ML21200A101
| ML21200A101 | |
| Person / Time | |
|---|---|
| Issue date: | 07/20/2021 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| Beall, Robert | |
| References | |
| 10 CFR Part 53, NRC-2019-0062, RIN 3150-AK31 | |
| Download: ML21200A101 (75) | |
Text
Advisory Committee on Reactor Safeguards (ACRS) Future Plant Designs Subcommittee 10 CFR Part 53 Licensing and Regulation of Advanced Nuclear Reactors Licensing Modernization Project Approach Subpart F - Requirements for Facility Operations: Emergency Preparedness (EP)
Status/Update on Technology-Inclusive Content of Application Project (TICAP) /
Advanced Reactor Content of Application Project (ARCAP) Development July 21, 2021 1
Agenda 9:30am - 9:40am Welcome / Introductions / Logistics / Goals 9:40am - 1:00pm Licensing Modernization Project (LMP)
Approach - Refresher Discussion 1:00pm - 2:00pm Lunch Break 2:00pm - 3:00pm Subpart F - Requirements for Facility Operations: Emergency Preparedness 3:00pm - 5:00pm Status/Update on TICAP/ARCAP Guidance Document Development 5:00pm - 6:00pm Discussion/Closing Remarks 2
NRC Staff Plan to Develop Part 53 Subpart B Subpart C Subpart D Subpart E Subpart F Subpart G Project Life Cycle Design and Siting Construction Operation Retirement Requirements Definition Analysis
- Safety Objectives External Facility Safety
- Safety Criteria System Hazards Construction/ Program
- Safety Functions & Component Manufacturing Design Site Surveillance Characteristics Ensuring Maintenance Analysis Capabilities/
Requirements Environmental Reliabilities Configuration Considerations Control Safety Change Control Categorization Staffing &
& Special Environmental Human Factors Treatment Considerations Programs Security, EP Other Plant/Site (Design, Construction, Configuration Control)
Clarify Subpart A General Provisions Analyses (Prevention, Mitigation, Compare to Criteria) Controls and Subpart J Admin & Reporting Plant Documents (Systems, Procedures, etc.) Distinctions Between Other 10 CFR Parts LB Documents (Applications, SAR, TS, etc.) Subparts H & I 3
NRC Staff Engagement Plan ACRS Interactions Concept/Introduction Discussion Interim Staff Resolution 4
LMP Approach 5
Licensing Modernization Risk-Informed, Technology-Inclusive Framework An Evolution 6
Nuclear Energy Institute (NEI) 18-04 General Approach
- Licensing Basis Events (LBEs) Selection
- Probabilistic Risk Assessment (PRA)
- Deterministic
- Structure, System, and Component (SSC)
Classification
- Function and Risk Considerations
- Safety Related (SR)
- Non-Safety Related with Special Treatment (NSRST)
- Defense-in-Depth (DID) Assessment
- SSCs
- Personnel
- Programmatic Controls
- Integrated Decision-making Process (IDP) 7
Key Considerations
- Integrated methodology consisting of three primary elements
- LBE Selection and Analyses
- SSC safety classification and performance requirements
- Assessing DID adequacy
- Uses existing regulatory criteria, primarily the guidelines for offsite dose and NRC safety goals
- Assessments performed using risk-informed and deterministic approaches, including IDP
- Includes methodology for assessing DID provided by plant capabilities and programmatic controls 8
Event Selection & Analysis
- Introduction of an actual frequency-consequence (F-C) curve as part of the regulatory process (vs. general relationship of decreased consequences expected for more frequent events) 9
Event Selection (F-C Curve)
- Purpose is to evaluate risk significance of individual LBEs and SSCs and to help define the required safety functions (RSFs)
- Derived from the next generation nuclear plant (NGNP) F-C Target and frequency bins for event categories
- F-C Target anchor points based on:
- 10 CFR 20 annual dose limits used to define iso-risk contour in AOO region
- Avoidance of offsite protective actions for lower frequency AOOs
- 10 CFR 50.34 dose limit for lowest frequency DBEs
- Consequences based on 30-day total effective dose equivalent (TEDE) dose at EAB
- EAB dose target for BDBEs related to NRC safety goal for limiting possibility of prompt fatality 10
Event Selection (F-C Curve)
- LBEs compared to F-C target based on mean, and upper (95th percentile) and lower (5th percentile) bound estimates of LBE frequency and dose
- When the uncertainty bands defined by the 5th percentile and 95th percentile of the frequency estimates straddles a frequency boundary, the LBE is evaluated in both LBE categories.
