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December 7, 2023 Advanced Reactor Stakeholder Meeting Slides
	ML23335A104
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Issue date:	12/07/2023
From:	Subbaratnam R, Katie Wagner; NRC/NRR/DANU/UARP
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	Download: ML23335A104 (118)
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Advanced Reactor Stakeholder Public Meeting December 7, 2023 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 977 164 568 #

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Time Agenda Speaker 10:00 am - 10:10 am Opening Remarks NRC 10:10 am - 10:15 am Advanced Reactor Integrated Schedule NRC 10:15 - 10:25 am Advanced Reactor Digital Instrumentation & Control Workshop Announcement NRC 10:25 - 10:55 am Micro-Reactor Next Steps NRC 10:55 - 11:05 am BREAK 11:05 - 11:20 am Overview and Recent Experience with the NRCs Application Acceptance Review NRC Processes 11:20 - 11:50 am Final Rule on Emergency Preparedness for Small Modular Reactors (EPSMR) and NRC Other New Technologies 11:50 am - 1:00 pm LUNCH BREAK 1:00 - 2:00 pm Selection of a Seismic Scenario for an EPZ Boundary Determination NEI/NRC 2 2

Time Agenda (Continued) Speaker 2:00 - 2:30 pm Options for Optimizing Hearing Opportunities Associated with Two-Part NRC Applications 2:30 - 2:40 pm Closing Remarks NRC 2:45 pm Adjourn 3

Advanced Reactor Integrated Schedule of Activities (Slide 1 of 2)

Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory - SECY paper publication expected soon NRC and Nuclear Energy Institute (NEI) discussion on Draft NEI 23-01, Operator Cold License Training Plan for Advanced Nuclear Reactors - public meeting on Dec. 14 Draft Regulatory Guide (DG)-4034 (RG 4.7, Rev. 4), General Site Suitability Criteria for Nuclear Power Stations Publication on Oct. 12 (ML23123A090) & related public meeting on Oct. 27 Federal Register Notice for public comment published on Oct. 18, public comment period closed on Nov. 17, 2023 Staff reviewing public comments received from NEI (ML23326A031), NuScale Power, LLC (ML23326A030), and The Breakthrough Institute (ML23326A032) 4

Advanced Reactor Integrated Schedule of Activities (Slide 2 of 2)

Public Outreach Meeting for the Forthcoming TerraPower Natrium Demonstration Reactor Construction Permit Application held on Nov. 7 in Kemmerer, Wyoming Advanced Reactor Content of Application Project (ARCAP)/Technology Inclusive Content of Application Project (TICAP) Guidance Documents -

Advisory Committee on Reactor Safeguards (ACRS) briefing on Dec. 6 Material Compatibility Interim Staff Guidance - publication of final version expected this calendar year 5

NEW INTEGRATED SCHEDULE PUBLIC RELEASE PENDING 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 QUESTIONS 21

Advanced Reactor Digital I & C Workshop Announcement

Public Workshops on the I&C Licensing Framework for Advanced Reactors o Workshop 1 held on 23-Feb-2023 & 16-Mar-2023 o Workshop 2 held on 04-Apr-2023

Workshop meeting summaries available on newly added webpage to the advanced reactor rulemaking and guidance website dedicated to I&C o https://www.nrc.gov/reactors/new-reactors/advanced/modernizing/rulemaking-and-guidance/digital-instrumentation-and-control.html

Proposed agenda for next Workshop in Jan - Feb 2024 timeframe:

o SRM-SECY-22-0076, expansion of existing policy for digital I&C common-cause failures (SRM-SECY-93-087) to allow the use of risk-informed approaches to demonstrate appropriate level of defense-in-depth Non-light-water reactors that follow the risk-informed and performance-based approach in RG 1.233 and Design Review Guides for Instrumentation and Controls o Codes & Standards for Advanced Reactors I&C Systems 22

SRM-SECY-22-0076 The Commission has approved the staffs recommendation to expand the existing policy for digital instrumentation and control (I&C) common-cause failures to allow the use of risk-informed approaches to demonstrate the appropriate level of defense-in-depth, subject to the enclosed edits. The staff should clarify in the implementing guidance that the new policy is independent of the licensing pathway selected by reactor licensees and applicants. Given the regulatory importance of this issue, the staff should complete the final implementing guidance within a year from the date of this Staff Requirements Memorandum.

(emphasis added) 23

SECY-23-0092 In SECY-23-0092, staff informed the Commission of plans for updating implementing guidance addressing new policy for non-LWR DI&C reviews, in parts:

- While the language used in the DRG does not clearly connect to the revisions of the four points in SRM-SECY-22-0076, the language does not preclude the reviewers from considering alternative approaches. Therefore, the NRC staff will use pre-application engagement to discuss use of the expanded policy with interested applicants to address any questions or concerns. The NRC staff plans to revise the DRG, and possibly RG 1.233, in the future. The revision will address the differences in language discussed above and reflect any additional clarifications or improvements based on lessons learned by the NRC staff and prospective applicants, input received from the stakeholders during the ongoing advanced reactor I&C public workshops, and other interactions.

