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 Regula on of Advanced Nuclear Reactors May 19, 2022
Agenda 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 2
Welcome and Introductions Welcome:
- Steve Lynch, Office of Nuclear Reactor Regulation (NRR)
NRC Speakers / Presenters:
- Bill Reckley, NRR
- Nan Valliere, NRR Meeting Slides:
- ADAMS Accession No. ML22125A001 3
Rulemaking Schedule Oct/Nov 2022 ACRS Interactions on Rulemaking Package for Proposed Rule 4
Part 53 Licensing Frameworks Subpart A - General Provisions Framework A Framework B
- Subpart B - Safety Requirements
- Subpart N - Purpose/Definitions
- Subpart C - Design Requirements
- Subpart D - Siting
- Subpart E - Construction
- Subpart O - Construction
- Subpart F - Operations
- Subpart P - Operations
- Subpart G - Decommissioning
- Subpart Q - Decommissioning Alternate Evaluation
- Subpart H - Licensing Processes
- Subpart R - Licensing Processes for Risk
- Subpart I - License Maintenance
- Subpart S - License Maintenance Insights
- Subpart J - Reporting
- Subpart T - Reporting
- Subpart K - Quality Assurance
- Subpart U - Quality Assurance 5
Part 53 Licensing Frameworks Framework B Framework A Emphasis Emphasis Design Criteria Risk metrics
- With addition of DBA used to set design criteria and
- Traditional approach represented by figure from IAEA guidance performance objectives for the design of Safety Related SSCs.
6
Consolidated Preliminary Rule Language (Including Second Iteration)
Summary & Changes 7
Framework A 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 8
Evolution of Part 53 & Stakeholder Feedback Topic Addressed in Preliminary Proposed Rule Language
- Quality Assurance (QA) requirements consolidated in Subpart K.
Duplicative/overlapping programs
- Added flexibility for licensees to organize and combine programs, as appropriate, to avoid duplication (Subparts F & K).
Expanded activities permitted under ML to include fabrication of entire reactor, Manufacturing license (ML) expansion 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.
Decoupled requirements for normal operation from those for licensing basis Normal operations events (LBEs) (Subparts B & C).
- The staff has removed references to advanced nuclear plant.
Use of advanced nuclear plant and
- No plans to expand applicability to research and test reactors (note that expansion beyond commercial reactors 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.
11
Subpart B - Safety Criteria
- Safety Objectives
- DBA Safety Criteria Safety Criteria What function(s)
- Non-DBA Safety Criteria (e.g., a barrier, cooling) are needed to satisfy safety
- Safety Functions criteria
- Licensing Basis Events Safety Functions
- Defense in Depth What design features (e.g., a structure, system) are provided to fulfill the
- Role of: safety function(s) o Structures, systems, and components (SSCs) Design Features (and Human Actions) What design criteria o Personnel (e.g., leak rate, cooling capacity) are needed for o Programs design feature Functional Design Criteria (Personnel; Concept of Operations) 12
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.
13
QHOs - Updated Preliminary Proposed Rule Language (May 2022)
§ 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.
14
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).
15
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 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.
18
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
§ 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.
20
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 (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.
23
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.
24
Event Selection & Analysis 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.
25
Event Selection & Analysis 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.
26
§ 53.450(f) - Design basis accidents (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.
27
Design Basis Accidents 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) 28
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
§ 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.
34
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).
35
Leveraging and Combining Existing Licensing Processes Commercial Operations Fuel Load Operating License Site (OL) selected Combined License (COL)
Use OL or custom COL to develop a CP based on Site subsequent DC SDA or DC selected Standard Design Manufacturing Design Certification Approval (SDA) License (ML) (DC)
Part 50 Site selected Part 52 CP and COL may reference Early Site Permit (ESP) Part 53 Construction Permit (CP) 36
Subpart I - Maintaining and Revising Licensing Basis Information
§ 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.
37
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).
38
§ 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.
39
§ 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.
40
Subpart J - Reporting and Other Administrative Requirements
§ 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.