- Events considered per plant-year to address multi-unit and multi-source event sequences
- Separate cumulative risk targets used to ensure that 10 CFR 20 and QHOs met for the total risk over all the LBEs 11
Event Selection & Analysis AOOs Anticipated event sequences expected to occur one or more times during the life of a nuclear power plant, which may include one or more reactor modules. Event sequences with mean frequencies of 1x10-2/plant-year and greater are classified as AOOs. AOOs take into account the expected response of all SSCs within the plant, regardless of safety classification.
12
Event Selection & Analysis DBEs Infrequent event sequences that are not expected to occur in the life of a nuclear power plant, which may include one or more reactor modules, but are less likely than AOOs. Event sequences with mean frequencies of 1x10-4/plant-year to 1x10-2/plant-year are classified as DBEs. DBEs take into account the expected response of all SSCs within the plant regardless of safety classification.
13
Event Selection & Analysis BDBEs Rare event sequences that are not expected to occur in the life of a nuclear power plant, which may include one or more reactor modules, but are less likely than a DBE. Event sequences with mean frequencies of 5x10-7/plant-year to 1x10-4/plant-year are classified as BDBEs. BDBEs take into account the expected response of all SSCs within the plant regardless of safety classification.
14
- PRA capabilities to systematically identify an exhaustive set of design specific events are key to reproducible selection of candidate LBEs
- Although not required, early introduction of PRA into design process is encouraged to incorporate risk insights into design decisions
- Scope and level of detail consistent with scope and level of detail of design and site information and fit for purpose in risk-informed and performance-based decisions o All radiological sources o All plant operating states o All internal and external hazards
- Limitations and uncertainties identified in PRA addressed in the LMP approach to evaluating DID adequacy
- Uncertainties identified by PRA processes
- American Society of Mechanical Engineers (ASME) / American Nuclear Society (ANS) non-light-water reactor (LWR) PRA standard specifically designed to support LMP PRA applications to ensure PRA technical acceptability
- NRC endorsement of PRA standard will facilitate LMP implementation 15
Cumulative Risk Metrics
- The total frequency of exceeding an offsite boundary dose of 100 mrem shall not exceed 1/plant-year to ensure that the annual exposure limits in 10 CFR 20 are not exceeded.
- The average individual risk of early fatality within the area 1 mile of the EAB shall not exceed 5x10-7/plant-year to ensure that the NRC Safety Goal QHO for early fatality risk is met
- The average individual risk of latent cancer fatalities within the area 10 miles of the EAB shall not exceed 2x10-6/plant-year to ensure that the NRC safety goal QHO for latent cancer fatality risk is met.
16
RSFs RSF: A PRA Safety Function that is required to be fulfilled to maintain the consequence of one or more DBEs or the frequency of one or more high-consequence BDBEs inside the F-C Target Provides connection to Safety-Related Classification 17
RSF Example
- Modular High-Temperature Gas-Cooled Reactor (MHTGR) RSFs 18
Design Basis Accidents (DBAs)
DBAs Postulated event sequences that are used to set design criteria and performance objectives for the design of SR SSCs. DBAs are derived from DBEs based on the capabilities and reliabilities of Safety-Related SSCs needed to mitigate and prevent event sequences, respectively. DBAs are derived from the DBEs by prescriptively assuming that only SR SSCs are available to mitigate postulated event sequence consequences to within the 10 CFR 50.34 dose limits.
19
LBEs - LWR Summary ANSI/ANS-51.1-1983; nuclear safety criteria for the design of stationary pressurized water reactor plants (withdrawn 1989) 20
Functional Containment (SECY-18-0096) 21
External Events
- Incorporation of External Events in to LBEs
- PRAs introduced at early stage of design are limited in scope and level of detail commensurate with design development
- A technically adequate at-power internal events PRA may be used for the initial selection of LBEs, selection of SR SSCs and definition of DBAs.