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Licensing and Deployment Considerations for Micro-Reactors:

Priorities and Next Steps Advanced Reactor Stakeholder Public Meeting December 7, 2023 William Kennedy Amy Cubbage Advanced Reactor Policy Branch U.S. Nuclear Regulatory Commission 25

Background

Updated NRC Staff Draft White Paper on Micro-Reactor Licensing and Deployment Considerations (ADAMS Accession No. ML23264A802)

Updated NRC Staff Draft White Paper on Micro-Reactor Licensing and Deployment Considerations - Enclosure (ML23264A803)

SECY-20-0093: Policy and Licensing Considerations Related to Micro-Reactors (ML20254A363)

NRC Staff Draft White Paper, Micro-reactors Licensing Strategies (ML21328A189) 26

NRC Staff Draft White Paper

Enhance regulatory clarity, predictability, and efficiency to address increasing stakeholder interest and novel licensing and deployment strategies

Describes regulatory approaches the NRC staff is developing for consideration by the Commission related to three topics:

1. Features to preclude criticality

2. Fuel loading at a factory

3. Operational testing at a factory

Includes an enclosure with information on other licensing and deployment topics and potential near-term strategies and next steps the NRC staff is considering 27

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Timeframe for authorization to operate at the deployment site

For licensing under 10 CFR Part 52, the NRC staff plans to clarify the circumstances under which the schedule for intended operation and initial fuel load can be accelerated and is considering ways to streamline public notifications, hearings, and the authorization to operate, as appropriate

For licensing under 10 CFR Part 50, the NRC staff is considering opportunities to expedite steps in the processing and review of applications for facility operating licenses, such as acceptance review and docketing, milestones for hearings, and the supplement to the environmental impact statement 28

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Licensing replacement reactors

The NRC staff previously addressed similar concepts and considered licensing options for multi-module facilities in SECY-11-0079 - License Structure for Multi-Module Facilities Related to Small Modular Nuclear Power Reactors (ML110620459)

The NRC staff is considering approaches under 10 CFR Part 50 and Part 52 where the construction permit application or combined license application would cover all reactors envisioned to be operated at the deployment site and each reactor would be authorized to begin operation under its own facility operating license or combined license once the Commission had made the required findings 29

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Autonomous and remote operations

The NRC staff plans to further develop its understanding of the industry deployment models for factory-fabricated micro-reactors with respect to industry plans for remote and autonomous operations, identify any gaps in the existing human factors engineering review needed to address the deployment models, and develop the technical bases for any new guidance that may be needed

The NRC staff is working with Idaho National Laboratory to organize a Remote Operations Workshop for industry participants to be held January 31 and February 1, 2024, at NRC Headquarters in Rockville, MD

- Please contact Stephanie Morrow (Stephanie.Morrow@nrc.gov) or Niav Hughes Green (Niav.Hughes@nrc.gov) with related inquiries.

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Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Transportation of fueled reactors

The NRC staff intends to use the existing regulatory framework (primarily 10 CFR Part 71) to review transportation of fueled commercial micro-reactors in the near term, which may include the use of the alternate test criteria in 10 CFR 71.41(c), the special package authorization option in 10 CFR 71.41(d), or exemptions, as appropriate 31

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Storage of fuel after irradiation in a power reactor

In order to use an independent spent fuel storage installation to store irradiated power reactor fuel withdrawn from a reactor that had undergone decay for less than a year, the licensee would be required to apply for a specific license under 10 CFR Part 72 and request and justify exemptions addressing the one-year decay time requirement in the regulations

The NRC staff intends to engage with stakeholders as they further develop their strategies for handling and storage of irradiated and spent fuel generated in factory-fabricated micro-reactors 32

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Decommissioning process and decommissioning funding assurance

In a scenario in which the reactor module is decommissioned away from the deployment site, the deployment site licensee would need to establish decommissioning funding assurance that considers the cost of removing the reactor from the site and decommissioning it elsewhere in addition to the cost of decommissioning activities at the deployment site.

The NRC staff may consider site-specific decommissioning cost estimates that appropriately account for all activities at both locations and all waste disposal costs 33

Other Licensing and Deployment Topics:

Potential Near-Term Strategies and Next Steps Siting in densely populated areas

In the near term, the staff will continue its effort to revise Regulatory Guide (RG) 4.7, General Site Suitability Criteria for Nuclear Power Stations, Revision 3, issued March 2014 (ML12188A053) and will review license applications in accordance with current Commission policy that allows alternative population-related criteria but precludes siting a commercial power reactor, no matter the size or type of reactor, within a population center of 25,000 residents or more 34

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Commercial maritime applications

The NRC staff will continue to engage with stakeholders and monitor developments related to commercial maritime applications and assess the need for future Commission direction Commercial space applications

If developers engage the NRC staff on terrestrial activities related to commercial space applications of factory-fabricated micro-reactors, the NRC staff intends to apply the established regulatory framework, as informed by the potential licensing approaches and strategies outlined in this presentation 35

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Commercial mobile applications

The NRC staff will monitor developments in the commercial sector related to deployment models and the demand for commercial mobile micro-reactor licensing.

