ML22125A001
| ML22125A001 | |
| Person / Time | |
|---|---|
| Issue date: | 05/19/2022 |
| From: | Robert Beall NRC/NMSS/DREFS/RRPB |
| To: | |
| Beall, Robert | |
| References | |
| 10 CFR Part 53, NRC-2019-0062, RIN 3150-AK31 | |
| Download: ML22125A001 (55) | |
Text
Advisory Committee on Reactor Safeguards (ACRS)
Regulatory Rulemaking, Policies and Practices:
Part 53 Subcommittee 10 CFR Part 53 Licensing and Regulaon of Advanced Nuclear Reactors May 19, 2022
Agenda 2
Topic Staff Introduction
- Introduction to Part 53
- Framework A Subparts o
A - General Provisions o
B - Technology-Inclusive Safety Requirements o
C - Design and Analysis Requirements o
D - Siting Requirements o
E - Construction and Manufacturing Requirements o
G - Decommissioning Requirements o
H - Licenses, Certifications, and Approvals o
I - Maintaining and Revising Licensing Basis Information o
J - Reporting and Other Administrative Requirements o
K - Quality Assurance Criteria Key Guidance
Welcome and Introductions 3
Welcome:
- Steve Lynch, Office of Nuclear Reactor Regulation (NRR)
NRC Speakers / Presenters:
- Bill Reckley, NRR
- Nan Valliere, NRR Meeting Slides:
- ADAMS Accession No. ML22125A001
Rulemaking Schedule 4
Oct/Nov 2022 ACRS Interactions on Rulemaking Package for Proposed Rule
Part 53 Licensing Frameworks 5
- Subpart B - Safety Requirements
- Subpart C - Design Requirements
- Subpart D - Siting
- Subpart E - Construction
- Subpart F - Operations
- Subpart G - Decommissioning
- Subpart H - Licensing Processes
- Subpart I - License Maintenance
- Subpart J - Reporting
- Subpart K - Quality Assurance
- Subpart N - Purpose/Definitions
- Subpart O - Construction
- Subpart P - Operations
- Subpart Q - Decommissioning
- Subpart R - Licensing Processes
- Subpart S - License Maintenance
- Subpart T - Reporting
- Subpart U - Quality Assurance Alternate Evaluation for Risk Insights Framework B Framework A Subpart A - General Provisions
Part 53 Licensing Frameworks 6
With addition of DBA used to set design criteria and performance objectives for the design of Safety Related SSCs.
Framework B Emphasis Design Criteria Framework A Emphasis Risk metrics Traditional approach represented by figure from IAEA guidance
Consolidated Preliminary Rule Language (Including Second Iteration)
Summary & Changes 7
Framework A 8
Subpart A General Provisions (Definitions)
Subpart B Safety Requirements (Including QHOs, ALARA)
Subpart C Design and Analysis Subpart D Siting Subpart E Construction and Manufacturing Subpart F Operation (Including Engineering Expertise, Operator Licensing)
Subpart G Decommissioning Subpart H Licenses, Certifications and Approvals Subpart I Maintaining Licensing Basis Subpart J Reporting and Administrative Subpart K Quality Assurance
Evolution of Part 53 & Stakeholder Feedback Topic Addressed in Preliminary Proposed Rule Language Duplicative/overlapping programs
- Quality Assurance (QA) requirements consolidated in Subpart K.
- Added flexibility for licensees to organize and combine programs, as appropriate, to avoid duplication (Subparts F & K).
Manufacturing license (ML) expansion Expanded activities permitted under ML to include fabrication of entire reactor, including fuel loading (Subparts E & H).
Safety criteria structure Eliminated two-tiered approach to safety criteria (Subpart B).
Codes and standards Enabled flexibility in using codes and standards.
Normal operations Decoupled requirements for normal operation from those for licensing basis events (LBEs) (Subparts B & C).
Use of advanced nuclear plant and expansion beyond commercial reactors
- The staff has removed references to advanced nuclear plant.