- Design Basis External Hazard Levels (DBEHLs) are selected to design the protections against area events, e.g., internal fires and floods, and external hazards, e.g., seismic events, external flooding, high winds and missiles
- When SR SSCs are required to be protected against the DBEHLs with appropriate design margins, the DBAs derived from the internal events PRA are expected to be stable
- As external hazards and area events are incorporated into the PRA there will be new AOOs, DBEs, and BDBEs and risk insights to incorporate; but no new DBAs are expected 22
Safety Classification and Performance Criteria Safety Classification Categories
- SR
- SSCs credited in the fulfillment of RSFs and are capable to perform their RSFs in response to any DBEHL
- NSRST
- Non-safety-related SSCs that perform risk-significant functions or perform functions that are necessary for DID adequacy
- Non-Safety Related with No Special Treatment
- All SSCs within a plant that are neither Safety-Related SSCs nor Non-Safety-Related SSCs with Special Treatment SSCs 23
Safety-Significant SSCs
- An SSC that performs a function whose performance is necessary to achieve adequate DID or is classified as Risk-Significant (see Risk-Significant SSC).
Summary 24
Special Treatment Requirements
- SR SSCs
- Required Functional Design Criteria derived from RSFs; may be used with Advanced Reactor Design Criteria (ARDC) in formulating principal design criteria
- SSC level Safety Related Design Criteria developed from RSFs
- SSC reliability and capability performance targets
- Focus on prevention and mitigation functions identified in LBEs
- IDP to derive additional specific special treatment requirements 25
Assessing DID 26
Layers of Defense Adapted from International Atomic Energy Agency (IAEA) 27
Risk-Significant LBEs 28
Integrated Decision Making
- The reactor designer is responsible for ensuring that DID is achieved through the incorporation of DID features and programs in the design phases and in turn, conducting the evaluation that arrives at the decision of whether adequate DID has been achieved
- The reactor designer uses an IDP to ensure there is an input from multiple functional areas
- Later, the reactor designer or plant operator may confirm DID adequacy through the use of an Integrated Decision Making Process Panel for the reference baseline confirmation 29
Tabletop Exercise (MHTGR; Xe-100)
Report: ADAMS Accession No. ML18228A779 30
Part 50 and Part 53 Comparing Licensing Frameworks
- Safety criteria o Same safety criteria in Parts 50 and 53 o QHOs used in guidance under Part 50
- Design and Analyses o DBAs Part 50: Assessed using prescriptive, highly conservative analyses Including single failure criterion Part 53: Assessed methodically considering event frequencies and assuming only safety-related SSCs are available o BDBEs Part 50: Identified & assessed by largely ad-hoc, prescriptive approach with uncertainties addressed through conservatisms Part 53: Derived methodically using event frequencies with explicit consideration for uncertainties Including combinations of various equipment failures
- Special Treatment for Non-Safety-Related but Risk-Significant SSCs o Part 50: Ad-hoc (e.g., § 50.69 programs, Reliability Assurance Programs) o Part 53: Systematic approach to control frequencies and consequences of the LBEs in relation to safety criteria 31
Mechanistic Source Term
Background
- TID-14844, "Calculation of Distance Factors for Power and Test Reactors (1962)
- MHTGR proposes a mechanistic (scenario-specific) source term (1980s)
- SECY-93-092, Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, AND PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory Requirements and related Staff Requirements Memorandum (SRM)
- A mechanistic source term is the result of an analysis of fission product release based on the amount of cladding damage, fuel damage, and core damage resulting from the specific accident sequences being evaluated. It is developed using best-estimate phenomenological models of the transport of the fission products from the fuel through the reactor coolant system, through all holdup volumes and barriers, taking into account mitigation features, and finally, into the environs.
- NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants (1995)
- SECY-03-0047, Policy Issues Related to Licensing Non-Light-Water Reactor Designs, and related SRM
- SECY-16-0012, ACCIDENT SOURCE TERMS AND SITING FOR SMALL MODULAR REACTORS AND NON-LIGHT WATER REACTORS
- SECY-19-0117, Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors, and related SRM (See also Regulatory Guide NEI 18-04 and 1.233) 32
Mechanistic Source Term
- Regulatory Guide 1.233, Guidance for a Technology-Inclusive Methodology to Inform the Licensing Basis
- An evaluation of events, plant features and programs, and related uncertainties must address the state of knowledge related to the behavior of reactor systems, fuel, and the way in which radionuclides may move within and be released from a facility.