The staff will assess the need for future Commission direction and rulemaking in this area.

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SECY-20-0093 Summary

SECY-20-0093 laid out several topics related to micro-reactor licensing and deployment, including information on the current regulations, applicability to micro-reactors, stakeholder perspectives, and NRC staff considerations

Some topics have been or are being addressed in rulemakings and guidance development, and some are topics considered in the white paper previously discussed in this presentation 37

SECY-20-0093 Summary

Security Requirements

Emergency Preparedness

Staffing, Training, and Qualification Requirements

Autonomous and Remote Operations

Regulatory Oversight

Aircraft Impact Assessment

Annual Fee Structure

Manufacturing Licenses and Transportation

Population-Related Siting Considerations

Environmental Considerations 38

Micro-reactor Licensing Strategies

NRC issued a draft white paper titled, Micro-reactors Licensing Strategies, to facilitate the development of optional strategies to streamline the licensing of micro-reactors Enhanced standardization of the design and operational programs

- Manufacturing license may provide flexibility for design and fabrication in a factory and reduce site-specific inspections and verifications

- Use of bounding values for external hazards and site characteristics could reduce NRC staff review effort

- Generic Environmental Impact Statement for Advanced Nuclear Reactors (ANR GEIS) rulemaking 39

Discussion Items

Are there other topics of interest that are not described in this presentation?

What do stakeholders see as the highest priority topics to address next?

Which regulatory topics pose the greatest risks to micro-reactor deployment?

Other feedback or questions 40

10 min BREAK 4141

Overview and Recent Experience with the NRCs Application Acceptance Review Processes December 7, 2023 Stephanie Devlin-Gill, Senior Project Manager Advanced Reactor Licensing Branch 1 Division of Advanced Reactors and Non-power Production and Utilization Facilities 42

NRC Application Acceptance Reviews:

Regulations and Guidance License applications are reviewed for completeness and acceptability for docketing consistent with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 2.101, Filing of application

10 CFR 2.101 Paragraph (a)(2) describes that

An application will be initially treated as a tendered, then the NRC performs an acceptance review to determination whether the application is complete and acceptable for docketing.

Generally, the determination on acceptability for docketing will be made within 30 days. In selected applications, the determination will be made within 60 days.

NRR has established procedures for conducting acceptance reviews in Office Instruction LIC-117, Acceptance Review Process for New Nuclear Facility Licensing Applications, (ML20283A188) 43 www.nrc.gov/reading-rm/doc-collections/cfr/part002/part002-0101.html

NRC Application Acceptance Reviews:

Possible Outcomes Possible outcomes which determine docketing and schedules:

- Application Acceptable for Docketing

A schedule is developed

- Application Not Acceptable for Docketing

Does not contain sufficient information to conduct the technical review or to develop a schedule and missing/incorrect information cannot be supplemented within 6 months.

- Acceptance Contingent on Receipt of Supplemental Information

Does not contain sufficient information to conduct the technical review or to develop a schedule and missing/incorrect information can likely be supplemented within 6 months. A provisional review schedule may be developed LIC-117, ML20283A188 44

NRC Application Acceptance Reviews:

Factors Influencing Schedule Development

- NEIMA Generic Schedules

- Quality of the application

- Standardization of the design

- Degree of pre-application engagement

- Custom regulatory approach

- Unresolved policy issues

- Risk-profile of the facility 45

NRC Application Acceptance Reviews:

High-Quality Applications

The NRC prioritizes the review of high-quality applications. Low-quality or incomplete applications can consume significant NRC and applicant resources and could divert attention and resources away from high-quality applications, resulting in potential unnecessary schedule delays.

Robust pre-application engagement, including participation in application readiness assessments, can assist in the preparation of high-quality applications.

Draft DANU-ISG-2022-01, Appendix A, Pre-Application Engagement Guidance, (ML22048B546)

NRR Office Instruction, LIC-116, Preapplication Readiness Assessment, (ML20104B698) 46

Thank You Questions?

