alter the results of the comparison with the acceptance guidelines. In Section 7 of Enclosure 2 of letter NRC 2018-0014, the licensee provided qualitative arguments for the at-power analysis bounding shutdown conditions. In that section, the licensee stated that a shutdown condition with an unisolated containment barrier would be bounded by the analysis for full-power operations due to the extremely low probability of occurrence of a thermal event when the containment barrier could be breached for maintenance and because operation in such a situation would restrict reactor coolant system temperature to less than or equal to 200 °F. The licensee further stated that the impact of a LOCA during mid-loop operation on the truss would be bounded by the full-power operation event and that MSLB cannot occur during mid-loop operation. The licensee explained that very little time was spent in mid-loop operation. The licensee further stated that the seismic risk during shutdown was minimized by the hazard exposure being a small fraction of the full-power operation exposure per year. The NRC staff finds that the licensee has provided sufficient qualitative arguments for not explicitly considering the shutdown mode for the risk assessment supporting this application. The NRC staff further finds those qualitative arguments to be acceptable for this application because the impact of the thermal and seismic initiators on the trusses under shutdown conditions will be bounded by the corresponding impacts at full power, which have been included in the risk assessments by the licensee.

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION In Section 1.6 and Attachment C of Enclosure 4 of the LAR, the licensee cited Part 6 of the 2009 ASME/American Nuclear Society (ANS) PRA Standard (ASME RA-Sa-2009) and stated that the bounding and demonstrably conservative analyses showed that the change in risk was acceptably low for hazards challenging the containment truss. According to Section 6-2.1 of the 2009 ASME/ANS Standard, Part 6 provides a process whereby an external hazard is excluded from further consideration in the risk assessment. Further, Section 6-1.2 of the 2009 ASME/ANS Standard clarifies that 'other external hazard' refers to external hazards other than earthquakes. In RAl-7.a, the NRC staff sought to clarify the intent of citing Part 6 of the 2009 ASME/ANS PRA Standard in the context of this application. In letter NRC 2018-0014, the licensee stated that Part 6 of the 2009 ASME/ANS PRA Standard was cited not to screen out hazards but rather to define the terms "bounding" and "demonstrably conservative analyses" and the corresponding methodology. The licensee further stated that reference to Part 6 of the 2009 ASME/ANS PRA Standard was removed in the revised risk assessment in Enclosure 2 of letter NRC 2018-0014. Based on its review, the NRC staff finds that the licensee's reference to Part 6 of the 2009 ASME/ANS PRA Standard does not impact this application because the licensee has not screened out hazards based on Part 6 and has removed the reference to Part 6 in the revised risk assessment.

Based on the review of the submittal, the supplement, and the responses to RAl-3 and RAl-7.a, the NRC staff finds that (1) the licensee has adequately considered and modeled the operating modes and initiating events of interest for the changes proposed in this application, and (2) demonstrated that the risk contribution from the events and modes that are not explicitly modeled does not affect this application.

3.5.2 Level of Detail of Risk Analysis Table 111-1 of SRP Section 19.2, summarizes the general guidance for establishing the cause-effect relationship for the PRA model and in identifying elements of the PRA model that may be affected by an application. As stated earlier, the licensee has performed two risk assessments, the BA and DCA. The licensee has performed the BA and DCA for the initiating events identified to be relevant to this application; namely, the seismic and thermal-initiating events.

(a) Bounding Analysis (i) Core Damage Frequency Evaluation for Seismic Initiator The BA for the seismic initiator assumes that any truss overstress always leads to core damage and does not credit mitigating systems or operator actions. Therefore, the cause under such ah assumption is the seismic-initiating event. The effect is the likelihood of structural failure of the truss captured by the fragility of the truss (i.e., conditional probability of failure of the truss based on the seismic acceleration) leading directly to core damage, thereby assuming a conditional core damage probability of 1.0. In response to RAl-12.a, the licensee stated that containment sprays were not credited for mitigation after a postulated overstress event due to the seismic initiator.

As noted in Section 3.1.1 of this SE, the licensee has performed a plant-specific seismic hazard reevaluation in response to NTTF Recommendation 2.1 and the NRC staff has previously reviewed the submitted re-evaluated hazard and concluded that "the licensee conducted the OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION seismic hazard reevaluation using present-day methodologies and regulatory guidance, appropriately characterized the site given the information available, and met the intent of the guidance for determining the reevaluated seismic hazard." In Section 2.2.1 of Enclosure 2 of letter NRC 2018-0014, the licensee stated that the mean seismic hazard curve as well as the different percentiles used to support this application were obtained from the above mentioned seismic hazard reevaluation. The NRC staff verified that the seismic hazard curves used to support this application, including the mean curve as well as the different percentiles, are consistent with those previously submitted by the licensee and reviewed by the NRC staff.

Therefore, the NRC staff finds that a separate review of the seismic hazard curves used for this application is not warranted and the seismic hazard curves used by the licensee to support this application are acceptable.

The licensee used the 'plant fragility' from its IPEEE seismic PRA to capture the impact of seismically induced failures of SSCs, which are independent of the postulated truss failure, on the CDF and change in CDF (hereafter referred to as £1CDF). Such seismically induced failures of SSCs can lead to core damage independent of the behavior of the truss during a seismic event. In response to RAl-9.b, the licensee stated that the BA assumed that the postulated failure of the truss from overstress due to a seismic-initiating event always led to core damage and during such an event other SSCs independent of the truss could also fail and result in core damage. The licensee further stated that integrating those concurrent failures would result in higher CDF and LERF values than considering the postulated truss failure alone and that the

£1CDF would decrease because the independent SSC failures partially overlapped with the impact from the postulated truss failure.

In RAl-9.a, the NRC staff requested the licensee to justify the applicability of the IPEEE based

'plant fragility' and in RAl-1 O.c, the NRC staff requested the licensee to justify that use of the

'plant fragility' approach bounds the impact of concurrent failures of SSCs independent of the truss. In response to RAl-10.c, the licensee stated that the 'plant level' seismic fragility curve for the risk assessment documented in Enclosure 2 of letter NRC 2018-0014, was derived from the HCLPF value from calculations performed as part of Generic Issue (Gl)-199, "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants: Safety/Risk Assessment" (Reference 34). The licensee stated that the IPEEE

'plant fragility' was derived from a quantification of the licensee's IPEEE seismic PRA that included a systems logic model, component seismic fragility data, random failure probabilities, and human error probabilities (HEPs). The licensee demonstrated that the IPEEE 'plant level' fragility curve was close to that used in the Gl-199 calculation and further stated that the Gl-199 seismic 'plant level' fragility appropriately accounted for seismic failures of SSCs not directly impacted by postulated failure of the truss. The NRC staff reviewed the information provided by the licensee and finds the licensee's consideration of the seismic failure of other SSCs independent of the truss for the CDF and £1CDF determination to be valid because of the nature of the plant-wide impact and resultant plant response during a seismic-initiating event.

Seismically induced failures of SSCs can lead to core damage independent of the behavior of the truss during a seismic event and exclusion of such failures provides a refined estimate of the impact of only the postulated truss failure under the assumptions of the BA for this application.

The NRC staff also finds the licensee's approach for capturing the impact of such seismically induced failures of SSCs independent of the truss on the CDF and £1CDF determination to be acceptable because the use of the 'plant fragility' adequately captures the likelihood of such failures in the context of the BA used for this application.

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OFFICIAL USE ONLY SECURITY RELATED INFORMATION (ii) Core Damage Frequency Evaluation for Thermal Initiators The BA for the thermal events used the initiating frequencies for the identified thermal events of interest, the probability of the availability of containment thermal mitigation functions as well as the resulting temperature, and the truss thermal fragility (conditional probability of failure of the truss based on the containment temperature) to determine the frequency of core damage from truss failure.

Section 3.5.1 of this SE provides a discussion of the thermal-initiating events used for the BA and DCA. According to Section 2.2.2 of Enclosure 4 of the LAR, information from an Electric Power Research Institute (EPRI) report 302000079, "Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments," Revision 3, was used to determine the initiating frequency for SLBIC and FLBIC. The NRC staff questioned the applicability of the initiating frequencies for those breaks from the cited EPRI report as they appeared to be related to breaks outside containment. In response to RAl-12.c, the licensee cited the revised risk assessment in Enclosure 2 of letter NRC 2018-0014. Section 2.3.2 of that enclosure provided information, including the source, for the frequency for thermal-initiating events used for the revised BA and DCA. The licensee used the SLBIC-initiating frequency from the 2015 update to NUREG/CR-6928, "Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants," for initiating events (Reference 35) in the revised risk assessments. The licensee continued to use the initiating frequency for FLBIC from the above-mentioned EPRI report in the revised BA and DCA. In response to RAl-12.c-1, in of the letter dated August 30, 2018, the licensee stated that the frequency for FLBIC in the 2015 update to NUREG/CR-6928 did not distinguish between breaks inside or outside containment. The licensee also provided the results of sensitivity calculations using the FLBIC-initiating frequency from the 2015 update to NUREG/CR-6928 for the BA for thermal initiators. The sensitivity calculations did not show an appreciable impact on the results. The NRC staff also notes that the energy released to the containment, and resulting temperature rise, is lower for FLBIC as compared to SLBIC and that the licensee used the thermal fragility curve, determined based on the temperature response for an MSLB, for all thermal initiators.