- The NRC will validate analytical tools and computer codes by comparing results to information available from operating experience and experiments.
- In SRM-SECY-93-092, the Commission approved the NRC staffs recommendation that source terms for non-LWRs be based upon a mechanistic analysis and that the acceptability of the applicants analysis will rely on the staffs assurance that the following conditions are met:
- The performance of the reactor and fuel under normal and off-normal conditions is sufficiently well understood to permit a mechanistic analysis. Sufficient data should exist on the reactor and fuel performance through the research, development, and testing programs to provide adequate confidence in the mechanistic approach.
- The transport of fission products can be adequately modeled for all barriers and pathways to the environs, including the specific consideration of containment design. The calculations should be as realistic as possible so that the values and limitations of any mechanism or barrier are not obscured.
- The events considered in the analyses to develop the set of source terms for each design are selected to bound severe accidents and design-dependent uncertainties.
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First Principles (Mechanistic Source Term)
See: SECY-18-0096, Functional Containment Performance Criteria for Non-Light-Water-Reactors, and INL/EXT-20-58717, Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities 34
Related Activities 35
NGNP/HTGR Mechanistic Source Term 36
Integrated Activities Population-related related Functional considerations Containment (SECY-20-0045)
EP for SMRs (SECY-18-0096) and ONTs NEI 18-04 RG 1.233 Insurance and Liability Environmental Reviews Consequence Based Security (SECY-18-0076) 37
LMP Approach Discussion 38
MEETING BREAK Meeting to resume in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 39
Part 53 General Layout
- Subpart A, General Provisions
- Subpart B, Technology-Inclusive Safety Objectives
- Subpart C, Design and Analysis
- Subpart D, Siting Requirements
- Subpart E, Construction and Manufacturing Requirements
- Subpart F, Requirements for Operation
- Subpart G, Decommissioning Requirements
- Subpart H, Applications for Licenses, Certifications and Approvals
- Subpart I, Maintaining and Revising Licensing Basis Information
- Subpart J, Reporting and Administrative Requirements 40
Subpart F - Emergency Preparedness 41
Subpart F - § 53.820 Emergency Preparedness
- Develop and implement an EP Program for operations that is commensurate with the risks posed by the LBEs as analyzed in accordance with
§ 53.450.
- Preliminary rule language will leverage existing EP regulatory framework in 10 CFR 50.47 and Appendix E and 10 CFR 50.160 Emergency Preparedness for Small Modular Reactors and Other New Technologies rulemaking
- Allows applicants the flexibility to choose a prescriptive or performance-based EP regulatory framework 42
Subpart F - § 53.820 Emergency Preparedness
- 10 CFR 50.47 and Appendix E: Provides a familiar deterministic framework similar to IAEA and other international regulatory frameworks
- 10 CFR 50.160: Performance based EP regulatory framework proportional to the risk without unwarranted regulatory burden:
- Provisions:
Scalable emergency planning zone (EPZ) size Performance-based requirements Hazard analysis Emergency plan description of offsite ingestion response planning
- Schedule:
Previous ACRS meeting - Aug & Oct 2018 Future ACRS meetings prior to proposed Final Rule to Commission
- Dec 2021 43
Subpart F - Emergency Preparedness Discussion 44
TICAP/ARCAP Guidance Document Development 45
Advanced Reactors Overview of ARCAP Roadmap Interim Staff Guidance (ISG) and TICAP DG White Papers
ARCAP To ensure review readiness to regulate a new generation of advanced reactors, a key element of a flexible regulatory framework is to provide guidance for the development of content of an advanced reactor application.
NRC Staff Stakeholders Lessons Learned
- Ensures consistency of staff
- Makes information about
- Apply lessons learned from LWR application reviews, regulatory matters widely available, reviews,
- Presents a well-defined base for
- Improves communication and
- Technology Inclusive, scope and requirements of understanding of the staff review process by interested members of
- Risk-Informed Performance-Based.
reviews.
the public and the nuclear power industry.