- No plans to expand applicability to research and test reactors (note that NEIMA is directed at commercial reactors) (Subpart A).
Subpart A - General Provisions
- Selected definitions o Commercial nuclear plant o Commercial nuclear reactor o Manufactured reactor o Manufactured reactor module
- Methodology definitions o Event categories o Defense in depth 10
Subpart B - Technology-Inclusive Safety Requirements
§ 53.200 Safety objectives.
§ 53.210 Safety criteria for design basis accidents.
§ 53.220 Safety criteria for licensing basis events other than design basis accidents.
§ 53.230 Safety functions.
§ 53.240 Licensing basis events.
§ 53.250 Defense in depth.
§ 53.260 Normal operations.
§ 53.270 Protection of plant workers.
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Safety Objectives DBA Safety Criteria Non-DBA Safety Criteria Safety Functions Licensing Basis Events Defense in Depth Role of:
o Structures, systems, and components (SSCs) o Personnel o
Programs 12 Subpart B - Safety Criteria Safety Criteria Safety Functions Design Features (and Human Actions)
Functional Design Criteria (Personnel; Concept of Operations)
What function(s)
(e.g., a barrier, cooling) are needed to satisfy safety criteria What design features (e.g., a structure, system) are provided to fulfill the safety function(s)
What design criteria (e.g., leak rate, cooling capacity) are needed for design feature
Subpart B - Technology-Inclusive Safety Requirements
- Revised criterion related to NRC quantitative health objectives (QHOs) to address feedback.
- Revised criteria related to as low as reasonably achievable (ALARA) to address feedback.
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QHOs - Updated Preliminary Proposed Rule Language (May 2022) 14
§ 53.220 Safety criteria for licensing basis events other than design basis accidents.
Design features and programmatic controls must be provided to:
(a) Ensure plant structures, systems and components (SSCs), personnel, and programs provide the necessary capabilities and maintain the necessary reliability to address licensing basis events in accordance with § 53.240 and provide measures for defense-in-depth in accordance with § 53.250; and (b) Maintain overall cumulative plant risk from licensing basis events other than design basis accidents analyzed in accordance with § 53.450(e) such that the calculated risk to an average individual in the vicinity of the commercial nuclear plant of prompt fatalities remains below five in 10 million years, and the calculated risk the population in the area near a commercial nuclear plant of cancer fatalities remains below two in one million years.
QHOs - Basis Performance-based approaches use measurable or calculable performance metrics.
Risk-informed approach benefits from cumulative risk measure as well as success criteria for specific event sequences QHOs are well established and have been used in making regulatory decisions since they were developed as part of the NRCs Safety Goal Policy Statement.
Examples include:
o Regulatory Guide (RG) 1.174 (Using probabilistic risk assessments (PRA) in risk-informed decisions - licensing basis) o NUREG/BR-0058, Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission.
Supports risk-informed, performance-based approach as encouraged by NEIMA.
Provides predictability and stability in that acceptance criteria are defined and are used by both applicants and NRC during initial licensing reviews and maintenance of licensing basis information (Subpart I, Maintaining and Revising Licensing Basis Information).
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QHOs - Basis Methodologies available for performing risk assessments and comparing to QHOs.
o Supported by recently issued RG 1.247, TRIAL - Acceptability of Probabilistic Risk Assessment Results for Non-Light Water Reactor Risk-Informed Activities Applicants may choose to use surrogate measures to show that designs or plants satisfy the QHO-related criteria (e.g., core damage frequency for light-water reactors (LWRs))
Recent language change to calculated risk and to refer to analyzed in accordance with § 53.450(e) intended to address issues about uncertainties associated with estimating risks to the public from the release of radionuclides.
Revised to fatalities to maintain alignment with Safety Goal Policy Statement Rationale for using QHOs as a metric will be provided in Statement of Considerations for the proposed rule package.
16
ALARA - Updated Preliminary Proposed Rule Language
§ 53.260 Normal operations.