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NRC Emergency Preparedness for Small Modular Reactors and Other New Technologies Todd Smith, PhD Senior Level Advisor for Emergency Preparedness and Incident Response Office of Nuclear Security and Incident Response U.S. Nuclear Regulatory Commission 48

Radiological emergency preparedness (EP) ensures protective actions can and will be taken is an independent layer of defense in depth provides dose savings is risk-informed 49

The NRC applies a graded approach to EP A graded approach is a risk-informed process in which the safety requirements and criteria are set commensurate to facility hazards Existing NRC regulations use a graded approach to EP Power reactors (low-power testing, power operations, decommissioning)

Research and test reactors Fuel Fabrication Facilities Independent Spent Fuel Storage Installations Monitored Retrievable Storage 50

Final Rule published in Federal Register Final rule (88 FR 80050) published on November 16, 2023, effective December 18, 2023 Final rule provides alternative requirements for small modular reactors and other new technologies https://www.federalregister.gov/emergency-preparedness-for-small-modular-reactors-and-other-new-technologies RG 1.242, Performance-Based Emergency Preparedness for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or Utilization Facilities (ADAMS Accession No. ML23226A036) 51

Preparedness begins with a proven planning basis The consequences from a spectrum of accidents, tempered by probability considerations, should be considered to scope the planning efforts for the distance to which planning for predetermined protective actions is warranted [the emergency planning zone (EPZ)]

the time-dependent characteristics of a potential release the type of radioactive materials 52

Major provisions of alternative EP regulations 10 CFR 50.160 provides an alternative framework for small modular reactors and other new technologies:

regulatory framework proportional to facility risk required EP functions set commensurate to radiological risk technology inclusive, performance based performance demonstration in drills and exercises hazard analysis for contiguous facilities ingestion planning capabilities scalable EPZ according to planning needs 53

Hazard Analysis Address the impact on emergency plan implementation from:

contiguous or nearby facilities and other credible hazards potential impacts of industrial plants, other reactors, transportation systems, or combination of factors site-specific, credible hazards from other facilities that may require additional EP considerations 54

Emergency response functions provide capabilities Event classification and mitigation Protective actions Communications Command and control Staffing and operations Radiological assessment Re-entry Critiques and corrective actions 55

Planning activities ensure readiness Prepare and issue public information during emergencies Implement the emergency plan in conjunction with the licensees safeguards contingency plan voice and data communications with the NRC Identify emergency facilities where effective direction and control can be exercised in an emergency Site familiarization training for offsite support Maintain the emergency plan 56

Scalable EPZ to support planning needs The EPZ is a planning tool, not a design feature The EPZ determination considers form and function:

The area within which public dose is projected to exceed 10 mSv (1 rem) TEDE over 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> considering:

accident likelihood source term timing of the accident sequence meteorology The area within which predetermined, prompt protective measures are warranted 57

Ingestion Pathway Planning Emphasizes capabilities and readiness to respond Identification of major exposure pathways for ingestion Identify resources available at all levels of government to sample, assess, and implement a quarantine or embargo of food and water to prevent ingestion 58

EP > EPZ 59

Roles and Responsibilities FEMA - Technical NRC State / Local Hazards Division Review Set EP standards &

Responsible for emergency reviews and approves Reviews State and local planning and response for Review Findings licensee plans radiological emergency the community plans against planning Considers FEMA findings of Assess their own standards offsite plans in reasonable Assist capabilities and needs assurance determinations Can assist State & locals in Voluntarily participate in planning, if asked licensee planning NRC Licensee Develops and maintains radiological emergency plans.

(NRC review and approval) 60

Coordination with offsite response organizations Coordination with offsite response organizations ensures readiness to respond within and beyond the EPZ as conditions warrant.

To facilitate predetermined prompt protective actions offsite, planning activities for an EPZ beyond the site boundary include:

Contacts and arrangements Protective actions Evacuation time estimate within the EPZ Primary and backup offsite response facilities Making and communicating dose projections Periodic emergency planning information for public General re-entry plans after an emergency Drill and exercise programs with offsite response 61

EP involves the Whole Community Licensee State/Local Federal 62

Risk-informed, performance-based is adaptable Flexibility for how to meet an EP function Flexible frequency of inspection and oversight Scaled response capabilities commensurate with the facility hazards Recognizes diversity in design and enhanced safety potential for evolutionary technologies

Technology propels the future of EP Geographic Digital Twins Information Systems Integrated Public Alert and Warning System Artificial Intelligence Federal capabilities 64

EP is the answer to uncertainty 65

Todd Smith, PhD todd.smith@nrc.gov 301-287-3744 66

LUNCH 67 67

Selection of Seismic Scenario for EPZ Sizing Determination December 7, 2023

Goal, Objectives, Scope Goal - Develop a technology-inclusive approach to selection of seismic scenario for EPZ determination Objectives

Be consistent with the philosophy discussed in NUREG-0396, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants

Allow definition of site-specific plant damage state for performing the EPZ sizing analysis required by 10 CFR 50.33(g)(2) [per SMR & ONT EP Rule]

Avoid over-reliance on the highly uncertain tails of the hazard curves

Not require a site-specific PRA prior to selection of the scenario Scope - Cover all new reactor designs except the following

Large (gigawatt scale) designs

Per RG 1.242, Facilities that use a maximum hypothetical accident should ensure that the estimated release is bounding for any event at the facility

The Two Pieces of this Puzzle Selection of the earthquake size to be used (the EPZ Earthquake)

Definition of the plant damage state given the occurrence of that earthquake (the EPZ Scenario)

Selection of the EPZ Earthquake Can We Risk-Inform the EPZ Earthquake?

