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
July 21, 2021 1
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
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
Plant Documents (Systems, Procedures, etc.)
Analyses (Prevention, Mitigation, Compare to Criteria)
Plant/Site (Design, Construction, Configuration Control)
Retirement Staffing &
Human Factors Configuration Control Surveillance Maintenance Operation Construction/
Manufacturing Construction Siting Design and Analysis LB Documents (Applications, SAR, TS, etc.)
NRC Staff Plan to Develop Part 53 Project Life Cycle System
& Component Design Analysis Requirements Subpart B Subpart C Subpart D Subpart E Subpart G Subparts H & I Safety Categorization
& Special Treatment External Hazards Site Characteristics Environmental Considerations Ensuring Capabilities/
Reliabilities Change Control Environmental Considerations Programs Security, EP Facility Safety Program 3
Requirements Definition Safety Objectives Safety Criteria Safety Functions Other Subpart A General Provisions Subpart J Admin & Reporting Clarify Controls and Distinctions Between Other 10 CFR Parts Subpart F
NRC Staff Engagement Plan 4
Concept/Introduction Discussion Interim Staff Resolution ACRS Interactions
5 LMP Approach
Licensing Modernization Risk-Informed, Technology-Inclusive Framework 6
An Evolution
7 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)
8 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
9 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)
Event Selection & Analysis
10 Event Selection (F-C Curve)
LBEs are defined by event sequence families from design specific PRA 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
11 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
12 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.
13 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.
Event Selection & Analysis
14 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.
Event Selection & Analysis
15 PRA 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
16 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.
17 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 RSFs Provides connection to Safety-Related Classification
18 Modular High-Temperature Gas-Cooled Reactor (MHTGR) RSFs RSF Example
19 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.
Design Basis Accidents (DBAs)
20 LBEs - LWR Summary ANSI/ANS-51.1-1983; nuclear safety criteria for the design of stationary pressurized water reactor plants (withdrawn 1989)
21 Functional Containment (SECY-18-0096)
22 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 External Events
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 Classification and Performance Criteria
24 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
25 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
26 Assessing DID
27 Layers of Defense Adapted from International Atomic Energy Agency (IAEA)
28 Risk-Significant LBEs
29 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
30 Tabletop Exercise (MHTGR; Xe-100)
Report: ADAMS Accession No. ML18228A779
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
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
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
32 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)
33 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.
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
35 Related Activities
36 NGNP/HTGR Mechanistic Source Term
Consequence Based Security (SECY-18-0076)
EP for SMRs and ONTs Functional Containment (SECY-18-0096)
Insurance and Liability related Population-related considerations (SECY-20-0045)
Environmental Reviews NEI 18-04 RG 1.233 Integrated Activities 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 Emergency Preparedness 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
41 Subpart F - Emergency Preparedness
Subpart F - § 53.820 Emergency Preparedness 42
- 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
Subpart F - § 53.820 Emergency Preparedness 43
- 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
Subpart F - Emergency Preparedness Discussion 44
45 TICAP/ARCAP Guidance Document Development
Advanced Reactors Overview of ARCAP Roadmap Interim Staff Guidance (ISG) and TICAP DG White Papers
- Ensures consistency of staff
- reviews,
- Presents a well-defined base for scope and requirements of reviews.
- Makes information about regulatory matters widely available,
- Improves communication and understanding of the staff review process by interested members of the public and the nuclear power industry.
- Apply lessons learned from LWR application
- reviews,
- Technology Inclusive,
- Risk-Informed Performance-Based.
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 ARCAP
Purpose Provides a roadmap for developing a tech-inclusive, risk-informed application. Leverages existing guidance or guidance that is under development.
Need for Additional Guidance Roadmap also identifies areas where additional guidance is needed (i.e.: Technical Specifications [TS]).
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
Background
Regulatory Applicability (As applicable) 10 CFR Parts 50, 52, and informs 53.
Broad Encompasses industry-led TICAP.
Identifies all Adv.
Rx application topics.
Provides background and overview of expected information for each topic.
Provides endorsements, clarifications, supplements info, or points of emphasis.
