ML20231A632

Issuance of Amendment No. 224 Regarding Adoption of 10 CFR 50.69, Risk Informed Categorization and Treatment of Structures, Systems and Components of Nuclear Power Reactors

ML20231A632
Person / Time
Site:	Hope Creek
Issue date:	09/29/2020
From:	James Kim Plant Licensing Branch 1
To:	Carr E Public Service Enterprise Group
Kim J
References
EPID L-2019-LLA-0265
Download: ML20231A632 (37)

Text

September 29, 2020 Mr. Eric Carr President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION ISSUANCE OF AMENDMENT NO. 224 REGARDING ADOPTION OF 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS OF NUCLEAR POWER REACTORS (EPID L-2019-LLA-0265)

Dear Mr. Carr:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 224 to Renewed Facility Operating License No. NPF-57 for the Hope Creek Generating Station (Hope Creek) in response to your application dated November 25, 2019, as supplemented by letters dated June 25, 2020, and July 21, 2020.

The amendment adds a new license condition to the Hope Creek Renewed Facility Operating License to allow the implementation of the risk-informed categorization and treatment of structures, systems, and components of nuclear power reactors in accordance with Title 10 of the Code of Federal Regulations Section 50.69.

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

Sincerely,

/RA/

James S. Kim, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-354

Enclosures:

1. Amendment No. 224 to Renewed License No. NPF-57

2. Safety Evaluation cc: Listserv

PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 224 Renewed License No. NPF-57

1.

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

A.

The application for amendment filed by PSEG Nuclear LLC dated November 25, 2019, as supplemented by letters dated June 25, 2020, and July 21, 2020 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; D.

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

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

2.

Accordingly, by Amendment No. 224, Renewed Facility Operating License No. NPF-57 is hereby amended to authorize use of a risk-informed process for the categorization and treatment of structures, systems, and components as set forth in the licensees application dated November 25, 2019, as supplemented by letters dated June 25, 2020, and July 21, 2020, and evaluated in the NRC staffs safety evaluation enclosed with this amendment.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: September 29, 2020 James G.

Danna Digitally signed by James G. Danna Date: 2020.09.29 10:21:03 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 224 HOPE CREEK GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following pages of the Renewed Facility Operating License with the revised pages.

The revised pages are identified by amendment number and contain a marginal line indicating the area of change.

Remove Insert 15 15 16 16 17 17

a.

Submit a report to the NRC staff in accordance with 10 CFR 50.4 describing the final drain line configuration and summarizing the testing results that demonstrate drainage has been established for all four quadrants.

b.

Monitor penetration sleeve J13 daily for water leakage when the reactor cavity is flooded up. In addition, perform a walkdown of the torus room to detect any leakage from other drywell penetrations. These actions shall continue until corrective actions are taken to prevent leakage through J13 or through the four air gap drains.

c.

Perform UT measurements of the drywell shell between elevation 86'-11" (floor of the drywell concrete) and elevation 93'-0" (bottom of penetration J13) below penetration J13 area during the next three refueling outages. In addition, UT measurements shall be performed around the full 360 degree circumference of the drywell between elevations 86'-11" and 88'-0" (underside of the torus down comer vent piping penetrations). The results of the UT measurements will be used to identify drywell surfaces requiring augmented inspections in accordance with IWE requirements for the period of extended operation, establish a corrosion rate, and demonstrate that the effects of aging will be adequately managed such that the drywell can perform its intended function until April 11, 2046. Within 90 days of completion of each refueling outage, submit a report to the NRC staff in accordance with 10 CFR 50.4 summarizing the results from the UT measurements and if appropriate, corrective action.

(28)

PSEG is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using:

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in PSEG submittal letter dated November 25, 2019, and all its subsequent associated supplements, as specified in License Amendment No. 224 dated September 29, 2020.

PSEG will complete the implementation items listed in Attachment 1 of PSEG's letter to the NRC dated July 21, 2020, prior to crediting portable FLEX equipment for 10 CFR 50.69 categorization.

Renewed License No. NPF-57 Amendment No. 224

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from the alternate seismic approach (referenced above) to a seismic probabilistic risk assessment approach).

D.

The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. An exemption from the criticality alarm requirements of 10 CFR 70.24 was granted in Special Nuclear Material License No. 1953, dated August 21, 1985. This exemption is described in Section 9.1 of Supplement No. 5 to the SER. This previously granted exemption is continued in this renewed operating license. An exemption from certain requirements of Appendix A to 10 CFR Part 50, is described in Supplement No. 5 to the SER.

This exemption is a schedular exemption to the requirements of General Design Criterion 64, permitting delaying functionality of the Turbine Building Circulating Water System-Radiation Monitoring System until 5 percent power for local indication, and until 120 days after fuel load for control room indication (Appendix R of SSER 5). Exemptions from certain requirements of Appendix J to 10 CFR Part 50, are described in Supplement No. 5 to the SER. These include an exemption from the requirement of Appendix J, exempting main steam isolation valve leak-rate testing at 1.10 Pa (Section 6.2.6 of SSER 5); an exemption from Appendix J, exempting Type C testing on traversing incore probe system shear valves (Section 6.2.6 of SSER 5); an exemption from Appendix J, exempting Type C testing for instrument lines and lines containing excess flow check valves (Section 6.2.6 of SSER 5); and an exemption from Appendix J, exempting Type C testing of thermal relief valves (Section 6.2.6 of SSER 5).

These exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. These exemptions are hereby granted. The special circumstances regarding each exemption are identified in the referenced section of the safety evaluation report and the supplements thereto. These exemptions are granted pursuant to 10 CFR 50.12. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

E.

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, submitted by letter dated May 19, 2006 are entitled: Salem-Hope Creek Nuclear Generating Station Security Training and Qualification Plan, and Salem-Hope Creek Nuclear Generating Station Security Contingency Plan. The plans contain Safeguards Information protected under 10 CFR 73.21.

Renewed License No. NPF-57 Amendment No. 224

PSEG Nuclear LLC shall fully implement and maintain in effect all provisions of the Commission-approved Cyber Security Plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Salem-Hope Creek CSP was approved by License Amendment No. 189 as supplemented by changes approved by License Amendment Nos. 192, 197, and 204.

F.

DELETED G.

The licensees shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

H.

This renewed license is effective as of the date of issuance and shall expire at midnight on April 11, 2046.

FOR THE NUCLEAR REGULATORY COMMISSION

- original signed by E. J. Leeds -

Eric J. Leeds, Director Office of Nuclear Reactor Regulation

Enclosures:

1. Appendix A - Technical Specifications (NUREG-1202)

2. Appendix B - Environmental Protection Plan

3. Appendix C - Additional Conditions Date of Issuance: July 20, 2011 Renewed License No. NPF-57 Amendment No. 224

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 224 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-57 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION DOCKET NO. 50-354

1.0 INTRODUCTION

By letter dated November 25, 2019 (Reference 1), as supplemented by letters dated June 25, 2020 (Reference 2), and July 21, 2020 (Reference 3), PSEG Nuclear LLC (PSEG, the licensee) submitted a license amendment request (LAR) for the Hope Creek Generating Station (Hope Creek). The licensee proposed to add a new license condition to Renewed Facility Operating License No. NPF-57 to allow the implementation of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of structures, systems, and components (SSCs) subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation) based on a method of categorizing SSCs according to their safety significance.

To support its review, the NRC staff conducted an audit as described in the audit plan dated January 24, 2020 (Reference 35). Based on its review of the LAR and information reviewed during the audit, the NRC staff transmitted requests for additional information (RAIs) to the licensee dated June 15, 2020 (Reference 7), and no audit summary was needed.

The supplemental letters dated June 25, 2020, and July 21, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on January 28, 2020 (85 FR 5054).

2.0 REGULATORY EVALUATION

2.1 Risk-Informed Categorization and Treatment of SSCs The provisions of 10 CFR 50.69 allow adjustment of the scope of SSCs subject to special treatment requirements. Special treatment refers to those requirements that provide increased assurance beyond normal industry practices that SSCs perform their design-basis functions.

For SSCs categorized as low safety significance (LSS), alternative treatment requirements may be implemented in accordance with the regulation. For SSCs determined to be of high safety significance (HSS), requirements may not be changed.

Section 50.69 of 10 CFR contains requirements regarding how a licensee categorizes SSCs using a risk-informed process, adjusts treatment requirements consistent with the relative significance of the SSC, and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four risk-informed safety class (RISC) categories.

SSC categorization does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility.

Instead, 10 CFR 50.69 enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as HSS, existing treatment requirements are maintained or potentially enhanced. Conversely, for SSCs categorized as LSS that do not significantly contribute to plant safety on an individual basis, the regulation allows an alternative risk-informed approach to treatment that provides a reasonable level of confidence that these SSCs will satisfy functional requirements. Implementation of 10 CFR 50.69 allows licensees to improve focus on equipment that has HSS.

2.2 Licensees Proposed Changes The licensee proposed the addition of the following conditions to the renewed facility operating license for Hope Creek to allow the implementation of 10 CFR 50.69.

PSEG is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in PSEG submittal letter dated November 25, 2019, and all its subsequent associated supplements, as specified in License Amendment No. [XXX] dated [DATE].

PSEG will complete the implementation items listed in Attachment 1 of PSEGs letter to the NRC dated July 21, 2020, prior to crediting portable FLEX equipment for 10 CFR 50.69 categorization.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from the alternate seismic approach (referenced above) to a seismic probabilistic risk assessment approach).

The provisions of 10 CFR 50.69 allow adjustment of the scope of SSCs subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation) based on an integrated and systematic risk-informed process that includes several approaches and methods for categorizing SSCs according to their safety significance.1 2.3 Regulatory Guides and NRC Staff Review Plans The NRC staff considered the following regulatory guidance during its review of the proposed changes:

Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (Reference 8)

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 9)

RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference 10)

NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking (Reference 11)

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP), Chapter 19, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis:

General Guidance (Reference 12)

NRC-Endorsed Guidance The Nuclear Energy Institute (NEI) issued NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (Reference 13), as endorsed by RG 1.201, Revision 1, with clarifications, limitations, and conditions, which describes a process acceptable to the NRC for determining the safety significance of SSCs and categorizing them into the four RISC categories defined in 10 CFR 50.69.