(a) Maximum public dose. Licensees under this part must ensure that normal plant operations do not result in public doses or dose rates in unrestricted areas that exceed the limits provided in Subpart D to 10 CFR part 20.
(b) As low as reasonably achievable. A combination of design features and programmatic controls must be established such that the estimated total effective dose equivalent to individual members of the public from effluents resulting from normal plant operation are as low as is reasonably achievable in accordance with 10 CFR part 20.
(similar text for occupational exposures) 17
ALARA - Basis Consistent with current requirements in § 50.34a, Design objectives for equipment to control releases of radioactive material in effluents nuclear power reactors. Additional ALARA requirements tied to the initial design of a facility include Appendix I to Part 50; 10 CFR 20.1101, and 40 CFR Part 190 (EPA).
Consistent with previous design certification (DC) applications 18 10 CFR 50.34a, Design objectives for equipment to control releases of radioactive material in effluentsnuclear power reactors.
(e) Each application for a design approval, a design certification, or a manufacturing license under part 52 of this chapter shall include:
(1)A description of the equipment for the control of gaseous and liquid effluents and for the maintenance and use of equipment installed in radioactive waste systems, under paragraph (a) of this section; and (a) a description of the preliminary design of equipment to be installed to maintain control over radioactive materials in gaseous and liquid effluents the application shall also identify the design objectives, and the means to be employed, for keeping levels of radioactive material in effluents to unrestricted areas as low as is reasonably achievable.
ALARA - Basis
- Recognizes that plant design plays essential role in controlling releases and protecting plant workers
- Consistent with past Commission decisions (Part 20 rulemaking, Advanced Reactor Policy Statement)
- Many cost-effective solutions are most effectively identified and addressed at the design stage of a project
- Staff is proposing more performance-based approach to preparing applications and NRC review of ALARA during design reviews through issuing draft guidance (Advanced reactor content of application project [ARCAP])
- Rationale for maintaining ALARA requirementsfor both licensees and designers will be provided in Statement of Considerations for the proposed rule package 19
Subpart C - Design and Analysis Requirements 20
§ 53.400 Design features for licensing basis events.
§ 53.410 Functional design criteria for design basis accidents.
§ 53.420 Functional design criteria for licensing basis events other than design basis accidents.
§ 53.425 Design features and functional design criteria for normal operations.
§ 53.430 Design features and functional design criteria for protection of plant workers.
§ 53.440 Design requirements.
§ 53.450 Analysis requirements.
§ 53.460 Safety categorization and special treatment.
§ 53.470 Maintaining analytical safety margins used to justify operational flexibilities.
§ 53.480 Earthquake engineering.
Subpart C - Design and Analysis Requirements
- Clarified references to use of consensus codes and standards and requirement that they must be found acceptable by NRC.
- Added design requirements for:
o Aircraft impact o Chemical hazards related to licensed materials o Minimizing contamination 21
Subpart C - Design and Analysis Requirements
- Changes and Clarifications for Analysis Sections o Added need to define evaluation criteria for each event or specific categories of LBEs
Required for safety-related SSCs As appropriate for non-safety-related but safety significant SSCs 22
§ 53.450(e) - LBEs other than DBA 23 (e) Analyses of licensing basis events other than design basis accidents.
Analyses must be performed for licensing basis events other than design basis accidents. These licensing basis events must be identified using insights from a PRA in combination with other generally accepted approaches for systematically evaluating engineered systems to identify and analyze equipment failures and human errors.
The analysis of licensing basis events other than design basis accidents must include definition of evaluation criteria for each event or specific categories of licensing basis events to determine the acceptability of the plant response to the challenges posed by internal and external hazards. The analyses must address event sequences from initiation to a defined end state and be used in combination with other engineering analyses to demonstrate that the functional design criteria required by § 53.420 provide sufficient barriers to the unplanned release of radionuclides to satisfy the evaluation criteria defined for each licensing basis event, to satisfy the safety criteria of § 53.220, and provide defense in depth as required by § 53.250. The methodology used to identify, categorize, and analyze licensing basis events must include a means to identify event sequences deemed significant for controlling the risks posed to public health and safety.