Provides pointers to key guidance in support of application topic.
ARCAP Roadmap ISG - Outline
Purpose TICAP is industry-led guidance focused on describing the scope and level of detail for portions of an application consistent with the LMP.
LMP is described in NEI18-04, as endorsed by RG) 1.233.
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.
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
Background
Regulatory Applicability (As applicable) 10 CFR Parts 50, 52, and informs 53.
Endorses LMP-based NEI 21-xx TICAP document.
Provides additional clarifications, exceptions, points of emphasis from information described in NEI 21-xx.
Provides further information needed outside of LMP-based affirmative safety case (ASC) for first 8 chapters.
Includes appendices to key guidance in support of final safety analysis report (FSAR) development for first 8 chapters.
TICAP draft DG-Outline
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.
3 2
1 Initial Thoughts The guidance structure, not detailed content, is the focus of stakeholder interactions, Openness Main purpose of releasing draft documents is to solicit stakeholder feedback on proposal, Efficiency NRC ARCAP/TICAP guidance being developed in parallel to
- industry, 6
5 4
Supplements NRC TICAP white paper supplements, as appropriate, information not addressed in industrys TICAP document (i.e.: Fuel Qual and ASME Sec III, Div 5).
Endorsement NRC TICAP white paper endorses, as appropriate, industrys TICAP document, Adaptable ARCAP guidance includes placeholders for guidance under development (e.g.,
Applicability of Regs),
NRC ARCAP/TICAP Guidance Other Insights
ARCAP Roadmap ISG - Example 1
- Contents of application are still under discussion. List represents a draft outline Contents of an Advanced Reactor Applicati FSAR structure developed as a 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.
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.
Chapter 1 of NEI 21-xx (TICAP) as one acceptable method.
Supplements Key Guidance Endorses Clarifies Construction Permit Information in NEI 21-xx by including Appendix A for info outside LMP for first 8 chapters.*
Roadmap clarifies that guidance applicable to chapter 1 is described in NEI 21-xx - TICAP document.
RG 1.2xx Guidance For A Technology-inclusive Content Of Application Methodology To Inform The Licensing Basis And Content Of Applications For Licenses, Certifications, And 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.
Chapter 2 of NEI 21-xx (TICAP) as one acceptable method.
Supplements Key Guidance Endorses Clarifies Site Information draft ISG previously released.
Staff positions on additional considerations to document information.
RG 1.2xx Guidance For A Technology-inclusive Content Of Application Methodology To Inform The Licensing Basis And Content Of Applications For Licenses, Certifications, And Approvals For Advanced Reactors.
Roadmap clarifies that guidance applicable to chapter 2 is described in NEI 21-xx - TICAP document.
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).
RG 8.8 RG 8.10 ANSI/ANS 18.1-1999 NEI 07-08A Draft list released in prior public meeting.
Expected to evolve.
(MLxyz123).
Supplements Key Guidance Endorses Clarifies Guidance is included for chapters 9-12.
DANU-ISG-2021-XX, Control of Occupational Dose.
Released on prior ARCAP/TICAP public meeting.
ARCAP Roadmap ISG - Example 2
- Contents of application are still under discussion. List represents a draft outline Contents of an Advanced Reactor Application*
Ongoing Emergency Preparedness Requirements 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.
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.
DG-1350, Emergency Response Planning and Preparedness for Nuclear Power Reactors.
SECY-18-0103 Supplements Key Guidance Endorses Clarifies Ongoing rulemaking.
TICAP draft RG, ARCAP Draft roadmap ISG, and ARCAP selected chapters (e.g., site information, TS) released as white-paper to solicit stakeholder feedback. Further iterations expected.
Key Messages Whats Next?
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.