Sections 2 through 10 of NEI 00-04 describe the following steps and elements of the SSC categorization process for meeting the requirements of 10 CFR 50.69:

Sections 3.2 and 5.1 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(i).

Sections 3, 4, 5, and 7 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(ii).

Section 6 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iii).

Section 8 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iv).

1 RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, dated May 2006 (Reference 8), describes the SSC categorization process in its entirety as an overarching approach that includes multiple approaches and methods identified for a PRA hazard and non-PRA methods.

Section 2 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(v).

Sections 9 and 10 provide specific guidance corresponding to 10 CFR 50.69(c)(2).

Additionally, Section 11 of NEI 00-04 provides guidance on program documentation and change control related to the requirements of 10 CFR 50.69(f). Section 12 of NEI 00-04 provides guidance on the periodic review related to the requirements in 10 CFR 50.69(e). Maintaining change control and periodic review provides confidence that all aspects of the program reasonably reflect the current as-built, as-operated plant configuration and applicable plant and industry operational experience as required by 10 CFR 50.69(c)(1)(ii).

3.0 TECHNICAL EVALUATION

3.1 Method of NRC Staff Review An acceptable approach for making risk-informed decisions about proposed licensing basis (LB) changes, including both permanent and temporary changes, is to show that the proposed LB changes meet the five key principles stated in Section C of RG 1.174, Revision 3Error!

Reference source not found.. These key principles are:

Principle 1: The proposed licensing basis change meets the current regulations unless it is explicitly related to a requested exemption.

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

Principle 3: The proposed licensing basis change maintains sufficient safety margins.

Principle 4: When the proposed licensing basis change results in an increase in risk, the increases should be small and consistent with the intent of the Commissions policy statement on safety goals for the operations of nuclear power plants.

Principle 5: The impact of the proposed licensing basis change should be monitored using performance measures strategies.

3.2 Traditional Engineering Evaluation The traditional engineering evaluation below addresses the first three key principles of RG 1.174, Revision 3, and is pertinent to (1) compliance with current regulations, (2) evaluation of defense in depth, and (3) evaluation of safety margins.

3.2.1 Key Principle 1: Licensing Basis Change Meets the Current Regulations Paragraph 50.69(c) of 10 CFR requires licensees to use an integrated decision-making process to categorize safety-related and non-safety-related SSCs according to the safety significance of the functions they perform into one of the following four RISC categories, which are defined in 10 CFR 50.69(a), as follows:

RISC-1: SSCs mean safety-related SSCs that perform safety significant functions.2 RISC-2: SSCs mean nonsafety-related SSCs that perform safety significant functions.

RISC-3: SSCs mean safety-related SSCs that perform low safety significant functions.

RISC-4: SSCs mean non-safety-related SSCs that perform low safety significant functions.

The SSCs are classified as having either HSS functions (i.e., RISC-1 and RISC-2 categories) or LSS functions (i.e., RISC-3 and RISC-4 categories). For HSS SSCs, 10 CFR 50.69 maintains current regulatory requirements for special treatment (i.e., it does not remove any requirements from these SSCs). For LSS SSCs, licensees can implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d). For RISC-3 SSCs, licensees can replace special treatment with an alternative treatment. For RISC-4 SSCs, 10 CFR 50.69 does not impose new treatment requirements.

Paragraph 50.69(b)(3) of 10 CFR states that the Commission will approve a licensees implementation of this section by issuance of a license amendment if the Commission determines that the SSC categorization process satisfies the requirements of 10 CFR 50.69(c).

As stated in 10 CFR 50.69(b), after the NRC approves an application for a license amendment, a licensee may voluntarily comply with 10 CFR 50.69 as an alternative to compliance with the following requirements for LSS SSCs:

(i) 10 CFR Part 21 (ii) a portion of 10 CFR 50.46a(b)

(iii) 10 CFR 50.49 (iv) 10 CFR 50.55(e)

(v) specified requirements of 10 CFR 50.55a (vi) 10 CFR 50.65, except for paragraph (a)(4)

(vii) 10 CFR 50.72 (viii) 10 CFR 50.73 (ix)

Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50 (x) specified requirements for containment leakage testing (xi) specified requirements of Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants, to 10 CFR Part 100 The NRC staff reviewed the licensees SSC categorization process against the categorization process described in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1. The NRC staffs review, as documented in this safety evaluation (SE), used the framework provided in RG 1.174, Revision 3, and NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1.

2 NEI 00-04, Revision 0 (Reference 13), uses the term high-safety-significant to refer to SSCs that perform safety-significant functions. The NRC understands HSS to have the same meaning as safety-significant (i.e., SSCs that are categorized as RISC-1 or RISC-2), as used in 10 CFR 50.69.

Section 2 of NEI 00-04, Revision 0, in part, states that the SSC categorization process includes eight primary steps:

1. Assembly of Plant-Specific Inputs (Section 3 of NEI 00-04, Revision 0)

2. System Engineering Assessment (Section 4 of NEI 00-04, Revision 0)

3. Component Safety Significance Assessment (Section 5 of NEI 00-04, Revision 0)

4. Defense-In-Depth Assessment (Section 6 of NEI 00-04, Revision 0)

5. Preliminary Engineering Categorization of Functions (Section 7 of NEI 00-04, Revision 0)

6. Risk Sensitivity Study (Section 8 of NEI 00-04, Revision 0)

7. Integrated Decisionmaking Panel Review and Approval (Section 9 of NEI 00-04, Revision 0)

8. SSC Categorization (Section 10 of NEI 00-04, Revision 0)

The licensee stated in Section 3.1.1 of the LAR that it will implement the SSC categorization process in accordance with NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1. In LAR Sections 3.1.1 and 3.2, the licensee described that the SSC categorization process uses probabilistic risk assessments (PRAs) to assess risks from internal events (includes internal floods) and internal fires. For the other risk contributors, the licensees process uses non-PRA methods for the risk characterization as follows:

Electric Power Research Institute (EPRI) alternative approach (Tier 1 criterion provided in EPRI Report 3002012988 (Reference 14)) to assess seismic risk; Individual plant examination of external events (IPEEE) screening to assess the risk from other external hazards (e.g., high winds, external floods) (Reference 5);

Shutdown events are assessed using the shutdown safety program described in Nuclear Management and Resources Council (NUMARC) 91-06, Guidelines for Industry Actions to Assess Shutdown Management (Reference 15); and Passive components are assessed using ANO-2 passive categorization methodology (Reference 4).

The NRC staff notes that use of the EPRI alternative approach for seismic hazard and the ANO-2 passive categorization methodology for passive components are deviations from the NEI 00-04 guidance, as endorsed by the NRC. A more detailed NRC staff review of these alternative methods is provided in Section 3.3.1.2 of this SE.

The licensee provided further discussion of specific elements within the SSC categorization process that is delineated in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1.

The regulatory requirements in 10 CFR 50.69 and 10 CFR Part 50, Appendix B, and the monitoring outlined in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, ensure that the SSC categorization process is sufficient to assure that the SSC functions continue to be met and that any performance deficiencies will be identified and appropriate corrective actions taken. The NRC staff reviewed the licensees SSC categorization program and concludes that it includes the appropriate steps and elements prescribed in NEI 00-04, Revision 0, to assure that SSCs specified in the technical specifications are appropriately categorized consistent with 10 CFR 50.69. The NRC staff performed a more detailed review of the specific steps/elements of the licensees SSC categorization process where necessary to confirm consistency with the NEI 00-04 guidance, as endorsed. As discussed further below, the NRC staff concludes that the proposed licensing basis is acceptable under the current 10 CFR 50.69 regulatory requirements, and therefore, meets the first key principle for risk-informed decision making prescribed in RG 1.174, Revision 3.

3.2.2 Key Principle 2: Licensing Basis Change is Consistent with the Defense-In-Depth Philosophy The defense-in-depth philosophy for the LB change is maintained if the following occurs:

Preserve a reasonable balance among the layers of defense.

Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

Preserve adequate defense against potential common-cause failures.

Maintain multiple fission product barriers.

Preserve sufficient defense against human errors.

Continue to meet the intent of the plants design criteria.

RG 1.201, Revision 1, endorses the guidance in Section 6 of NEI 00-04, Revision 0, but notes that the containment isolation criteria in this section of the guidance are separate and distinct from those set forth in 10 CFR 50.69(b)(1)(x). The criteria in 10 CFR 50.69(b)(1)(x) are to be used in determining which containment penetrations and valves may be exempted from the Type B and Type C leakage testing requirements in both Option A and Option B of Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, to 10 CFR Part 50. The criteria provided in 10 CFR 50.69(b)(1)(x) are not to determine the proper RISC category for containment isolation valves or penetrations.

In Section 3.1.1 of the LAR, the licensee clarified that it will require an SSC to be categorized as HSS based on the defense-in-depth assessment performed in accordance with NEI 00-04, Revision 0. In light of the above, the NRC staff concludes that the proposed change is consistent with the defense-in-depth philosophy, and therefore, satisfies the second key principle for risk-informed decision making prescribed in RG 1.174, Revision 3. The NRC staff reviewed the LAR and finds that the licensees process is consistent with the NRC-endorsed guidance in NEI 00-004, Revision 0; therefore, key principle 2 of risk-informed decision making is met and fulfills the 10 CFR 50.69(c)(1)(iii) criterion that requires DID to be maintained.

3.2.3 Key Principle 3: Licensing Basis Change Maintains Sufficient Safety Margins Section 50.69(c)(1)(iv) of 10 CFR requires the evaluations to provide reasonable confidence that for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in core damage frequency and large early release frequency resulting from changes in treatment are small. The engineering evaluation that will be conducted by the licensee under 10 CFR 50.69 for SSC categorization will assess the design function(s) and risk significance of the SSCs to assure that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring the categorization of the SSC does not adversely affect any assumptions or inputs to the safety analysis, or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist.

Consistent with the guidance provided in NEI 00-04 for review of safety margins, and in accordance with the implementation of the SSC categorization program, the only requirements that are relaxed for LSS SSCs (includes RISC-3) are those related to treatment. The SSCs design-basis function as described in the plants LB, including the updated final safety analysis report and technical specifications bases, do not change and should continue to be met.

Similarly, there is no impact to safety analysis acceptance criteria as described in the plant LB.