41
Subpart K - Quality Assurance Criteria
§ 53.1800 General Provisions 10 CFR Part 50, Appendix B
§ 53.1805 Organization (Criterion I)
§ 53.1810 Quality Assurance Program (Criterion II)
§ 53.1815 Design Control (Criterion III)
§ 53.1820 Procurement Document Control (Criterion IV)
§ 53.1825 Instructions, Procedures and Drawings (Criterion V)
§ 53.1830 Document Control (Criterion VI)
§ 53.1835 Control of Purchased Material, Equipment and Services (Criterion VII)
§ 53.1840 Identification and Control of Materials, Parts and Components (Criterion VIII)
§ 53.1845 Control of Special Processes (Criterion IX)
§ 53.1850 Inspection (Criterion X)
§ 53.1855 Test Control (Criterion XI)
§ 53.1860 Control of Measuring and Test Equipment (Criterion XII)
§ 53.1865 Handling, Storage and Shipping (Criterion XIII)
§ 53.1870 Inspection, Test and Operating Status (Criterion XIV)
§ 53.1875 Nonconforming Materials, Parts or Components (Criterion XV)
§ 53.1880 Corrective Action (Criterion XVI)
§ 53.1885 Quality Assurance Records (Criterion XVII)
§ 53.1890 Audits (Criterion XVIII) 42
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)
Key Guidance Coordination Under Development Near-Term Part 53 Existing Future
- Non-LWR PRA Std
- TICAP/ARCAP (NEI 21-
- Qualitative Risk
- Licensing
- Analytical Margin
- 07) Estimate/Insights Modernization
- Chemical Hazards
- High Temp Materials (Alternative Evaluation of Project (NEI 18-
- Manufacturing (ASME III-5) Risk Insights [AERI])
04 & RG. 1.233)
- Technical
- Reliability & Integrity Mgt
- Operator licensing Exam
- Human Factors Engineering 4.7)
- Facility Safety
- Fuel Qualification
- Concept of Operations/
- Fuel Qualification Program (technology-specific) Staffing Framework
- Contents of
- PRA Level of Detail (NEI-
- Fitness for Duty (NUREG-2246) led) Applications for
- Access Authorization
- Seismic Design/Isolators Framework B
- Emergency Planning
- Physical Security
- Change Process (SNC-led)
- Materials Compatibility ISG
- QA Alternatives (NEI-led)
- Operator Training Program
Existing Guidance Existing Future
- Existing guidance documents currently exist
- Licensing and will be referenced in the Part 53 Modernization rulemaking package as key guidance.
Project (NEI 18-04 & RG. 1.233)
- Conforming changes will be needed to ensure
- Siting Criteria (RG they are applicable to Part 53.
4.7)
- Fuel Qualification
- Revision will occur between proposed rule Framework and final rule stages.
Guidance Under Development
- Near-term guidance documents are Near-Term Part 53 currently under development and will Future
- Non-LWR PRA Std
- Initiating Events be referenced as key guidance.
- TICAP/ARCAP (NEI 21-
- Qualitative Risk
- These will be issued prior to the
- 07) Estimate/Insights (AERI) finalization of Part 53 to support
- High Temp Materials
- Operator licensing Exam near-term applicants and will need (ASME III-5)
- Human Factors Engineering conforming changes to ensure they
- Reliability & Integrity Mgt
- Concept of Operations/ are applicable to Part 53.
(ASME XI-2) Staffing
- Revision will occur between
- Fuel Qualification
- Fitness for Duty proposed rule and final rule stages.
(technology-specific)
- Access Authorization
- PRA Level of Detail (NEI-
- Cyber Security led)
- Physical Security
- Part 53-specific guidance
- Seismic Design/Isolators
- Materials Compatibility ISG documents are currently under
- Emergency Planning development and are expected to be included with the Part 53 rulemaking
- Change Process (SNC-led) package as key guidance.
- QA Alternatives (NEI-led)
- Operator Training Program
Future Guidance Future
- Future guidance documents are identified as future guidance that may need to be developed to support Part
- Analytical Margin
- 53.
- Chemical Hazards
- Manufacturing
- These guidance documents may be referenced in the Part
- Technical 53 rulemaking document as under development and are expected to be completed to support the final rule. Specifications
- Facility Safety
- Additional operational program guidance and reporting Program requirements guidance may be provided with the final
- Contents of rule. Applications for Framework B
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 DEC Design extension condition Agencywide Documents Access and EAB Exclusion area boundary ADAMS Management System EPA U.S. Environmental Protection Agency AERI Alternative Evaluation of Risk Insights ESP Early site permit ALARA As Low As Reasonably Achievable F-C Frequency-consequence AOO Anticipated operational occurrence IAEA International Atomic Energy Agency ARCAP Advanced reactor content of application project ISG Interim staff guidance ASCE American Society of Civil Engineers LBE Licensing basis event ASME American Society of Mechanical Engineers LWR Light-water reactor BDBE Beyond design basis event ML Manufacturing license CFR Code of Federal Regulations NEI Nuclear Energy Institute COL Combined license Nuclear Energy Innovation and Modernization CP Construction permit NEIMA Act DBA Design basis accident NO Normal operations DBE Design basis event NRC U.S. Nuclear Regulatory Commission DC Design certification NRR Office of Nuclear Reactor Regulation 54
Acronyms and Abbreviations U.S. Nuclear Regulatory Commission technical SDA Standard design approval NUREG report designation SEI Structural Engineering Institute OL Operating license SNC Southern Nuclear Company PAG Protective action guide SRM Staff requirements memorandum PRA Probabilistic risk assessment SSCs Structures, systems, and components QA Quality assurance Technology-inclusive content of application QHO Quantitative health objective TICAP project REM Roentgen equivalent man UFSAR Updated final safety analysis report RG Regulatory guide 55