The initiating frequencies used by the licensee for the fire-induced SLOCA and MLOCA, which were additional initiators used in the revised risk assessment, were based on the licensee's NFPA 805 fire PRA model. Further, the licensee used the initiating event frequencies for SLOCA, MLOCA, and LLOCA from its internal events model. The NRC staff's approval of the licensee's transition to NFPA 805, in the letter dated September 8, 2016, found the licensee's internal events, including internal flooding, and fire PRA models to have sufficient technical adequacy. The NRC staff's conclusion on the technical acceptability of the internal events, including internal flooding, and fire PRA models from the NFPA 805 review is applicable to this submittal because those models are used as the base for the risk assessments supporting this application. The NRC staff compared the SLOCA, MLOCA, and LLOCA initiating frequencies used by the licensee against those in the 2015 update to NUREG/CR-6928 and found the values used by the licensee to be similar or higher.

Based on its review, the NRC staff finds that the licensee has used appropriate initiating event frequencies for all thermal initiators of interest for this application because (1) the initiating frequencies are derived from acceptable sources, and (2) the licensee demonstrated that using the selected FLBIC-initiating frequency did not appreciably impact the results.

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OFFICIAL USE ONLY SECURITY RELATED INFORMATION The licensee considered the impact of the operation and failures of systems that mitigate containment temperature transients in the BA for thermal events. The mitigating systems considered were the containment safeguards actuation trains, the containment fan coolers, the containment spay, the feedwater isolation valve, and the feedwater regulating valve. In response to RAl-12.a, the licensee explained that if the containment spray failed to mitigate the overstress, the truss was assumed to be overstressed and to generate debris. The licensee used the results of thermal-hydraulic simulation of an MSLB to determine the temperature rise in the containment under different combinations of mitigating system availability. According to the response to RAl-19.a, the thermal-hydraulic simulation used by the licensee was the same as that referenced in Chapter 14 of the Point Beach UFSAR (Reference 28). The NRC staff finds that the MSLB represents the limiting thermal loading challenge to the truss for this application because of the magnitude as well as the rate of temperature rise during that event compared to the other thermal initiators included in the risk assessment.

The licensee used the predicted containment temperature for each combination of mitigating system availability to determine the corresponding conditional probability of truss failure using the thermal fragility curve developed for this application. According to the response to RAl-6.d, the licensee developed logic models (fault trees) based on the licensee's internal events and NFPA 805 fire PRA model to determine the failure probability of each mitigation system. The licensee provided details of the logic model development in Attachment E to Enclosure 2 of letter NRC 2018-0014. The licensee then combined those failure probabilities with the thermal fragility curve for the truss and the initiating event frequency of the thermal initiators to determine the frequency of core damage from truss failure due to thermal events. Such a calculation was performed for each combination of mitigating system availability, including the containment sprays as clarified by the response to RAl-12.a. Based on its review, the NRC staff finds the licensee's above-described approach for determining the frequency of core damage from truss failure due to thermal events to be acceptable for this application because it captures the impact of containment temperature rise in a probabilistic manner consistent with state-of-practice approaches and the intended use in the BA for thermal initiators. The NRC staff notes that the licensee's approach does not differentiate between the containment response for all the thermal-initiating events of interest which is conservative for certain initiators such as SLOCA and FLBIC. The NRC staff's review of the thermal fragility curve developed for the risk assessments supporting this application is provided in conjunction with the response to RAl-5 in Section 3.1.2 of this SE.

(iii) Large Early Release Frequency Evaluation In the LAR, the licensee used a conditional large early release probability (CLERP) of 0.2 based on the information from its PRAs for other hazards, such as internal events as well as fire, and applied that value to determine the LERF for this application. The NRC staff requested additional information in RAl-14 to justify the determination of the chosen CLERP value. In letter NRC 2018-0014, the licensee stated that the method used to assess LERF had been revised and accounted for potential damage to containment penetrations from postulated falling truss debris and independent containment failures due to seismic events. In Section 2.1.3 of Enclosure 2 of letter NRC 2018-0014, the licensee stated that a CLERP value of 0.5 was used for the revised BA for both the thermal and seismic initiators and explained that:

• Containment penetrations were grouped and spanned less than 33 percent of the containment circumference. Further, penetrations below the ((

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)), were unlikely to be damaged by postulated falling truss debris because they were protected by the floor and other robust barriers above that elevation.

• Containment penetration barriers had minimal exposure to falling debris. Postulated truss debris would tend to fall away from the containment wall and avoid striking the containment penetrations. Barriers in the direct path of debris that may target these penetrations would deflect the debris or absorb energy. Discussion of the seven penetrations above the operating floor as well as justification for their robustness and/or their failures being inconsequential in the context of LERF included:

o Two conical steel seals for the main stem line containment penetrations that were welded to the containment liner. The seals were considered unlikely to be struck and perforated by postulated truss debris because of their location and construction. The release path was blocked by the main steam isolation valve located outside containment.

o Two main steam vent lines were considered to be fragile and likely to be damaged by postulated truss debris. Based on the UFSAR, the vent line piping was normally isolated outside the containment by isolation valves which would block the release path.

o The personnel hatch had a robust cover and seal inside and outside containment. In the unlikely event that the inside cover was perforated the outside cover would block the release pathway.

o Two containment vent purge supply and exhaust penetrations were not utilized during power operation because the system is not normally operational during power operation. There was a normally closed valve outside the containment and the inside was blank flanged. The flanges and valves inside the containment were unlikely to be subjected to direct impact because of their minimal protrusion as well as shielding by the polar crane ring girder.

In response to RAl-14, the licensee also explained that instrument air was not required for containment isolation valves to perform their isolation function and that the valves returned to their desired position for containment isolation upon loss of instrument air.

The NRC staff's review finds the approach used to estimate CLERP due to containment penetrations for the revised BA to be acceptable for this application because the approach considers the potential damage to containment penetrations from falling objects. The NRC staff also finds that the presence of isolation valves outside the containment results in large early releases from postulated truss failure being highly unlikely. The NRC staff notes that for the BA, the licensee's CLERP approach, when combined with the approach used for calculating the CDF, results in large early release assumed for half of postulated truss failures and that mitigation is not credited.

For calculation of LERF and change in LERF (hereafter referred to as .6LERF) values due to the seismic initiator, the licensee considered SSCs failures independent of the truss during a seismic-initiating event. The licensee used the dominant LERF contributor from the licensee's IPEEE seismic PRA model to capture the failures during a seismic event independent of the OFFICIAL lJSE ONLY SECYRITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION postulated truss failure that can lead to LERF. The licensee stated that the dominant LERF contributor from the licensee's IPEEE seismic PRA model was the failure of containment isolation. The NRC staff's SE report for the licensee's IPEEE seismic PRA ("Staff Evaluation Report on Individual Plant Examination of External Events Submittal for Point Beach Nuclear Plant, Units 1 and 2"), dated September 15, 1999, about the containment performance under seismic conditions, confirms that "dominant contributors were failures of containment isolation."

The licensee re-created the containment isolation failure logic model (i.e., fault tree) from its IPEEE seismic PRA model to capture the failure of key components and the operator action needed to isolate the containment during a seismic event. The licensee used fragility data as well as the HEP from the IPEEE for the components and action in the recreated logic model.

The licensee also included post-lPEEE modifications that impact containment isolation into the logic model. The licensee used the fragility of containment isolation with the contribution of post-lPEEE modifications to capture the impact of failures independent of the postulated truss failure on LERF and LiLERF for seismic-initiating events.

The NRC staff's SE report for the licensee's IPEEE seismic PRA stated that "the fragility analysis employed in the Point Beach seismic IPEEE is consistent with the guidelines of NUREG-1407" (Reference 36). Based on its review, the NRC staff finds the licensee's use of fragilities from its IPEEE seismic PRA acceptable for the BA used for this application because (1) the BA does not alter the approach used to determine those values, (2) the approach was previously found to be acceptable and based on staff guidelines, and (3) the resulting values are sufficiently representative of the as-built, as-operated plant.

According to the licensee's IPEEE submittal, manual containment isolation was an in-control room action and was assigned an HEP that varied with seismic acceleration. With regard to the HEPs used in the licensee's IPEEE seismic PRA, the technical evaluation report states that "HEPs were based on engineering judgment." However, the NRC staff notes that assuming the operator action to be failed would not appreciably impact LiLERF determined using BA. Such an outcome is due to the logic model for containment isolation failure during seismic initiator in the context of the BA. Based on its review, the NRC staff finds that the HEP assigned for the operator action to manually isolate the containment used for the BA has minimal impact on the outcome and resulting decision.

In summary, based on the review of the submittal and the supplements, the NRC staff finds the licensee's consideration of the seismic failure of other SSCs independent of the truss for the LERF and LiLERF determination to be valid because of the nature of the plant-wide impact and resultant plant response during seismic initiating. The NRC staff also finds the licensee's approach for capturing the impact of such failures on the LERF and LiLERF determination to be acceptable for this application, in the context of the BA, because the licensee has selected and appropriately modeled the dominant containment failure contributor for seismic events from the licensee's IPEEE seismic PRA.

(iv) Conclusion on Level of Detail for Bounding Analysis Based on the NRC staff's review discussed in this section, in conjunction with the staff's conclusion on the scope of the risk assessment discussed in Section 3.5.1 of this SE, the staff finds that the licensee's assumptions used for the BA support a bounding risk assessment of a postulated truss failure due to the initiators of interest. Further, the NRC staff concludes that the OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION licensee's BA approach and implementation has adequate level of detail to (1) capture the cause-effect relationship under the assumptions used for the BA, and (2) support decisionmaking for this application. The NRC staff's review of the quantification of aCDF and aLERF from the BA are provided in Section 3.5.5 of this SE.