24 Event Selection & Analysis Licensing Modernization Project 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.
25 Licensing Modernization Project DBEs Part 53: Unlikely Event Sequences 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
26 Licensing Modernization Project BDBEs Part 53: Very Unlikely Event Sequences 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
(f) Analysis of design basis accidents.
The analysis of licensing basis events required by § 53.240 must include analysis of design basis accidents that address possible challenges to the safety functions identified in accordance with § 53.230. The events selected as design basis accidents must be those that, if not terminated, have the potential for exceeding the safety criteria in § 53.210. The design basis accidents selected must be analyzed using deterministic methods that address event sequences from initiation to a safe stable end state and assume only the safety-related SSCs identified in accordance with § 53.460 and human actions addressed by the requirements of Subpart F are available to perform the safety functions identified in accordance with § 53.230. The analysis must conservatively demonstrate compliance with the safety criteria in § 53.210.
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§ 53.450(f) - Design basis accidents
28 Licensing Modernization Project DBAs Postulated event sequences that are used to set design criteria and performance objectives for the design of Safety Related 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 Safety Related SSCs are available to mitigate postulated event sequence consequences to within the 10 CFR 50.34 dose limits.
(Part 53: Safety Criteria in Subpart B)
Design Basis Accidents
Subpart C - Design and Analysis Requirements
- Added § 53.480 Earthquake engineering.
o Intended to support the more flexible and graded seismic design approaches afforded by performance-based standards such as the American Society of Civil Engineers (ASCE)/Structural Engineering Institute (SEI) 4-19, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities o Future interactions with ACRS expected on this topic through development of guidance documents 29
Subpart D - Siting Requirements
§ 53.500 General siting.
§ 53.510 External hazards.
§ 53.520 Site characteristics.
§ 53.530 Population-related considerations
§ 53.540 Siting interfaces.
- Added QA requirement for siting activities.
- Made changes related to the earthquake engineering section in Subpart C 30
Subpart E - Construction and Manufacturing Requirements
§ 53.600 Scope and purpose.
§ 53.605 Reporting of defects and noncompliance.
§ 53.610 Construction
§ 53.620 Manufacturing
- Made changes reflecting consolidation of QA requirements in Subpart K
- Added § 53.605 to capture requirements in § 50.55(e).
- Clarified requirements for MLs allowing fuel loading.
o References to 10 CFR Part 70 31
Subpart F - Requirements for Operation To be discussed during June meeting 32
Subpart G - Decommissioning Requirements
§ 53.1000 Scope and purpose.
§ 53.1010 Financial assurance for decommissioning.
§ 53.1020 Cost estimates for required decommissioning funds.
§ 53.1030 Annual adjustments to cost estimates for decommissioning.
§ 53.1040 Methods for providing financial assurance for decommissioning.
§ 53.1045 Requirements for decommissioning trust funds.
§ 53.1050 NRC oversight.
§ 53.1060 Reporting and recordkeeping requirements.
§ 53.1070 Termination of license.
§ 53.1080 Release of part of a commercial nuclear plant for unrestricted use.
33
Subpart H - Licenses, Certifications and Approvals 34
§ 53.1100 - 53.1121 General/common requirements.
§ 53.1124 Relationship between sections.
§ 53.1130 Limited work authorizations.
§ 53.1140 Early site permits.
§ 53.1200 Standard design approvals.
§ 53.1230 Standard design certifications.
§ 53.1270 Manufacturing licenses
§ 53.1300 Construction permits.
§ 53.1360 Operating licenses.
§ 53.1410 Combined licenses.
§ 53.1470 Standardization of commercial nuclear power plant designs:
licenses to construct and operate nuclear power reactors of identical design at multiple sites.
Subpart H - Licenses, Certifications and Approvals
- Formatted using early site permits (ESPs) for siting-related content and DCs for design-related content
- Added existing provisions exempting U.S. Department of Defense reactors from NRC licensing (§ 53.1120).