What would a risk-informed framework look like?

What risk information do we have?

What insights can this information provide?

How can those insights be shaped into a framework?

Can that framework be applied to light-water and non-light-water SMRs?

What Would a Risk-informed Framework Look Like?

It would be anchored to the site GMRS (which serves as the plants safe shutdown earthquake).

It would lead to the selection of a significant earthquake beyond the GMRS.

It would encourage the reactor designers to suppress the contribution from the moderate earthquakes.

It would give credit for design features that address the highest risk contributors from the legacy fleet.

It would put more emphasis on lower release frequency as preferable to a greater dependence on emergency response.

What Risk Information Do We Have?

SPRAs for a small set of Gen II plants from the NTTF 2.1 submittals

These SPRAs have all been peer reviewed.

They are of high quality.

Arguably, they do represent the highest potential seismic risk plants relative to the rest of the fleet.

There is extremely limited SPRA information for the SMRs.

Mostly margins assessments.

Paper designs that lack seismic maturity.

Difficult to place in a true risk context.

What Insights Can This Information Provide?

What can we learn? There are insights to be gained from the current fleet SPRAs if we keep a few things in mind.

The new designs utilize a lot of passive systems, which eliminate many support system dependencies.

This will suppress (but not eliminate) that contribution of earthquakes that moderately exceed the design basis.

Design dependent (e.g., instrumentation and control panels and power)

For earthquakes that significantly exceed the design basis, there wont be that much difference. The dominant contributors will be structural and RCS component failures.

How Can Those Insights be Shaped Into a Framework?

How can we use this information?

Build a framework Given what we do know, and the NRC EPZ guidance, do the legacy plant SPRAs provide a path to risk-informing the overall approach?

Once the framework is formulated, can it be applied in a technology-inclusive context?

Look at the big picture and the insights, not each individual piece.

Is that picture inclusive, despite the risk profile differences?

Consider the Margin Plant core damage HCLPFs range from around 0.15g to 0.4g.

When compared to the SSEs for these plants, the HCLPFs range up to about 2.5 times the SSE.

When compared to the GMRS for these plants, the HCLPFs range up to about 2 times the GMRS.

The comparison to GMRS may be more informative, because these high hazard plants have been subjected to various upgrades over the years (i.e., they are no longer just designed for the SSE).

A Non-probabilistic Consideration Seismic design criteria for plant safety SSCs are not applicable to EP facilities, systems, and equipment.

The local infrastructure required for implementation of an offsite emergency response (e.g., roads and bridges, emergency operations centers, communications towers) are also not designed to NPP safety systems and are not within licensee control.

Therefore, at the HCLPF (0.01 CCDP) the conditional probability that emergency response capabilities would be severely degraded likely would be much higher.

This conditional probability is likely to be close to 1.0 at 1g.

Using This Risk Information Margin provided by the legacy plants that can be determined by looking at their seismic PRAs.

This is a conservative estimate since these are the plants with highest potential for seismic risk.

This provides a basis for a framework for risk-informed selection of the EPZ earthquake.

If we were creating a risk-informed framework for selection of the EPZ earthquake for a current plant, it would look like.

Proposed Risk-informed Framework for Selection the EPZ Earthquake This information provides useful insights that suggest we should be able to specify:

A multiplier to the site-specific GMRS (i.e., an event severity some factor above the design basis).

A reasonable upper bound cutoff PGA to select the seismic event to be evaluated in dose calculations for determining the size of the EPZ.

No magic formula or algorithm for this. Need to step back and take it all in holistically.

Our conclusion - the EPZ consequence calculation for any specific site should be based on the lower of:

An earthquake of two (2) times the site-specific GMRS, or

1.0g PGA

Can That framework be applied to LWR and non-LWR SMRs?

It is not suggested that the newer plants would exhibit the same risk profile as these legacy plants

HCLPF and HCLPF/GMRS ratios will be higher.

Risk profile will be dominated by failures of major structures and RCS vessels, piping and internals.

However, if the basis for this framework is a logical extension of the available seismic risk information for an existing plant, then it should be more than adequate for advanced plants. ©2023 Nuclear Energy Institute 81

Definition of the EPZ Plant Damage State (the EPZ Scenario)

How to Decide What Fails and What Succeeds One acceptable approach is to implement the concept of Mitigating Strategies Assessment for New Seismic Hazard Information (i.e., the Seismic MSA, Appendix H of NEI 12-06).

Established a precedent: The use of a 90% probability of success is equivalent to a 10% probability of unacceptable performance. This use of the 10% probability of unacceptable performance has been used in the past as a criteria for demonstrating seismic adequacy for beyond design basis seismic performance reviews in standards such as ASCE 43-05 and in commercial criteria such as ATC-63.

All SSCs whose C10% is less than the EPZ earthquake assumed to fail.

All SSCs whose C10% is greater than that EPZ earthquake assumed to succeed.

Applies to both safety, DID/RTNSS, and non-safety SSCs.

