ML22131A001
| ML22131A001 | |
| Person / Time | |
|---|---|
| Issue date: | 05/25/2022 |
| From: | Robert Beall NRC/NMSS/DREFS/RRPB |
| To: | |
| Beall, Robert | |
| References | |
| 10 CFR Part 26, 10 CFR Part 53, 10 CFR Part 73, NRC-2019-0062, RIN 3150-AK31 | |
| Download: ML22131A001 (43) | |
Text
10 CFR Part 53 Risk-Informed, Technology-Inclusive Regulatory Framework For Commercial Nuclear Plants Public Meeting on Second Iteration of the Consolidated Rule Text May 25, 2022
Agenda 2
Topic Staff Introduction
- Part 53 Overview
- Framework A Subparts
- Fitness for Duty (Part 26)
- Physical Security, Cyber Security, Access Authorization (Part 73)
Welcome and Introductions 3
Welcome: Steve Lynch, Office of Nuclear Reactor Regulation NRC Speakers / Presenters:
- Office of Nuclear Material Safety and Safeguards
- Office of Nuclear Reactor Regulation
- Office of Nuclear Security and Incident Response Meeting Slides:
- ADAMS Accession No. ML22131A001
Rulemaking Schedule 4
Part 53 Licensing Frameworks 5
Framewor k B
- Subpart B - Safety Requirements
- Subpart C - Design & Analysis
- Subpart D - Siting
- Subpart E - Construction & Manufacturing
- 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 & Manufacturing
- 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 A Subpart A - General Provisions Framework B
Consolidated Preliminary Rule Language Second Iteration (May 2022)
(ADAMS Accession No. ML22125A000)
Summary & Changes 6
Framework A 7
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 Operations (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
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 8
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.
9
Safety Objectives Design basis accident (DBA)
Safety Criteria Non-DBA Safety Criteria Safety Functions Licensing Basis Events (LBEs)
Defense in Depth Role of:
o Structures, systems, and components (SSCs) o Personnel o
Programs 10 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 11
QHOs - Updated Preliminary Proposed Rule Language (May 2022) 12
§ 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.
Subpart C - Design and Analysis Requirements 13
§ 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 Minimizing contamination
- Reordered and clarified design requirements 14
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
- § 53.480 Earthquake engineering o Applicability to non-safety-related but safety significant SSCs commensurate with their safety significance 15
§ 53.450(e) - LBEs other than DBA 16 (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.
17 Event Selection & Analysis Licensing Modernization Project Anticipated Operational Occurrences (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.
18 Licensing Modernization Project Design Basis Events (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
19 Licensing Modernization Project Beyond Design Basis Events (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.
20
§ 53.450(f) - Design basis accidents
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 special treatment, including where appropriate quality assurance (QA) criteria from Subpart K, for siting activities.
21
Subpart E - Construction and Manufacturing Requirements
§ 53.600 Construction and manufacturing - scope and purpose.
§ 53.605 Reporting of defects and noncompliance.
§ 53.610 Construction.
§ 53.620 Manufacturing.
- Added § 53.605 to capture requirements in § 50.55(e).
- Clarified requirements for manufacturing licenses (MLs) allowing fuel loading.
o References to 10 CFR Part 70 22
Subpart F - Requirements for Operation To be discussed during June ACRS meeting 23
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.
24
Subpart H - Licenses, Certifications and Approvals 25
§ 53.1100 - 53.1121 General/common requirements.
§ 53.1124 Relationship between sections.
§ 53.1130 Limited work authorizations.
§ 53.1140 - 53.1188 Early site permits.
§ 53.1200 - 53.1221 Standard design approvals.
§ 53.1230 - 53.1263 Standard design certifications.
§ 53.1270 - 53.1295 Manufacturing licenses.
§ 53.1300 - 53.1348 Construction permits.
§ 53.1360 - 53.1405 Operating licenses.
§ 53.1410 - 53.1461 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
- Added existing provisions exempting U.S.
Department of Defense reactors from NRC licensing (§ 53.1120).
- Removed allowance for construction permits to reference MLs.
26
27 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 28
§ 53.1500 Licensing basis information.