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36 Site selected Part 50 Part 52 Part 53 Leveraging and Combining Existing Licensing Processes Operating License (OL)
CP based on SDA or DC Construction Permit (CP)
Commercial Operations Site selected Site selected Fuel Load Combined License (COL)
Manufacturing License (ML)
Standard Design Approval (SDA)
Use OL or custom COL to develop a subsequent DC Design Certification (DC)
CP and COL may reference Early Site Permit (ESP)
Subpart I - Maintaining and Revising Licensing Basis Information 37
§ 53.1500 Licensing basis information.
§ 53.1505 Changes to licensing basis information requiring prior NRC approval.
§ 53.1510 - 53.1520 License amendments.
§ 53.1525 - 53.1535 Specific provisions
§ 53.1540 - 53.1545 Other licensing information
§ 53.1550 Evaluating changes to facility as described in final safety analysis reports.
§ 53.1555 - 53.1565 Program-related documents
§ 53.1570 Transfer of licenses or permits.
§ 53.1575 Termination of license.
§ 53.1580 Information requests.
§ 53.1585 Revocation, suspension, modification of licenses, permits, and approvals for cause.
§ 53.1590 Backfitting.
§ 53.1595 Renewal.
Subpart I - Maintaining and Revising Licensing Basis Information
- Added existing change control requirements for individual programs (QA, Emergency Preparedness, security).
- Added change control provisions for DBAs and aircraft impact.
- Added existing generic license conditions (§ 53.1502).
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§ 53.1550 Evaluating changes to facility as described in final safety analysis reports (1 of 2)
(a) A licensee may make changes in the facility as described in the UFSAR and make changes in the procedures as described in the UFSAR without obtaining a license amendment pursuant to § 53.1510 only if:
(1) A change to the technical specifications incorporated in the license is not required and (2) The change meets all of the following criteria:
(i) Does not result in an increase to the frequency or consequences of an event sequence such that an event sequence previously deemed not risk significant becomes risk significant by the analyses performed in accordance with § 53.450(e).
(ii) Does not result in an increase to the frequency or consequences of an event sequence such that an event sequence deemed risk significant in accordance with § 53.450(e) has a decrease of 10 percent or more in the calculated margins to the LBE evaluation criteria required to be established in accordance with § 53.450(e).
(iii) Does not result in an increase to the frequency or consequences of one or more event sequences such that the margin between the calculated cumulative risks posed by the commercial nuclear plant and the safety criteria of § 53.220 decreases by 10 percent or more.
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§ 53.1550 Evaluating changes to facility as described in final safety analysis reports (2 of 2)
(iv) Does not involve a departure from a method of evaluation described in the UFSAR used in assessing margins in accordance with § 53.450(e) unless the results of the analysis under § 53.450(e) are conservative or essentially the same, the revised method of evaluation has been previously approved by the NRC for the intended application, or the revised method of evaluation can be used in accordance with an NRC endorsed consensus code or standard.
(v) For commercial nuclear plants licensed under this part for which alternative evaluation criteria are adopted in accordance with § 53.470, does not result in a change to the frequency or consequences of event sequences such that the calculated margins between the results for event sequences evaluated in accordance with § 53.450(e) and the alternative evaluation criteria decreases by 25 percent or more.
(vi) Does not result in the identification of a new design basis accident in accordance with § 53.450(f).
(vii) Does not result in a decrease by 10 percent or more in the margin between the consequence of any design basis accident and the safety criteria in § 53.210.
(viii) Does not prevent meeting the design requirements in § 53.440(j) to limit the release of radionuclides from reactor systems, waste stores, or other significant inventories of radioactive materials assuming the impact of a large, commercial aircraft.
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Subpart J - Reporting and Other Administrative Requirements 41
§ 53.1600 General information.
§ 53.1610 Unfettered access for inspections.
§ 53.1620 Maintenance of records, making of reports.
§ 53.1630 Immediate notification requirements for operating commercial nuclear plants.