In Section 3.1.1 of the LAR, the licensee states that it will implement the risk categorization process in accordance with NEI 00-04, Revision 0, as endorsed by RG 1.201. On this basis, the NRC staff concludes that sufficient safety margins are maintained by the proposed methodology (i.e., NEI 00-04), and the third key principle of RG 1.174, Revision 3, is satisfied and fulfills the requirements set forth in 10 CFR 50.69(c)(1)(iv).

3.3 Risk-Informed Assessment 3.3.1 Key Principle 4: Change in Risk is Consistent with the Safety Goals The risk-informed considerations prescribed in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, address the fourth key principle of RG 1.174, Revision 3, pertaining to the assessment for change in risk and for monitoring the impact of the LB change.

A summary of how the licensees SSC categorization process is consistent with the guidance and methodology prescribed in NEI 00-04, Revision 0, and RG 1.201, Revision 1, is provided in the sections below.

3.3.1.1 Assembly of Plant-Specific Inputs (NEI 00-04, Revision 0, Section 3)

As outlined in NEI 00-04, the assembly of plant-specific inputs involves the collection of plant-specific risk information, including design and licensing information, PRA analyses, and other relevant plant data sources. The NRC staff acknowledges that elements of the SSC categorization process are not always performed in chronological order and may be performed in parallel such that the systematic process for evaluating the plant-specific PRA may include other aspects of the categorization process (e.g., system selection, system boundary definition, identification of system functions, and mapping of components to functions). The licensees SSC categorization process uses PRAs to assess risks from internal events (includes internal floods) and internal fires. The SSC categorization process uses non-PRA methods to assess risks from the other hazards (i.e., seismic, other external hazards, shutdown events, and passive components) as discussed in Section 3.3.1.2 of this SE.

Paragraph 50.69(c)(1)(v) of 10 CFR requires that SSC categorization be performed for entire systems and structures, not for selected components within a system or structure. The NRC staff finds the process described in the LAR, as supplemented, for collecting and organizing information at the system level for defining boundaries, functions, and components, is consistent with NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, and therefore, meets the requirements set forth in 10 CFR 50.69(b)(2)(ii) and (iii).

3.3.1.2 System Engineering Assessment (NEI 00-004, Revision 0, Section 4)

The system engineering assessment involves the identification and development of the base information necessary to perform the risk-informed categorization. Section 50.69(c)(1)(ii) of 10 CFR requires licensees to determine SSC functional importance using an integrated, systematic process for addressing initiating events (internal and external), SSCS, and plant operating modes, including those not modeled in the plant-specific PRA. Functions to be identified and considered include design bases functions and prevention of severe accidents. In Section 2.2 of the LAR, the licensee stated, [t]he safety functions include the design-basis functions, as well as functions credited for severe accidents (including external events).

Section 3.1.1 of the LAR summarizes the different initiating events and plant operating modes for which functional and risk-significant information will be collected. In Section 3.1.1 of the LAR, the licensee confirmed that the SSC categorization process documentation will include, among other items, system functions identified and categorized with the associated bases and the mapping of components to support function(s).

The NRC staff finds that the process described in the LAR, as supplemented, for system engineering assessment is consistent with NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, and therefore, meets the requirements set forth in 10 CFR 50.69(c)(1)(ii) and 10 CFR 50.69(c)(1)(v).

3.3.1.3 Component Safety Significance Assessment (NEI 00-04, Section 5)

This step in the licensees SSC categorization process assesses the safety significance of components using quantitative or qualitative risk information from a modeled PRA hazard, other hazards that can be screened, and non-PRA methods. In NEI 00-04, Revision 0, component risk significance is assessed separately for the following hazard groups:

internal events (includes internal floods) internal fires seismic hazard other external hazards (e.g., high winds, external floods) shutdown events passive component categorization In Sections 3.1.1 and 3.2 of the LAR, the licensee described that the SSC categorization process uses PRAs to assess risks from internal events (includes internal floods) and internal fires. The SSC categorization process uses the following non-PRA methods to assess risks from the other risk contributors:

Seismic Hazard: EPRI Report 3002012988 (Reference 14), Tier 1 alternative approach to assess seismic risk.

Other External Hazards: Screening analysis performed for IPEEE (Reference 5) updated using criteria from Part 6 of ASME/ANS RA-Sa-2009 (Reference 6), as endorsed by the NRC.

Shutdown Events: Shutdown safety program described in NUMARC 91-06 (Reference 15).

Passive Components: ANO-2 passive categorization methodology (Reference 4).

The approaches and methods proposed by the licensee to address internal events (includes internal floods), internal fires, other external hazards, defense in depth, and shutdown events are consistent with the approaches and methods in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1. The non-PRA method for the categorization of passive components is consistent with the ANO-2 methodology for passive components (Reference 4) approved for risk-informed safety classification and treatment for repair/replacement activities in Class 2 and Class 3 moderate-and high-energy systems. A detailed NRC staff review of use of the ANO-2 methodology in the SSC categorization process is provided in Section 3.3.1.2 of this SE. To address the seismic hazard, the licensee proposed to use an alternative method not specified in NEI 00-04 guidance as endorsed by the NRC. A detailed NRC staff review of the licensees proposed approach for the use of the EPRI alternative Tier 1 seismic approach is provided in Section 3.3.1.2 of this SE.

3.3.1.3.1 Scope of the PRA The Hope Creek PRA is comprised of a full-power, internal events PRA (IEPRA) and internal fires PRA (FPRA) that evaluate the core damage frequency (CDF) and large early release frequency (LERF) risk metrics. The licensee discussed in Section 3.3 of the LAR that the IEPRA (includes internal floods) model has been assessed against RG 1.200, Revision 2.

Furthermore, LAR Section 3.3 states that a finding closure review was conducted on the IEPRA model in August 2017 and on the FPRA in September 2018 and in 2019, using the NRC-accepted process documented in the NEI letter to the NRC entitled, Final Revision of Appendix X to NEI 05-04/07-12/12-[13], Close-Out of Facts and Observations (F&Os), dated February 21, 2017 (Reference 16).

The NRC staff finds that the LAR, as supplemented, provides sufficient information to support the NRC staff review of the IEPRA (includes internal floods) and FPRA for technical acceptability, and therefore, meets the requirements set forth in 10 CFR 50.69(b)(2)(ii) and (iii).

Aspects considered by the NRC staff to evaluate the scope of the PRA include: (1) peer-review history and results (includes open F&Os), (2) the Appendix X, F&O closure process, (3) credit for mitigating strategies (FLEX) in the PRA, and (4) assessment of assumptions and approximations. In e-mail correspondence to the licensee dated June 15, 2020 (Reference 7),

the NRC staff issued RAls to further assess the acceptability of the Hope Creek IEPRA (includes internal floods) and FPRA for consistency with RG 1.200, Revision 2; NEI 00-04, Revision 0; and RG 1.201, Revision 1. The NRC staffs review of these aspects of the PRA and supplemental responses to the applicable RAls are provided below.

Internal Events PRA (Includes Internal Floods) Peer-Review History The LAR states the full-power IEPRA model credited in the request is the same PRA model credited in the NRCs issuance of Amendment No. 215 for an inverter allowed outage time (AOT) extension, dated March 27, 2019 (Reference 17). Therefore, the NRC staffs review of the IEPRA (includes internal floods) is based on the results provided in the LAR and the inverter AOT extension application.

As stated in the LAR and in the inverter AOT extension application (Reference 17), a full-scope peer review was performed in March 2009 for the IEPRA (includes internal floods) against the requirements of the ASME/ANS 2007 PRA standard and RG 1.200, Revision 1. The licensee provided a gap assessment during the NRC staffs review of the inverter AOT extension application where the NRC staff concluded the licensee identified and addressed all applicable differences between the ASME/ANS 2007 PRA standard and ASME/ANS RA-Sa-2009 with the applicable regulatory positions contained in Appendix A of RG 1.200, Revision 2.

In Section 3.2 of the LAR for the IEPRA (includes internal floods), the licensee stated, in part, there are no PRA upgrades that have not been peer reviewed. Also, LAR Section 3.3 states an F&O closure review of the finding-level F&Os was conducted in 2017 and concluded that all IEPRA (includes internal floods) F&Os have been closed. A detailed NRC staff review of this F&O closure review is included below. The NRC staff concluded that all IEPRA F&Os were appropriately assessed by the F&O closure review team to assure that no new methods or upgrades were inadvertently incorporated into the IEPRA without a peer review in accordance with ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2, and the peer review process requirements in 10 CFR 50.69(c)(1)(i).

In LAR Attachment 6, as supplemented by the response to APLA RAI 05.a (Reference 2), the licensee identifies as a source of uncertainty the failure probabilities associated with the digital feedwater system (DFWS) and clarifies that the DFWS modeling incorporates the industry generic values in NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants (Reference 18). The NRC staff finds that the DFWS failure probabilities were developed and applied consistent with ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2.

The NRC staff reviewed the LAR and finds that the IEPRA (includes internal floods) conforms to the applicable technical elements in ASME/ANS RA-Sa-2009 (Reference 6), as endorsed by RG 1.200, Revision 2, and is acceptable to the extent needed to support the Hope Creek 10 CFR 50.69 program. Therefore, the Hope Creek 10 CFR 50.69 program uses an IEPRA that is of sufficient quality to meet the requirements set forth in 10 CFR 50.69(c)(1)(i) regarding PRA quality.

Internal Fire PRA Peer-Review History The LAR states the FPRA model credited in the request is the same FPRA model credited in the NRCs issuance of Amendment No. 215 for an inverter AOT extension, and subsequently subjected to two F&O closure reviews and a focused-scope peer review. Therefore, the NRC staffs review of the FPRA was based on the results provided in the LAR and inverter AOT extension application.

The licensee evaluated the technical acceptability of the Hope Creek FPRA model by conducting a full-scope peer review in October 2010 using NEI 07-12, Revision 1 (Reference 19), and ASME/ANS RA-Sa-2009, as qualified by RG 1.200, Revision 2. A subsequent focused-scope peer review was performed for several FPRA upgrades in September 2018 using ASME/ANS RA-Sa-2009, as qualified by RG 1.200, Revision 2. Based on LAR Section 3.2 for the FPRA, there are no FPRA upgrades that have not been peer reviewed, and the FPRA was developed consistent with NUREG/CR-6850 (Reference 20) and only utilized NRC-approved methods.

