ML22165A114
| ML22165A114 | |
| Person / Time | |
|---|---|
| Issue date: | 06/16/2022 |
| From: | Robert Beall NRC/NMSS/DREFS/RRPB |
| To: | |
| Beall, Robert | |
| References | |
| 10 CFR Part 53, NRC-2019-0062, RIN 3150-AK31 | |
| Download: ML22165A114 (43) | |
Text
P u b l i c M e e t i n g 6 / 1 6 / 2 2 Part 53 Framework B Overview
Agenda
- Overview of Part 53 Structure
- Comparison of Part 53 Frameworks
- Framework B Development Approach
- Framework B Subparts Overview
- Alternative Evaluation of Risk Insights (AERI)
- Next steps 2
Welcome and Introductions Welcome:
Rob Taylor, Office of Nuclear Reactor Regulation NRC Speakers / Presenters:
Office of Nuclear Material Safety and Safeguards
- Bob Beall Office of Nuclear Reactor Regulation
- Bill Jessup
- Marty Stutzke
- Boyce Travis Meeting Slides:
ADAMS Accession No. ML22165A114 3
Purpose of Todays Meeting
- Overview of the Part 53 proposed Framework B rulemaking effort.
- Todays meeting is a Comment-Gathering meeting, which means that public participation is actively sought in the discussion of the regulatory issues during the meeting.
- The meeting is being transcribed and the transcription will be available with the meeting summary by July 16, 2022.
- No regulatory decisions will be made at todays meeting.
4
Part 53 Licensing Frameworks Framework A o Probabilistic risk assessment (PRA)-led approach o Functional design criteria Framework B o Traditional use of risk insights o Principal design criteria o Includes an AERI approach Subpart A - General Provisions Subpart B - Safety Requirements Subpart C - Design Requirements Subpart D - Siting Subpart E - Construction/Manufacturing Subpart F - Operations Subpart G - Decommissioning Subpart H - Application Requirements Subpart I - License Maintenance Subpart J - Reporting Subpart K - Quality Assurance Subpart N - Definitions Subpart O - Construction/Manufacturing Subpart P - Operations Subpart Q - Decommissioning Subpart R - Application Requirements Subpart S - License Maintenance Subpart T - Reporting Subpart U - Quality Assurance 5
Part 53 Subpart Comparison Subpart Title Framework A Subpart Framework B Subpart General Provisions Subpart A (Common)
Technology-Inclusive Safety Requirements Subpart B Design and Analysis Requirements Subpart C Siting Requirements Subpart D (Part 100)
Definitions Subpart N Construction and Manufacturing Requirements Subpart E Subpart O Requirements for Operation Subpart F Subpart P Decommissioning Requirements Subpart G Subpart Q Licenses, Certifications, and Approvals Subpart H Subpart R Maintaining and Revising Licensing Basis Information Subpart I Subpart S Reporting and Other Administrative Requirements Subpart J Subpart T Quality Assurance Criteria Subpart K Subpart U 6
Framework B Development Approach 7
Subpart N - Definitions Definitions specific to Framework B o
Anticipated operational occurrence (AOO) o Design bases o
Reactor coolant pressure boundary o
Safety-related structures, systems, and components (SSCs)
Common definitions remain in Subpart A (§ 53.020) 8
Subpart O - Construction and Manufacturing Requirements Parallel structure and content to Framework A Subpart E Variations largely limited to conforming changes needed to adapt Framework A provisions to Framework B 9
Subpart P - Requirements for Operation
§ 53.4210 Maintenance, repair, and inspection programs.
§ 53.4213 Technical specifications.
§§ 53.4220 - 53.4299 General staffing, training, personnel qualifications, and human factors requirements.
§ 53.4300 Programs.
§ 53.4310 Programs: Radiation protection.
§ 53.4320 Programs: Emergency preparedness.
§ 53.4330 Programs: Security programs.
§ 53.4340 Programs: Quality assurance.
§ 53.4350 Programs: Fire protection.
§ 53.4360 Programs: Inservice inspection/inservice testing.
