ML24261B974
| ML24261B974 | |
| Person / Time | |
|---|---|
| Issue date: | 09/17/2024 |
| From: | Office of Nuclear Reactor Regulation, Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| Download: ML24261B974 (122) | |
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Advanced Reactor Stakeholder Public Meeting September 18, 2024 Microsoft Teams Meeting Bridge line: 301-576-2978 Conference ID: 425 132 082#
Time Agenda Speaker 9:50 - 10:00 am Advanced Reactor Ready Slides 10:00 - 10:15am Opening Remarks NRC 10:15 - 10:35 am Safety Evaluation Template Development Initiative for the LMP-Based Applications NRC 10:35 - 11:00 am Alternative Risk-Informed, Technology-Inclusive Approaches to Advanced Reactor Regulation SECY Paper NRC 11:00-11:10 am Announce newly issued proposed rule for power reactor security NRC 11:10-11:30 Present NEI 24-05: Risk-informed Performance-based Emergency Planning ANL LUNCH 1:00 - 1:30 pm ADVANCE Act Report to Congress involving Nuclear Reactor Application Environmental Reviews NRC 1:30 - 1: 45 pm ADVANCE Act Advanced Reactor Topics NRC 2
Time Agenda Speaker 1:45-2: 00 pm NEI draft technical report on Fire Brigade Staffing NEI 2:00- 2: 15 pm Update on NEI 22-04/ISO-9001 and AR Codes and Standards NEI 2:15 - 2:30 pm Public Comment Period NRC 2:30 - 2:35 pm Closing Remarks NRC 3
Opening Remarks
Advanced Reactor Program Recent Highlights
Advanced Reactor Program Highlights (Continued)
- The revised Part 53 proposed rulemaking package was transmitted to the Commission on September 4, 2024, for their review before publication in the Federal Register, in accordance with Commission direction in SRM-SECY-23-0021.
- On Monday, September 16, the Office of Nuclear Reactor Regulation (NRR) issued a construction permit (CP) and associated safety evaluation (SE) to Abilene Christian University (ACU) for its Molten Salt Research Reactor (MSRR) to be located on its campus in Abilene, Texas. The CP and SE were issued ahead of the September 30, 2024, public milestone and were completed within the published resource estimate.
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Highlights (continued)
September 25, 2024 -2024 NRC Standards Forum
- This will be a hybrid meeting:
- In-person at the TWFN Auditorium
- Online via MS Teams
- The meeting notice and registration page are available:
- Meeting Notice (https://www.nrc.gov/pmns/mtg?do=details&Code=20240927)
- Registration Page (https://events.gcc.teams.microsoft.com/event/3da8da0e-f4ed-4ec0-92e8-1a8e75daffcb@e8d01475-c3b5-436a-a065-5def4c64f52e) 7
Development of Advanced Reactor Safety Evaluation Templates Ian Jung NRR/DANU 8
Introduction
- NRC is implementing strategies to improve the effectiveness and efficiency of its licensing reviews while maintaining its safety focus.
- The use of templates to standardize and streamline the safety evaluation (SE) writing process is essential to managing an increasing workload.
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Background===
- SE templates can be used for the review of new and advanced reactor applications and topical reports
- Initiatives on the more safety-focused reviews led to new or revised SE templates and associated staff guidance
- Consistent, clear, complete, and concise SEs support our principles of efficiency, openness, and reliability
- Based on lessons learned from past reviews, SE structure reflects scope and depth of review, commensurate with the risk or safety significance of the issues 10
ARCAP/TICAP
- ARCAP/TICAP provides guidance on content of application for the LMP-based applications
- ARCAP Interim staff guidance (ISG) documents developed to cover the remaining portions of the content of applications not addressed by the LMP.
- Chapters (1-12) and associated contents are very different from those of NUREG-0800.
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ARCAP Chapters Chapter 1 - General Plant Information, Site Description, and Overview of the Safety Analysis Chapter 2 - Methodologies, Analyses, and Site Evaluations Chapter 3 - Licensing Basis Events Chapter 4 - Integrated Evaluations Chapter 5 - Safety Functions, Design Criteria, and SSC Categorization Chapter 6 - Safety-Related SSC Criteria and Capabilities Chapter 7 - Non-Safety-Related with Special Treatment (NSRST) SSC Criteria and Capabilities Chapter 8 - Plant Programs Chapter 9 - Control of Routine Plant Radioactive Effluents, Plant Contamination, and Solid Waste Chapter 10 - Control of Occupational Dose Chapter 11 - Organization and Human-Systems Considerations Chapter 12 - Post Construction Inspection, Testing, and Analysis Program.
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Template Structure Example 3.1 Probabilistic Risk Assessment 3.1.1 Introduction
[Briefly describe the primary subject (e.g., system, function, program, and plan) for the Chapter.
The staff can usually find this information in the application such as Preliminary Safety Analysis Report (PSAR) or Final Safety Analysis Report (FSAR). The description should be purposely concise and provide the reader with a basic understanding of the subject. It should point to the application section for details to minimize unnecessary duplication.]
Section 3.1 of the Kemmerer 1 PSAR summarizes the probabilistic risk assessment (PRA) used as the primary tool for implementing the risk-informed and performance-based Licensing Modernization Project (LMP) methodology in NEI 18-04, Revision 1, as endorsed in RG 1.233, Revision 0, which includes identifying licensing basis events (LBEs), determining the classification of structures, systems and components (SSCs) and their special treatments, and evaluating defense-in-depth (DID) adequacy. Review of Section 3.1 for a CP application should acknowledge the potentially preliminary nature of the design and PRA.
3.1.2 Regulatory Evaluation 3.1.3 Technical Evaluation 3.1.4 Conclusion 3.1.5 References
- Template contains pre-populated languages, partly based on application content, and guidance for the staff use.