Cliff Edge Check Such effects have not generally been seen in seismic assessments and PRAs, as the risk and consequences tend to increase smoothly as earthquake severity increases.

Sensitivity analysis to be performed as a check.

SSCs whose C10% is within 10% higher than the EPZ earthquake should be added to the EPZ seismic scenario (i.e.,

assumed to fail).

This will determine whether the damage state is significantly altered (as it relates to the size of potential release).

Risk-informed Insights vis--vis NUREG-0396

Risk-informed Insights vis--vis NUREG-0396 The EPZ scenario frequency would include EPZ Seismic EPZ

The frequency of the 2 x GMRS earthquake Earthquake Scenario Exceedance Frequency

The probabilities of the failures (i.e., the list of Frequency (/year)

SSCs with C10% less than the EPZ earthquake). Plant (/year)

In a practical sense, we would expect that A B

2.3E-05 1.1E-05

<2.3E-06

<1.1E-06 the product of those SSC failure probabilities C D

1.4E-05 9.7E-06

<1.4E-06

<9.7E-07 would be at least 0.1, and likely less. E 1.0E-05 <1.0E-06 F 2.8E-05 <2.8E-06 Assuming 0.1, the EPZ seismic scenario G 1.4E-05 <1.4E-06 using this approach at these sites is less H I

1.1E-05 9.1E-06

<1.1E-06

<9.1E-07 than somewhere between 9E-7/yr and 3E- J 1.6E-05 <1.6E-06 K 1.2E-05 <1.2E-06 6/yr (below NUREG-0396 goal of 5E-5/yr.

Conclusions Conclusions Despite differences in seismic risk profiles, existing SPRA result (when carefully interpreted) provide useful insights for establishing the EPZ seismic scenario.

Since the framework is based on the NTTF 2.1 submittals, it likely bounds the understanding of what constitutes an adequate EPZ size.

Consistent with the guidance in RG 1.242, Appendix A, the seismic event used for the EPZ sizing determination will be a BDBE.

Consistent with the guidance in RG 1.242, Appendix A, and Appendix I to NUREG-0396, the scenario used for the dose assessment will be below 5x10-5 /yr, likely by an order of magnitude or more including uncertainty.

Conclusions The EPZ earthquake should be 2 x the site-specific GMRS, capped at 1.0g.

The EPZ seismic scenario should assume failure of all SSCs with C10%

less than the EPZ earthquake and success of all other SSCs.

The framework avoids dependence on the highly uncertain tails of the hazard curves.

Overall, the framework

ensures that a cliff edge effect is assessed,

incentivizes vendors to push high consequence seismic scenarios well out beyond the GMRS since this will lead to a smaller EPZ, and

benefits public safety by focusing on prevention over EP.

NRC Staff Feedback on NEIs draft White Paper, Selection of Seismic Scenario for EPZ Sizing Determination Eric Schrader, Clifford Munson and John Segala U.S. Nuclear Regulatory Commission 90

NRC Staff Feedback on NEIs draft White Paper

Based on the preliminary information provided, the staff has the following initial observations:

The concept of using a single BDB seismic event for design- and site-specific EPZ sizing could be acceptable.

Other approaches may also be found acceptable

Need additional discussion and detail in several areas discussed on the following slides

Future submittal would benefit from a demonstration/tabletop to help the staff understand the implementation of the proposed methodology.

91

Establishing Beyond Design Basis Ground Motion GMRS developed using ASCE 43 is a design basis ground motion that is risk-informed and performance-based GMRS as required by 10 CFR 100.23 is a free-surface and free-field ground motion Using two times the GMRS with a cap at 1.0g PGA for the beyond design basis ground motion results in a ground motion that is not risk-consistent from site to site It is not clear that capping the GMRS PGA at 1g at 100 Hz is appropriate.

100 Hz ground motions are not damaging to SSCs 92

Structural Response and C10 Capacity

Once the GMRS is established then the motion at the foundation levels of critical structures, referred to as the Foundation Input Response Spectrum (FIRS) is determined

Foundation level motion is then transferred into the plant structures to determine in-structure response spectra at various elevations in the plant facility

NEI approach uses the 2xGMRS 100 Hz value directly and not an in-structure motion for comparison with SSC seismic capacities

Use of C10 capacity for failure determination is made only at a single spectral frequency of 100 Hz

Most SSC resonance frequencies generally between 1 to 10 Hz

NEI approach does not cover how C10 capacity values will be determined and for which SSCs 93

Additional NRC Staff Feedback

The proposal is silent on how the source term will be determined from or assigned to the single BDB seismic sequence.

It is unclear how the result will be used to compare against criteria in 10 CFR 50.160 and how the methodology interfaces with the remainder of the rule.

NEI should address how potential cliff edge effects would be handled.

NEI should address how changes in the facility during the life of the plant would be addressed to assess any changes needed to the emergency plan.

How does this proposed approach coordinate with other ongoing activities regarding low frequency hazards for design and EPZ sizing for non-LWRs using LMP?