§ 53.4380 Programs: Environmental qualification of electric equipment
§ 53.4390 Programs: Procedures and guidelines.
§ 53.4400 Programs: Integrity assessment program.
§ 53.4410 Programs: Primary containment leakage rate testing program.
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Subpart P - Requirements for Operation
- Maintenance, repair, and inspection programs generally aligned with § 50.65
- Technical specifications generally aligned with § 50.36
- Programs o Security, Emergency Preparedness, Radiation Protection requirements aligned with Framework A o Environmental qualification of electrical equipment derived from § 50.49 o Scope of SSCs in Integrity Assessment Program aligned more closely with
§ 53.4210(b) (§ 50.65(b))
o Containment leak rate requirements from Part 50 (§ 50.54(o))
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Subpart P - Requirements for Operation Staffing, Training, Personnel Qualifications, and Human Factors
- Framework B adopts most requirements from Framework A through cross-references or copying requirements with some minor changes
- Staffing plan requirements in § 53.4226(f) include the need for engineering expertise availability to support on-shift operating personnel o Must be familiar with facility operation and meet at least one educational or credential requirement in §§ 53.4226(f)(1)(i) through (f)(1)(iii) o Developed in response to ACRS feedback on blanket removal of shift technical advisor position
- Framework A's provisions for alternatives to the use of licensed Reactor Operators and Senior Reactor Operators are not currently translated to Framework B; the staff will continue to evaluate options in this area 12
Subpart P - Requirements for Operation Fire Protection
- Combination of § 50.48, Appendix R, and NFPA 805 Chapter 3 o All requirements are contained in in-line rule text No appendices in Part 53 No cross-references back to Parts 50 or 52
- No fire PRA required, but may be useful in performance-based justifications o Provision for performance-based alternatives to detailed requirements with NRC approval (like § 50.48(c)(2)(vii) and § 50.48(c)(4))
- Technology neutral o Designers must define the safe and stable state for their design o Designers must determine the safe shutdown functions to achieve and maintain safe and stable state 13
Subpart Q - Decommissioning Requirements Parallel structure and content to Framework A Subpart G Variations largely limited to conforming changes needed to adapt Framework A provisions to Framework B 14
Subpart S -
Maintaining and Revising Licensing Basis Information
- Parallel structure and content to Framework A Subpart I
- Notable differentials o § 53.6010, Application for amendment of license o § 53.6040, Updating licensing basis information and determining the need for NRC approval o § 53.6045, Updating final safety analysis reports o § 53.6050, Evaluating changes to facility as described in final safety analysis reports o § 53.6052, Maintenance of risk evaluations
- Remaining variations largely limited to conforming changes to adapt Framework A provisions to Framework B 15
Subpart T - Reporting and Other Administrative Requirements
- Parallel structure and content to Framework A Subpart J
- Notable differentials o § 53.6320(e) added to align with state-of-practice policy initiative on reporting requirement for fee purposes o § 53.6330, Immediate notification requirements for operating commercial nuclear plants, aligned with § 50.72 o § 53.6340, Licensee event report system, aligned with § 50.73
- Remaining variations largely limited to conforming changes to adapt Framework A provisions to Framework B 16
Subpart U - Quality Assurance
- Subpart U parallels structure and content of Framework A Subpart K
- Closely aligned with 10 CFR Part 50 Appendix B (18 criteria)
- Exception: § 53.6635, Control of Purchased Material, Equipment and Services (10 CFR Part 50 Appendix B Criterion VII) o Commercial nuclear plant used in lieu of nuclear power plant o Ensures consistency with terminology throughout Part 53 17
Subpart R - Licenses, Certifications, and Approvals
§ 53.4700 General Provisions.
§ 53.4725 Standards for review.
§ 53.4730 General technical requirements.
§ 53.4731 Risk-informed classification of structures, systems, and components.
§ 53.4740 Limited work authorizations.
§ 53.4750 Early site permits.
§ 53.4800 Standard design approvals
§ 53.4830 Standard design certifications.
§ 53.4870 Manufacturing licenses.