(b) Demonstrably Conservative Analysis (i) Context of Staff's Review of Demonstrably Conservative Analysis In letter NRC 2018-0014, the licensee stated that the results of the DCA will be included in the baseline risk of the plant for future RI applications to determine the cumulative impact of changes to the LB per the guidance in RG 1.174. In response to RAl-7 .c in Enclosure 1 of letter NRC 2018-0014, the licensee stated that "[i]t is the demonstrably conservative analysis that is requested for review by NRC staff."

As stated in Section 3.0 of this SE, the fourth principle of RI decisionmaking is that any increase in risk from proposed LB changes should be small and consistent with the intent of the Commission's Safety Goal Policy Statement. The risk acceptance guidelines in RG 1.174, Revision 2, are one acceptable approach to demonstrate that the fourth principle is met. As stated above, the NRC staff finds that the licensee's assumptions used for the BA support a bounding risk assessment of a postulated truss failure due to the initiators included in the analysis. As noted in Section 3.5.5 of this SE, the results of the BA meet the risk acceptance guidelines in RG 1.174, Revision 2, for aCDF and aLERF. Therefore, those results can support the NRC staff's integrated RI decision.

Based on the preceding discussion, the context of the NRC staff's review of the approach, assumptions, and inputs of DCA was to ensure that the analysis appropriately captured the cause-effect relationship for supporting a conservative analysis.

The DCA involved a probabilistic treatment of postulated truss failure due to seismic and thermal initiators which included the failure of mitigation systems both due to and independent of postulated falling truss debris. The licensee performed a target assessment to examine the effect of a postulated truss failure and its consequences. The DCA incorporated system dependencies and operator actions. It included consideration of random and seismically induced failures of credited mitigation systems as well as the impact of the seismic event on human actions. The remainder of this section discusses the target assessment and the approach in the DCA to evaluate acDF and aLERF.

(ii) Target Assessment In order to support DCA, the licensee performed a containment truss target assessment and submitted a report (hereafter referred to as the target assessment report; Reference 3 [publicly available] and Reference 4 [non-publicly available]) documenting the assessment in response to RAl-15.a. According to the target assessment report, the aspects of a postulated failure of the containment truss and its consequences that were considered in the assessment included how the truss would deconstruct when overstressed, the trajectory of debris from the deconstructed truss, the vulnerability of SSCs to postulated falling truss debris, and the location and robustness of the barriers that could protect such SSCs.

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION In order to assess the truss deconstruction when overstressed and the trajectory of postulated falling truss debris, the licensee assumed that bolted connections fail before welded connections. Further, it was assumed that all bolted elements, after postulated failure, would fall unrestrained, without rotation, and without striking any obstructions in their pathway towards the target (i.e., SSC). In addition, the licensee also assumed that the postulated falling truss debris would strike the target at a normal orientation (i.e., perpendicular to the target). In response to RAl-15.d.i, the licensee stated that the missile (i.e., postulated truss debris) orientation that maximized damage to the targeted SSC was a normal orientation that aligned with the smallest missile contact area and with the center of gravity of the missile. In response to RAl-15.d.ii, the licensee explained that assuming a normal orientation that maximized the damage to a targeted SSC bounded other orientations due to random detachment and disassembly of the truss elements, contact with obstacles, and rotation and spinning of the detached elements at random angles and speeds. In Section 3.3.1 of the target assessment report, the licensee clarified that the assessment assumed the worst-case consequence of being contacted by the heaviest truss member (termed the T1 truss) at a normal orientation. In response to RAl-15.b, the licensee clarified that in addition to the perforation and penetration failure modes, crushing of a fragile object was also considered as a failure mode from falling debris. The licensee cited the auxiliary feedwater (AFW) supply pipe connecting to the main feedwater pipe as an example of a failure due to crushing by postulated falling construction truss debris.

The NRC staff's review finds that, due to the configuration of the truss, assuming bolted connections to fail before welded connections is conservative because it presumes that the truss elements remain intact as very heavy missiles for subsequent consideration in the assessment. The NRC staff concludes that there is reasonable assurance that the licensee's approach of assuming the heaviest truss member as well as a normal orientation for use in the target assessment is conservative because of (i) the configuration of the truss, (ii) the random nature in which the truss elements would detach, (iii) the subsequent behavior (e.g., rotation),

and (iv) likely contact with obstacles leading to deflection as well as energy dissipation. The NRC staff also concludes that the licensee's approach bounds possible orientations of a detached truss element during contact with SSCs and addresses the likely failure modes due to such contact.

The licensee developed a list of key SSCs for the assessment along with their location in .

Section 3.4 of the target assessment report. The licensee determined the impact of postulated truss deconstruction and falling truss debris on the SSCs using empirical equations to evaluate potential damage to concrete and steel from missile impacts as well as qualitative arguments based on design drawings, three-dimensional reconstruction of the containment layout, photographs, and in-containment walkdown observations. The equations used by the licensee were derived from Bechtel Power Corporation's topical report, "Design of Structures for Missile Impact," BC-TOP-9A, Revision 2, September 1974 (Reference 37). The licensee identified and calculated the smallest contact area that would also be aligned with the center of gravity of the truss elements for use in the equations. The licensee also calculated the probability of being aligned in the orientation that maximizes the damage. In response to RAl-15.d.iii, the licensee stated that the probability of a particular orientation was qualitatively studied to develop a more realistic understanding of target damage. The NRC staff notes that the probability of a particular orientation is not used in the calculations presented in the target assessment report. In response to RAl-15.c, the licensee stated that the equations considered the ability of a target to absorb energy based on the dynamic properties of the target, support conditions, and other OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION imposed loads at the time of impact. The NRC staff, with respect to the topical report used by the licensee, has previously found that "design criteria and procedures described by this report are acceptable." Therefore, the NRC staff finds the use of the equations from the cited topical report to be acceptable for the target assessment performed to support the DCA for this application.

The NRC staff's key observations and findings from the review of the licensee's target assessment report relevant to the DCA are:

• In its assessment of the pressurizer PORVs, the licensee stated that a reinforced concrete missile shield completely covered the top of pressurizer cubicle. The licensee demonstrated that the missile shield had sufficient thickness to prevent perforation under the assumptions used for unimpeded missile impact. The licensee also explained that the underside of the missile shield had an integrated metal plate which was not accounted for in the equation and that the metal plate provided additional protection and also eliminated the potential concrete debris impacting the PORVs due to spalling.

In its response to RAl-11.e and in its letter dated August 30, 2018, the licensee stated that the modifications for the backup nitrogen supply system were completed in both units. In addition, the proposed modifications supporting this application included rerouting the existing control cables for a Unit 2 PORV and its associated block valve (Reference 7) to ensure that they were adequately protected from postulated falling debris. The NRC staff finds that there is high likelihood that the PORVs will be available to mitigate an accident under a postulated containment truss failure because of the presence of the missile shield and the integrated metal plate as well as the completed modifications for the backup nitrogen supply system.

• The (( 11 was credited by the licensee for protecting

((

11- Using design drawings and three-dimensional reconstruction of the containment layout, the licensee explained that the ((

11 that were not occupied by other SSCs.

The licensee stated that the perforation of the (( 11 was highly unlikely due to significant amount of equipment and many obstructions above the floor that would prevent an unimpeded impact by postulated falling truss debris. The licensee used three-dimensional reconstruction of the containment layout to support its argument and demonstrated that the (( 11 would not be perforated for an orientation of the falling truss debris different from the one assumed to maximize damage. The licensee further stated that the floor was supported by I-beams. The NRC staff finds that the perforation of the (( 11 and subsequent damage to the

(( 11 between the ((

11 is unlikely because of the presence of obstacles above the floor. The NRC staff further notes that even if perforation is assumed, unimpeded travel of the truss element to damage equipment below the (( 11 is highly unlikely due to the tapered shape of the truss and energy dissipation caused by the perforation.

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• The licensee performed a review of ((

)) to determine if they were vulnerable to postulated falling truss debris. Based on its review, the licensee stated that all components for ((

11 or under missile shields. According to the licensee, the same was true for the components in the other (( 11 with the exception of a ((

)). The licensee explained that redundant injection paths would remain available under a postulated failure of the conduit. The NRC staff finds that there is high likelihood that the ((

)), will be available to mitigate an accident under a postulated containment truss failure because of the location of the majority of the (( 11 and the NRC staff's conclusion for the impact on the (( )).

In RAl-15.d.iv, the NRC staff requested the licensee to provide a discussion, with justification, of the potential impact of the postulated falling truss debris on the cable holding the polar crane hook. In response to RAl-15.d.iv, the licensee provided details on the positioning of the polar crane, its trolley, and the main hoist drum during normal power operations in relation to the containment truss. The licensee stated that the main hoist bottom block for the polar crane weighed more than the heaviest truss element considered in the target assessment. The licensee provided details of the capacity and configuration of the wire rope. The licensee further explained that the load from the impact of postulated falling debris would likely be distributed across several wraps of the wire rope. The NRC staff's review of the licensee's response finds that the location of the drum with respect to the containment truss makes an unobstructed and direct impact highly unlikely. Further, the energy dissipation from spreading apart of the wire rope, the relative height of the containment truss and the crane, and the protected portion of the wire rope in the groove of the drum will further diminish the likelihood of failure of the wire rope, and therefore, the crane hook. The NRC staff concludes that the not considering the failure of the wire rope and, therefore, the polar crane hook due to postulated falling truss debris is acceptable for this application.

In Attachment G of Enclosure 2 of letter NRC 2018-0014, the licensee described the approach used to consider the potential failure of the main steam line (A and 8), AFW (A and 8) supply pipes, and the in-core instrumentation seal table from falling truss debris in the DCA. The licensee used qualitative assumptions to determine the likelihood of occurrences necessary for truss debris falling and impacting each component.