Section 3.3 of the LAR states an F&O closure review of the finding-level F&Os was conducted in September 2018 and in 2019 and concluded that all FPRA F&Os have been closed. A detailed NRC staff review of this F&O closure review is included below. The NRC staff concluded that all FPRA F&Os were appropriately assessed by the F&O closure review team to assure that no new methods or upgrades were inadvertently incorporated into the FPRA without a peer review in accordance with ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2.

In LAR Attachment 6, the licensee identifies as a source of uncertainty the floor value (i.e.,

1E-06 or 5E-07) applied to dependent combinations of joint human error probabilities (JHEPs).

However, industry guidance, such as NUREG-1792 (Reference 21), recommend a JHEP floor value of 1E-05. In response to APLA RAI 05.b, the licensee provided results of a sensitivity study where the JHEP floor value was set to 1E-05 and demonstrated the floor value did not impact the SSC categorization results.

In LAR Attachment 6, the licensee identifies as a source of uncertainty the truncation limit for quantifying the FPRA results. The FPRA was quantified for CDF and LERF using a truncation limit of 1E-11/year. While this truncation limit is adequate for quantifying CDF, this limit may not be adequate for quantifying LERF and is considered a source of uncertainty. In response to APLA RAI 05.c, the licensee provided results of an analysis that demonstrated a truncation value of 1E-12 would reasonably meet the convergence requirement specified in ASME/ANS RA-Sa-2009. The response also stated the current FPRA LERF model will use a truncation value of 1E-12 for the SSC categorization process, and the truncation value used for future updates to the FPRA model will be evaluated consistent with ASME/ANS RA-Sa-2009.

The NRC staff reviewed the LAR and finds that the FPRA conforms to the applicable technical elements in ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2, and is acceptable to the extent needed to support the Hope Creek 10 CFR 50.69 program. Therefore, the Hope Creek 10 CFR 50.69 program uses an FPRA that is of sufficient quality to meet the requirements set forth in 10 CFR 50.69(c)(1)(i) regarding PRA quality.

Appendix X, Independent Assessment Process for F&O Closure Review Section X.1.3 of Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13 (Reference 16), as accepted by the NRC in a letter dated May 3, 2017 (Reference 22), provides guidance to perform an independent assessment for the closure of F&Os identified from a full-scope or focused-scope peer review.

An F&O closure review was conducted on the IEPRA model in 2017. Finding-level F&Os were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13, as accepted by the NRC in the letter dated May 3, 2017. After the F&O closure review, there were no remaining open F&Os. In response to APLA RAI 01, the licensee confirmed that none of the F&O resolutions closed during the F&O closure review constituted a PRA upgrade as defined in ASME/ANS RA-Sa-2009.

An F&O closure review was conducted on the FPRA model in September 2018. Finding-level F&Os were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13, as accepted by the NRC in the letter dated May 3, 2017. The F&O closure review team determined three F&O resolutions were PRA upgrades (regarding human reliability analysis, main control room abandonment, and fire modeling). Therefore, the same closure review team conducted a focused-scope peer review of the PRA upgrades in accordance with NEI 07-12, Revision 1, and ASME/ANS RA-Sa-2009, as qualified by RG 1.200, Revision 2. A second F&O closure review was conducted on the FPRA in 2019 in accordance with Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13. All open F&Os were closed.

Based on the discussion above, the NRC staff concludes that the licensee has adequately implemented the F&O closure process for the closed F&Os for the IEPRA and FPRA.

Credit for FLEX Equipment The NRC memorandum dated May 30, 2017, entitled, Assessment of the Nuclear Energy Institute 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Guidance for Risk-Informed Changes to Plants Licensing Basis (Reference 23), provides the NRC staffs assessment of challenges to incorporating FLEX (diverse and flexible coping strategy) equipment and strategies into a PRA model in support of risk-informed decision making in accordance with the guidance of RG 1.200, Revision 2.

of the LAR identifies several sources of uncertainty associated with the PRA modeling of FLEX equipment. In response to APLA RAI 04 (Reference 2), the licensee stated that FLEX equipment and associated operator actions will not be credited in the IEPRA and FPRA models used for SSC categorization until appropriately peer reviewed in accordance with RG 1.200, Revision 2. In its supplement to the LAR dated July 21, 2020, the licensee specified an implementation item that PSEG will not credit portable FLEX equipment in the PRA models used for 10 CFR 50.69 categorization until focused-scope peer reviews are performed on the model changes associated with incorporating mitigating strategies, and until associated F&Os are resolved to Capability Category II in accordance with the current revision of RG 1.200.

Identification and Treatment of Key Assumptions and Sources of Uncertainty In the LAR, the licensee confirmed that NUREG-1855, Revision 1 (Reference 11), was used to identify, screen, and characterize those sources of model uncertainty and related assumptions in the base Hope Creek IEPRA (includes internal floods) and FPRA that are relevant to this LAR. Substep E-1.4 of NUREG-1855, Revision 1, is a qualitative screening process that involves identifying and validating whether consensus3 models have been used in the PRA to evaluate identified model uncertainties. The licensee confirmed in the LAR that for the Hope Creek 10 CFR 50.69 uncertainty analysis, some PRA uncertainties and assumptions were screened from further consideration based on the use of a consensus method (i.e., step E-1 of NUREG-1855, Revision 1). Furthermore, the licensee reviewed nonconservative treatments in the PRA to determine their impact on this application. In response to APLA RAI 03 (Reference 2), the licensee clarified that all sources of PRA modeling uncertainty, including modeling conservatisms, were reviewed for their impact on this application. The NRC staff finds that the assessment performed by the licensee to identify the key assumptions and sources of modeling uncertainty for the IEPRA (includes internal floods) and FPRA is consistent with the guidance provided in NUREG-1855, Revision 1, and therefore is acceptable.

The guidance in Section 5 of NEI 00-04, Revision 0, specifies sensitivity analyses to be conducted to address PRA uncertainties that could impact the SSC categorization results. In LAR Attachment 6, the licensee provided a list of key assumptions and sources of modeling uncertainties associated with the IEPRA (includes internal floods) and FPRA, and the licensee dispositioned each. The licensees conclusion of this evaluation is that no additional sensitivity analyses beyond those required under Section 5 of NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, are needed to address Hope Creek PRA model-specific assumptions or sources of uncertainty. The licensee confirmed that sensitivity analyses will be performed consistent with Tables 5-2 and 5-3 of NEI 00-04 for the IEPRA (includes internal floods) and FPRA, respectively. As described in the LAR, in accordance with NEI 00-04, the results of the sensitivity analyses will be given to the integrated decision-making panel (IDP) for consideration in the final risk characterization for components initially classified as LSS that may be reclassified to HSS. The NRC staff finds that the licensee will perform sensitivity analyses consistent with Tables 5-2 and 5-3 of NEI 00-04 for the IEPRA (includes internal floods) and 3 Per NUREG-1855, Revision 1, a consensus model is a model that has a publicly available published basis and has been peer reviewed and widely adopted by an appropriate stakeholder group.

FPRA to address the identified key assumptions and sources of uncertainty in the context of the decision making under consideration for the categorization of SSCs.

In addition, the NRC staff recognizes that the licensee will perform routine PRA changes and updates to assure the PRA continually reflects the as-built, as-operated plant, in addition to changes made to the PRA to support the context of the analysis being performed (e.g.,

sensitivity analyses). To address 10 CFR 50.69(e) and (f), which stipulate the process for feedback and adjustment to assure configuration control is maintained for routine changes and updates to the PRAs, Section 3.5 of the LAR states that after a PRA model update, an SSC categorization review will be performed to assess its impact.

Since the evaluation of seismic and other external hazards use non-PRA methods, there is no quantifiable uncertainty.

Based on the discussion above, the NRC staff finds that the identification and treatment of the key assumptions and sources of modeling uncertainty for the IEPRA (includes internal floods) and FPRA are acceptable because it is consistent with the guidance provided in NUREG-1855, Revision 1, and Section 5 of NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, and therefore, meet the requirements set forth in 10 CFR 50.69(b)(2)(ii) and (c)(1)(ii).

PRA Importance Measures and Integrated Importance Measures Pursuant to 10 CFR 50.69(c)(1)(ii), the licensees SSC characterization process must determine SSC functional importance using an integrated systematic process for addressing internal and external initiating events. NEI 00-04, Section 5, provides guidance on the risk importance assessment process. The scope of modeled hazards for Hope Creek includes the IEPRA (includes internal floods) and FPRA. The NRC staff reviewed the LAR and finds that the licensees use and treatment of importance measures is consistent with the guidance in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1. The detailed NRC staff review of alternative methods for assessing the risk for seismic, other external hazards, shutdown events, and passive components is provided in Section 3.3.1.2 of this SE.

PRA Acceptability Conclusions Pursuant to 10 CFR 50.69(c)(1)(i), the SSC categorization process must consider results and insights from a plant-specific PRA. The use of the IEPRA (includes internal floods) and FPRA to support SSC categorization is endorsed by RG 1.201, Revision 1. The PRAs must be acceptable to support the SSC categorization process and must be subjected to a peer-review process assessed against a standard that is endorsed by the NRC. Revision 2 of RG 1.200 provides guidance for determining the acceptability of the PRA by comparing the PRA to the relevant parts of ASME/ANS RA Sa-2009 using a peer-review process.

The licensee has subjected the IEPRA and FPRA to the peer-review processes and closed all F&Os associated with these peer reviews in accordance with an NRC-accepted process. The NRC staff reviewed the peer-review history, F&O closure review history, and the identification and disposition of key assumptions and sources of uncertainty. The NRC staff concludes:

(1) the licensees IEPRA (includes internal floods) and FPRA are acceptable to support the categorization of SSCs using the process endorsed by the NRC staff in RG 1.201, Revision 1; and (2) the key assumptions and sources of model uncertainty for the PRAs have been identified consistent with the guidance in RG 1.200, Revision 2 and NUREG-1855, Revision 1, and addressed appropriately for this application.

The NRC staff finds the licensee provided the required information and the IEPRA (includes internal floods) and FPRA are acceptable, and therefore, meet the requirements set forth in 10 CFR 50.69(c)(1)(i) and (ii).