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Status
- Generic templates, based on the LMP, under development
- Key LMP Chapters (e.g., Chapters 2-5) developed early
- Preliminary draft complete and to be updated when CP application is submitted 14
Path forward
- Create generic SE template for CP applications reflecting experience from Kemmerer 1 and Project Long Mott
- Continue refinement as needed 15
Summary
- The LMP has introduced a new licensing framework for applications and SE templates are being developed consistent with the ARCAP/TICAP guidance and LMP structure.
- The use of templates to standardize and streamline SE writing is an essential strategy to manage an increased licensing workload.
- Staff will continue to refine the templates and expand their use to additional application types, such as OLs and COLs.
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Acronym ARCAP Advanced Reactor Content of Application Project COL combined license CP construction permit DID defense-in-depth LBE licensing basis event LMP Licensing Modernization Project NEI Nuclear Energy Institute NRC U.S. Nuclear Regulatory Commission NSRST non-safety-related with special treatment OL operating license PRA probabilistic risk assessment PSAR preliminary safety analysis report RG Regulatory Guide SE safety evaluation SSC structure, system, and component TICAP Technology Inclusive Content of Application Project 17
Alternative Risk-Informed, Technology-Inclusive Approaches to Advanced Reactor Regulation Advanced Reactor Stakeholders Meeting September 18, 2024 Rebecca Ober, Advanced Reactor Policy Branch Angelica Gheen, Advanced Reactor Policy Branch U.S. Nuclear Regulatory Commission https://www.nrc.gov/reactors/new-reactors/advanced.html
Contents
- Goals of this presentation
- Background
- Proposed Options
- Attributes
- Next steps 19
Goals of this Presentation
- Seek stakeholder perspectives and input on alternative approaches to risk-informed, technology-inclusive approaches to advanced reactor regulation.
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Background===
- In 2019, the Nuclear Energy Innovation and Modernization Act (NEIMA) was signed into law which required the NRC to prepare the regulatory infrastructure to support the development and commercialization of advanced nuclear reactors.
- In response, the staff delivered to the Commission in March 2023 a draft proposed rule known as "Part 53" for advanced reactor regulation.
The draft proposed rule consisted of two distinct frameworks, known as Framework A and Framework B.
- The approach in Framework A highlights the role of PRA in risk-informed and performance-based approaches to identifying enhanced safety margins that can be used to justify operational flexibilities.
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Background===
- During initial development of Part 53, stakeholders indicated that some designers might find the use of PRA unduly restrictive. To address this feedback, the NRC developed an alternate approach to licensing in part 53, which became Framework B in the draft proposed rule.
- Framework B largely replicates the existing licensing approach in parts 50 and 52 but modifies it to be technology-inclusive.
- In addition, Framework B would require applicants to use risk insights from a PRA, or an alternative evaluation for risk insights (AERI), in a confirmatory role to support a largely deterministic safety analysis.
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Background===
- Staff delivered the draft proposed rule to the Commission in SECY 00211.
- In the resulting SRM2, the Commission disapproved the inclusion of the proposed Framework B in Part 53 and directed staff to develop a notation vote paper for Commission consideration that proposes options for the use of Framework B outside of the Part 53 rulemaking.
23 1SECY-23-0021, Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors, dated March 1, 2023 (ML21162A095).
2SRM-SECY-23-0021, Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors, dated March 4, 2024 (ML24064A047)
Potential Alternative Approaches
- 1. Update 10 CFR Parts 50 and 52 to include technology-inclusive provisions for advanced reactors.
- 2. Use a separate part in 10 CFR for Framework B.
- 3. Create a less prescriptive regulation where methods of compliance similar to Framework B could be located in guidance.
- 4. No action.
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Criteria for Evaluating Approaches
- In order to ensure each approach is appropriately developed and evaluated, staff is using the Principles of Good Regulation (Efficiency, Openness, Clarity, Reliability, and Independence) to determine applicable attributes.
- These attributes will be used to evaluate the advantages and disadvantages of each approach and help shape staffs recommendation to the Commission.
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Applicable Attribute Examples
- Efficiency
- Potential costs for both NRC and Stakeholders
- Timeframe for Completion
- Openness
- Stakeholder Engagement
- Clarity
- Predictability and Consistency of Reviews
- Enhancing the ability to implement a flexible approach
- Reliability
- Able to account for future changes in technology
- Independence
- Supports unbiased assessments of all information during decision-making process 26
Next Steps
- Identify additional opportunities for stakeholder engagement, as needed.
- Publish a draft white paper, expected in late 2024, to support more detailed discussion.
- Develop a Commission paper on alternative technology-inclusive, risk-informed approaches for advanced reactors where risk analyses are used in a supporting or complementary role 27
Discussion Items
- What licensing needs should an alternative framework address that cannot be met by either the existing Parts 50 and 52 frameworks or the proposed Part 53 framework??
- Are there other attributes that the NRC staff should consider when evaluating the potential alternatives?
- Other feedback or questions for the NRC staff?
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Draft Proposed Rule 10 CFR Parts 50, 52, and 73 Alternative Physical Security Requirements for Advanced Reactors August 9, 2024 Dennis Andrukat U.S. Nuclear Regulatory Commission Rulemaking Project Manager NMSS/REFS/RRPB
Meeting Specifics
- Notify stakeholders of the date and time of public meeting for the proposed rule
-September 19th from 1 PM - 4 PM
- Provide information to help stakeholders prepare comments on the Alternative Physical Security Requirements for Advanced Reactors proposed rule and draft regulatory guidance 30
Background and Status
- The NRC decided to pursue this rulemaking due to the emergence of new reactor designs, which may warrant different methods for meeting the NRCs physical security requirements.
- The NRC conducted extensive public outreach including soliciting comments on a regulatory basis document and hosting public meetings on the preliminary proposed rule language.
- The proposed rule was published in the Federal Register on August 9, 2024 (89 FR 65226). The 75-day comment period ends October 23, 2024.