The licensee used fault trees to determine the failure probabilities for main steam line (A and 8),

AFW (A and 8) supply pipes, and the in-core instrumentation seal table from postulated falling truss debris and calculated three different failure probabilities for each component, termed the "realistic," "conservative," and "bounding," by changing certain assumptions. The "bounding" value was determined assuming that all truss elements (i.e., T1 and T2) always dislodge completely. In response to RAl-10.c, the licensee explained that while the probability that a target would be damaged when hit by postulated truss debris was independent of the seismic acceleration, the probability of truss debris generation was based on the truss fragility which did vary with seismic acceleration. Based on the discussion in Section 9.1.3 of Enclosure 2 of letter NRC 2018-0014, the licensee used the "conservative" failure probability in the revised DCA for seismic initiator. The NRC staff's review finds that the licensee's approach and assumptions OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION are based on acceptable engineering judgement supported by three-dimensional reconstruction of the containment layout and walkdown photographs. The NRC staff further finds that the impact of variability of the assumptions as well as the change in likelihood of failure with seismic acceleration is captured adequately by the assumptions used in the calculation of the "bounding" value for each component. The licensee has used the "bounding" values in sensitivity studies documented in Sections 9.1.3 of Enclosure 2 of the supplement to the submittal. The NRC staff's discussion of those sensitivities is in Section 3.5.5 of this SE.

(iii) Core Damage Frequency Evaluation The licensee provided its CDF evaluation in the original DCA approach in Enclosure 4 to the LAR. During its review, the NRC staff requested additional information in RAl-10 on multiple aspects of the DCA approach. In response to RAl-10.a, the licensee stated that the logic used to develop the event trees, including the top events, had been completely revised to address the NRC staff's requests. As noted previously, the licensee's revised risk assessment, including the revised DCA, is documented in Enclosure 2 of letter NRC 2018-0014. As stated in response to RAl-10.a, Attachment F of Enclosure 2 of letter NRC 2018-0014, provided detailed description of the inputs, assumptions, and approach used to develop the revised DCA for seismic initiator.

The revised DCA for thermal-initiating events was described in Attachment E of Enclosure 2 of letter NRC 2018-0014.

In response to RAl-10.b, the licensee stated that the event trees in the revised DCA were integrated into the licensee's internal events PRA model. Similarly, in response to RAl-6.d, the licensee stated that the revised DCA was based on the licensee's internal events and NFPA 805 fire PRAs for Units 1 and 2. The licensee explained that those PRAs were refined and augmented to address the SSCs and operator actions that could potentially be impacted by the postulated failure of the truss and that the changes included seismic hazard data, SSC fragilities, and operator actions. Discussion of the PRA models used for integration of the revised DCA can be found in Attachment F of the revised risk assessment. That attachment stated that the modification commitments made in the licensee's NFPA 805 application were included in the PRA models used for the integration of the revised DCA.

The licensee provided the event tree for the revised DCA for the seismic initiator in Section F.4 and described each top event in the event tree in Section F.5 of Attachment F of Enclosure 2 of letter NRC 2018-0014. Based on its review, key NRC staff observations on the event tree are:

• Steam line breaks (a separate top events for each steam line) model the failure of the steam lines due to impact from postulated falling debris. In response to RAl-15.a-1 in Attachment 1 of letter NRC 2018-0041, dated August 30, 2018, the licensee stated that the revised DCA for the seismic initiator assumed that the faulted steam generator was failed and not recoverable. The licensee explained that the assumption of the non-recovery of the faulted steam generator precluded the need to consider the probability and consequences of a steam vent line rupture based on thermal-hydraulic analyses for the rupture of a single and both steam line vents. The licensee stated that the thermal-hydraulic analyses confirmed that for the rupture of a single or both steam line vents, the faulted steam generator(s) remained capable of cooling the reactor coolant system for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with AFW availability.

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• The licensee included the seismic failure of the condensate storage tanks as a failure mode for the AFW. The licensee stated that the failure mode caused by block walls falling on condensate storage tank level transmitters, which was modeled in the licensee's IPEEE seismic PRA model, was no longer valid because a structural shield had been installed over the level transmitters to prevent the block walls from failing them.

• SLOCA was modeled to occur due to either the failure, during a seismic event, of the in-core instrumentation seal table from the falling truss debris or due to piping failures independent of postulated truss failure. The failure probability for the in-core instrumentation seal table in Appendix G of Enclosure 2 of letter NRC 2018-0014, and discussed above was used by the licensee. As noted earlier, the licensee used the fragility value from NUREG/CR-4840 for modeling SLOCA independent of postulated truss failure. In response to RAl-10.a-1 in Attachment 1 of letter NRC 2018-0041, dated August 30, 2018, the licensee explained that thermal-hydraulic analyses showed that a very small LOCA did not require the PORVs for F&B cooling to mitigate core damage when AFW was available for at least one steam generator.

• The licensee stated that if F&B was used for decay heat removal then recirculation would eventually be required to maintain decay heat removal. The licensee explained that switchover to recirculation would be required several hours after F&B was initiated and that the open pressurizer PORVs essentially created a SLOCA. Therefore, the licensee used the success criterion for high pressure recirculation for long term F&B to be the same as that for SLOCA.

• The licensee included, as appropriate, both seismically induced as well as random failures of mitigating systems. In addition, HEPs were also considered by the licensee.

Section F.7 of Attachment F to Enclosure 2 of the letter NRC 2018-0014, provided a review of the top events from the licensee's IPEEE seismic PRA and a basis for inclusion or exclusion of the corresponding event from the revised DCA for seismic initiator. In response to RAl-10.c, the licensee clarified that the fragilities and HEPs in the revised DCA for seismic initiator varied with seismic acceleration whereas the random failure probabilities did not.

• The licensee stated that based on a thermal-hydraulic analysis of an SLBIC, which represented the most limiting F&B case, the internal events model did not require containment heat removal for success. of F&B scenarios.

In Section F.3 of Attachment F of Enclosure 2 of letter NRC 2018-0014, the licensee provided a list of the component fragility values used in the revised DCA for seismic initiator along with the corresponding information source. The majority of the fragility values used by the licensee were from the licensee's IPEEE seismic PRA except for containment isolation and the refueling water storage tank. The licensee stated that using the IPEEE fragilities was conservative because of the structural or configuration changes since the IPEEE that increased the ability of certain SSCs to withstand structural failure during a seismic event. The licensee included seismically induced loss of offsite power in the revised DCA for seismic initiator as discussed in Section F.3 of Attachment F of Enclosure 2 of letter NRC 2018-0014. Section F.7 of Attachment F of to letter NRC 2018-0014, provided the basis for exclusion of certain top events that OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL YSE ONLY SECYRITY RELATED INFORMATION were present in the licensee's IPEEE seismic PRA from the revised DCA for seismic initiator.

As stated in its response to RAl-8.d, the licensee, in Section F.6 of Attachment F to Enclosure 2 of letter NRC 2018-0014, provided a brief description of each sequence that resulted from the revised DCA event tree for the seismic initiator. In response to RAl-12.a, the licensee stated that containment sprays were not credited for mitigation after a postulated overstress event due to the seismic initiator.

The NRC staff reviewed the event trees for loss of offsite power, SLBIC, and SLOCA as well as the fault tree for F&B in the Standardized Plant Analysis Risk model for Point Beach.

Based on its review of the information submitted by the licensee, the NRC staff finds that the licensee's revised DCA event tree that is used to evaluate the CDF associated with the seismic initiator is appropriate for this application because:

• it includes the necessary mitigation systems and actions,

• the excluded top events from the licensee's IPEEE seismic PRA do not impact this application, and

• the licensee has considered seismically induced and random failures of the modeled systems as well as HEPs, including the impact of the seismic event on HEPs, as discussed below.

The NRC staff's review finds that the licensee has integrated the event tree for the revised DCA for seismic initiator with its internal events model. The NRC staff finds the licensee's use of fragilities from its IPEEE seismic PRA acceptable for the DCA for this application because (1) the DCA does not alter the approach used to determine those values, (2) the approach was previously found to be acceptable and based on staff guidelines, and (3) the resulting values are sufficiently representative of the as-built, as-operated plant.

In response to RAl.9-a, the licensee stated that the integration of the 'plant' fragility to capture the impact of independent failures of SSCs was not applied to DCA. The NRC staff reviewed the licensee's response and finds that not considering the 'plant' fragility for the revised DCA for seismic initiator is appropriate because of the dependence between the systems modeled in the event tree and those used for the determination of the 'plant' fragility.

In Section 5.3 of Enclosure 2 of letter NRC 2018-0014, the licensee stated that the revised DCA for thermal initiators credited only F&B for decay heat removal. The licensee provided details of the modeling for the revised DCA for thermal initiators in Attachment E to Enclosure 2 of letter NRC 2018-0014. The licensee also noted that the revised DCA for thermal initiators did not credit AFW or main feedwater for decay heat removal which were expected to be available for mitigation. In response to RAl-12.a, the licensee explained that if the containment spray failed to mitigate the overstress, the truss was assumed to generate debris. The NRC staff's review finds that the licensee's approach for the CDF evaluation for the revised DCA for thermal initiators is appropriate for this application because it crediting F&B as the only mitigation approach for decay heat removal is conservative and the mitigation is only applied to those thermal initiators where such mitigation can be effective.

OFFICIAL YSE ONLY SECYRITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION (iv) Human Reliability Analysis The licensee provided details of the human reliability analysis performed to support the DCA for seismic and thermal initiators in Attachment B of Enclosure 2 to letter NRC 2018-0014. The discussion included the approach followed by the licensee to determine the HEPs that needed to be evaluated, the screening of HEPs, and consideration of the impact of seismic events in the evaluation of HEPs.