3.3.1.3.2 Evaluation of the Use of Non-PRA Methods in SSC Categorization The licensees SSC categorization process uses the following non-PRA methods:

Seismic Hazard: EPRI Report 3002012988 (Reference 14), Tier 1 alternative approach to assess seismic risk.

Other External Hazards: Screening analysis performed for IPEEE (Reference 5) updated using criteria from Part 6 of ASME/ANS RA-Sa-2009, as endorsed by the NRC.

Shutdown Events: Shutdown safety program described in NUMARC 91-06 (Reference 15).

Passive Components: ANO-2 passive categorization methodology (Reference 4).

The NRC staffs review of these methods is discussed below.

Seismic Risk As part of its proposed process to categorize SSCs according to safety significance, the licensee proposed to use a non-PRA method to consider seismic hazards. Paragraphs 50.69(b)(2)(ii) and 50.69(c)(1)(ii) of 10 CFR permit the use of non-PRA methods in the SSC categorization process.

The licensee provided a description of its proposed alternative seismic approach for considering seismic risk in the categorization process and described how the proposed alternative seismic approach would be used in the categorization process in Section 3.2.3 of the LAR and its supplement dated June 25, 2020. In addition, the licensee based the acceptability of its proposed alternative seismic approach on the conclusions gained from case studies performed in EPRI Report 3002012988 (Reference 14), and therefore, indirectly, on the acceptability of the PRAs used for the case studies. The licensee incorporated relevant information provided in its supplements to the Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) 10 CFR 50.69 LAR as part of its proposed approach in its supplement dated June 25, 2020. The information presented in the LAR and its supplement, including the information incorporated by the licensee from the Calvert Cliffs 10 CFR 50.69 LAR, as well as in EPRI Report 3002012988, taken together, provides sufficient detail on the description of the licensees proposed alternative seismic approach, how the proposed alternative seismic approach would be used in the categorization process, and the measures for assuring that the quality and level of detail of the licensees proposed alternative seismic approach are adequate for the categorization of SSCs. Therefore, the NRC staff finds that the requirements in 10 CFR 50.69(b)(2)(ii) for the proposed alternative seismic approach are met.

EPRI Report 3002012988 includes the results from case studies performed to determine the extent and type of unique HSS SSCs from seismic PRAs (SPRAs). The NRC staffs review confirmed that the case studies in EPRI Report 3002012988 used by the licensee to support its proposed alternative seismic approach, as well as the information from its supplements to the Calvert Cliffs 10 CFR 50.69 LAR that was incorporated by reference by the licensee for its proposed alternative seismic approach, are identical to that reviewed by the staff for the Calvert Cliffs 10 CFR 50.69 LAR. The information presented in the LAR and supplements, taken together, provides a sufficient description of, and basis for acceptability of, the evaluations to be conducted to satisfy 10 CFR 50.69(c)(1)(iv) for the alternative seismic approach. Therefore, the NRC staff finds that the requirements in 10 CFR 50.69 (b)(2)(iv) are met for the proposed alternative seismic approach.

Evaluation of Technical Acceptability of the PRAs Used for Case Studies Supporting the Proposed Alternative Seismic Approach The NRC staff reviewed and evaluated the technical acceptability of the PRAs used in the case studies for Plants A, C, and D in EPRI Report 3002012988 as part of the Calvert Cliffs 10 CFR 50.69 LAR. The NRC staffs review and conclusions on the technical acceptability of the PRAs used in the case studies for Plants A, C, and D in EPRI Report 3002012988 are discussed in the SE for Calvert Cliffs 10 CFR 50.69 LAR (Reference 24).

As discussed in the Calvert Cliffs 10 CFR 50.69 SE, the NRC staff evaluated the peer-review process and resolution of peer-review findings and key assumptions and sources of uncertainty for Plants A, C, and D. The Calvert Cliffs 10 CFR 50.69 SE discussed the NRC staffs review of the approach, implementation, and results of the mapping performed for the Plants A, C, and D case studies to ensure that the mapping was technically justified. The NRC staff also reviewed the conclusions on the determination of unique HSS SSCs from SPRAs in EPRI Report 3002012988 from the Plants A, C, and D case studies in the Calvert Cliffs 10 CFR 50.69 SE.

Based on the information in Section 3.2.3 of the LAR and the supplement dated June 25, 2020, the case studies, mapping approach, and conclusions on the determination of unique HSS SSCs from the case studies used by the licensee to support its proposed alternative seismic approach are identical to those reviewed by the staff for the Calvert Cliffs 10 CFR 50.69 LAR.

Therefore, the NRC staffs conclusions on the technical acceptability of PRAs used for the Plants A, C, and D case studies in EPRI Report 3002012988, the mapping approach used in those case studies, and conclusions on the determination of unique HSS SSCs from the case studies in the Calvert Cliffs SE are directly applicable to this licensees proposed alternative seismic approach. Consequently, the NRC staff finds that the Plants A, C, and D PRAs were technically acceptable for use in the corresponding case studies supporting the proposed alternative seismic approach; the mapping of SSCs between the SPRA, the IEPRA, and as applicable, the FPRA for the Plants A, C, and D case studies, was performed in a technically justifiable manner; and that the conclusions on the determination of unique HSS SSCs from SPRAs in the Plants A, C, and D case studies in EPRI Report 3002012988, and therefore, in the proposed alternative seismic approach, are valid.

Evaluation of the Criteria for the Proposed Alternative Seismic Approach The licensee proposed the following criteria for applicability and use of the proposed alternative seismic approach:

The GMRS [Ground Motion Response Spectrum] peak acceleration is at or below approximately 0.2g or where the GMRS is below or approximately equal to the SSE [Safe Shutdown Earthquake] between 1.0 Hz and 10 Hz.

The licensees proposed criteria are identical to that reviewed by the NRC staff for the Calvert Cliffs 10 CFR 50.69 LAR. Therefore, the NRC staffs conclusions on the criteria in the Calvert Cliffs SE are directly applicable to this licensees proposed criteria. Consequently, licensees proposed criteria of GMRS peak acceleration at or below approximately 0.2 gram (g) or the GMRS below or approximately equal to the safe SSE between 1.0 and 10 hertz (Hz) to determine the applicability and use of the proposed alternative seismic approach is acceptable.

Evaluation of Applicability of Criteria for this Application The licensee compared the GMRS from the reevaluated seismic hazard for Hope Creek developed in response to Near-Term Task Force Recommendation 2.1 against the sites design-basis SSE, as shown in Figure 1 of Attachment 4 to the LAR, to demonstrate that the site meets the criteria for application of the proposed alternative seismic approach. The NRC staff previously evaluated the licensees response to the 10 CFR 50.54(f) letter associated with Near-Term Task Force Recommendation 2.1 in which the licensee submitted its reevaluated seismic hazard (Reference 25). The NRC staffs previous assessment of the licensees reevaluated seismic hazard (Reference 26) indicates that the licensees methodology was acceptable and that the GMRS determined using the reevaluated hazard adequately characterized the site. Since the same reevaluated hazard is used for comparison against the criteria for use of the proposed alternative seismic approach, the NRC staffs previous assessment on the reevaluated hazard is applicable to this review. The NRC staffs evaluation of the licensees reevaluated seismic hazard determined that the GMRS is below the plant SSE between 1 Hz to 10 Hz, although the GMRS peak acceleration is higher than 0.2g. Based on its review, the NRC staff finds that the licensees seismic hazard meets the criteria for the proposed alternative seismic approach.

In Section 3.2.3 of the LAR, the licensee stated that the small percentage contribution of seismic to total plant risk makes it unlikely that an integral importance assessment for a component, as defined in NEI 00-04, would result in an overall HSS determination from seismic risk considerations. In its supplement to the LAR dated June 25, 2020, the licensee provided an estimate of the seismic CDF and demonstrated that the seismic risk is low relative to the overall plant risk.

The NRC staff verified the licensees estimate by convolving the median seismic capacity with composite uncertainty provided in the supplement to the LAR dated June 25, 2020, and the reevaluated seismic hazard for the mean peak ground acceleration (Reference 25). Based on its evaluation, the NRC staff determined that the seismic CDF is approximately 2 percent of the total plant CDF, and the seismic LERF is approximately 3 percent of total plant LERF.

Therefore, based on its evaluation and review, the NRC staff concludes that the seismic risk contribution for the licensee is not expected to solely result in an SSC being categorized as HSS.

In summary, the NRC staffs review finds that the licensees basis for applying the proposed alternative seismic approach is acceptable because: (1) the reevaluated hazard meets the criteria for use of the proposed alternative seismic approach, and (2) the seismic risk contribution would not solely result in an SSC being categorized as HSS.

Evaluation of the Implementation of Conclusions from the Case Studies The categorization conclusions from the EPRI Report 3002012988 case studies, performed for GMRS to SSE ratios significantly higher than Hope Creek, indicated that seismic-specific failure modes resulted in HSS categorization uniquely from SPRAs. Therefore, these seismic-specific failure modes, such as correlated failures, relay chatter, and passive component structural failure mode, can influence the categorization process. The NRC staff reviewed the proposed alternative seismic approach to evaluate whether the categorization-related conclusions from EPRI Report 3002012988 were appropriately included and implemented.

Section 3.2.3 of the LAR discussed the licensees proposed alternative seismic approach. The licensee stated that the proposed categorization approach for seismic hazards will include qualitative consideration of the mitigation capabilities of SSCs during seismically-induced events and seismic failure modes based on insights obtained from prior seismic evaluations performed for Hope Creek. Additional information on the proposed alternative seismic approach, including examples of prior seismic evaluations that would act as sources of the insight, is discussed by the licensee in Sections 3.1.1 and 3.2.3 of the LAR.

In its supplement to the LAR dated June 25, 2020, the licensee stated that it will follow the same alternative seismic approach as approved for the Calvert Cliffs 10 CFR 50.69 LAR. The licensee also incorporated by reference relevant information provided in its supplements to the Calvert Cliffs 10 CFR 50.69 LAR to support the proposed alternative seismic approach.

Therefore, the NRC staffs review of and conclusions on the implementation of the alternative seismic approach from the Calvert Cliffs 10 CFR 50.69 LAR are directly applicable to this licensees proposed alternative seismic approach. Consequently, the NRC staffs review of the proposed alternative seismic approach, in conjunction with the requirements in 10 CFR 50.69 and the corresponding statement of considerations, finds that the proposed alternative seismic approach provides reasonable confidence in the evaluations required by 10 CFR 50.69(c)(1)(ii),

as well as 10 CFR 50.69(c)(1)(iv), because:

1. It includes qualitative consideration of seismic events at several steps of the categorization process, including documentation of the information for presentation to the IDP as part of the integrated, systematic process for categorization.