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Proposed Rule and Related Documents
- Proposed Rule
- Citation: 89 FR 65226 (August 9, 2024)
- Web version (ML24178A370)
- Supporting & Related Material
- Draft Regulatory Analysis (ML24178A372)
- Draft Environmental Assessment (ML24178A374)
- Draft Supporting Statements for Information Collections (ML21334A009; ML22131A161; ML22131A167) 32
Guidance Documents
- DG-5072 / RG 5.90, Rev 0 (ML20041E037)
- Guidance for Alternative Physical Security Requirements for Small Modular Reactors and Non-Light-Water Reactors
- Early version posted by NRC for awareness on 02-05-24
- DG-5071 / RG 5.81, Rev 2 (ML22021B529) (Official Use Only)
- Target Set Identification and Development for Nuclear Power Reactors
- Withheld from public disclosure and can be made available upon request to those members of the public with a need to know.
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Proposed Rule Language FOCUS AREAS:
- 73.55(b)(3) - Added requirements specific to small modular reactors (SMRs) and non-light-water reactors (non-LWRs)
- 73.55(s) - Alternative physical security requirements
- 73.55(s)(1) - General requirements
- 73.55(s)(2) - Specific alternative physical security requirements 34
Tips for Preparing Comments 35
- Regulations.gov: Comment Form or
- Email: Rulemaking.Comments@nrc.gov or
- Mail: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 Attn: Rulemakings and Adjudications Staff Applies to all public comments on the proposed rule and DG-5072 (comments on DG-5071 follow a separate process, contact Lou Cubellis for more information) 36 How to submit a comment
- Commenters Checklist link available on this comment submission form webpage:
https://www.regulations.gov/commenton/NRC-2017-0227-0038
- Also available in a printable format (also referred to Tips for submitting comments) 37 Review the Commenters Checklist on Regulations.gov
- Public comment period ends: October 23, 2024
- Final rule to the Commission: September 9, 2025 (estimated)
- Final rule publication: March 2026 (estimated) 38 Next Steps
Thank You!
Dennis Andrukat Project Manager -NMSS/REFS Email: Dennis.Andrukat@nrc.gov Regulations.gov docket ID: NRC-2017-0227 Please provide feedback on this public meeting using this link: https://www.nrc.gov/public-involve/public-meetings/contactus.html 39
NEI 24-05:
AN APPROACH FOR RISK-INFORMED PERFORMANCE-BASED EMERGENCY PLANNING DAVE GRABASKAS Manager, Licensing and Risk Assessments Group Ben Chen, David Young, Bob Kahler, Karl Fleming, Amir Afzali, Dennis Henneke, Brandon Chisholm, Partha Chandran
MOTIVATION AND BACKGROUND EPZ GUIDANCE EPZ EXAMPLE PREVIOUS QUESTIONS EMERGENCY PLAN GUIDANCE
MOTIVATION AND BACKGROUND EPZ GUIDANCE EPZ EXAMPLE PREVIOUS QUESTIONS EMERGENCY PLAN GUIDANCE
MOTIVATION Emergency Preparedness:
Historically, the LWR fleet has utilized large, uniform emergency planning zones (EPZs)
- Plume exposure pathway EPZ: 10 miles
- Ingestion exposure pathway EPZ: 50 miles The advanced reactor industry seeks to size EPZs and develop emergency plans that are commensurate with plant risk to provide cost savings and operational simplicity The NRC recently finalized a new emergency planning rulemaking (new 50.160 pathway), which provides such flexibility While NRC guidance in RG 1.242 provides high-level methodologies, industry is seeking more detailed guidance, specifically associated with use of the Licensing Modernization Project (LMP) 43
Goal Establish an approach that leverages the insights from technology-inclusive RIPB design and licensing methods to develop an EP strategy that provides reasonable assurance of adequate protection of the public health and safety while allocating resources for dose savings in an efficient and effective manner.
Objective Develop an approach for deriving the plume exposure pathway EPZ and associated emergency plan elements (actions, resources, etc.) which integrates information from the LMP-based safety case.
Outcome Development of a guidance document that is submitted by industry to the NRC for review and endorsement PROJECT OVERVIEW NEI 24-05, Rev 0 44
PROJECT TEAM 45 Argonne Team (PI)
LMP Developers Amir Afzali (Aalo Atomics)
Karl Fleming (KNF Consulting)
Industry Dennis Henneke (GE-Vernova)
Partha Chandran (GE-Vernova)
Brandon Chisholm (Southern Company)
Emergency Preparedness Experts Bob Kahler (Former Branch Chief of NRC Emergency Preparedness Policy and Oversight Branch 2001-2021)
David Young (NEI, Senior Technical Advisor - Security and Incident Preparedness)
Expert Reviewers
- Mark Cunningham (Former Director of NRC Division of Risk Assessment)
- Keith Woodard (Radiological consequence expert)
- ANS Risk-Informed Emergency Preparedness Working Group 45
Scope Utilization of the Licensing Basis Events (LBEs) and associated attributes (such as frequency, timing, consequence, etc.) identified through the LMP approach as a comprehensive spectrum of potential events to inform:
- 1) The determination of the PEP EPZ.
- 2) The development of appropriate emergency plans (actions, resources, coordination, etc.), including consideration of the ingestion pathway.
PROJECT SCOPE AND BENEFITS 46 46
PROJECT SCOPE AND BENEFITS 47 Benefits
- Alignment with new 50.160 pathway alleviates need for exemptions
- Integration with LMP provides:
- A plant-wide analysis that can include all sources of radioactivity and all initiators (internal and external)
- A structured and comprehensive approach for credible event selection to inform EPZ sizing and development of the emergency plan
- Event frequency and consequence information from LBEs
- A consistent DID framework where emergency planning is part of DID adequacy analysis performed by the integrated decision-making process 47
MOTIVATION AND BACKGROUND EPZ GUIDANCE EPZ EXAMPLE PREVIOUS QUESTIONS EMERGENCY PLAN GUIDANCE
Plume Exposure Pathway (PEP) EPZ Determination:
The new 50.160 pathway retains the plume exposure pathway (PEP) EPZ but removes the ingestion pathway EPZ, which is addressed by referencing available capabilities The new EPZ regulation provided in 50.33(g)(2)(i) has two criteria:
NRC EP RULEMAKING (2023) 50.33(g)(2)(i): The plume exposure pathway EPZ is the area within which:
(A)Public dose, as defined in § 20.1003 of this chapter, is projected to exceed 10 mSv (1 rem) total effective dose equivalent over 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> from the release of radioactive materials from the facility considering accident likelihood and source term, timing of the accident sequence, and meteorology; AND (B) Pre-determined, prompt protective measures are necessary.