§ 53.4900 Construction permits.
§ 53.4960 Operating licenses.
§ 53.5010 Combined licenses.
18
- Subpart R developed to parallel Subpart H in Framework A o Covers all application types (e.g., Construction Permit (CP), Operating License (OL), Combined License (COL))
o Process-related requirements (e.g., duration of a license) similar or the same between frameworks o Technical contents of application structures derived from Parts 50 and 52 and represent primary differentiator between Subparts H and R o Includes § 53.4731 that parallels § 50.69 regarding risk-informed SSC classification Subpart R - Licenses, Certifications, and Approvals 19
- Section § 53.4730, General technical requirements, consolidates technical content of application requirements for the various application types o COL technical contents of application (§ 52.79) used as a starting point o Each application type references back to § 53.4730
Reduces rule length
Minimizes the potential for requirements to diverge between application types Subpart R - Licenses, Certifications, and Approvals COL OL CP ML DC SDA ESP (1)
X X
X X
X X
X (2)
X X
X X
X X
X (3)
X X
X X
X X
(37)
X X
X X
X X
X Application Type 53.4730(a)
Requirement 20
Accident Analyses and Initiating Event Requirements Requirements in § 53.4730(a)(5) derived from previous Part 5X work undertaken in 2021 that proposed technology-inclusive alternatives to some requirements in Parts 50 and 52 AOOs (§ 53.4730(a)(5)(iii)): Requirements consistent with existing requirements in traditional frameworks with Part 20 acceptance criteria Design Basis Accidents (DBAs) (§ 53.4730(a)(5)(ii)): New technology-neutral requirements for DBA analyses and SSC classification based loosely on §§ 50.34(a)(4) and 50.46 Beyond Design Basis Events (BDBEs) (§ 53.4730(a)(5)(iv)): Provides technology-inclusive requirements for relevant BDBEs and analysis requirements for other BDBEs, drawn from Anticipated Transient Without Scram/Station Black Out rulemakings; similar to international defense in depth requirements.
Severe Accidents (§ 53.4730(a)(5)(v)): Derived from current requirements in § 52.79(a)(38), with modifications made to support technology-inclusiveness Chemical Hazards (§ 53.4730(a)(5)(vi)): Requirements based on language proposed in Framework A to address potential chemical hazards associated with licensed material Subpart R - Licenses, Certifications, and Approvals 21
Risk insights support or complement deterministic analyses, consistent with traditional approach Includes requirement to provide a description of the plant-specific PRA and its results translated to Framework B
§ 52.79(a)(44) § 53.4730(a)(34)(i)
Optional alternative risk evaluation for applicants that meet the criteria in § 53.4730(a)(34)(ii) o No PRA required o Implicitly demonstrates that quantitative health objectives (QHOs) are met, searches for severe accident vulnerabilities, and provides risk insights without a requirement for a PRA o Inherently addresses the mitigation of beyond-design-basis events requirements when AERI entry criteria are met o Cannot implement risk-informed applications if AERI approach is used Risk evaluations (PRA or AERI) must be maintained consistent with requirements in Subpart S
(§ 53.6052, informed by § 50.71(h))
Subpart R - Licenses, Certifications, and Approvals Assessing Risk in Framework B 22
- Evolved from the staffs graded PRA initiative starting in Spring 2021 o Grade the technical content of the PRA o Grade the uses of the PRA in the design and licensing process PRA in an enhanced/leading role PRA in a supporting/confirmatory/traditional role
- Various names have been used to describe the concept:
o Dose/consequence-based approach o Technology-inclusive, risk-informed maximum accident (TIRIMA) approach o Part 53-BE (bounding event) o AERI Alternative Evaluation for Risk Insights 23
Uses of PRA
- The Policy Statement on the Regulation of Advanced Reactors (73 FR 60612; October 14, 2008) references three PRA-related policy statements:
o Safety Goals for the Operation of Nuclear Power Plants (51 FR 28044; August 4, 1986 as corrected and republished at 51 FR 30028; August 21, 1986) o Severe Reactor Accidents Regarding Future Designs and Existing Plants (50 FR 32138; August 8, 1985) o Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities (60 FR 42622; August 16, 1995)
- The AERI approach and two pre-decisional draft regulatory guides (PDGs) have been developed to:
o Provide sufficient risk information to inform licensing decisions o Address related ACRS recommendations Search for severe accident vulnerabilities Identify risk insights Meet the QHOs 24
- October 7, 2019 - Letter concerning review of draft SECY paper, "Population - Related Siting Considerations for Advanced Reactors," ML19277H071:
o Need to examine new designs with a clean sheet of paper.