In order to determine the HEPs that needed to be evaluated, the licensee set all except three of the HEPs in the internal events PRA to a high value. The licensee excluded the HEPs for F&B at low and high seismic acceleration because a detailed calculation for those HEPs had been performed by the licensee. In addition, the licensee also excluded the HEP related to manual operation of valves for the motor driven AFW because of a modification that had been installed which supplied pneumatic supply to the valves for the PRA mission time (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) such that the action to manually control the discharge valves was no longer required. The licensee chose the HEPs that subsequently appeared in the quantified cut-sets for further evaluation of the impact of the seismic event on the HEPs. The licensee explained that additional HEPs were not identified from the LERF logic or the revised DCA. The licensee cited plant modifications that rendered two modeled HEPs unnecessary and screened those HEPs out from further evaluation. In addition, the licensee also screened out restoration errors, which occur following SSC maintenance because those actions take place before the seismic event.

In RAl-11, the NRC staff requested information on the modeling of HEPs in general and the HEP for F&B in particular for the revised DCA. In response to RAl-11.b, the licensee stated that the HEPs for the revised DCA were developed based on the detailed quantification methodology outlined in EPRI report 3002008093, "An Approach to Human Reliability Analysis for External Events with a Focus on Seismic," dated December 2016. In response to RAl-11.a, the licensee explained that the time available to initiate F&B used in the human reliability analysis was based on a thermal-hydraulic analysis of an SLBIC with AFW failure and F&B success. In response to RAl-11.c the licensee also stated that the baseline internal events HEP for F&B was not used across all seismic accelerations in the revised DCA. The information in Attachment B of Enclosure 2 to letter NRC 2018-0014, demonstrated that the licensee's approach to capture the impact of the seismic event on the internal events HEPs was dependent on the seismic acceleration range. The licensee identified all but one of the HEPs that were modified to consider the impact of the seismic event as being in-control room actions.

The single action outside the control room was required more than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the event and the action was to be performed in a seismically qualified building. The licensee also stated that several of the HEPs, identified in Attachment B of Enclosure 2 to letter NRC 2018-0014, were not risk significant and those HEPs were assumed to always fail.

In response to RAl-11.d, the licensee explained that the baseline HEP for F&B in the revised DCA did not credit instrument air availability for PORV operation because it was assumed that instrument air was unavailable during a seismic event. The licensee provided additional details of the modification to provide backup nitrogen supply for the PORVs in response to RAl-11.e.

The licensee explained that the modifications to install the nitrogen supply to the PORVs were completed in both units and that the nitrogen supply system and the instrument air system did not depend on each other or common components thereby providing redundant means for operating the PORVs. In response to RAl-11.f, the licensee stated that the nitrogen supply was automatically aligned to the PORVs upon reduced instrument air pressure to the PORVs. The OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION licensee also explained how the implementation of the backup nitrogen system modification allowed the operator to attempt opening the PORVs earlier and with fewer required actions as compared to using the instrument air system.

The NRC staff reviewed the information provided by the licensee on the HRA for this application and finds the licensee's approach for performing the HRA for the revised DCA to support this application to be acceptable because the licensee followed a systematic approach to identify the HEPs important to the application and included the impact of the seismic event on those actions. The NRC staff's acceptance of the licensee's approach for considering the impact of seismic events on the HEPs for this application is limited to this application and does not constitute a generic acceptance or endorsement of the methodology in the EPRI report cited by the licensee for any other application. The NRC staff also finds that the evaluation of the HEP for PORV operation to support F&B to be acceptable because of the installation of backup nitrogen supply for the PORVs and the resulting impact on PORV availability as well as the corresponding operator actions.

(v) Large Early Release Frequency Evaluation In Section F.2 of Attachment F of Enclosure 2 to letter NRC 2018-0014, the licensee stated that the revised DCA LERF evaluation for the seismic initiator considered the contribution from containment isolation and the containment structure. The seismic fragility of containment isolation and the containment structure were used in conjunction with the outcome of the revised DCA CDF evaluation for the seismic initiator discussed above. According to Section F.3 of Attachment F of Enclosure 2 to letter NRC 2018-0014, the containment isolation fragility value with the contribution of post-lPEEE modifications was used by the licensee for the revised DCA for seismic initiators.

The NRC staff's review finds that the licensee's approach for LERF evaluation for the revised DCA for seismic initiator is appropriate for the assessment of the postulated truss failure because it includes the necessary mitigation systems and actions, and the licensee has considered seismically induced and random failures of the modeled systems. The NRC staff also finds that the licensee has appropriately integrated the LERF evaluation with the CDF evaluation for the revised DCA for seismic initiator. The NRC staff notes that the licensee has performed a sensitivity study to determine the impact of using IPEEE fragility for containment isolation instead of the one that considers the post-lPEEE modifications. The NRC staff's discussion of that sensitivity is provided in Section 3.5.5 of this SE. As noted earlier, the integration of the plant fragility and construction truss fragility results was not applied to the revised DCA, which the NRC staff finds acceptable.

Based on Section 5.3 of Enclosure 2 to letter NRC 2018-0014, the licensee used a CLERP of 0.5 for the LERF evaluation for the revised DCA for thermal initiators. The same CLERP value was used for the LERF evaluation for the BA. The NRC staff's review and conclusions of that CLERP value are provided earlier in this section in the context of the revised BA and are applicable to the revised DCA for thermal initiators.

(vi) Conclusion on Level of Detail in Demonstrably Conservative Analysis Based on the NRC staff's review discussed in this section, in conjunction with the staff's conclusion on the scope of the risk assessment discussed in Section 3.5.1 of this SE, the NRC OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION staff finds that the licensee's revised DCA approach for seismic and thermal initiators as well as its implementation has adequate level of detail to (1) capture the cause-effect relationship under the assumptions used for DCA, and (2) support decisionmaking for this application. The NRC staff notes conservatisms in the assumptions used for revised DCA such as the assumption of failure and collapse of truss elements immediately following overstress and the collapse of the heaviest truss member in the most damaging orientation for target assessment. The NRC staff's review of to the quantification of 6CDF and 6LERF from the revised DCA are provided in Section 3.5.5 of this SE.

3.5.3 Technical Elements of the Probabilistic Risk Assessment According to the guidance in RG 1.200, Revision 2, the technical acceptability of the PRA depends on the assurance that the portions of the PRA relied upon for the application have been performed in a technically correct manner and that the assumptions and approximations used in developing the PRA are appropriate. Typically, the NRC staff relies on the PRA peer review results to obviate the need for an in-depth review of the base PRA. This allows the staff to focus on application-specific impacts, key assumptions, and areas identified by peer reviewers and self-assessments as being of concern that are relevant to the application.

In Section 3.1.2 of Enclosure 1 of the LAR, the licensee stated that a "peer review of the PRA analysis as well as the seismic and thermal fragility analyses was conducted. Peer reviews are typically performed based on NRC staff-endorsed guidance to confirm acceptability with respect to documented requirements, such as the ASME/ANS PRA Standard. In response to RAl-6.a,

• the licensee clarified that the "peer review" stated in the submittal was an independent review contracted to consultants with expertise in the methodologies applied in the analyses supporting the submittal. The licensee explained that the same independent reviewers were contracted to review the revised analyses performed in response to the staff's RAls. The licensee clarified that the ASME/ANS PRA Standard was not implemented. The NRC staff does not accept independent reviews as an alternative to a formal peer review which would follow the guidance in RG 1.200. Therefore, the NRC staff did not rely on the independent reviews performed by the licensee.

According to Section 111.2.2.4.1 of SRP Section 19.2, the NRC staff performs a focused-scope evaluation that concentrates on application-specific attributes of the PRA and on the assumptions and elements of the PRA model that drive the results and conclusions. The NRC staff reviewed the assumptions, inputs, approach, and implementation for the revised BA and DCA for thermal and seismic initiators supporting this application. The discussion of the NRC staff's review and conclusions on scope and level of detail of the revised BA and DCA for thermal and seismic initiators is provided in Sections 3.5.1 and 3.5.2, respectively, of this SE.

The NRC staff concludes that under the assumptions used for the BA for the seismic initiator, a PRA as defined by RG 1.200, Revision 2, is not necessary for this application.

The revised BA for thermal initiators and the revised DCA for thermal and seismic initiators used the licensee's internal events and NFPA 805 fire PRAs. During its review of the licensee's NFPA 805 application, the NRC staff found the licensee's internal events, including internal flooding, and fire PRAs to meet the guidance in RG 1.200, Revision 2, and to be technically acceptable. The NRC staff's conclusion on the technical acceptability of the internal events PRA model from the NFPA 805 review is applicable to this submittal because those PRAs are used as the base for this application. In Attachment A of Enclosure 2 to letter NRC 2018-0014, OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION the licensee listed and provided dispositions for finding level facts and observations, in the context of this application, from the licensee's internal events, including internal flooding, and fire PRA models. The NRC staff's review of the information in Attachment A of Enclosure 2 to letter NRC 2018-0014, found that the finding level facts and observations either do not impact this application or have been appropriately dispositioned by the licensee for this application.