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Plume Exposure Pathway (PEP) EPZ Determination:
The new 50.160 pathway retains the plume exposure pathway (PEP) EPZ but removes the ingestion pathway EPZ, which is addressed by referencing available capabilities The new EPZ regulation provided in 50.33(g)(2)(i) has two criteria:
NRC EP RULEMAKING (2023) 50.33(g)(2)(i): The plume exposure pathway EPZ is the area within which:
(A)Public dose, as defined in § 20.1003 of this chapter, is projected to exceed 10 mSv (1 rem) total effective dose equivalent over 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> from the release of radioactive materials from the facility considering accident likelihood and source term, timing of the accident sequence, and meteorology; AND (B) Pre-determined, prompt protective measures are necessary.
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Plume Exposure Pathway (PEP) EPZ Determination:
The new 50.160 regulation is structured for three possible outcomes of the EPZ determination process:
Outcome Additional Emergency Plan Requirements1 Additional Requirement Description PEP EPZ > SB
- 50.160(b)(1)(iv)(B)
- 50.160(b)(3)
- Discuss offsite response
- 50.160(b)(3)
- Describe the PEP EPZ No PEP EPZ 1 In addition to the emergency plan requirements provided in 50.160(a), (b)(1)(i) - (iv)(A), (b)(2), (b)(4), and (c).
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52 Started with a clean-slate but used precedent as guide Taking a broader look at appropriate criteria while looking for simplifications and how LMP-derived information can be used to predictably provide data regarding the major considerations for 50.33(g)(2)(i)(A) (accident likelihood, source term, timing, meteorology)
Remembering the goal Establish an approach for EP that provides reasonable assurance of adequate protection while allocating resources in an efficient and effective manner for dose savings to workers and the public.
The EPZ is just one part of EP EPZ the area for predetermined, prompt protective actions.
The identified LBEs will also inform other aspects of EP.
PEP EPZ DETERMINATION PROCESS 52
53 PEP EPZ DETERMINATION PROCESS 53
54 PEP EPZ DETERMINATION PROCESS 54
55 PEP EPZ: SPECTRUM OF EVENTS The process starts with those LBEs (AOOs, DBEs, and BDBEs) with radionuclide release.
Preliminary screening is possible dependent on:
- Dose size (very small releases)
- Timing:
- Time from accident initiation to radionuclide release (including recognition of need for actions, which may not occur at time zero)
- Time from radionuclide release to when protective actions are necessary The preliminary screening reduces the effort necessary for subsequent analyses If a hazard is analyzed outside of the PRA/LMP, it is also included in the spectrum of events 55
56 PEP EPZ DETERMINATION PROCESS 56
57 PEP EPZ: EVENT AND DOSE EVALUATION For those non-screened LBEs with radionuclide release, a probabilistic dose aggregation is performed and dose-versus-distance curves* are created for 1 rem and 200 rem.
Aligns with Appendix A of RG 1.242 and the historic approach utilized in NUREG-0396.
Analyses are performed utilizing cumulative dose-versus-distance curves (a plant-holistic perspective).
- Applicants may select pre-determined distance for analysis, such as EAB.
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58 PEP EPZ: EVENT AND DOSE EVALUATION Aggregation Individual LBEs Cumulative Mean values of frequency and dose are utilized for the analysis 58
59 Determination of Frequency Criteria:
In NUREG-0396, probabilistic dose aggregation curves based on WASH-1400 were used to partially derive the 10-mile plume exposure EPZ A distance of 10 miles corresponds to:
Note: WASH-1400 only considered full power internal events and the radionuclide source from one reactor.
PEP EPZ: EVENT AND DOSE EVALUATION 59
60 Based on NUREG-0396, two frequencies were selected for the probabilistic dose aggregation curve comparison:
Criterion A:
- Aligns with 50.33(g)(2)(i)(A/B) and consistency with historic criteria.
Criterion B:
- Consistency with historic criteria
- Provides additional confidence regarding the need for predetermined, prompt actions for low frequency, potentially high consequence events The criteria guide decision-making, they are not strict quantitative thresholds PEP EPZ: EVENT AND DOSE EVALUATION 60
61 Following the assessment, an uncertainty analysis and cliff-edge evaluation is performed.
Aligns with Appendix B of RG 1.242 Specific approach left to vendor, given the diverse nature of uncertainties and methods for addressing, but could include:
- Assessment at 95th percentile
- Bounding/conservative analysis
- Sensitivity analysis The goal is to identify cases such as the following PEP EPZ: EVENT AND DOSE EVALUATION 61
62 Metric just below criterion Small change in metric causes significantly different result PEP EPZ: EVENT AND DOSE EVALUATION 62
63 PEP EPZ DETERMINATION PROCESS 63
64 If no distance is derived from the dose evaluation, then no PEP EPZ is established.
If a distance is derived within the SB, then the PEP EPZ is set at the SB, in accordance with the requirements in §50.160(b)(1)(iii)(B) for onsite protective actions (more on this later).
If a distance is derived beyond the SB, then a protective measures evaluation is performed to determine the need for prompt, predetermined protective measures.