ARCAP
Background
Purpose Provides a roadmap for developing a tech-inclusive, risk-informed application. Leverages existing guidance or guidance that is under development.
Broad Encompasses industry-led TICAP.
Need for Additional Guidance Roadmap also identifies areas where additional guidance is needed (i.e.: Technical Specifications [TS]).
Regulatory Applicability (As applicable) 10 CFR Parts 50, 52, and informs 53.
Streamlined Review Process ARCAP guidance document not intended to replicate NUREG-0800, Standard Review Plan for LWRs.
Previous Discussions ARCAP overview discussed at August 2020, October 2020, and February 2021 public meetings.
ARCAP Roadmap ISG - Outline Identifies all Adv. Provides background Provides endorsements, Provides pointers to Rx application and overview of clarifications, key guidance in topics. expected information supplements info, or support of application for each topic. points of emphasis. topic.
TICAP
Background
Purpose
- TICAP is industry-led guidance focused on describing the scope and level of detail for portions of an application consistent with the LMP.
- Industry-led TICAP guidance only applicable to portions of first 8 safety analysis report (SAR) chapters.
- Aims to minimize burden of generating and supplying non-safety significant information.
Regulatory Applicability (As applicable) 10 CFR Parts 50, 52, and informs 53.
Methodology Scope is governed by the LMP-based safety case. LMP process is one approach to select LBEs, develop SSC categorization and ensures DID is considered.
TICAP draft DG- Outline Endorses LMP- Provides additional Provides further Includes appendices to based NEI 21- clarifications, exceptions, information needed key guidance in xx TICAP points of emphasis from outside of LMP-based support of final safety document. information described in affirmative safety case analysis report (FSAR)
NEI 21-xx. (ASC) for first 8 chapters. development for first 8 chapters.
ARCAP and TICAP - Nexus
- Additional contents of application outside of SAR are still under discussion. The above list is draft and for illustration purposes only.
NRC ARCAP/TICAP Guidance Other Insights Efficiency Adaptable NRC ARCAP/TICAP guidance being developed in parallel to 1 4 ARCAP guidance includes placeholders for guidance under development (e.g.,
industry, Applicability of Regs),
Openness Endorsement Main purpose of NRC TICAP white paper releasing draft documents is to solicit 2 5 endorses, as appropriate, industrys stakeholder feedback TICAP document, on proposal, Initial Thoughts Supplements The guidance structure, NRC TICAP white paper not detailed content, is the focus of stakeholder 3 6 supplements, as appropriate, information interactions, not addressed in industrys TICAP document (i.e.: Fuel Qual and ASME Sec III, Div 5).
ARCAP Roadmap ISG - Example 1
- FSAR structure developed as a Contents of an Advanced Reactor Applicati result of extensive stakeholder engagement.
- Consists of 12 main chapters.
- Provides the most safety-significant information at the forefront (affirmative safety case).
- Focus on the most relevant safety information while removing unnecessary details.
- Additional information/background is available for audit/inspection by NRC.
- Contents of application are still under discussion. List represents a draft outline
Note: SAR Chapters 1-8 addressed by TICAP. SAR Chapters 9-12 addressed by ARCAP.
Our Ch. 1- General Plant Information, site description, and overview of safety case (TICAP)
Information should provide an understanding of the overall facility (type of application, the number of plant units, a brief description of the proposed plant location, and the type of advanced reactor being proposed). The site description should provide an overview of the actual physical, environmental and demographic features of a site, and how they relate to the ASC.
Clarifies Endorses Key Guidance Supplements
- Roadmap clarifies
- RG 1.2xx Guidance For
- Chapter 1 of NEI
- Construction that guidance A Technology-inclusive 21-xx (TICAP) as Permit Information applicable to Content Of Application one acceptable in NEI 21-xx by chapter 1 is Methodology To Inform method. including Appendix The Licensing Basis And described in NEI Content Of Applications A for info outside 21-xx - TICAP For Licenses, LMP for first 8 document. Certifications, And chapters.*
Approvals For Advanced Reactors.