o Think carefully about the failures and combinations of failures that could occur.
o Must remain vigilant and remember that nature provides surprises.
o Creative thinking will be required to identify such unique situations, to thoroughly identify the scenarios that will be the basis of the safety analysis and the source of releases, and to evaluate the suitability of sites.
- October 20, 2020 - Letter concerning 10 CFR Part 53, ML20091L698:
o Compensate for novel designs with uncertainties due to incompleteness in the knowledge base by performing systematic searches for hazards, initiating events, and accident scenarios with no preconceptions that could limit the creative process.
- May 5, 2021 - Letter concerning Part 53, ML21140A354:
o Compensate for novel designs with uncertainties due to incompleteness in the knowledge base by performing systematic searches for hazards, initiating events, and accident scenarios with no preconceptions that could limit the creative process.
- October 26, 2021 - Letter concerning RG 1.247, ML21288A018:
o Include guidance that the initial search for initiating events and scenarios should be done without preconceptions or using existing lists.
ACRS Recommendations 25
26 26 Perform transient and accident analyses Perform design basis accident radiological consequences analyses Identify and analyze the bounding event Finish PRA development Select LBEs Select DBAs Classify SSCs Continue design and licensing activities Evaluate defense-in-depth Comprehensive and systematic initiator search and event sequence delineation without preconceptions or reliance on predefined lists Select licensing events Select licensing framework Perform transient and accident analyses Perform design basis accident radiological consequences analyses Elect to develop PRA Finish PRA development AERI Q1 - Develop demonstrably conservative risk estimate using the bounding event Q3 - Develop risk insights by reviewing all event sequences Q2 - Search all event sequences for severe accident vulnerabilities Continue design and licensing activities Continue design and licensing activities A
Parts 50 and 52 with LMP Part 53 Framework A Parts 50 and 52 without LMP Part 53 Framework B B
C D
E F
G H
I J
K L
M N
O yes no Applicant decision PDG-1413, Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants PDG-1414, Alternative Evaluation for Risk Insights (AERI) Framework Licensing Modernization Project (LMP) guidance - NEI 18-04, Rev. 1, as endorsed in RG 1.233 AERI entry conditions met?
P yes no Q
Licensing Frameworks - Risk Evaluation Perspective AERI ONLY for Part 53 Framework B Notes:
1)
Each step builds on all of the preceding steps (considers all information available at that point) 2)
Feedback loops (e.g., the impact of design revisions) are not shown 26
Proposed AERI Entry Conditions 53.4730(a)(34) Description of risk evaluation.
A description of the risk evaluation developed for the commercial nuclear plant and its results. The risk evaluation must be based on:
(i)
A PRA, or (ii)
An AERI, provided that the dose from a postulated bounding event to an individual located 100 meters (328 feet) away from the commercial nuclear plant does not exceed 1 rem total effective dose equivalent (TEDE) over the first four days following a release, an additional 2 rem TEDE in the first year, and 0.5 rem TEDE per year in the second and subsequent years.
Provides plants with flexibility in establishing their exclusion area boundaries if the bounding events source term is small.
The 100-meter criterion was back-calculated from a scoping consequence model:
o 50-year dose at 100 meters = 27.5 rem TEDE o
Conditional individual latent cancer fatality risk = 2 x 10-6 per event o
Meet the QHO without developing a PRA to credit accident frequency in the risk estimate Some stakeholders have confused the AERI entry conditions with safety/siting criteria.