§ 53.1640 Licensee event report system.
§ 53.1650 Facility information and verification.
§ 53.1655 Reporting of defects and noncompliance.
§ 53.1660 Financial requirements.
§ 53.1670 Financial qualifications.
§ 53.1680 Annual financial reports.
§ 53.1690 Licensees change of status; financial qualifications.
§ 53.1700 Creditor regulations.
§ 53.1710 Financial protection.
§ 53.1720 Insurance required to stabilize and decontaminate plant following an accident.
§ 53.1730 Financial protection requirements.
Subpart K - Quality Assurance Criteria
§ 53.1800 General Provisions
§ 53.1805 Organization
§ 53.1810 Quality Assurance Program
§ 53.1815 Design Control
§ 53.1820 Procurement Document Control
§ 53.1825 Instructions, Procedures and Drawings
§ 53.1830 Document Control
§ 53.1835 Control of Purchased Material, Equipment and Services
§ 53.1840 Identification and Control of Materials, Parts and Components
§ 53.1845 Control of Special Processes
§ 53.1850 Inspection
§ 53.1855 Test Control
§ 53.1860 Control of Measuring and Test Equipment
§ 53.1865 Handling, Storage and Shipping
§ 53.1870 Inspection, Test and Operating Status
§ 53.1875 Nonconforming Materials, Parts or Components
§ 53.1880 Corrective Action
§ 53.1885 Quality Assurance Records
§ 53.1890 Audits 42 10 CFR Part 50, Appendix B (Criterion I)
(Criterion II)
(Criterion III)
(Criterion IV)
(Criterion V)
(Criterion VI)
(Criterion VII)
(Criterion VIII)
(Criterion IX)
(Criterion X)
(Criterion XI)
(Criterion XII)
(Criterion XIII)
(Criterion XIV)
(Criterion XV)
(Criterion XVI)
(Criterion XVII)
(Criterion XVIII)
Discussion 43
Additional Topics for Discussion 44
Rulemaking Coordination
- Emergency Planning for Small Modular Reactors and Other New Technologies
- Decommissioning
- Part 50-52 Lessons Learned
- Financial Qualifications Requirements for Reactor Licensing 45
Key Guidance 46
Rulemaking Plan - SECY-20-0032 and SRM The staff should accelerate its timeline while balancing the need to produce a high-quality, thoroughly vetted regulation to achieve publication of the final rule by October 2024.
- Staffs response to SRM identified timing of guidance document development to support the Part 53 rulemaking as an uncertainty in meeting the accelerated schedule o Focus resources on developing the proposed rule language o Possible need to publish proposed rule before completing draft supporting guidance o Continue engaging external stakeholders to ensure common prioritization of guidance documents o Support early applications under Parts 50/52 (e.g., U.S. Department of Energys Advanced Reactor Demonstration Program)
- Siting Criteria (RG 4.7)
- Fuel Qualification Framework (NUREG-2246)
Existing
- Analytical Margin
- Chemical Hazards
- Manufacturing
- Technical Specifications
- Facility Safety Program
- Contents of Applications for Framework B Future Under Development Near-Term
- Non-LWR PRA Std
- TICAP/ARCAP (NEI 21-07)
- High Temp Materials (ASME III-5)
- Reliability & Integrity Mgt (ASME XI-2)
- Fuel Qualification (technology-specific)
- PRA Level of Detail (NEI-led)
- Seismic Design/Isolators
- Emergency Planning
- Change Process (SNC-led)
- QA Alternatives (NEI-led)
- Operator Training Program Part 53
- Qualitative Risk Estimate/Insights (Alternative Evaluation of Risk Insights [AERI])
- Operator licensing Exam
- Human Factors Engineering
- Concept of Operations/
Staffing
- Access Authorization
- Physical Security
- Materials Compatibility ISG Part 53
- Qualitative Risk Estimate/Insights (Alternative Evaluation of Risk Insights [AERI])
- Operator licensing Exam
- Human Factors Engineering
- Concept of Operations/
Staffing
- Access Authorization
- Physical Security
- Materials Compatibility ISG Key Guidance Coordination
- Siting Criteria (RG 4.7)
- Fuel Qualification Framework (NUREG-2246)
Existing Future Existing Guidance
- Existing guidance documents currently exist and will be referenced in the Part 53 rulemaking package as key guidance.