Note: Construction Permit information for all other portions of the application are described in Appendix E of the ARCAP roadmap ISG)
Our Ch. 2- Methodologies and Analyses (TICAP)
Certain analyses are common to several licensing-basis event analyses. Information should describe the process and methods used to develop baseline information related to the PRA (overview of the PRA), source-term analysis, and design-basis accidents (DBAs) analytical methods.
Clarifies Endorses Key Guidance Supplements
- Roadmap clarifies
- RG 1.2xx Guidance For
- Chapter 2 of NEI
- Site Information that guidance A Technology-inclusive 21-xx (TICAP) as draft ISG applicable to Content Of Application one acceptable previously chapter 2 is Methodology To Inform method. released.
The Licensing Basis And described in NEI Content Of Applications
- Staff positions on 21-xx - TICAP For Licenses, additional document. Certifications, And considerations to Approvals For Advanced document Reactors. information.
Our Ch. 10 - Control of Occupational Dose Information should include facility and equipment design, radiation sources, and operational programs that are necessary to ensure that the occupational radiation protection standards set forth in 10 CFR Part 20 are met.
The information should also include any commitments made by the applicant to develop the management policy and organizational structure necessary to ensure occupational radiation exposures are as low as (is) reasonably achievable (ALARA).
Clarifies Endorses Key Guidance Supplements
- Guidance is
- DANU-ISG-2021-XX,
- RG 8.8 included for Control of
- RG 8.10 chapters 9-12. Occupational Dose.
- Released on prior
- NEI 07-08A ARCAP/TICAP public
- Draft list released in meeting. prior public meeting.
Expected to evolve.
(MLxyz123).
ARCAP Roadmap ISG - Example 2
- Ongoing Emergency Preparedness Requirements Contents of an Advanced Reactor Application*
for Small Modular Reactors and Other New Technologies rulemaking.
- Rule would amend the NRCs regulations to add new emergency preparedness requirements for SMRs, non-LWR and non-power production or utilization facilities.
- Rule would adopt a scalable plume exposure pathway EPZ approach that is performance-based, consequence-oriented, and technology-inclusive.
- Contents of application are still under discussion. List represents a draft outline
Our Emergency Preparedness Plan This rulemaking would develop a dose-based, consequence-oriented framework for future SMR applicants and licensees with respect to offsite EP that would reduce the need for exemptions related to regulations associated with large LWRs.
Clarifies Endorses Key Guidance Supplements
- Ongoing
- DG-1350, rulemaking. Emergency Response Planning and Preparedness for Nuclear Power Reactors.
Key Messages TICAP draft RG, ARCAP Draft roadmap ISG, and Whats ARCAP Next?
selected chapters (e.g., site information, TS) released as white-paper to solicit stakeholder feedback. Further iterations expected.
Draft documents provided in Table 2 of ARCAP/TICAP public webpage https://www.nrc.gov/reactors/new-reactors/advanced/details.html#advRx ContentAppProj Some sections are primarily aligned with the LMP, however:
- the concepts and general information may be used to inform the review of an application submitted using other methodologies (as applicable) such as a maximum hypothetical accident, or deterministic approaches.