94

Options for Optimizing Hearing Opportunities for Two-Part Applications NRR/DANU/UARP Advanced Reactor Stakeholder Meeting December 7, 2023 95

Options for Optimizing Hearing Opportunities for Two-Part Applications Purpose

The purpose of the presentation is to inform stakeholders of staff initiative to develop regulatory options for Commission consideration for optimizing hearing opportunities associated with two-part applications

Have a dialogue with stakeholders on the topic 96

Options for Optimizing Hearing Opportunities for Two-Part Applications - Background

10 CFR 2.101(a)(5) allows for a construction permit (CP) or combined license (COL) application to be submitted in two parts, if each part is submitted within six months of the other

An exemption would be needed to submit an application in two parts where more than six months would elapse between filing each part

- There is precedent for applications being submitted in two parts with a greater than six-month gap between each part (e.g., Unistars Calvert Cliffs COL application) 97

Options for Optimizing Hearing Opportunities for Two-Part Applications - Options

The staff is considering providing the following two options to the Commission to address hearing opportunities for CP and COL applications submitted in two parts:

- Option 1: issue one Notice of Hearing after the entirety of the application is submitted with the direction that environmental contentions be submitted on the draft environmental impact statement (DEIS) if it is available instead of the environmental report

This would require the Commission to address the requirements in 10 CFR 2.309(f)(2) that contentions be submitted based on the environmental report 98

Options for Optimizing Hearing Opportunities for Two-Part Applications - Options (continued)

- Option 2: issue two Notices of Hearing:

(a) one when the environmental report is docketed and one when the other part is docketed

(b) An alternative to this option would be to issue one Notice of Hearing after the environmental report is received that also explains in detail the process for filing contentions on the second part upon its docketing 99

Options for Optimizing Hearing Opportunities for Two-Part Applications - Background

Commission Policy Statement on Conduct of New Reactor Licensing Proceedings Issued in April of 2008 (73 FR 20963)

- With two exceptions the Commissions policy is to issue a Notice of Hearing only when the entire application is submitted

- Neither of the exceptions involve circumstances that the NRC staff expects to encounter in the near future 100

Options for Optimizing Hearing Opportunities for Two-Part Applications - Discussion Items

Stakeholders perspectives on options

Stakeholders perspectives on proposal that environmental contentions be based on the NRC staffs DEIS if it is available versus the applicants environmental report.

- Publishing a Notice of Hearing after issuance of the DEIS and limiting contentions to the DEIS rather than the applicants ER would involve the Commission addressing the requirement in 10 CFR. § 2.309(f)(2) that environmental contentions must be filed on the ER.

101

Options for Optimizing Hearing Opportunities for Two-Part Applications - Discussion Items

Should approaches other than the environmental report being submitted first be considered?

- Submitting the ER first appears to be the more complicated scenario because the staffs review of the ER could lead to the DEIS being issued prior to the second hearing opportunity

- Possibility of the need for exemptions if the preliminary safety analysis report submitted with the environmental report does not contain all the information required by 10 CFR 2.101(a)(5)

Example of a recently issued exemption for this requirement can be found in a November 21, 2023, letter to Tennessee Valley Authority for the Clinch River Nuclear Site CP application (see: ML23045A008 and ML23114A098) 102

Future Meeting Planning

The next periodic stakeholder meetings are scheduled for January 24, 2024.

Potential topics for our next meeting include topics that might come up at the stakeholder meeting and other Stakeholder feedback.

If you have suggested topics, please reach out to Ramachandran Subbaratnam at Ramachandran.Subbaratnam@nrc.gov.

103

How Did We Do?

Click link to NRC public meeting information:

https://www.nrc.gov/pmns/mtg?do=details&Code=20230812

Then, click link to NRC public feedback form:

104

Back-up Slides Plant Margins What Margin do These Plants Have?

HCLPF (0.01 conditional probability of core damage) generally ranges from 0.15g to 0.4g.

There are a few outliers.

How does this relate to plant design basis?

What Margin do These Plants Have?

Margins generally up to about 2.5 times SSE.

Again, there are a few outliers.

While informative, this may paint an incomplete picture because of seismic upgrades that have been implemented.

What Margin do These Plants Have?

Updated seismic design criteria going forward - GMRS.

These plants have performed IPEEE, ESEP, seismic MSA in addition to their seismic PRAs.

These plants have effectively addressed the higher hazard associated with the current GMRS through plant improvements designed to reduce risk and/or increase margins.

Many such improvements have been implemented.

While these plants are not specifically designed to the GMRS (their design basis is still the original SSE), they have been improved relative to the GMRS.

What Margin do These Plants Have?

Margins up to about 2 times GMRS.

Decided downward trend in ratio as the GMRS increases.

Not surprising given that HCLPF values seem to be relatively insensitive to the site-specific design level.

Emergency Response Effectiveness Example: Offsite Power Grid Typically considered to have median failure acceleration of 0.3g.

High failure probability at low acceleration.