27
Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants (Pre-decisional DG-1413)
- Formatted like a regulatory guide; currently a pre-decisional draft regulatory guide
- Section A: Applies to light water reactors (LWRs) and non-LWRs licensed under Parts 50, 52, and 53 (Frameworks A and B)
- Section B:
o Identifies licensing events for each licensing framework o Provides historical perspectives (early licensing, development of the standard review plan) o Addresses ACRS recommendations to start with a blank sheet of paper (10/7/2019, 10/21/2020, 5/30/2021, and 10/26/2021)
- Section C provides an integrated approach for:
o Conducting a systematic and comprehensive search for initiating events o Delineating a systematic and comprehensive sets of event sequences o Grouping the lists of initiating events and event sequences into licensing events
- Appendix:
o Recommends the use of one inductive method and one deductive method when searching for initiating events o Points the user to helpful references (NRC, IAEA, IEC, ASME/ANS, AIChE, EPRI, open literature) o Does not endorse or recommend any specific method 28
Alternative Evaluation for Risk Insights (AERI) Framework (Pre-decisional DG-1414)
- Formatted like a regulatory guide; currently a pre-decisional draft regulatory guide
- Section A: Only applies to LWRs and non-LWRs licensed under Part 53 Framework B
- Sections B & C: Components of the AERI approach:
o Identification and characterization of the bounding event o Definition of a bounding event Multiple events may need to be considered as bounding events o Determination of a consequence estimate for the bounding event to confirm that the reactor design meets the AERI entry conditions o Determination of a demonstrably conservative risk estimate for the bounding event to demonstrate that the QHOs are met Assumed frequency of 1/yr consistent with frequency of all event sequences for LWRs Applicant may use a lower frequency with justification o Search for severe accident vulnerabilities for the entire set of licensing events Definitions of severe accident and severe accident vulnerability o Identification of risk insights for the entire set of licensing events o Assessment of defense-in-depth adequacy for the entire set of licensing events 29
Framework B Guidance Development Many Framework A and B guidance development activities are linked May involve updates or supplements to existing guidance covering existing regulatory frameworks Guidance for technical content of application requirements now part of Advanced Reactor Content of Application Project effort 30
Areas of Focus for Merger of Frameworks A and B Ensure consistency between parallel provisions
- Siting
- Seismic Design Criteria
- Requirements for Operation Evaluate other provisions for potential alignment
- Definitions
- General Provisions Commonalities in Subpart A Continue consideration of stakeholder feedback 31
Next Steps Advisory Committee on Reactor Safeguards
- Subcommittee:
June 23 - 24, 2022
- Full Committee:
July 6 - 9, 2022 Advanced Reactor Public Stakeholder Meeting:
June 30, 2022 Commission Meeting:
July 21, 2022 32
Additional Information Additional information on the 10 CFR Part 53 rulemaking is available at https://www.nrc.gov/reactors/new-reactors/advanced/rulemaking-and-guidance/part-53.html For information on how to submit comments go to https://www.regulations.gov and search for Docket ID NRC-2019-0062 For further information, contact Robert Beall, Office of Nuclear Material Safety and Safeguards, telephone: 301-415-3874; email:
Robert.Beall@nrc.gov 33
ACRS Advisory Committee on Reactor Safeguards ADAMS Agencywide Documents Access and Management System AERI Alternative evaluation of risk insights AIChe American Institute of Chemical Engineers ANS American Nuclear Society AOO Anticipated operational occurrence ASME American Society of Mechanical Engineers BDBE Beyond design basis event BE Bounding event CFR Code of Federal Regulations COL Combined license CP Construction permit DBA Design basis accident DC Design certification EPRI Electric Power Research Institute ESP Early site permit FR Federal Register Acronyms IAEA International Atomic Energy Agency IEC The Incident and Emergency Centre LBE Licensing basis event LMP Licensing Modernization Project LWR Light water reactor ML Manufacturing license NEI Nuclear Energy Institute NFPA National Fire Protection Association NRC U.S. Nuclear Regulatory Commission OL Operating license PDG Pre-decisional draft regulatory guide PRA Probabilistic risk assessment QHO Quantitative health objective RG Regulatory guide SDA Standard design approval SSCs Structures, systems, and components TEDE Total effective dose equivalent TIRIMA Technology-inclusive, risk-informed maximum accident 34
Backup Slides 35
Regulatory Framework Options
- 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 and Insights
- Traditional approach represented by figure from IAEA guidance.