PEP EPZ: PROTECTIVE MEASURES EVALUATION 64
65 Although doses may exceed the EPA PAGs beyond the SB, there may be situations where prompt, predetermined proactive actions are not necessary. Considerations include:
PEP EPZ: PROTECTIVE MEASURES EVALUATION LBE Characteristics
- Timing of release, including time for ad hoc actions
- Release characteristics, such as types and forms of radionuclides
- Initiating event type, such as external hazards Site Characteristics
- Population distribution, such as remote sites
- Release pathways and direction, such as spatial dose assessment results
- Presence of co-located facilities Effectiveness
- Evaluation of effectiveness of protective action strategies for dose savings
- Comparison of doses with and without actions
- Evaluation of capabilities of local organizations for ad hoc actions The findings of this analysis are reviewed by the IDPP to assess DID adequacy when considering uncertainties, model limitations, etc.
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66 PEP EPZ DETERMINATION PROCESS 66
67 PEP EPZ: DETERMINATION 67
MOTIVATION AND BACKGROUND EPZ GUIDANCE EPZ EXAMPLE PREVIOUS QUESTIONS EMERGENCY PLAN GUIDANCE 68
69 Example Plant Simplified example based on advanced reactor design and PRA experience Assume a uniform SB at a distance of 500m LMP Analysis and LBEs 38 LBEs identified, including 26 that involve radionuclide release No preliminary screening performed PEP EPZ: EXAMPLE 69
70 Probabilistic Dose Aggregation Aggregation PEP EPZ: EXAMPLE 70
71 Probabilistic Dose Aggregation Aggregation PEP EPZ: EXAMPLE Distance ~1000m 71
72 Uncertainty & Cliff-Edge Analysis PEP EPZ: EXAMPLE 72
73 Uncertainty & Cliff-Edge Analysis PEP EPZ: EXAMPLE Distance ~1250m 73
74 Protective Measures Evaluation Since the derived distance is beyond the SB, a protective measures evaluation is performed, which focuses on those LBEs contributing to the curve.
Only LBE-32 contributes to the 1 rem exceeding the PAGs beyond the SB; however, the protective measures evaluation determines prompt, predetermined protectives actions are not warranted.
PEP EPZ: EXAMPLE 74
75 Result PEP EPZ: EXAMPLE Analysis Step Assessment Spectrum of Events LBEs identified through LMP approach, no alternative hazard event selection considerations.
Event Evaluation and Dose assessment The LBE assessment resulted in the following findings:
Criterion A: 1 rem curve - Distance of 750m to 1000m Criterion B: 200 rem curve - No distance derived Uncertainty/Cliff-Edge Assessment: 1 rem curve may extend to 1250m due to uncertainty Protective Measures Evaluation Beyond the SB:
- One LBE contributed to doses exceeding the EPA PAGs beyond the SB; however, there is adequate time for OROs to implement protective measures.
- Predetermined, prompt protective measures are not warranted.
Within the SB:
- Protective measures were developed for onsite personnel, given the nature of the releases.
PEP EPZ Determination The analysis determined that doses exceeding the EPA PAGs were possible beyond the SB but predetermined, prompt protective measures are not warranted. Within the SB, protective measures are warranted. Therefore, the PEP EPZ is established at the SB (500m).
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MOTIVATION AND BACKGROUND EPZ GUIDANCE EPZ EXAMPLE PREVIOUS QUESTIONS EMERGENCY PLAN GUIDANCE 76
77 Question: Why not lower frequency criteria?
Consistent with NUREG-0396 and demonstrates that predetermined, prompt protective actions from the plant and local response organizations are not warranted as they are not an efficient and effective use of their resources:
Criterion A:
- If U.S. operating 100 reactors, 0.1% chance of a reactor event requiring protective actions outside the EPZ in a given year (NUREG-0396 consideration).
Criterion B:
- The likelihood of an event resulting in early health effects beyond the EPZ is less than 1 in 1,000,000 plant years.
PEP EPZ: DETERMINATION 77
78 PEP EPZ: DETERMINATION Question: What are the implications of a SB EPZ?
50.33(g)(2)(i) contains two criteria for EPZ determination:
- 1) Exceed 1 rem over 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> (with considerations of timing, source term, etc.) and
- 2) Prompt, predetermined protective measures are necessary However, predetermined, prompt protective measures are those actions taken by offsite response organizations (OROs) to protect the public in offsite locations.
Therefore, within the developed approach, protection of the public onsite is the responsibility of the licensee under §50.160(b)(1)(iii)(B) and included within the site response plan.
Further discussion with the NRC likely necessary to ensure consistent understanding and that regulations are being addressed properly.
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79 Question: What about Design Basis Accidents (DBAs)?
NUREG-0396 examined the consequence associated with a spectrum of potential accidents, derived from environmental reports, DBAs, and WASH-1400.
Within the developed approach, the LMP analysis includes a PRA that is far more comprehensive than WASH-1400.
- Event sequences that are analogous to DBAs are included in the PRA and addressed at their appropriate frequency level.
- Historically, DBAs were primary driver for plant design with PRA providing supplemental information. In LMP, the roles are essentially reversed, with PRA leading and DBAs providing supplemental insights.
Why not include DBAs without a consideration of frequency?
- The goal of the approach is to allocate resources in the most efficient and effective manner for dose savings. Utilizing realistic risk information is the best avenue for accomplishing this goal. Adding postulated accident sequences could distort the findings and cause a misallocation of resources
PEP EPZ: DETERMINATION 79
80 Question: What about security events?
The new NRC EP rulemaking does not include security events and they are not evaluated as part of LMP. However, the consideration of such events is important for a comprehensive EP strategy.
An applicant should state that security events are removed from detailed consideration in the facilitys PEP EPZ technical basis. This decision should be supported by documenting:
The LBEs that were used to establish the basis for the EPZ size, and Compliance with regulatory requirements to protect against applicable design-basis and beyond-design-basis threats.
The basis should also discuss the facilitys security-by-design features and available capabilities for mitigating beyond-design-basis events.