- Conforming changes will be needed to ensure they are applicable to Part 53.
- Revision will occur between proposed rule and final rule stages.
Future Near-Term
- Non-LWR PRA Std
- TICAP/ARCAP (NEI 21-07)
- High Temp Materials (ASME III-5)
- Reliability & Integrity Mgt (ASME XI-2)
- Fuel Qualification (technology-specific)
- PRA Level of Detail (NEI-led)
- Seismic Design/Isolators
- Emergency Planning
- Change Process (SNC-led)
- QA Alternatives (NEI-led)
- Operator Training Program Part 53
- Qualitative Risk Estimate/Insights (AERI)
- Operator licensing Exam
- Human Factors Engineering
- Concept of Operations/
Staffing
- Access Authorization
- Physical Security
- Materials Compatibility ISG Part 53
- Qualitative Risk Estimate/Insights (AERI)
- Operator licensing Exam
- Human Factors Engineering
- Concept of Operations/
Staffing
- Access Authorization
- Physical Security
- Materials Compatibility ISG Guidance Under Development
- Near-term guidance documents are currently under development and will be referenced as key guidance.
- These will be issued prior to the finalization of Part 53 to support near-term applicants and will need conforming changes to ensure they are applicable to Part 53.
- Revision will occur between proposed rule and final rule stages.
- Part 53-specific guidance documents are currently under development and are expected to be included with the Part 53 rulemaking package as key guidance.
- Analytical Margin
- Chemical Hazards
- Manufacturing
- Technical Specifications
- Facility Safety Program
- Contents of Applications for Framework B Future Future Guidance
- Future guidance documents are identified as future guidance that may need to be developed to support Part 53.
- These guidance documents may be referenced in the Part 53 rulemaking document as under development and are expected to be completed to support the final rule.
- Additional operational program guidance and reporting requirements guidance may be provided with the final rule.
Final Discussion and Questions 52
Closing Remarks Rulemaking Contacts Robert.Beall@nrc.gov 301-415-3874 Nanette.Valliere@nrc.gov 301-415-8462 Regulations.gov docket ID: NRC-2019-0062 Please provide feedback on this public meeting using this link:
https://www.nrc.gov/public-involve/public-meetings/contactus.html 53
Acronyms and Abbreviations ACRS Advisory Committee on Reactor Safeguards ADAMS Agencywide Documents Access and Management System AERI Alternative Evaluation of Risk Insights ALARA As Low As Reasonably Achievable AOO Anticipated operational occurrence ARCAP Advanced reactor content of application project ASCE American Society of Civil Engineers ASME American Society of Mechanical Engineers BDBE Beyond design basis event CFR Code of Federal Regulations COL Combined license CP Construction permit DBA Design basis accident DBE Design basis event DC Design certification DEC Design extension condition EAB Exclusion area boundary EPA U.S. Environmental Protection Agency ESP Early site permit F-C Frequency-consequence IAEA International Atomic Energy Agency ISG Interim staff guidance LBE Licensing basis event LWR Light-water reactor ML Manufacturing license NEI Nuclear Energy Institute NEIMA Nuclear Energy Innovation and Modernization Act NO Normal operations NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation 54
Acronyms and Abbreviations NUREG U.S. Nuclear Regulatory Commission technical report designation OL Operating license PAG Protective action guide PRA Probabilistic risk assessment QA Quality assurance QHO Quantitative health objective REM Roentgen equivalent man RG Regulatory guide SDA Standard design approval SEI Structural Engineering Institute SNC Southern Nuclear Company SRM Staff requirements memorandum SSCs Structures, systems, and components TICAP Technology-inclusive content of application project UFSAR Updated final safety analysis report 55