36
Derivation of AERI Entry Conditions (1 of 7)
Risk, R, is the sum of the products of frequency,, and consequence,,
over the set of delineated event sequences.
Suppose we can identify a bounding event.
max,,,
= sum of the initiating event frequencies 1/plant-year, based on large LWR history Then we can bound the risk.
1 2
3 4
This demonstrably conservative approach eliminates the need to estimate the individual event sequence frequencies by developing a PRA.
37
Derivation of AERI Entry Conditions (2 of 7)
There are two quantitative health objectives (QHOs):
Individual early fatality risk (IEFR)
Individual latent cancer fatality risk (ILCFR)
Justification for these values is provided in NUREG-0880, Rev. 1, pp. 30-31.
Focus on ILCFR:
Part 53, Framework B has been developed to provide the same level of safety as currently operating plants.
The State-of-the-Art Reactor Consequence Analysis (SOARCA) studies indicate that IEFR is essentially zero.
5 6
7 5 10 2 10
= conditional latent cancer fatality risk,
, of the bounding event
1 2 10
= expected number of latent cancer fatalities within 10 miles of the site over 50 years following occurrence of the bounding event
= total population within 10 miles of the site CILCFR
38
Derivation of AERI Entry Conditions (3 of 7)
Assume that the plume is confined to one of sixteen 22.5-degree sectors.
= expected number of latent cancer fatalities in the 10-mile, 22.5° sector over 50 years following occurrence of the bounding event 8
9 Assume a uniform population density,.
This assumption eliminates the need to consider the wind direction 1
1 10 22.5° 8
39
Derivation of AERI Entry Conditions (4 of 7) 10 11 Apply the linear no-threshold model, which relates cumulative radiation exposure to fatality risk.
The Commission affirmed the NRCs use of the LNT model in SRM-SECY-19-0008, July 16, 2021.
On a differential basis, the number of latent cancer fatalities is a random variable that is characterized by a binomial probability distribution:
~,
Accordingly, the expected (mean) value is:
= probability that an individual located at distance r dies within 50 years
= differential number of individuals in the 22.5° sector that are located between r and r + dr
= risk coefficient (per rem) 6 10according to BEIR-VII*
= 50-year dose at distance r (rem)
- National Research Council. 2006. Health Risks from Exposure to Low Levels of Ionizing Radiation: BEIR VII Phase 2.
Washington, DC: The National Academies Press.
https://doi.org/10.17226/11340.
40
Derivation of AERI Entry Conditions (5 of 7) 12 13 Assume a power-law dose vs. distance model:
Consistent with NUREG-0396, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants, November 1978.
- 1 16
- 2
4 10
The subscript 0 refers to an arbitrary reference location and dose.
Apply the uniform population density, LNT, and power-law dose vs. distance assumptions.
Integrate over the 10-mile area surrounding the site.
= expected number of latent cancer fatalities in the 10-mile, 22.5° sector over 50 years following occurrence of the bounding event 41
Derivation of AERI Entry Conditions (6 of 7) 14 15 The total population in the 10-mile area is:
Apply the uniform population density assumption.
80 10
5 Scoping consequence model.
Note: decreases as increases.
- 10
205 16
.400
- 10 0.422 Upper bound of the scoping consequence model
.10 400
Criterion for the reference point 42
Derivation of AERI Entry Conditions (7 of 7) 17 Dose (rem TEDE)
Condition 1
First 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 2
Additional dose during the 1st year 0.5 x 49 = 24.5 Additional dose during the second and subsequent years 27.5 TOTAL Note: The reference location is not necessarily the same as the EAB 43