The applicant should conclude that, based on the above information, the consequences from security-related events are adequately considered in the determination of the PEP EPZ.
PEP EPZ: DETERMINATION 80
81 Question: Why integration with LMP?
1)
The DID adequacy reviews within LMP, including assessments by the IDPP, provide a structured, comprehensive framework that can be leveraged for EP decision-making, such as the protective measures evaluation 2)
The LMP integrated risk metric results (the QHOs) may depend on the execution of protective measures. Therefore, the results of the EPZ determination process must feed back to the LMP analyses to ensure consistency.
PEP EPZ: DETERMINATION 81
MOTIVATION AND BACKGROUND EPZ GUIDANCE EPZ EXAMPLE PREVIOUS QUESTIONS EMERGENCY PLAN GUIDANCE 82
83 Emergency Plan Guidance In general, only high-level guidance is provided given the diverse nature of advanced reactors, unlike existing NEI LWR EP guidance docs Focus areas include:
- Emergency classification levels (ECLs)
- Emergency action levels (EALs)
- Initiating conditions (ICs)
- Protective actions
- Hazard analysis EMERGENCY PLAN: OVERVIEW
§50.160 Description Comment (a)
Definitions No Additional Guidance Provided (b)(1)(i)
Maintenance of Performance Supplemental Guidance Provided (b)(1)(ii)
Performance Objectives No Additional Guidance Provided (b)(1)(iii) Emergency
Response
Supplemental Guidance Provided (b)(1)(iv) Planning Activities Supplemental Guidance Provided (b)(2)
Hazard Analysis Supplemental Guidance Provided (b)(3)
PEP EPZ Supplemental Guidance Provided (b)(4)
Ingestion Pathway Supplemental Guidance Provided (c)
Implementation No Additional Guidance Provided 83
84 Emergency Classification Levels (ECLs)
The ECL definitions in NEI 99-01 and 07-01 were revised to remove reactor design-specific attributes and to align with LMP terminology and structure A plant may not need all four ECLs depending on the characteristics of the derived LBEs. For example, a general emergency level may not be needed if no LBEs lead to offsite doses exceeding the PAGs EMERGENCY PLAN: ECLS Level Technology-inclusive Description Notification of Unusual Event Events are in progress or have occurred which indicate a potential degradation to a capability to perform a RSF, or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of capabilities providing RSFs occurs.
Alert Events are in progress or have occurred which involve an actual or potential substantial degradation in the capability to perform a RSF or a security event that involves probable life-threatening risk to site personnel or damage to safety significant SSCs because of hostile action. Any radionuclide releases are expected to be limited to small fractions of the EPA PAG exposure levels.
Site Area Emergency Events are in progress or have occurred which involve actual or likely failure of SSCs, or the capability, to perform a RSF or hostile action that results in intentional damage or malicious acts; 1.toward site personnel or equipment that could lead to the actual or likely failure or; 2.that prevent effective access to, equipment needed for the protection of the public.
Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.
General Emergency Events are in progress or have occurred which result in the failure to perform a RSF and involve actual or imminent release of radioactive material that would be reasonably expected to exceed EPA PAG exposure levels offsite. This includes degradation resulting from hostile actions.
84
85 Initiating Conditions and Emergency Action Levels Within LMP, required safety functions (RSFs) are discretized into different levels of design criteria for SR and non-safety related with special treatment (NSRST) SSCs.
This structure is leveraged to identify monitoring attributes for initiating conditions into different emergency action levels for specific LBEs. For example, monitoring pump performance levels as part of a loss-of-flow LBE that could lead to eventual radionuclide release.
EMERGENCY PLAN: IES AND EALS 85
86 Hazards Analysis
§50.160(b)(2) requires a hazard analysis of any contiguous or nearby facilities, including any credible hazards that could adversely impact the implementation of the emergency plan.
The developed approach outlines three types of scenarios to be considered:
External hazard:
o An external hazard, such as natural phenomena, impacts the nuclear plant and the co-located/nearby facility simultaneously.
o Example: A seismic event that results in damage at the nuclear plant and the release of toxic material from a co-located/nearby chemical facility.
Nuclear plant event initiated by an event at the contiguous/nearby facility:
o An event at the co-located/nearby facility and the resulting hazard causes a condition at the nuclear plant that may jeopardize RSF performance (i.e., result in an EAL threshold exceedance).
o Example: The release of toxic gas for a co-located/nearby chemical facility impacts the operation of the nuclear plant.
Event at the contiguous/nearby facility initiated by a nuclear plant event:
o An event at the nuclear power plant impacts a co-located/nearby facility and results in an additional hazard.
o Example: A release of radioactive material from the nuclear plant results in operational disruptions at a co-located/nearby chemical facility and the release of toxic gases.