Draft documents provided in Table 2 of ARCAP/TICAP public webpage https://www.nrc.gov/reactors/new-reactors/advanced/details.html#advRx ContentAppProj
Timeline for TICAP Guidance and ARCAP Guidance (rev 7/13/2021)
Legend Industry Action NRC Staff Action Industry/NRC Joint Action 2022 Jan Mar May Jul Sep Nov 2022 Mar Southern Revision B of TICAP Guidance Document 4/15/2021 Southern Revision C of TICAP Guidance Document 8/3/2021 NEI Revision 0 of TICAP Guidance Document 8/27/2021 NEI Revision 1 of TICAP Guidance Document 1/19/2022 NRC Comments based on TICAP Workshops 6/10/2021 NRC TICAP Regulatory Guide (Draft) 9/10/2021 NRC TICAP RG 3/25/2022 NRC/Industry update ACRS Subcommittee on status of ARCAP/TICAP guidance documents 7/21/2021 NRC/Industry brief ACRS Subcommittee on ARCAP/TICAP guidance documents (NEI, Rev0 and Staff Draft RG) 10/12/2021 NRC/Industry brief ACRS Subcommittee on final ARCAP/ TICAP guidance 2/9/2022 NRC/Industry brief ACRS Full Committee on final TICAP guidance 3/3/2022 ARCAP Application Outline Updated to be Consistent with TICAP outline 1/30/2021 Draft ARCAP Roadmap ISG, ARCAP ISG for "Site Information," and ARCAP Chapters 9, 10, 11, 12, and Technical Specifications issued 9/10/2021 2/1/2021 TICAP Tabletop Exercises 4/2/2021 5/2/2021 TICAP Workshops 5/26/2021 62
Next Steps - Future Milestones TICAP Near-Term Milestones Target Date Southern Revision C to TICAP Guidance Document Early August 2021 NEI Revision 0 of TICAP Guidance Document Late August 2021 Update of NRC Draft Guidance Documents September 2021 ACRS Future Plant Subcommittee Meeting on ARCAP/TICAP Guidance Documents October 2021 63
Backup Slides
A-General Provisions B-Tech-Incl Safety Requirements C-Design and Analysis Req.
D-Siting E-Const. and Manufacturing F-Operations G-Decommissioning H-Licenses, Cert, and Approvals I-Maintaining/Revising LB Info J-Administrative requirements Part 53-Proposed Structure Note: The illustrated content structure for Part 53 (including Subpart H) is part of ongoing work and subject to change.
Technology-Inclusive Content of Application (TICAP) and Advanced Reactors Content of Application (ARCAP)- Nexus to Part 53 Subpart H - Licenses, Cert, and Approvals This subpart is envisioned to address requirements for initial applications for licenses, certifications, or approvals. The subpart will support either licensing under the Part 50 or Part 52 frameworks. Assessment and update of manufacturing licenses is possible.
ARCAP-Guidance for Content of Application Guidance (Roadmap)
TICAP - LMP-based portions of FSAR that are related to:
LBEs Safety classification DID ARCAP specific chapter guidance - examples:
Site information ARCAP chapters 9, 10, 11, and 12 Guidance under Development examples:
Fuel Qualification ASME Section III, Division 5 EP rulemaking Security Rulemaking Existing Regulatory Guidance 65
Industry-led TICAP
- Focused on portions of the license application SAR for non-LWR designs related to the Licensing Modernization Project (LMP)-based ASC.
NRC-led TICAP
- NRC plans to issue a RG endorsing TICAP that also focuses on providing exceptions and/or clarifications on TICAP.
- Include supplemental TICAP guidance for areas outside of the LMP for the first 8 SAR chapters. Examples includes site information, ASME Section III, Division 5 NRC-led SAR Guidance
- 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 FSAR Chapters 1-8 Chapters 9-12
- For example:
o TS, o QA Plan, o Fire Protection, etc.