Section 9.3 of Enclosure 2 to letter NRC 2018-0014, described the licensee's approach to identify model uncertainties and related assumptions that were "key to the decision" for this application. The licensee's approach included identifying sources of model uncertainty and related assumptions through a review of the base PRA as well as identifying new sources of model uncertainty and related assumptions that may be introduced by the application. The licensee reviewed the identified sources for their relevance to this application and to determine if they were key to the decision. Each identified key source of uncertainty was dispositioned by performing a quantitative sensitivity analysis to determine the impact on this application. The licensee provided the results of its review in Section 9.3 of Enclosure 2 to letter NRC 2018-0014. The sensitivities performed for the identified key sources or assumptions did not show an appreciable impact on this application. Based on its review, the NRC staff finds that the licensee searched for, identified, and evaluated sources of uncertainty in the base PRA consistent with RG 1.200, Revision 2, and performed appropriate sensitivity studies as input to the decisionmaking associated with this application. The NRC staff concludes that the licensee provided sufficient information to appropriately disposition the identified key assumptions and sources of uncertainty in the context of the revised DCA for this application.

The licensee provided "statements of compliance" against high-level requirements and supporting requirements from the 2009 ASME/ANS PRA Standard, Parts 5 (seismic PRA) and 6 (external hazard screening), in Attachment C of Enclosure 2 to letter NRC 2018-0014. In response to RAl-6.b, the licensee stated that the "statements of compliance" was the "in-house assessment of the seismic supporting requirements relative to the analyses and methodology applied in the PRA and structural calculations." Therefore, the information provided in Attachment C of Enclosure 2 to letter NRC 2018-0014, regarding "compliance" against the high-level requirements and supporting requirements for Parts 5 and 6 in the 2009 ASME/ANS PRA Standard was not reviewed by the NRC staff or considered in the staff's decisionmaking.

In summary, based on its review, the NRC staff concludes that the revised BA for thermal initiators and the revised DCA for thermal and seismic initiators, under the corresponding assumptions and in conjunction with the NRC staff's previous review of the licensee's internal events and fire PRAs, conform to technical elements to support decisionmaking for this application.

3.5.4 Conclusion on Technical Elements of the Probabilistic Risk Assessment The NRC staff's focused-scope application specific review, in accordance with the guidance in SRP Section 19.2 and discussed in detail in Section 3.3.3.3 of this SE, finds that:

• Under the assumptions used for the BA for the seismic initiator, a PRA as defined by RG 1.200, Revision 2, is not necessary for this application and the licensee's approach captures the cause-effect relationship adequately.

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• The PRA developed to support the revised BA for thermal initiators as well as the revised DCA for seismic and thermal initiators adequately captures the cause-effect relationships under the corresponding assumptions.

• The PRA used to support the revised BA for thermal initiators as well as the revised DCA for seismic and thermal initiators is acceptable to support decisionmaking for this application.

• The scope and quality of the engineering analyses performed to justify the proposed changes, including those supporting the revised BA and DCA for thermal and seismic initiators, are appropriate for the nature as well as scope of the changes, and are sufficiently plant-specific.

3.5.5 Comparison Against Acceptance Guidelines The licensee has used CDF and LERF, which are the risk metrics used in RG 1.174, Revision 2, to quantify the incremental risk from the proposed change and compared the same against the acceptance guidelines in Section 2.4 of RG 1.174, Revision 2.

The licensee calculated 6CDF due to the seismic initiator using a base case which was based on the configuration of the Units 1 and 2 trusses that included all modifications necessary to meet ANSI/AISC N690-1994 (R2004) limits without exception. Therefore, the base case represented a plant that was hypothetically compliant with the limits of the proposed alternate code. The change in CDF for the seismic initiator was determined using the proposed configuration of the trusses, which included the proposed modifications for the Unit 1 truss. This approach for calculating 6CDF for the seismic initiator was used for the revised BA and DCA.

In response to RAl-8.b, the licensee provided additional information on the base case used for the calculations in the submittal. The licensee summarized all the modifications that would have been necessary to meet ANSI/AISC N690-1994 (R2004) limits. The licensee explained that those modifications would have eliminated contact with the containment liner under seismic loading and that the resulting capacity was used to support the determination of the change in risk as compared to the proposed configuration.

The 6CDF calculation for the thermal initiators used a base case CDF of 0.0 (zero) because the configuration of the Units 1 and 2 trusses that included all modifications necessary to meet ANSI/AISC N690-1994 (R2004) limits would not result in any contact with the liner and therefore, thermal overstress of the trusses. The use of a CDF of 0.0 to calculate the increase in risk due to the thermal initiators can also be considered to be conservative. Due to this approach, the increase in the CDF for thermal initiators was identical to the CDF for the thermal initiators, which included the proposed modifications for the Unit 1 truss. This approach for calculating 6CDF for the thermal initiators was used for the revised BA and DCA.

The licensee's approach to calculate LERF for the revised BA and DCA for seismic and thermal initiators is described in Section 3.3.3.2 of this SE. The 6LERF calculation for the seismic and thermal initiators for BA and DCA uses the same base cases as those described above for the 6CDF calculations the revised.

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION The NRC staff's review of the licensee's approach for calculating the change in risk for this application finds the base cases used for the f1CDF and f1LERF calculations to be appropriate for this application because they capture the incremental risk under seismic and thermal loading conditions due to the proposed configuration (i.e., not performing all the modification necessary to the limits of the proposed alternate code without exception). Further, the base cases use the same alternate code as that proposed by the licensee in the submittal resulting in consistent calculation of the incremental risk.

3.5.6 Bounding Analysis As noted in its response to RAl-8.e, the licensee provided the results from the revised BA for the seismic and thermal initiators in Section 3 of Enclosure 2 to letter NRC 2018-0014. Key results are reproduced in the Table 2 below.

The guidance in RG 1.174, Revision 2, states that if the change in CDF and LERF due to the proposed change to the LB falls in Region II (which corresponds to an increase in CDF greater than 1x10-6 per year but less than or equal to 1x10-5 per year or an increase in LERF greater than 1x10-1 per year but less than or equal to 1x10-6 per year) the total baseline CDF and LERF should also be provided. In Table 1 and Section 6 of Enclosure 4 of the LAR, the licensee included the CDF and LERF values from different hazards to determine the total baseline CDF and LERF. However, neither the sources of the values nor the status of the peer-review of the sources was specified.

Table 2: Key Results from the Bounding Analysis Units 1 and 2 Initiator ACDF (/year) ALERF (/year)

Seismic 1.87x10-6 7.30x10-7 Thermal 4.83x10-a 2.42x10-a Seismic + Thermal 1.93x10-6 7.55x10-7 In response to RAl-6.c, the licensee provided the total baseline CDF and LERF values including the source as well as the CDF and LERF values, for individual hazards that contribute to the total baseline values. The licensee stated that there have not been subsequent modifications to the associated PRAs that were considered "upgrades." The NRC staff compared the information presented in the licensee's response to RAl-6.c against that in Table 3.4.6 of the staff's September 8, 2016, letter approving the licensee's transition to NFPA 805 as well as a separate application from the licensee (Reference 38). The NRC staffs September 8, 2016, letter approving the licensee's transition to NFPA 805 found the licensee's internal events, including internal flooding, and fire PRAs to be acceptable to support that application. The NRC staff notes that the conclusion on the technical acceptability of the licensee's PRA models from previous review is applicable to this submittal because the licensee's internal events and fire PRAs form the base for the risk assessments supporting this application. The NRC staff finds that the baseline CDF and LERF presented by the licensee are acceptable for this application because they represent the most current baseline risk of each unit.

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OFFICIAL USE ONLY SECURITY RELATED INFORMATION ln response to RAl-13.a, RAl-13.b, and in Section 9 of Enclosure 2 to letter NRC 2018-0014, the licensee provided additional information on sensitivity studies performed for revised BA. The licensee performed a sensitivity by using the 5th and the 95th percentile seismic hazard curves for BA for seismic initiator. The licensee also performed a sensitivity using the probability distributions for the failure probability of SSCs and human actions for the revised BA for thermal initiators which addressed parametric uncertainty. As noted in Section 3.1.2 of this SE and in Section 9.1.2.2 of Enclosure 2 to letter NRC 2018-0014, the licensee performed a sensitivity to quantitatively demonstrate the impact of changing the failure probability as a function of temperature by an order of magnitude on the results of the revised BA for thermal initiators.

The NRC staff's review finds that the results of the revised BA when considered with the baseline risk are in Region II of the RG 1.174, Revision 2, acceptance guidelines and principle 4 of RI decisionmaking is met. The NRC staff also finds that the licensee has performed sensitivities for important inputs to BA for seismic and thermal initiators and the sensitivities demonstrate that results of the revised BA for seismic and thermal initiators, as compared with the acceptance guidelines in RG 1.174, Revision 2, are robust in the context of this application.

The NRC staff's review of the scope, level of detail, and risk impact of the BA, in conjunction with the supporting engineering evaluations (see Sections 3.1.1 and 3.1.2 of this SE) as well as defense-in-depth and safety margins considerations (see Sections 3.3.1 and 3.3.2 of this SE),

finds the proposed modifications to the Unit 1 truss and no modification to the Unit 2 truss to be acceptable because the incremental risk from performing only the modifications included in the implementation items in the proposed license conditions meets the acceptance guidelines, and therefore principle 4, in RG 1.174, Revision 2.

3.5. 7 Demonstrably Conservative Analysis As noted in its response to RAl-8.e, the licensee provided the results from the revised DCA for seismic and thermal initiators Section 6 of Enclosure 2 to letter NRC 2018-0014. Key results are reproduced in the Table 3 below.

The guidance in RG 1.174, Revision 2, states that if the change in CDF and LERF due to the proposed change to the LB falls in Region Ill (which corresponds to an increase in CDF less OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION than 1x10-6 per year or an increase in LERF less than 1x10-7 per year the total) the baseline CDF and LERF need not be considered.