EMERGENCY PLAN: HAZARDS ANALYSIS 86
LUNCH BREAK MEETING WILL RESUME AT 1:00 PM EDT Microsoft Teams Meeting Bridge line: 301-576-2978 Conference ID: 425 132 082#
88
ADVANCE Act Congressional Report on Environmental Reviews of Nuclear Reactor Applications (#ADVANCENRC)
Lance Rakovan Senior Environmental Project Manager Ted Smith Branch Chief 89
ADVANCE Act
- On 7/9/2024, the president signed into law the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy (ADVANCE) Act - bipartisan legislation to provide a major boost to the future of nuclear energy in America
- Section 506 - Modernization of Nuclear Reactor Environmental Reviews 90
ADVANCE Act Section 506 Under Section 506 of the ADVANCE Act, Congress has directed the NRC with developing a report on the agencys efforts to facilitate efficient, timely, and predictable environmental reviews of nuclear reactor applications under section 103 of the Atomic Energy Act (AEA) of 1954 (42 U.S.C. 2133), including expanded use of categorical exclusions, environmental assessments, and generic environmental impact statements 91
ADVANCE Act Section 506 The NRC staff, led by the NRCs Environmental Center of Expertise (ECOE), will leverage the Fiscal Responsibility Act (FRA) amendments to the National Environmental Policy Act (NEPA) and the ADVANCE Act itself to enhance ongoing efforts to improve environmental review cost, timeliness, and predictability 92
ADVANCE Act Section 506 The scope of the NRCs report will include modernization of environmental reviews for all types of nuclear reactor applications (e.g., advanced reactors, license renewals, power uprate amendments, etc.) but will not include other licensing actions unrelated to AEA Section 103 nuclear reactor applications 93
ADVANCE Act Section 506
- Section 506 of the Act contains several items that NRC has been directed to consider as part of its report
- The NRC staff is examining lessons learned from recent environmental reviews across business lines, as well as stakeholder feedback, to achieve efficiencies beyond the new NEPA requirements 94
Public Meeting
- The NRC staff is seeking input from external stakeholders as it prepares a report to Congress on efforts to facilitate efficient, timely, and predictable environmental reviews for nuclear reactor applications
- Wednesday, September 25th at 1:00 pm ET
- Meeting details: ML24247A101 or https://www.nrc.gov/pmns/mtg?do=details&Code=20241112 95
Report Contacts
- Lance Rakovan, Congressional Report Lead, Office of Nuclear Material Safety and Safeguards (NMSS)
- Sarah Lopas, Back-up Lead, NMSS
Next Steps NRC will
- Issue a meeting summary for the public meeting
- Analyze the input received during the meeting to inform the report to Congress
- Issue the report to Congress by 1/5/2025 97
ADVANCE Act Advanced Reactor Topics
ADVANCE Act Advanced Reactor Topics Sec. 203, "Licensing considerations relating to use of nuclear energy for nonelectric applications" Sec. 206, "Regulatory issues for nuclear facilities at brownfield sites" Sec. 207, "Combined license review procedures" Sec. 208, "Regulatory requirements for micro-reactors" Sec. 401, "Report on advanced methods of manufacturing and construction for nuclear energy projects" 99
©2024 Nuclear Energy Institute 100 Overview of Technical Report
Background
Regulation and guidance document review Regulatory submittals and approval process Advancements in technology and analysis techniques Risk-informed, performance-based fire response evaluation process Fire Brigade Staffing
©2024 Nuclear Energy Institute 101 Overview of RI-PB Fire Response Evaluation
©2024 Nuclear Energy Institute 102 Plant specific input:
- Fire Safe Shutdown Analysis
- Fire PRA (if performed)
- Process
- Address need for onsite response
- Address potential for onsite incipient brigade
- Assess the acceptability of the offsite response capability Fire Response Evaluation Process
©2024 Nuclear Energy Institute 103 Micro-reactors are more like research and test reactors as recognized in the Supplementary Information in the publication of the final rule for Emergency Preparedness for Small Modular Reactors and Other New Technologies. Emergency Planning Zone (EPZ) criteria is outlined in 10 CFR 50.33(g)(2)(i)
Any fire-induced damage will not result in a public dose to exceed 10 mSv (1 rem) total effective dose equivalent at the site boundary for 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, and Pre-determined, prompt protective measures are established.
Ensure adequacy of offsite responders Offsite Responders Only
©2024 Nuclear Energy Institute 104 The incipient stage fire is defined in NFPA 600 as A fire which is in the initial or beginning stage and which can be controlled or extinguished by portable fire extinguishers, Class II standpipe, or small hose systems without the need for protective clothing or breathing apparatus.
Onsite Incipient Fire Brigade
©2024 Nuclear Energy Institute 105 The types of fires expected in the fire areas of concern.
Whether the types of fire can propagate beyond the ignition source.
Is the fire expected to be identified in the incipient stage?
If the fire is expected to progress, evaluate the detection and/or automatic suppression to address uncertainty.
Onsite Incipient Fire Brigade
©2024 Nuclear Energy Institute 106 Identify the desired fire response time What is the distance of the facility from response organization and response time?
Does the postulated fires require specialized training beyond offsite fire response organizations capabilities?
Can the reliability of the offsite responders be verified?
MOU to guarantee response and service level?
Acceptability of Offsite Responders
©2024 Nuclear Energy Institute 107 The site fire protection program plan should identify:
An individual with overall responsibility for the fire protection program.
An individual with the necessary level of understanding of the plant be available to oversee the fire response.
A process for maintaining configuration control of the fire protection program.
A method for ongoing demonstration of capability of the offsite fire response to effectively respond to fire events.
Maintain Basis for Acceptability
©2024 Nuclear Energy Institute 108 NEI 22-04/ISO-9001 and AR Codes and Standards Status Update NRC Advanced Reactor Stakeholder Meeting Mark Richter Technical Advisor Nuclear Energy Institute September 18, 2024
©2024 Nuclear Energy Institute 109 TR-CS-01: Alignment and Improvement of Codes and Standards ACTION for 2024: Identify additional gaps in, and any adjusted timelines for, advanced reactor codes and standards Consolidate and update prior advanced reactor codes and standards gap analysis Define development timelines for commercial relevance Prioritize gaps and associated actions Secure resources to address gaps in and timelines for advanced reactor codes and standards development Action Owners: ARCSC, SDOs, NEI, EPRI, AR Vendors Need Date: Gaps identified by end of 2024 NEI/EPRI North American Advanced Reactor Roadmap Assigned Actions to ARCSC
©2024 Nuclear Energy Institute 110 TR-CS-02 Risk-Informed and Performance-Based Approach
Demonstrate Risk-Informed and Performance-Based Approach Standard
Develop and execute a pilot project that applies Risk-Informed and Performance-Based (RIPB) methods in development of a new AR standard jointly with US and Canada-based SDOs (potential cross-cut with International Harmonization actions).