Essentially guaranteed failure by 1g.

No power to homes and businesses, no power to communications systems, no power for traffic controls, etc.

SOARCA Looked at ER for Seismic Considered a severe earthquake in the range of 0.5g-1.0g.

Concluded that there could be substantial damage to the emergency response infrastructure, the ability to take protective actions would not be significantly impeded.

Didnt really provide a basis for this - engineering judgment.

Did not consider the higher levels of damage (and additional failure modes) that would occur with earthquakes above 1.0g.

Not suggesting that there is a cliff at 1.0g; rather, it would be difficult to show effectiveness at higher levels.

More Risk-informed Insights vis--vis NUREG-0396

Risk-informed Insights vis--vis NUREG-0396 Using this approach, the GMRS GMRS and EPZ Earthquake earthquake frequencies GMRS Exceedance Frequency EPZ Earthquake EPZ Earthquake Exceedance for the sites in this study Plant A

(PGA) 0.19

(/year) 1.1E-04 (PGA) 0.38 Frequency (/year) 2.3E-05 are shown on the right. B 0.25 6.4E-05 0.50 1.1E-05 C 0.31 4.4E-05 0.61 1.4E-05 The resultant EPZ D 0.39 6.8E-05 0.78 9.7E-06 earthquake frequencies E F

0.26 0.18 6.0E-05 8.7E-05 0.52 0.36 1.0E-05 2.8E-05 are in the range of 9E- G 0.57 4.9E-05 1.00 0.74 1.4E-05 H 0.37 5.6E-05 1.1E-05 6/yr to 3E-5/yr. I 0.44 1.3E-04 0.88 9.1E-06 J 0.41 7.3E-05 0.82 1.6E-05 K 0.40 5.3E-05 0.80 1.2E-05

Risk-informed Insights vis--vis NUREG-0396 The EPZ scenario frequency would then EPZ EPZ Seismic Scenario Earthquake include the probabilities of the failures (i.e., Exceedance Frequency

(/year) the list of SSCs with C10% less than the EPZ Plant Frequency

(/year) earthquake). A B

2.3E-05 1.1E-05

<2.3E-06

<1.1E-06 In a practical sense, we would expect that C 1.4E-05 <1.4E-06 D 9.7E-06 <9.7E-07 the product of those SSC failure probabilities E 1.0E-05 <1.0E-06 would be at least 0.1, and likely less. G F 2.8E-05 1.4E-05

<2.8E-06

<1.4E-06 Assuming 0.1, the EPZ seismic scenario H 1.1E-05 <1.1E-06 I 9.1E-06 <9.1E-07 using this approach at these sites is less J 1.6E-05 <1.6E-06 than somewhere between 9E-7/yr and 3E- K 1.2E-05 <1.2E-06 6/yr.

Risk-informed Insights vis--vis NUREG-0396 How does that scenario frequency range comport with NUREG-0396?

R.G. 1.242, Section A-3.7 states.

The likelihood of exceeding a TEDE of 10 mSv (1 rem) at the proposed EPZ boundary should be consistent with the evaluation in Appendix I to NUREG-0396, which provides relative probabilities of exceeding certain critical doses as a function of distance from the facility for a spectrum of severe accidents. For example, NUREG-0396 examined the conditional probability of exceeding a variety of dose levels of interest, given a core melt accident with a stated frequency of 5x10-5 per reactor year.

Substitute release for core melt.

Using the proposed approach, the EPZ seismic scenarios would be at least an order of magnitude below this (i.e., a BDB accident state with frequency of approximately 3E-6/yr or less).

Risk-informed Insights vis--vis NUREG-0396 NRC guidance recommends that uncertainty also be 84%-tile EPZ 84%-tile EPZ 95%-tile EPZ 95%-tile EPZ considered. Earthquake Exceedance Seismic Scenario Earthquake Exceedance Seismic Scenario Table on right shows EPZ Frequency Frequency

(/year)

Frequency Frequency

(/year)

Plant Earthquake (/year) (/year) uncertainty in the hazard for A 0.38 3.8E-05 <3.8E-06 7.4E-05 <7.4E-06 B 0.50 1.7E-05 <1.7E-06 2.8E-05 <2.8E-06 the sites. C 0.61 2.9E-05 <2.9E-06 3.7E-05 <3.7E-06 D 0.78 1.4E-05 <1.4E-06 4.0E-05 <4.0E-06 Assuming, again, that the E 0.52 1.7E-05 <1.7E-06 n/a n/a scenario frequency is at F 0.36 4.3E-05 <4.3E-06 8.7E-05 <8.7E-06 G 1.00 2.2E-05 <2.2E-06 n/a n/a least an order of magnitude H I

0.74 0.88 1.8E-05 1.3E-05

<1.8E-06

<1.3E-06 n/a n/a n/a n/a lower, even the 95%-tile J K

0.82 0.80 2.8E-05 1.8E-05

<2.8E-06

<1.8E-05 n/a 3.5E-05 n/a

ML23335A104

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