Table 3: Key Results from the Demonstrably Conservative Analysis Unit 1 Initiator flCDF (/year) ~LERF (/year)

Seismic 2.15x10-7 5.15x10-s Thermal 1.48x10-9 7.40x10-10 Seismic + Thermal 2.17x10-7 5.22x1Q-8 Unit2 Initiator ~CDF (/year) ~LERF (/year)

Seismic 2.17x10-7 5.19x10-s Thermal 1.48x10-9 7.40x10-10 Seismic + Thermal 2.19x10-7 5.26x10-s In response to RAl-13.c and RAl-13.d, the licensee performed sensitivities using the probability distributions for the failure probability of SSCs and human actions for the revised DCA for seismic and thermal initiators. Section 9 of Enclosure 2 to letter NRC 2018-0014 provided details of additional sensitivities performed by the licensee to determine the resulting impact on DCA for the seismic initiator. Those sensitivities included:

• Decreasing the median fragility of the proposed and the hypothetically compliant truss configurations (i.e., increasing the failure probability of the truss at various seismic accelerations).

• Using fragility values from the licensee's IPEEE wherever updated fragilities were used for components.

• Using different failure probabilities, specifically the values referred to as realistic and bounding by the licensee, for main steam line (A and B), AFW (A and B) supply pipes, and the in-core instrumentation seal table failure from falling truss debris (see discussion on target assessment in Section 3.3.3.2 of this SE).

• Increasing the likelihood of failure of containment penetrations used in the revised DCA by a factor of 5.

• Assuming that failure of both main steam lines cannot be mitigated by F&B and leads directly to core damage.

Using the fragility values from the licensee's IPEEE wherever updated or generic fragilities were used for components and increasing the likelihood of failure of containment penetrations used in the revised DCA by a factor of 5 showed noticeable impact on the results of DCA for seismic initiator. However, the results continued to remain within Region Ill of RG 1.174, Revision 2, acceptance guidelines. The licensee performed the sensitivities using different failure probabilities for main steam line (A and B), AFW (A and B) supply pipes, and the in-core instrumentation seal table failure from falling truss debris individually. The NRC staff notes that OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION combining the 6CDF and 6LERF from the individual sensitivities for the "bounding" values of the failure probabilities causes the result to slightly exceed the 6CDF and 6LERF thresholds for Region 111 of the RG 1.174, Revision 2, but do not exceed any of the acceptance guidelines. As noted in Section 3.1.2 of this SE and in Section 9.1.2.2 of Enclosure 2 to letter NRC 2018-0014, the licensee performed a sensitivity to quantitatively demonstrate the impact of changing the failure probability as a function of temperature by an order of magnitude on the results of the revised DCA for thermal initiators. The licensee also performed a truncation sensitivity for the revised DCA for seismic and thermal initiators to demonstrate convergence at the truncation selected for quantification of results from the revised DCA.

The NRC staff's review finds that the results of DCA are in Region Ill of the RG 1.174, Revision 2, acceptance guidelines and principle 4 of RI decisionmaking is met. The NRC staff also finds that the licensee has performed sensitivities for important inputs to DCA for seismic and thermal initiators and the sensitivities demonstrated that results of DCA for seismic and thermal initiators, as compared with the acceptance guidelines in RG 1.174, Revision 2, are acceptable in the context of this application.

As stated in the discussion in the beginning of the Enclosure 1 to letter NRC 2018-0014, as well as the response to RAl-7.b in the same enclosure, the licensee will include the CDF and LERF values from the DCA for seismic and thermal initiators for the proposed configuration of the truss (i.e., the configuration following the modifications in the proposed license condition) in the plant's baseline risk for future RI submittals. Based on the results presented in Section 6 of to the Enclosure 2 to letter NRC 2018-0014, the CDF and LERF from the revised DCA for the configuration of Unit 1 truss subsequent to the completion of the relevant modifications included in the implementation items in the proposed license conditions considering both the seismic and thermal initiators are 1.23x10-6 per year and 5.92x10-7 per year, respectively. The corresponding values for the Unit 2 truss subsequent to the completion of the relevant modifications included in the implementation items in the proposed license conditions are 1.24x10-6 per year and 5.95x10-7 per year, respectively.

3.6 Performance Measurement Strategies - Principle 5 The licensee stated that the station structural monitoring program, procedure NP 7.7.9, would continue to perform routine visual examinations of the construction trusses and their supports.

Further, one of the proposed modifications included the implementation of proposed new seismic and thermal operating limits applicable to both units. The NRC staff's review of the acceptability of the proposed new operating limits in the context of this application is provided in Section 3.1.5 of this SE. In response to RAl-2, the licensee stated that existing site procedures did not require a dual-unit backdown to hot shutdown based on any specific seismic criteria.

The proposed new seismic and thermal operating limits establish a threshold whereby both units were positioned into a condition that would permit inspection and evaluation of the trusses, equipment supported by the trusses, and the containment liner, as necessary, for any event exceeding the proposed new operating limits. The inspection and evaluation would evaluate the affected structures/components before returning the affected units to power operation. The NRC staff finds that the proposed new seismic and thermal operating limits, in conjunction with the existing structural monitoring program, provide reasonable assurance that performance degradation will be detected and corrective actions will be taken to ensure that defense-in-depth is preserved and safety margins are maintained.

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USI! ONLY S&CURITY R&LAT&D INFORMATION 3.7 Integrated Decisionmaking RG 1.174, Revision 2, supports integrated decision making during the evaluation of RI changes to the LB. According to SRP Section 19.2, integrated decisionmaking uses the results of traditional engineering evaluations supported by insights, derived through the use of PRA methods, on the risk significance of the proposed changes, and may build upon qualitative factors.

The licensee, via its submittal and supplements, has addressed all five principles of RI decisionmaking. Based on its integrated review, as documented in this SE, the NRC staff finds that:

• The licensee has requested an acceptable regulatory action and sufficient information was provided to support the conclusions regarding the acceptability of the proposed changes.

• Pertinent information on the change to the LB will be included in UFSAR subsequent to the completion of the implementation items in the proposed license conditions.

• The licensee has used results from traditional engineering analyses and risk assessments to demonstrate that the principles for RI decisionmaking have been met.

• The licensee has used a systematic process to integrate traditional and probabilistic engineering evaluations for RI decisionmaking.

• The licensee performed engineering evaluations and risk assessments that are representative of the plant subsequent to the completion of the implementation items in the proposed license conditions.

• The licensee has used risk-management practices and included monitoring strategies that, subsequent to the completion of the implementation items in the proposed license conditions, will provide confidence that the results of the underlying engineering analyses remain valid.

• The licensee has demonstrated that defense-in-depth and safety margins will continue to be maintained following the proposed changes as well as after the completion of the implementation items in the proposed license conditions.

4.0 LICENSE CONDITIONS The NRC staffs findings on the RI approach proposed by the licensee in its submittal and supplements is conditioned on the completion of the modifications that were included in the engineering evaluations and risk assessments supporting this application.

The licensee proposed license conditions regarding the implementation of the modifications prior to the change to its LB allowed by the approval of this application. The modifications, as well as the corresponding completion schedule, are listed in Attachment 3 of letter NRC 2019-0007, dated March 13, 2019.

OFFICIAL US& ONLY S&CURITY R&LAT&D INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION The following license conditions related to this application is included in the Point Beach, Units 1 and 2 operating licenses as Paragraphs 4. I and 4. H, respectively:

4.1 Containment Building Construction Truss NextEra Energy Point Beach shall complete implementation items 1, 2, 3, 5, and 6 included in Attachment 3 of licensee letter NRC 2019-0007, dated March 13, 2019, by the completion date specified in the Attachment.

4.H Containment Building Construction Truss NextEra Energy Point Beach shall complete implementation items 3, 4, 5, and 6 included in Attachment 3 of licensee letter NRC 2019-0007, dated March 13, 2019, by the completion date specified in the Attachment.

Based on its review and evaluation discussed in this SE, the NRC staff concludes that the proposed license conditions and the implementation items, as well as schedule therein, are acceptable because they adequately implement the modifications to the units as well as the changes to the LB consistent with the staff's review. The staff notes that, due to the license condition, prior NRC approval is required for any changes to the implementation items listed in Attachment 3 of licensee letter NRC 2019-0007, dated March 13, 2019.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Wisconsin State official was notified of the proposed issuance of the amendments on November 30, 2018. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

These 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. 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 that these amendments involve no significant hazards consideration and there has been no public comment on such finding (82 FR 27890). Accordingly, these 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 these amendments.

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7.0 CONCLUSION

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

8.0 REFERENCES

1. Coffey R., NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, License Amendment Request 278, Risk-Informed Approach to Resolve Construction Truss Design Code Nonconformances," NRC 2017-0017, dated March 31, 2017 (Agencywide Document Access and Management System (ADAMS) Accession No. ML17090A511 ).

2. Craven R., NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, NextEra Energy Point Beach, LLC, Construction Truss License Amendment Request 278, Response to Request for Additional Information," NRC 2018-0014, dated April 12, 2018 (ADAMS Accession No. ML181028164).

3. Craven R., NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, NextEra Energy Point Beach, LLC, Construction Truss License Amendment Request Document Transmittal,"

NRC 2018-0001, dated April 12, 2018 (ADAMS Accession No. ML181028173).

4. NextEra Energy Point Beach, LLC, "Point Beach Construction Truss Target Assessment," Report PBN-BFJR-17-020, Revision 1, dated April 6, 2018; Enclosure 1 to NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission NRC 2018-0001 (non-publicly available; security-related information).

5. Craven R., NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, NextEra Energy Point Beach, LLC, Construction Truss License Amendment Request 278, Draft Updated Final Safety Analysis Report (UFSAR) Revision," NRC 2018-0030, dated May 29, 2018 (ADAMS Accession No. ML18149A466).