Action Owners: ARCSC Need Date: 2025 NEI/EPRI North American Advanced Reactor Roadmap Assigned Actions to ARCSC
©2024 Nuclear Energy Institute 111 ARCSC Activities to Date Fall 2022 December 1, 2022
- Kickoff Workshop Mar - Oct 2023
- Charter Developed
- Process Developed
- Collection of relevant standards
- Survey initiated
- Survey Data Collected November 30, 2023
- Annual Workshop to share preliminary survey results Winter/Spring 2024
- ARCSC Website deployed in February https://arcsc.nei.org/
- SDOs received Survey results and evaluate survey data.
Summer 2024
- Survey feedback from SDO members collected including identification of research needs.
- Formation of ARCSC
©2024 Nuclear Energy Institute 112 SDO committee questions:
- 1. Is there a gap identified? Y/N
- 2. [If Y] Committee disposition of gap: Persuasive (P) l Non-persuasive (NP) l Non-germane (NG) l Needs more investigation (NMI)
- 3. [If Persuasive] Proposed action to address gap: New standard l Update l Other solution
- 4. [Optional] Anticipated timeline for action to address gap: Start date l Completion date
- 5. [Optional] Anticipated resources needed for action to address gap: Liaisons lTechnical basis l Funding Process to Translate SDO Committee Responses of Master Spreadsheet SDO Designation Title Status Applicable to ARs?
Relevant topical area Gap identified from survey?
SDO input:
gap disposition (P, NP, NG, NMI)
SDO input:
proposed action to address gap SDO input:
timeline to address SDO input: resources needed (liaisons/input from other SDOs, R&D, RIB, funding)
©2024 Nuclear Energy Institute 113 ASME Priorities from ARCSC Gap Assessment Survey 18 September 2024
BPV Section III, Division 1: Seismic Analysis
BPV Section III, Division 5: High Temperature Materials for SMRs
BPV Section III, Division 5: Graphite Materials
BPV Section XI, Division 2: Inspection Protocols for Graphite and RIM for Sodium Fast Reactor
Operations & Maintenance Code: Update for advanced reactors, including non-LWRs
Qualification of Mechanical Equipment Standard:
Update for advanced reactors, including non-LWRs and risk-informed qualification processes
Nuclear Quality Assurance: Considerations for graded QA applications
©2024 Nuclear Energy Institute 114 Gap Analysis Example - ASME OM Code SDO Document Title Gap Description Proposed Action Resources Needed Priority ASME OM Code Operations &
Maintenance or Nuclear Power Plants Language in currently available OM Code still specifies water or LWR applications. OM-2's release and subsequent NRC endorsement will fill all gaps. OM-2 Rules for Inservice Testing Requirements for Pumps, Valves, and Dynamic Restraints at Nuclear Facilities OM 2 is being developed to address all types of advanced reactors. The ASME Operation and Maintenance (OM) Code Committee is presently addressing gaps between the existing fleet and new reactors. The new OM Code seeks to incorporate component level testing and inspection requirements for pumps, snubbers, and valves that are compatible with all of the new reactor designs represented by stakeholders in the ASME OM Code New Reactor Subcommittee. The present effort additionally seeks to align the QME and the OM Codes in a way that assists the regulatory authority and owners in the USA with the transition from construction to operation.
NRC endorsement High
©2024 Nuclear Energy Institute 115 ARCSC Website - Launched in February 2024 http://arcsc.nei.org NEI NEA, 10 September 2024
©2024 Nuclear Energy Institute 116
Builds on NRCs SECY-03-0117 conclusion that ISO-9001 offers viable path to meet Appendix B
Aligns with North American Advanced Reactor Roadmap supply chain action
NEI guidance document primarily for the purchaser/customer use
Content Process description for implementation Purchaser performs screening process for potential suppliers Identifies differences in requirements for ISO-9001 compliance with Appendix B Purchaser identifies supplier actions to address differences in requirements and achieve Appendix B compliance Part 21 compliance remains with the purchaser
Submit to NRC for review and endorsement by end of Q2 2024.
©2024 Nuclear Energy Institute 117 Goal: Expand the existing community of nuclear suppliers ISO-9001 focus is about meeting customers requirements (Appendix B)
ISO 9001 adaptable to any industry with programmatic rigor commensurate with industry demands There is a significant population of suppliers with ISO 9001 programs now Many ISO 9001 suppliers have robust QA programs and supply reliable products of high quality to other industries Nuclear suppliers already use ISO 9001 suppliers through commercial grade dedication U.S. suppliers qualified to globally accepted quality assurance programs will be more competitive Why focus on ISO 9001?
©2024 Nuclear Energy Institute 118 Implement NEI Policy Process to seek NEI member executive leadership insights and inform the final version for submittal Submit to NRC for review and endorsement (Q2 2025)
Develop a long-term strategy for using ISO-9001, ISO-19443 or other commercial programs which may include:
Implementation of pilot exercises (Evaluating component performance by making an identical part from both augmented ISO-9001and NQA-1 Appendix B programs to validate that they are equivalent)
Future expansion of NEI 22-04 (e.g., taking credit for the ISO accreditation process to eliminate the need for purchaser audits/surveys)
Develop Rulemaking accepting ISO-9001(and/or ISO-19443) directly for use for safety related SSCs (Long-term aspirational goal)
Whats Next?
©2024 Nuclear Energy Institute 119 Commercial quality programs as well as codes and standards are valuable aids in growing an efficient, competitive supply chain Global acceptance of codes and standards along with commercial quality programs that address advanced reactor needs will support design, licensing, procurement and construction The nexus of commercial quality programs and risk-informed codes and standards is now evident, for example, by emergent needs to construct civil structures at nuclear facilities commensurate with actual safety risk, or new advanced manufacturing methods Design once and build everywhere is our aspirational goal!
Future - Design Once, Build Everywhere ARCSC Workshop 2023 - IAEA Technical Meeting - December 14, 2023
©2024 Nuclear Energy Institute 120 Questions?
mar@nei.org
Public Comments 121
Closing Remarks October 30, 2024, Periodic Advanced Reactor Stakeholder Public Meeting