Timeline for TICAP Guidance and ARCAP Guidance (rev 7/13/2021)
NRC Comments based on TICAP Workshops 6/10/2021 NRC/Industry update ACRS Subcommittee on status of ARCAP/TICAP guidance documents 7/21/2021 TICAP Tabletop Exercises NEI Revision 1 of TICAP Guidance 2/1/2021 4/2/2021 Southern Revision C of TICAP Document Guidance Document 1/19/2022 8/3/2021 NRC/Industry brief ACRS Subcommittee NEI Revision 0 of TICAP Guidance on final ARCAP/ TICAP guidance Legend TICAP Workshops Document 2/9/2022 5/26/2021 8/27/2021 Industry Action 5/2/2021 NRC/Industry brief ACRS Full NRC TICAP Regulatory Guide (Draft) Committee on final TICAP 9/10/2021 guidance NRC Staff Action 3/3/2022 NRC/Industry brief ACRS Subcommittee on Industry/NRC Southern Revision B of TICAP Guidance ARCAP/TICAP guidance documents (NEI, NRC TICAP RG Joint Action Document Rev0 and Staff Draft RG) 4/15/2021 10/12/2021 3/25/2022 Jan Mar May Jul Sep Nov 2022 Mar 2022 1/30/2021 9/10/2021 Draft ARCAP Roadmap ISG, ARCAP ISG for ARCAP Application Outline Updated to be "Site Information," and ARCAP Chapters 9, Consistent with TICAP outline 10, 11, 12, and Technical Specifications issued 62
Next Steps - Future Milestones TICAP Near-Term Milestones Target Date Southern Revision C to TICAP Guidance Early August 2021 Document NEI Revision 0 of TICAP Guidance Late August 2021 Document Update of NRC Draft Guidance Documents September 2021 ACRS Future Plant Subcommittee Meeting October 2021 on ARCAP/TICAP Guidance Documents 63
Backup Slides
Technology-Inclusive Content of Application (TICAP) and Advanced Reactors Content of Application (ARCAP)- Nexus to Part 53 Part 53- Proposed Structure Subpart H - Licenses, Cert, and Approvals This subpart is envisioned to address requirements for initial applications for licenses, certifications, or approvals. The A- General Provisions subpart will support either licensing under the Part 50 or Part 52 frameworks. Assessment and update of manufacturing licenses is possible.
B- Tech-Incl Safety Requirements C- Design and Analysis Req.
D- Siting ARCAP-Guidance for Content of Application Guidance (Roadmap)
E- Const. and Manufacturing TICAP - LMP-based Guidance under F- Operations portions of FSAR that are Development examples:
related to:
- Fuel Qualification
- G- Decommissioning
- Safety classification ASME Section III, Division 5
- EP rulemaking H- Licenses, Cert, and Approvals
- Security ARCAP specific chapter Rulemaking guidance - examples:
I- Maintaining/Revising LB Info
- Site information Existing
- ARCAP chapters 9, Regulatory 10, 11, and 12 Guidance J- Administrative requirements Note: The illustrated content structure for Part 53 (including Subpart H) is part of ongoing work and subject to change. 65
Industry-led TICAP NRC-led TICAP
- Focused on portions of the
- NRC plans to issue a RG license application SAR for non- endorsing TICAP that also Chapters 1-8 LWR designs related to the focuses on providing exceptions Licensing Modernization Project and/or clarifications on TICAP.
TICAP (LMP)-based ASC.
- Include supplemental TICAP guidance for areas outside of the LMP for the first 8 SAR chapters. Examples includes FSAR site information, ASME Section III, Division 5 Chapters 9-12 NRC-led SAR Guidance ARCAP*
- Focused on remaining portions of the license application SAR not related to LMP.
- ARCAP ISGs under development that include an overall roadmap ISG and separate ISGs for FSAR Chapters 2, 9, 10, 11, and 12 Additional
- For example:
Contents of o TS, o QA Plan, o Fire Protection, etc.
Application
- Staff plans to issue an ARCAP Roadmap ISG that would provide pointers to various guidance documents developed/issued.
66
Key Part 53 Guidance by Subpart Subpart A: General Provisions Existing / Ongoing Guidance Additional Guidance N/A Subpart B: Safety Criteria Existing / Ongoing Guidance Additional Guidance Further explanation of criteria and N/A structure in the Statements of Consideration Subpart C: Design and Analysis Existing / Ongoing Guidance Additional Guidance NEI 18-04 & RG 1.233 (LMP)
ANS/ASME-RA-S-1.4 (Non-LWR PRA ISG on PRA for Initial Licensing Standard)
RG 1.247 Endorsing Non-LWR PRA Industry PRA Peer Review Guidance for Standard and NEI Peer Review Guidance Non-LWRs (NEI 20-09)
Application of Analytical Margins ANS/ASME Standards Treatment of Chemical Hazards Fuel Qualification RG 1.232 (ARDCs)
Subpart D: Siting Requirements Existing / Ongoing Guidance Additional Guidance SECY-20-0045/RG 4.7 External Hazard Updates N/A 67 Risk-Informed Seismic Design; ANS 2.26
Key Part 53 Guidance by Subpart Subpart E: Construction and Manufacturing Existing / Ongoing Guidance Additional Guidance Manufacturing Guidance N/A QA Alternatives Subpart F: Operations SSCs Existing / Ongoing Guidance Additional Guidance TS Special Treatment NEI 18-04 & RG 1.233 (LMP)
Maintenance, Repair & Inspection Facility Safety Program Personnel Existing / Ongoing Guidance Additional Guidance DRO Paper/preliminary ISG Concept of Operations Programs Existing / Ongoing Guidance Additional Guidance Emergency Preparedness EPZ Draft Final Rule, RG 1.242 Security Programs Radiation Protection (ARCAP)
Integrity Assessment Program 68
Key Part 53 Guidance by Subpart Subpart G: Decommissioning Existing / Ongoing Guidance Additional Guidance N/A N/A Subpart H: Licensing Existing / Ongoing Guidance Additional Guidance TICAP Manufacturing Licenses ARCAP Subpart I: Maintaining Licensing Basis Existing / Ongoing Guidance Additional Guidance 50.59 Equivalent N/A FSAR/PRA Updates Subpart J: Administrative/Misc.