Additional Contents of Application TICAP ARCAP*
- 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 67 Subpart A: General Provisions Existing / Ongoing Guidance Additional Guidance N/A Subpart B: Safety Criteria Existing / Ongoing Guidance Additional Guidance N/A Further explanation of criteria and 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 Standard)
Industry PRA Peer Review Guidance for Non-LWRs (NEI 20-09)
ANS/ASME Standards Fuel Qualification RG 1.232 (ARDCs)
ISG on PRA for Initial Licensing RG 1.247 Endorsing Non-LWR PRA Standard and NEI Peer Review Guidance Application of Analytical Margins Treatment of Chemical Hazards Subpart D: Siting Requirements Existing / Ongoing Guidance Additional Guidance SECY-20-0045/RG 4.7 External Hazard Updates Risk-Informed Seismic Design; ANS 2.26 N/A
Key Part 53 Guidance by Subpart 68 Subpart E: Construction and Manufacturing Existing / Ongoing Guidance Additional Guidance N/A Manufacturing Guidance QA Alternatives Subpart F: Operations SSCs Existing / Ongoing Guidance Additional Guidance NEI 18-04 & RG 1.233 (LMP)
TS Special Treatment 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 EPZ Draft Final Rule, RG 1.242 Radiation Protection (ARCAP)
Emergency Preparedness Security Programs Integrity Assessment Program
Key Part 53 Guidance by Subpart 69 Subpart G: Decommissioning Existing / Ongoing Guidance Additional Guidance N/A N/A Subpart H: Licensing Existing / Ongoing Guidance Additional Guidance TICAP ARCAP Manufacturing Licenses Subpart I: Maintaining Licensing Basis Existing / Ongoing Guidance Additional Guidance N/A 50.59 Equivalent FSAR/PRA Updates Subpart J: Administrative/Misc.
Existing / Ongoing Guidance Additional Guidance N/A Reporting Requirements Financial/Liability
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 Generation of Proposed Rule Package Present to April 2022 (9 months)
Submit Draft Proposed Rule Package to Commission May 2022 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 Package January 2023 to February 2024 (14 months)
Submit Draft Final Rule Package to Commission March 2024 Office of Management and Budget and Office of the Federal Register Processing July 2024 to September 2024 Publish Final Rule and Key Guidance October 2024 72
Acronyms and Abbreviations 73 ACRS Advisory Committee on Reactor Safeguards ADAMS Agencywide Documents Access and Management System ANS American Nuclear Society AOO Anticipated operational occurrence ARCAP Advanced reactor content of application project ARDC Advanced reactor design criteria ASC Affirmative safety case ASME American Society of Mechanical Engineers BDBE Beyond design basis event CANDU3 Canadian deuterium-uranium-3 reactor (remove)
CFR Code of Federal Regulations CR Control rod withdrawal CT Circulator trip DBA Design basis accident DBE Design basis event DBEHL Design basis external hazard levels DID Defense-in-depth DG Draft guidance DRO Division of Reactor Oversight EAB Exclusion area boundary EP Emergency preparedness EPA U.S. Environmental Protection Agency EPZ Emergency planning zone F-C Frequency-consequence FSAR Final safety analysis report FW Steam generator feedwater pump trip IAEA International Atomic Energy Agency IDP Integrated decision-making process
Acronyms and Abbreviations 74 ISG Interim staff guidance ISI Inservice inspection IST Inservice testing ITAAC Inspections, tests, analyses and acceptance criteria LB Licensing basis LBE Licensing basis event LD Large helium depressurization LF Loss of primary flow LMP Licensing Modernization Project LO Loss of offsite power LWR Light-water-reactor MC&A Material control and accounting MD Medium helium depressurization MHTGR Modular High-Temperature Gas-Cooled Reactor NEI Nuclear Energy Institute NGNP Next Generation Nuclear Plant NRC U.S. Nuclear Regulatory Commission NSRST Non-Safety Related with Special Treatment NST Non-Safety Related with no Special Treatment NUREG U.S. Nuclear Regulatory Commission technical report designation ONT Other new technologies PIUS Process Inherent Ultimate Safety PRA Probabilistic risk assessment PRISM Power reactor innovative small module QA Quality assurance
Acronyms and Abbreviations 75 QHO Quantitative health objective Rem Roentgen-equivalent man RG Regulatory guidance RSF Required safety function RT Reactor trip SAR Safety analysis report SD Small helium depressurization SG Steam generator tube rupture SMR Small modular reactor SR Safety related SRM Staff requirements memorandum SSC Structure, system, and component TEDE Total effective dose equivalent TICAP Technology-inclusive content of application project TS Technical specifications TT Turbine trip