2. It presents system-specific seismic insights to the IDP for consideration as part of the IDP review process as each system is categorized, thereby providing the IDP a means to consider potential impacts of seismic events in the categorization process.

3. The insights presented to the IDP include potentially important seismically-induced failure modes, as well as mitigation capabilities of SSCs during seismically-induced design-basis and severe accident events, consistent with the conclusions on the determination of unique HSS SSCs from SPRAs in EPRI Report 3002012988. The insights will use prior plant-specific seismic evaluations, and therefore, in conjunction with performance monitoring for the proposed alternative seismic approach, reasonably reflect the current plant configuration. Further, the recommendation for categorizing civil structures in the alternative seismic approach provides appropriate consideration of such failures from a seismic event.

4. It presents the IDP with the basis for the proposed alternative seismic approach, including the low seismic hazard for the plant and the criteria for use of the proposed alternative seismic approach.

5. It includes qualitative consideration and insights related to the impact of a seismic event on SSCs for each SSC that is categorized and does not limit the scope to SSCs from the case studies supporting this application.

Consideration of Changes to Seismic Hazard An important input to the NRC staffs evaluation of the proposed alternative seismic approach is the current knowledge of the seismic hazard at the plant. The possibility exists for the seismic hazard at the site to increase such that the criteria for use of the proposed alternative seismic approach are challenged. In such a situation, the categorization process may be impacted from a seismic risk perspective either solely due to the seismic risk or by the integrated importance measure determination.

In Section 3.2.3 of the LAR, the licensee stated that U.S. nuclear power plants that utilize the 50.69 Seismic Alternative (EPRI 3002012988) will continue to compare GMRS to SSE. Since the alternative seismic approach explicitly cites and is based on EPRI Report 3002012988, the continued comparison of GMRS to SSE applies to Hope Creek. The licensee also stated that the seismic hazard at the plant is subject to periodic reconsideration as new information becomes available through industry evaluations.

The NRC staffs review finds that consideration of changes to seismic hazard in the licensees proposed alternative seismic approach is the same as the Calvert Cliffs 10 CFR 50.69 submittal (Reference 24). Therefore, the NRC staffs evaluation of the consideration of changes to the seismic hazard against the requirements in 10 CFR 50.69(e)(1), 10 CFR 50.69(e)(3), and 10 CFR 50.69(d)(2)(ii), as well as the resulting conclusion on consideration of changes to the seismic hazard in the Calvert Cliffs 50.69 SE, is directly applicable to this licensees proposed alternative seismic approach. Consequently, the NRC staff finds that the consideration of changes to the seismic hazard at Hope Creek that exceeds the criteria for use of the proposed alternative seismic approach is acceptable because: (1) the criteria for use of the proposed alternative seismic approach is clear and traceable, (2) the proposed alternative seismic approach includes periodic reconsideration of the seismic hazard as new information becomes available, (3) the proposed alternative seismic approach satisfies the requirements in 10 CFR 50.69 discussed above, and (4) the licensee has included a proposed license condition in the LAR.

Monitoring of Inputs to and Outcome of Proposed Alternative Seismic Approach In Section 3.5 of the LAR and the supplement dated June 25, 2020, the licensee stated that its configuration control process ensures that changes to the plant, including a physical change to the plant and changes to documents, are evaluated to determine the impact on design bases, licensing documents, programs, procedures, and training. The licensee further stated that its performance monitoring process required periodic review to assess changes that could impact the categorization results and to provide the IDP with an opportunity to recommend categorization and treatment adjustments due to such changes.

The NRC staff found that consideration of the feedback and adjustment process in the licensees proposed alternative seismic approach is the same as the Calvert Cliffs 10 CFR 50.69 submittal. Therefore, the NRC staffs review of and conclusion on the feedback and adjustment process for the alternative seismic approach in the Calvert Cliffs 10 CFR 50.69 SE is applicable to this application. Consequently, the NRC staff finds that (1) the licensees programs provide reasonable assurance that the existing seismic capacity of LSS components would not be significantly impacted, and (2) the monitoring and configuration control program ensures that potential degradation of the seismic capacity would be detected and addressed before significantly impacting the plant risk profile. Therefore, the NRC staff finds that reasonable confidence exists that the potential impact of the seismic hazard on the categorization is maintained acceptably low, and the requirements in 10 CFR 50.69(c)(1)(iv) are met for the proposed alternative seismic approach.

Other Non-Seismic External Hazards The licensee discussed its consideration of other non-seismic external hazards in Section 3.2.4 of the LAR. Non-seismic external hazards include high winds, external flood hazards, and other hazards listed in Appendix 6-A of ASME/ANS RA-Sa-2009. The licensee evaluated all non-seismic external hazards for the 10 CFR 50.69 application using a plant-specific evaluation in accordance with Generic Letter 88-20 (Reference 27) and the criteria in ASME/ANS RA-Sa-2009.

The NRC staff reviewed the licensees evaluation, which was provided in Attachments 4 and 5 to the LAR.

The NRC staffs assessment of the licensees focused evaluation for its reevaluated external flood hazard (Reference 28 and Reference 29) determined that local intense precipitation elevations exceeded the plants current design basis. In Attachment 4 to the LAR, the licensee stated, Watertight doors are the only active flood protection features. Therefore, since certain watertight doors are credited with screening of this hazard, these SSCs will be considered HSS during categorization of systems containing these doors. Based on the review of the information provided by the licensee in Attachment 4 to the LAR and the NRC staffs previous assessment of the licensees focused evaluation for the reevaluated external flood hazard, the NRC staff finds that the licensees SSC categorization process will evaluate the safety significance of any SSCs for the external flooding hazard consistent with the guidance provided in NEI 00-04, as endorsed by the NRC.

In Attachment 4 to the LAR, the licensee discussed screening of extreme winds, tornado, and hurricane hazards. The licensee cited its updated final safety analysis report (Reference 30),

which provides the design basis for tornado hazard with a maximum wind speed of 360 miles per hour (mph). The NRC staffs review of Table 6-1 in NUREG/CR-4461, Tornado Climatology of the Contiguous United States (Reference 31), indicates that the tornado wind speed for 1E-6 per year occurrence frequency is 205 mph based on the Fujita scale and 166 mph based on the Enhanced Fujita scale. The NRC staff notes that, since tornados bound the extreme wind hazard, and the 1E-6 per year tornado wind speed is much less than the design-basis tornado maximum wind speed of 360 mph, damage due to the forces associated with high winds and tornadoes can be screened out. The hurricane hazard can also be screened out because its maximum probable wind speed of 132 mph is bounded by the tornado wind speed. The NRC staffs review further notes that the primary concern for high straight winds is loss-of-offsite power caused by the winds and that the IEPRA already includes loss-of-offsite power events due to severe weather, including high straight winds. Therefore, the NRC staff finds that the impact of high straight winds on plant response and the resulting categorization of SSCs is included in the categorization process.

The discussion provided for extreme winds or tornados in Attachment 4 to the LAR states that a categorization prerequisite has been added to Attachment 1 to the LAR to complete necessary actions (e.g., analyses, modifications, etc.) to screen tornado missile hazards prior to the adoption of 10 CFR 50.69. The licensee stated that if tornado missile protection vulnerabilities are discovered as part of assessing tornado missile protection in response to Regulatory Issue Summary 2015-06 (Reference 32), then this information will be used to update the screening process. In its supplement to the LAR dated June 25, 2020, the licensee confirmed that the categorization prerequisite in Attachment 1 to the LAR to complete necessary actions to screen tornado missile hazards has been completed. The licensee further stated that SSCs credited for screening of external hazards will be evaluated according to the flow chart in NEI 00-04, Figure 5-6. The NRC staffs review finds that the licensees SSC categorization process will evaluate the safety significance of SSCs for the extreme winds or tornado hazard consistent with the guidance provided in NEI 00-04, as endorsed by the NRC.

In summary, the NRC staffs review finds that the licensees SSC categorization process will evaluate the safety significance of SSCs for other non-seismic external hazards in Attachment 4 to the LAR consistent with the guidance provided in Figure 5-6 of NEI 00-04, Revision 0, as endorsed by the NRC in RG 1.201, Revision 1. Therefore, the staff concludes that the licensees treatment of other external hazards is acceptable and meets 10 CFR 50.69(c)(1)(ii).

Method for Assessing Shutdown Events Consistent with the guidance in NEI 00-04, Revision 0, the licensee in Section 3.2.5 of the LAR proposed using the shutdown safety program described in NUMARC 91-06 (Reference 15).

NUMARC 91-06 provides considerations for maintaining defense in depth for the five key safety functions during shutdown - namely, decay heat removal capability, inventory control, power availability, reactivity control, and containment-primary/secondary. NUMARC 91-06 also specifies that a defense-in-depth approach should be used with respect to each defined shutdown key safety function. This is accomplished by designating a running and an alternative system/train to accomplish the given key safety function.

The use of NUMARC 91-06 described by the licensee in the submittal is consistent with the guidance in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1. The approach uses an integrated and systematic process to identify HSS components consistent with the shutdown evaluation process. Therefore, the NRC staff finds that the licensees use of NUMARC 91-06 is acceptable and meets the requirements set forth in 10 CFR 50.69(b)(2)(ii) and (c)(1)(ii).

Method for Assessing Passive Components Passive components are not modeled in the PRA; therefore, a different assessment method is necessary to assess the safety significance of these components. Passive components are those components having only a pressure -retaining function. This process also addresses the passive function of active components such as the pressure/liquid retention of the body of a motor-operated valve.

In Section 3.1.2 of the LAR, the licensee proposed using a categorization method for passive components not cited in NEI 00-04, Revision 0, or RG 1.201, Revision 1, but that was approved by the NRC for ANO-2 (Reference 4). The ANO-2 methodology is a risk-informed safety classification and treatment program for repair/replacement activities for Class 2 and Class 3 pressure retaining items and their associated supports (exclusive of Class CC and Class MC items), using a modification of ASME Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1 (Reference 33).