Existing / Ongoing Guidance Additional Guidance Reporting Requirements N/A Financial/Liability 69
TICAP/ARCAP Guidance Document Development Discussion 70
Final Discussion and Questions 71
Part 53 Rulemaking Schedule Milestone Schedule Major Rulemaking Activities/Milestones Schedule Public Outreach, ACRS Interactions and Present to April 2022 Generation of Proposed Rule Package (9 months)
Submit Draft Proposed Rule Package to May 2022 Commission Publish Proposed Rule and Draft Key Guidance October 2022 Public Comment Period - 60 days November and December 2022 Public Outreach and Generation of Final Rule January 2023 to February 2024 Package (14 months)
Submit Draft Final Rule Package to Commission March 2024 Office of Management and Budget and Office of July 2024 to September 2024 the Federal Register Processing Publish Final Rule and Key Guidance October 2024 72
Acronyms and Abbreviations Advisory Committee on Reactor DBA Design basis accident ACRS Safeguards DBE Design basis event Agencywide Documents Access ADAMS DBEHL Design basis external hazard levels and Management System DID Defense-in-depth ANS American Nuclear Society DG Draft guidance Anticipated operational AOO DRO Division of Reactor Oversight occurrence Advanced reactor content of EAB Exclusion area boundary ARCAP application project EP Emergency preparedness ARDC Advanced reactor design criteria U.S. Environmental Protection EPA ASC Affirmative safety case Agency American Society of Mechanical EPZ Emergency planning zone ASME Engineers BDBE Beyond design basis event F-C Frequency-consequence Canadian deuterium-uranium-3 FSAR Final safety analysis report CANDU3 reactor (remove)
Steam generator feedwater pump CFR Code of Federal Regulations FW trip CR Control rod withdrawal IAEA International Atomic Energy Agency CT Circulator trip IDP Integrated decision-making process 73
Acronyms and Abbreviations Modular High-Temperature Gas-ISG Interim staff guidance MHTGR Cooled Reactor ISI Inservice inspection NEI Nuclear Energy Institute IST Inservice testing NGNP Next Generation Nuclear Plant Inspections, tests, analyses and U.S. Nuclear Regulatory ITAAC NRC acceptance criteria Commission Non-Safety Related with Special LB Licensing basis NSRST Treatment LBE Licensing basis event Non-Safety Related with no NST Special Treatment LD Large helium depressurization U.S. Nuclear Regulatory NUREG Commission technical report LF Loss of primary flow designation LMP Licensing Modernization Project ONT Other new technologies Process Inherent Ultimate LO Loss of offsite power PIUS Safety LWR Light-water-reactor PRA Probabilistic risk assessment MC&A Material control and accounting Power reactor innovative small PRISM module Medium helium MD QA Quality assurance depressurization 74
Acronyms and Abbreviations QHO Quantitative health objective Technology-inclusive content of TICAP application project Rem Roentgen-equivalent man TS Technical specifications RG Regulatory guidance RSF Required safety function TT Turbine trip RT Reactor trip SAR Safety analysis report SD Small helium depressurization SG Steam generator tube rupture SMR Small modular reactor SR Safety related Staff requirements SRM memorandum Structure, system, and SSC component TEDE Total effective dose equivalent 75