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OFFICIAL USE ONLY SECURITY RELATED INFORMATION

6. Craven R., NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Response to Second Round Request for Additional Information Regarding Point Beach Nuclear Plant, Units 1 and 2, License Amendment Request for Risk-Informed Approach to Resolve Construction Truss Design Code Nonconformances, (EPID L-2017-LLA-0209)," NRC 2018-0041, dated August 30, 2018 (ADAMS Accession No. ML18242A572).

7. Craven R., NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission, "Point Beach Nuclear Plant, Units 1 and 2, License Conditions and Implementation Items Required to Resolve Construction Truss Design Code Nonconformances," Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, NRC 2019-0007, dated March 13, 2019 (ADAMS Accession No. ML19072A179).

8. Cameron, J., U.S. Nuclear Regulatory Commission, letter to Eric McCartney, NextEra Energy Point Beach, LLC, "Point Beach Nuclear Plant, Units 1 and 2, NRC Integrated Inspection Report 05000266/2014004; 05000301/2014004; and 07200005/2014001,"

dated October 30, 2014 (ADAMS Accession No. ML14303A355).

9. U.S. Nuclear Regulatory Commission, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Regulatory Guide 1.174, Revision 2, May 2011 (ADAMS Accession No. ML100910006).

10. U.S. Nuclear Regulatory Commission, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Regulatory Guide 1.174, Revision 3, January 2018 (ADAMS Accession No. ML17317A256).

11. Chawla, M., U.S. Nuclear Regulatory Commission, letter to Robert Coffey, NextEra Energy Point Beach, LLC, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (CAC Nos. MF2372 and MF2373)," dated September 8, 2016 (ADAMS Accession No. ML16196A093).

12. Coffey R., NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, Periodic Update of the Final Safety Analysis Report- Chapter 1: Introduction and Summary," NRC 2016-0034, dated September 1, 2016 (ADAMS Accession No. ML16251A150).

13. U.S. Nuclear Regulatory Commission, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

Regulatory Guide 1.200, Revision 2, March 2009 (ADAMS Accession No. ML090410014).

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

14. U.S. Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," NUREG-0800, Section 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load," Revision 3, September 2012 (ADAMS Accession No. ML12193A107).

15. U.S. Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," NUREG-0800, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance," Revision 0, June 2007 (ADAMS Accession No. ML071700658).

16. U.S. Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," NUREG-0800, Section 3.8.4, "Other Seismic Category I Structures," Revision 4, September 2013 (ADAMS Accession No. ML13198A258).

17. U.S. Nuclear Regulatory Commission, "Safety Goals for the Operations of Nuclear Power Plants: Policy Statement," Federal Register, Volume 51, p. 30028 (51 FR 30028), August 4, 1986.

18. Chawla, M., U.S. Nuclear Regulatory Commission, letter to Robert Coffey, NextEra Energy Point Beach, LLC, "Point Beach Nuclear Plant, Units 1 and 2 - Audit Plan and Setup of Online Reference Portal for License Amendment Request - Risk-Informed Approach to Resolve Construction Truss Design Code Nonconformances (CAC Nos. MF9532 and MF9533)," dated July 25, 2017 (ADAMS Accession No. ML17202G591 ).

19. Chawla, M., U.S. Nuclear Regulatory Commission, letter to Robert Craven, R., NextEra Energy Point Beach, LLC, "Summary Report for NRC Audit for Point Beach Nuclear Plant, Units 1 and 2, License Amendment Request to Resolve Nonconformance Relating to Containment Dome Truss (EPID L-2017-LLA-0209)," dated March 6, 2018 (ADAMS Accession No. ML18038A785).

20. McCartney, E., NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, NextEra Energy Point Beach, LLC Seismic Hazard and Screening Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident,"

NRC 2014-0024, dated March 31, 2014 (ADAMS Accession No. ML14090A275).

21. Govan, T., U.S. Nuclear Regulatory Commission, letter to Eric McCartney, NextEra Energy Point Beach, LLC, "Point Beach Nuclear Plant, Units 1 and 2 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (TAC Nos. MF3959 and MF3960)," dated August 3, 2015 (ADAMS Accession No. ML15211A593).

OFFICIAL USE ONLY SECURITY RELATED INFORMATION

OFFICIAb USE ONbY SECURITY REbATED INFORMATION

22. U.S. Nuclear Regulatory Commission, "Evaluation of the Seismic Design Criteria in ASCE/SEI Standard 43-05 for Application to Nuclear Power Plants," NUREG/CR-6926, March 2007, available electronically from https://www.nrc.gov/reading-rm/doc-co1lections/nuregs/contracUcr6926/.

23. U.S. Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," NUREG-0800, Section 3. 7.1, "Seismic Design Parameters," Revision 4, December 2014 (ADAMS Accession No. ML14198A460).

24. U.S. Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," NUREG-0800, Section 3.7.2, "Seismic System Analysis," Revision 4, September 2013 (ADAMS Accession No. ML13198A223).

25. U.S. Nuclear Regulatory Commission, "Damping Values for Seismic Design of Nuclear Power Plants," Regulatory Guide 1.61, Revision 1, March 2007 (ADAMS Accession No. ML070260029).

26. Electric Power Research Institute, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," EPRI NP-6041-SL, Revision 1, August 1991.

27. Electric Power Research Institute, "Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1:

Seismic," EPRI 1025287, Final Report, February 2013.

28. Coffey R., NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, Periodic Update of the Final Safety Analysis Report- Chapter 14: Safety Analysis," NRC 2016-0034, dated September 1, 2016 (ADAMS Accession Nos. ML16251A136 and ML16251A166).

29. U.S. Nuclear Regulatory Commission, "Combined License Applications for Nuclear Power Plants," Regulatory Guide 1.206, June 2007, available electronically from https://www.nrc.gov/read ing-rm/doc-col lections/reg-g uides/power-reactors/rg/01-206/.

30. U.S. Nuclear Regulatory Commission, "Seismic Design Classification for Nuclear Power Plants," Regulatory Guide 1.29, Revision 5, July 2016 (ADAMS Accession No. ML16118A148).

31. Chawla, M., U.S. Nuclear Regulatory Commission, "Summary of July 12, 2018, Meeting with NextEra Energy Point Beach, LLC, to Discuss Supplemental Response to the Request for Additional Information for Containment Dome Truss License Amendment Request (CAC Nos. MF9532 and MF9533; EPID L-2017-LLA-0209)," dated August 13, 2018 (ADAMS Accession No. ML182206019).

OFFICIAb USE ONbY SECURITY REbATED INFORMATION

OfflCIAL USE ONLY SECURITY RELATED INFORMATION

32. U.S. Nuclear Regulatory Commission, "Procedures for the External Event Core Damage Frequency Analyses for NUREG-1150," NUREG/CR-4840, November 1990 (ADAMS Accession No. ML063460465).

33. U.S. Nuclear Regulatory Commission, "Risk Assessment of Operational Events Handbook Volume 2 - External Events," Revision 1.01, January 2008 (ADAMS Accession No. ML080300179).

34. Highland, P., memorandum to Brian Sheron, dated September 2, 2010, U.S. Nuclear Regulatory Commission, "Safety/Risk Assessment Results for Generic Issue 199, "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants - Appendix C: Plant-Level Fragility Data," August 2010 (ADAMS Accession Nos. ML100270598 and ML100270731 ).

35. U.S. Nuclear Regulatory Commission, "Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants," NUREG/CR-6928, Initiating Events 2015 Summary, available at https://nrcoe.inl.gov/resultsdb/publicdocs/AvqPerf/lnitiatingEvents2015.pdf.

36. U.S. Nuclear Regulatory Commission, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," NUREG-1407, June 1991 (ADAMS Accession No. ML063550238).

37. Bechtel Power Corporation, "Design of Structures for Missile Impact," BC-TOP-9A, Revision 2, September 1974 (ADAMS Accession No. ML14093A217).

38. Coffey R., NextEra Energy Point Beach, LLC, letter to U.S. Nuclear Regulatory Commission, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, License Amendment Request 287 Application to Adopt 10 CFR 50.69, 'Risk-informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants,"'

NRC 2017-0043, dated August 31, 2017 (ADAMS Accession No. ML17243A201).

Principal Contributors: Shilp K. Vasavada, NRR/DRA Ching Ng, NRR/DRA Dan V. Hoang, NRR/DE Kamal A. Manely, NRR/DE Fred Forsaty, NRR/DSS Date: March 26, 2019 OfflCIAL USE ONLY SECURITY RELATED INFORMATION

ML18340A088 (Security-Related);

ML18345A110 (non-Security-Related) *via email **via SE memo OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DRA/APLB/RILIT** NRR/DRA/APLB/TL **

NAME MChawla (RHaskell for) SRohrer (JBurkhardt for) SVasavada MReisiFard DATE 12/17/18 12/17/18 10/23/18 10/23/18 OFFICE NRR/DRA/APHB** NRR/DE/ESEB** NRR/DE/ESEB/SLA** NRR/DSS/SRXB**

NAME NChing DHoang KManoly FForsaty DATE 10/10/18 10/10/18 10/10/18 10/09/18 OFFICE NRR/DSS/SRXB/BC** NRR/DE/ESEB/BC** NRR/DSS/SCPB/BC* OGC (NLO)*

NAME JWhitman* BWittick* SAnderson BHarris DATE 10/05/18 10/12/18 11/13/18 03/05/19 OFFICE NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NRR/DORL/LPL3/PM NAME DWrona RHaskell MChawla DATE 03/26/19 03/26/19 03/26/19