The ANO-2 methodology relies on the conditional core damage and large early release probabilities associated with pipe ruptures. Safety significance is generally measured by the frequency and the consequence of the event, in this case, pipe ruptures. Treatment requirements (including repair/replacement) only affect the frequency of passive component failure. Categorizing solely based on consequences, which measure the safety significance of the pipe given that it ruptures, is conservative compared to including the rupture frequency in the categorization. The categorization will not be affected by changes in frequency arising from changes to the treatment. Therefore, the NRC staff finds that the use of the repair/replacement methodology is acceptable and appropriate for passive component categorization of Class 2 and Class 3 SSCs.

In Section 3.1.2 of the LAR, the licensee stated:

The passive categorization process is intended to apply the same risk-informed process accepted by the NRC in the ANO2-R&R-004 for the passive categorization of Class 2, 3, and non-class components. All ASME Code Class 1 SSCs with a pressure retaining function, as well as supports, will be assigned high safety-significant, HSS, for passive categorization which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP.

Because all Class 1 SSCs and supports will be considered HSS, and only Class 2 and Class 3 SSCs will be categorized using the ANO-2 passive categorization methodology consistent with previous NRC staff approval, the NRC staff finds the licensees proposed approach for passive categorization is acceptable for the SSC categorization process and meets the requirements set forth in 10 CFR 50.69(b)(2)(ii) and (c)(1)(ii).

3.3.1.4 Risk Sensitivity Study (NEI 00-04, Section 8)

Sections 50.69(c)(1)(iv) and (b)(2)(iv) of 10 CFR require that any potential increases in CDF and LERF resulting from changes to treatment are small and that a description and basis for acceptability of the evaluation be conducted. Section 3.2.7 of the LAR states that an unreliability factor of three will be used for the sensitivity studies described in Section 8, Risk Sensitivity Study, of NEI 00-04, Revision 0. Section 3.2.7 of the LAR further confirms that a cumulative sensitivity study will be performed where the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in PRAs for all systems that have been categorized are increased by a factor of three. The NRC staff finds the application of a factor of three for the sensitivities is consistent with the guidance in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1.

In Section 3.1.1, Overall Categorization Process, of the LAR, the licensee specifically noted that the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence and that all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1)(iv). This cumulative sensitivity study, together with the periodic review process discussed in Section 3.3.2 of this SE, assure that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study. The NRC staff finds that the licensee will perform the risk sensitivity study consistent with the guidance in Section 8 of NEI 00-04, Revision 0, and therefore, will assure that the potential cumulative risk increase from the categorization is maintained acceptably low, as required by 10 CFR 50.69(c)(1)(iv).

3.3.1.5 Integrated Decision Making Section 50.69(c)(2) of 10 CFR requires SSCs to be categorized by an Integrated Decision-Making Panel staffed with expert plant-knowledgeable members whose expertise includes, at a minimum, PRA, safety analysis, plant operation, design engineering, and system engineering.

Appendix B of NUREG-0800, Chapter 19, Section 19.2, provides guidance and the NRC staff expectations for the licensees integrated decision-making process. The appendix states in part, Risk-informed applications are expected to require a process to integrate traditional engineering and probabilistic considerations to form the basis for acceptance. NEI 00-04, Revision 0, identifies two steps in the SSC categorization process that are responsible for the integrated assessment of the traditional engineering analyses and the risk results from the PRA and non-PRA assessments that are performed to make a determination and approval of the safety significance of the SSCs for categorization. These two steps are: (1) preliminary engineering categorization of function, and (2) IDP review and approval. The NRC staffs review of these two steps to ensure the processes is well-defined, systematic, repeatable, and scrutable are provided as follows.

3.3.1.5.1 Preliminary Engineering Categorization of Function (NEI 00-04, Section 7)

Section 3.1.1 of the LAR states, [c]ategorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201, which includes the determination of safety significance through the various elements identified above [LAR Figure 3-1]. All the information collected and evaluated in the licensees engineering evaluations will be provided to the IDP as described in Section 7 of NEI 00-04, Revision 0. As outlined in the LAR, the IDP will make the final decision about the safety significance of SSCs based on guidelines in NEI 00-04, Revision 0; the information it receives; and its expertise.

In Section 3.1.1 of the LAR, the licensee stated, in part,... if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6), the associated system function(s) would be identified as HSS. The licensee also stated, Once a system function is identified as HSS, then all the components that support that function are preliminary HSS.

The guidance in NEI 00-04, Revision 0, includes consideration of interfacing functions.

Section 7.1 of NEI 00-04, Revision 0, states, in part, Due to the overlap of functions and components, a significant number of components support multiple functions. In this case, the SSC or part thereof should be assigned the highest risk significance for any function that the SSC or part thereof supports. Section 4 of NEI 00-04, Revision 0, provides guidance for circumstances when the categorization of a candidate LSS SSC within the scope of the system being considered cannot be completed because it also supports an interfacing system. The guidance states, in part, In this case, the SSC will remain uncategorized until the interfacing system is considered. Therefore, the SSC will remain uncategorized and continue to receive its current level of treatment requirements.

In its supplement to the LAR dated June 25, 2020, in response to APLA RAI 02, the licensee provided an option to categorize an SSC(s) that supports functions in an interfacing system(s) without completing the categorization of that interfacing system(s). The licensee will perform NEI 00-04, Section 7.1 assessments, including defense in depth, of the interfacing SSC(s) that will assign a safety categorization to those supporting functions. The interfacing SSC(s) will then be assigned the highest categorization based on this assessment. The results of this assessment will be provided to the IDP for its evaluation in accordance with NEI 00-04, Section 9. The NRC staff recognizes that an SSC being categorized using this option is evaluated for all functions that it supports (as identified in accordance with NEI 00-04, Section 4), including interfacing functions consistent with NEI 00-04, Revision 0.

The NRC staff finds that the above description provided by the licensee for the preliminary categorization of functions is consistent with NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, and is, therefore, acceptable.

3.3.1.5.2 IDP Review and Approval (NEI 00-04, Sections 9 and 10)

In Section 3.1.1 of the LAR, the licensee stated that the IDP will be composed of a group of at least five experts who collectively have expertise in plant operation, design (mechanical and electrical) engineering, system engineering, safety analysis, and PRA. Therefore, the IDP will comprise the required expertise outlined in 10 CFR 50.69(c)(2).

The guidance in NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, addresses the IDP review and approval process, including the panel makeup and training, to provide confidence that the IDP expertise is sufficient to perform the categorization and that the results of the different evaluations (PRA and non-PRA) are used in an integrated, systematic process as required by 10 CFR 50.69(c)(1)(ii). In Section 3.1.1 of the LAR, the licensee discussed that at least three members of the IDP will have a minimum of 5 years of experience at the plant, and there will be at least one member of the IDP who has a minimum of 3 years of experience in modeling and updating of the plant-specific PRA. The licensee further states that the IDP will be trained in the specific technical aspects and requirements related to the SSC categorization process. This training will address, at a minimum, the purpose of the categorization; present treatment requirements for SSCs including requirements for design-basis events; PRA fundamentals; details of the plant-specific PRA, including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and the defense-in-depth philosophy and requirements to maintain this philosophy. The NRC staff reviewed the LAR and finds that the licensees IDP areas of expertise meet the requirements in 10 CFR 50.69(c)(2) and the additional descriptions of the IDP characteristics, training, processes, and decision guidelines are consistent with NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1.

The IDP may change the categorization of an SSC from LSS to HSS based on its assessment and decision making. As outlined in Section 10.2, Detailed SSC Categorization, of NEI 00-04, Revision 0, and confirmed by the licensee in Section 3.1.1 of the LAR, the IDPs ability to re-categorize components supporting an HSS function from HSS to LSS is limited and only available to the IDP based upon the prescribed steps in the NEI 00-04 guidance, as endorsed by RG 1.201, Revision 1.

As discussed in NEI 00-04, Revision 0, the only LSS SSC requirements that are relaxed for RISC-3 SSCs are those related to treatment - not design or capability, and 10 CFR 50.69(d)(2)(i) requires that the licensee ensures with reasonable confidence that RISC-3 SSCs remain capable of performing their safety-related functions under design-basis conditions. Therefore, the NRC staff finds that the IDP for the SSC categorization process is consistent with the endorsed guidance in NEI 00-04, Revision 0, and therefore, fulfills 10 CFR 50.69(c)(2).

3.3.1.6 Key Principle 4 Conclusions In light of the above, the NRC staff review for (1) IEPRA (includes internal floods) and FPRA acceptability, (2) PRA importance measures and integrated importance measure, (3) evaluation of the use of non-PRA methods, (4) risk sensitivity study, and (5) integrated decision making, concludes that the proposed change satisfies the fourth key principle for the risk-informed decision making prescribed in RG 1.174, Revision 3.

3.3.2 Key Principle 5: Monitor the Impact of the Proposed Change The guidance under NEI 00-04, Revision 0, includes programmatic configuration control and a periodic review to ensure all aspects of the 10 CFR 50.69 program, including use of PRA models, continue to reflect the as-built, as-operated plant, and that updates to the PRA are continually incorporated.

Sections 11 and 12 of NEI 00-04, Revision 0, include discussions on periodic review, program documentation, and change control. Maintaining change control and periodic review will also maintain confidence that all aspects of the 10 CFR 50.69 program, including risk categorization of SSCs, continually reflect the Hope Creek as-built, as-operated plant. A more detailed NRC staff review of these aspects is provided as follows.

3.3.2.1 Periodic Review (NEI 00-04, Section 12)

Section 50.69(e) of 10 CFR requires the licensee to review changes to the plant, operational practices, applicable plant and industry operational experience, and, as appropriate, update the PRA and SSC categorization and treatment processes.

In Section 3.2.6 of the LAR, the licensee described the process for maintaining and updating the Hope Creek PRA models used for the SSC categorization process. Consistent with NEI 00-04, the licensee confirmed that the Hope Creek risk management process ensures the applicable PRA models used in this application continue to reflect the as-built and as-operated plant. The licensees process includes provisions for monitoring issues affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience); assessing the risk impact of unincorporated changes; and controlling the model and associated computer files. The process also includes reevaluating previously categorized systems to ensure the continued validity of the categorization.