§ 53.1502 Specific terms and conditions.
§ 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.
§ 53.1575 Termination of license.
§ 53.1580 Information requests.
§ 53.1585 Revocation, suspension, modification of licenses 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 (§ 53.1550(a)(2)(vi)-(viii)).
- Added existing generic license conditions (§ 53.1502).
29
Subpart J - Reporting and Other Administrative Requirements 30
§ 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.
§ 53.1645 Periodic reports.
§ 53.1650 Facility information and verification.
§ 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.
31 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 32
Part 26 - Fitness for Duty Revised Fitness for Duty (FFD) criteria from two criteria to one Enabled the use of hair specimens for pre-access drug screening and portal monitor passive drug and alcohol screening at the Protected Area access point Published proposed performance measures 1.
Conduct of the behavioral observation program 2.
Occurrence of FFD policy violations categorized by licensee employee, contractor/vendor, and labor category 3.
Occurrence of individuals with potentially disqualifying information, who possessed FFD prohibited items, or who were impaired while in a work status 4.
If drug and alcohol testing, performance measures must also include:
a) pre-access and random positive testing rates b) the number of subversion attempts categorized by licensee employee, contractor/vendor, and labor category 33
Part 26 - Fitness for Duty Minimum Staff Sizes and/or Geographically-remote Sites Upon further review of facilities licensed with minimal staff sizes (potentially in remote locations), the NRC staff strengthened the behavioral observation requirement to ensure that individuals who arrive onsite to monitor the facility at the controls are subject to behavioral observation by a video and audible capture at the central monitoring or security station 34
Part 26 - Fitness for Duty Minimum Staff Sizes and/or Geographically-remote Sites NRC staff intends to issue regulatory guidance to address the implementation of behavioral observation and random drug and alcohol testing programs at facilities that have relatively small staff sizes and/or are geographically remote o Small staff sizes have the potential to degrade the conduct of behavioral observation (e.g., the identification and mitigation of the insider threat) and challenge NRC Safety Culture o Geographically-remote locations may intensify the small staff size challenge 35
Discussion 36
Section 73.110 - Cyber Security
- The NRC staff made changes to enhance the consistency/clarity of the current rule language.
Examples include:
o Changed the term postulated radiological release in
§73.110(a)(1) to postulated fission product release to align with terminology used in other Part 53 sections.
o Changed §73.110 (e)(2)(ii) to clarify that the Cyber Security Plan should include the analysis described in paragraph (b)(1) dealing with the potential consequences resulting from cyber attacks.
37
Section 73.120 - Access Authorization
- The NRC staff made changes to enhance the consistency/clarity of the current rule language to be in alignment with the Part 52 rulemaking effort. For example:
o Revised Section 73.120(a)(1), Introduction and scope, to align with the security trigger point under Part 52 for requirements to establish, maintain, and implement an access authorization program before initial fuel load into the reactor.
o Deleted the reference to (ii)(A) self-reporting of legal actions, under (c)(2), and made it its own performance element as a (c)(3). This would be consistent with how the current operating power reactors implement self-reporting of legal actions under
§ 73.56.
38
Discussion 39
Final Discussion and Questions 40
Next StepsFuture Public Meetings
- June TBD Public Meeting on Framework B
- June 23rd and 24th ACRS Subcommittee Meeting on Framework B and Subpart F (Operations) of Framework A Stay informed! Subscribe to GovDelivery:
https://service.govdelivery.com/accounts/USNRC/
subscriber/new 41
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 42
Acronyms and Abbreviations ACRS Advisory Committee on Reactor Safeguards ADAMS Agencywide Documents Access and Management System ALARA As Low As Reasonably Achievable AOO Anticipated operational occurrence 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 EAB Exclusion Area Boundary EPA U.S. Environmental Protection Agency ESP Early site permit F-C Frequency-consequence FFD Fitness for Duty LBE Licensing basis event ML Manufacturing license NRC U.S. Nuclear Regulatory Commission OL Operating license PAG Protective Action Guide PRA probabilistic risk assessment QA Quality assurance QHO Quantitative Health Objective Rem Roentgen equivalent man SDA Standard design approval SSCs Structures, systems, and components 43