Routine PRA updates are performed every two refueling cycles at a minimum. The NRC staff finds the risk management process described by the licensee in the LAR is consistent with Section 12 of NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, and consistent with the requirements in 10 CFR 50.69(e). Furthermore, in light of the above, the NRC staff has determined that the proposed change satisfies the fifth key principle for risk-informed decision making prescribed in RG 1.174, Revision 3.

3.3.2.2 Program Documentation and Change Control (NEI 00-04, Section 11)

Section 50.69(f) of 10 CFR requires, in part, program documentation, change control, and records. In Section 3.2.6 of the LAR, the licensee stated that it will implement a process that addresses the requirements in Section 11 of NEI 00-04, Revision 0, pertaining to program documentation and change control records. Section 3.1.1 of the LAR states that the SSC categorization process documentation will include the following ten elements:

1.

Program procedures used in the categorization

2.

System functions, identified and categorized with the associated bases

3.

Mapping of components to support function(s)

4.

PRA model results, including sensitivity studies

5.

Hazards analyses, as applicable

6.

Passive categorization results and bases

7.

Categorization results including all associated bases and RISC classifications

8.

Component critical attributes for HSS SSCs

9.

Results of periodic reviews and SSC performance evaluations

10. IDP meeting minutes and qualification/training records for the IDP members The NRC staff also recognizes that for facilities licensed under 10 CFR Part 50, Appendix B, Criterion VI, Document Control, the procedures are considered formal plant documents that include requirements where [m]easures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality. The NRC staff finds that the elements provided in Section 3.1.1 of the LAR will be documented in formal licensee procedures consistent with Section 11 of NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, and are therefore, sufficient for meeting the 10 CFR 50.69(f) requirement for program documentation, change control, and records.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New Jersey State official was notified of the proposed issuance of the amendment on August 18, 2020. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 or changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, as published in the Federal Register (85 FR 5054; January 28, 2020), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

7.0 REFERENCES

1. PSEG Nuclear LLC, Letter to U.S. Nuclear Regulatory Commission, "Application to Adopt 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors,'" dated November 25, 2019 (Agencywide Documents Access and Management System (ADAMS Accession No. ML19330C961).

2. PSEG Nuclear LLC, Letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information Re: Adopt 10 CFR 50.69 LAR," dated June 25, 2020 (ADAMS Accession No. ML20177A535).

3. PSEG Nuclear LLC, Letter to U.S. Nuclear Regulatory Commission, "Supplement to License Amendment Request to Adopt 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors,'" dated July 21, 2020 (ADAMS Accession No. ML20203M297).

4. Markley, M. T., U.S. Nuclear Regulatory Commission, Letter to Vice President, Operations, Entergy Operations, Inc., "Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative AN02-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activites in Class 2 and 3 Moderate and High Energy Safety Systems," dated April 22, 2009 (ADAMS Accession No. ML090930246).

5. U.S. Nuclear Regulatory Commission, Generic Letter 88-20, Supplement 4, "Individual Plant Examination of External Events (IPEEEs) for Severe Accident Vulnerabilities -

10 CFR 50.54(f)," dated June 28, 1991.

6. American Society of Mechnical Engineers and American Nuclear Society, PRA Standard ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated February 2009, New York, NY (copyright).

7. U.S. Nuclear Regulatory Commission, Letter to PSEG Nuclear LLC, "Hope Creek - Final RAI RE: Adopt 10 CFR 50.69 LAR (L-2019-LLA-0265)," dated June 15, 2020 (ADAMS Accession No. ML20167A330).

8. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.201, Revision 1, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance," dated May 2006 (ADAMS Accession No. ML061090627).

9. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed Activities," dated March 2009 (ADAMS Accession No. ML090410014).

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

11. U.S. Nuclear Regulatory Commission, NUREG-1855, Revision 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, Final Report," published March 2017 (ADAMS Accession No. ML17062A466).

12. U.S. Nuclear Regulatory Commission, NUREG-0800, Chapter 19, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance," dated June 2007 (ADAMS Accession No. ML071700658).

13. Nuclear Energy Institute, NEI 00-04, Revsion 0, "10 CFR 50.69 SSC Categorization Guideline," dated July 2005 (ADAMS Accession No. ML052910035).

14. Electric Power Research Institute, Report 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," published July 26, 2018 (publicly available for free at https://www.epri.com/#/pages/product/000000003002012988/?lang=en-US.

15. Nuclear Management and Resources Council (NUMARC) 91-06, "Guidelines for Industry Actions to Assess Shutdown Management," dated December 1991 (ADAMS Accession No. ML14365A203).

16. Anderson, V. K., Nuclear Energy Institute, Letter to Stacey Rosenbergy, U.S. Nuclear Regulatory Commission, "Final Revision of Appendix X to NEI 05-04/07-12/12-[13],

Close-Out of Facts and Observations," dated February 21, 2017 (ADAMS Package Accession No. ML17086A431).

17. PSEG Nuclear LLC, Letter to U.S. Nuclear Regulatory Commission, "Hope Creek Generating Station - Issuance of Amendment No. 215 Re: Inverter Allowed Outage Time Extension (EPID L-2018-LLA-0101)," dated March 27, 2019 (ADAMS Accession No. ML19065A156).

18. U.S. Nuclear Regulatory Commission, NUREG/CR-6928, "Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants," Idaho National Laboratory, published February 2007 (ADAMS Accession No. ML070650650).

19. Nuclear Energy Institute, NEI 07-12, Revision 1, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines," dated June 2010 (ADAMS Accession No. ML102230070).

20. U.S. Nuclear Regulatory Commission, NUREG/CR-6850, "Electric Power Research Institute (EPRl)/NRC-RES Fire PRA Methodology for Nuclear Power Facilities" (ADAMS Accession Nos. ML052580075, ML052580118, and ML103090242).

21. U.S. Nuclear Regulatory Commission, NUREG-1792, "Good Practices for Implementing Human Reliability Analysis (HRA), Final Report," Science Applications International Corporation and Sandia National Laboratories, dated April 2005 (ADAMS Accession No. ML051160213).

22. Giitter, J., and Ross-Lee, M. J., U.S. Nuclear Regulatory Commission, Letter to Krueger, G., Nuclear Energy Institute, "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os)," dated May 3, 2017 (ADAMS Accession No. ML17079A427).

23. Reisi-Fard, M., U.S. Nuclear Regulatory Commission, Memorandum to Giitter, J. G.,

U.S. Nuclear Regulatory Commission, "Assessment of the Nuclear Energy Institute 16-06,

'Crediting Mitigating Strategies in Risk-Informed Decision Making,' Guidance for Risk-Informed Changes to Plants Licensing Basis," dated May 30, 2017 (ADAMS Accession ML17031A269).

24. Marshall, Jr., M. L., U.S. Nuclear Regulatory Commission, Letter to Hanson, B. C., Exelon Generation Company, LLC, "Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 332 and 210 Re: Risk-informed Categorization and Treatment of Systems, Structures, and Components for Nuclear Power Reactors" (EPID L-2018-LLA-0482)," dated February 28, 2020 (ADAMS Accession No. ML19330D909).

25. Davison, P. J., PSEG Nuclear LLC, Letter to U.S. Nuclear Regulatory Commission, Hope Creek Generating Station, Seismic Hazard and Screening Report (CEUS Sites), Response to 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," dated March 28, 2014 (ADAMS Accession No. ML14087A436).

26. Vega, F. G., U.S. Nuclear Regulatory Commission, Letter to Braun, R., PSEG Nuclear LLC, "Hope Creek Generating Station - Staff Assessment of Information Povided 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 and Staff Closure of Activities Associated with Recommendation, 2.1, 'Seismic' (CAC No. MF3924)," dated February 29, 2016 (ADAMS Accession No. ML16049A609).

27. U.S. Nuclear Regulatory Commission, "Review of Individual Plant Examination of External Events (IPEEE) Submittal for Hope Creek Generating Station (TAC NO. M83630)," dated July 26, 1999.

28. Carr, E., PSEG Nuclear LLC, Letter to U.S. Nuclear Regulatory Commission, Hope Creek Generating Station Focused Evaluation of External Flooding," dated June 27, 2017 (ADAMS Accession No. ML17178A307).

29. Boska, J., U.S. Nuclear Regulatory Commission, Letter to Sena, III, P. P., PSEG Nuclear LLC, "Hope Creek Generating Station - Staff Assessment of Flooding Focused Evaluation (CAC NO. MG0036)," dated October 17, 2017 (ADAMS Accession No. ML17275A945).

30. Hope Creek Generating Station Updated Final Safety Analysis Report, Revision 23.

31. U.S. Nuclear Regulatory Commission, NUREG/CR-4461, Revision 2, "Tornado Climatology of the Contiguous United States," Pacific Northwest National Laboratory, published February 2007 (ADAMS Accession No. ML070810400).

32. U.S Nuclear Regulatory Commission, "NRC Regulatory Issue Summary 2015-06 Tornado Missile Protection," dated June 10, 2015 (ADAMS Accession No. ML15020A419).

33. American Society of Mechanical Engineers, ASME Code Case N-660, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1," dated July 2002.

34. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan,"

Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance," dated June 2007 (ADAMS Accession No. ML071700658).

35. U.S. Nuclear Regulatory Commission, Letter to PSEG Nuclear LLC, Hope Creek Generating Station - Audit Plan and Setup of Online Reference Portal for License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, dated January 24, 2020 (ADAMS Accession No. ML20023A994).

Principal Contributors: T. Hilsmeier J. Circle D. Wu S. Park Date: September 29, 2020

ML20231A632

• by memorandum **by e-mail OFFICE DORL/LPL1/PM DORL/LPL1/LA DRA/APLA/BC(A)*

DRA/APLC/BC*

DEX/EICB/BC**

NAME JKim LRonewicz JCircle SRosenberg MWaters DATE 08/24/2020 08/24/2020 08/13/2020 08/13/2020 08/12/2020 OFFICE DSS/SNSB/BC** DEX/EEOB/BC** DEX/EMIB/BC**

DNRL/NVIB/BC**

DNRL/NPHP/BC**

NAME SKrepel BTitus ABuford HGonzalez MMitchell DATE 08/31/2020 08/28/2020 08/25/2020 08/27/2020 08/24/2020 OFFICE OGC - NLO**

DORL/LPL1/BC DORL/LPL1/PM NAME CCarson JDanna JKim DATE 09/21/2020 09/29/2020 09/29/2020