ML22172A091

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06/23-24/2022 ACRS Subcommittee Meeting-10 CFR Part 53 Licensing and Regulation of Advanced Nuclear Reactors: Overview of Framework B Preliminary Proposed Rule Language, Alternative Evaluation for Risk Insights, and Subpart F-Requirements F
ML22172A091
Person / Time
Issue date: 06/23/2022
From: Robert Beall
NRC/NMSS/DREFS/RRPB
To:
Beall, Robert
References
10 CFR Part 53, NRC-2019-0062, RIN 3150-AK31
Download: ML22172A091 (95)
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Text

1 0 C F R Pa r t 5 3 L i c e n s i n g a n d R e g u l a o n o f A d v a n c e d N u c l e a r R e a c t o r s J u n e 2 3 - 2 4, 2 0 2 2 Advisory Committee on Reactor Safeguards (ACRS) Regulatory Rulemaking, Policies and Practices:

Part 53 Subcommittee

A C R S S u b c o m m i t t e e M e e t i n g 6 / 2 3 / 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 Guidance Development Framework Merger Next Steps 3

  • Previously released preliminary proposed rule text (Part 5X) outlined technology-inclusive, risk-informed alternatives for using the traditional technical requirements in Parts 50 and 52
  • Including a traditional, technology-inclusive framework in Part 53 minimizes potential impact on existing requirements and centralizes alternatives for new commercial nuclear reactors
  • Dedicated staff to develop the traditional licensing framework are integrated with existing Part 53 team
  • Aligned with the established Part 53 rulemaking schedule

Background

4

Part 53 Licensing Frameworks Framework A o PRA-led approach o Functional design criteria Framework B o Traditional use of risk insights o Principal design criteria o Includes an Alternative Evaluation for Risk Insights (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

Quantitative Risk Information Bounding Approaches Traditional Use of PRA Risk-Informed Continuum Part 53 Licensing Frameworks 6

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 7

Framework B Development Approach 8

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) 9

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 10

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.

11

Subpart P - Requirements for Operation

  • Maintenance, repair, and inspection programs generally aligned with § 50.65
  • Programs o Security, Emergency Preparedness, Radiation Protection requirements aligned with Framework A o Environmental qualification of electrical equipment derived from § 50.49 o Integrity Assessment Program translated from Framework A, with changes to reflect SSC classification scheme in Framework B o Requirements for procedures and guidelines translated from Framework A o Inservice inspection/inservice testing requirements acknowledge existing requirements for light water reactors (LWRs); non-LWRs must meet Framework A requirements o Containment leak rate requirements from Part 50 (§ 50.54(o))

12

Subpart P - Requirements for Operation Technical Specifications

  • Requirements in § 53.4213 generally aligned with § 50.36
  • Limiting Conditions for Operation (LCOs) o Modifications made to Criteria 1 and 3 to ensure technology-inclusiveness Requirements for detection/indication of reactor coolant pressure boundary degradation maintained for water-cooled reactors Functional containment requirements used to drive LCOs for other technologies o No changes to Criteria 2 and 4 from § 50.36(c)(2)(ii)
  • No material changes to other requirements in § 53.4213 that were translated from § 50.36 (e.g., surveillance requirements, design features, administrative controls, decommissioning) 13

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
  • Provision for performance-based alternatives to detailed requirements with NRC approval o Similar to § 50.48(c)(2)(vii) and § 50.48(c)(4) o Must demonstrate achievement of safe and stable state, safety margin, defense-in-depth
  • No fire PRA required, but may be useful in performance-based justifications 14

Subpart P - Requirements for Operation Fire Protection

  • 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
  • Written for new designs and implementations o Compromises needed to license plants that existed prior to the rule have been removed No allowance for repairs to needed safe shutdown equipment No allowance for operator actions to achieve safe and stable state Some of this flexibility may be available through performance-based alternatives 15

Subpart P - Requirements for Operation Staffing, Training, Personnel Qualifications, and Human Factors

  • Analysis by staff identified that essentially all Framework A, Subpart F provisions for staffing, human factors engineering (HFE), and the licensing of senior reactor operators (SROs) and reactor operators (ROs) were suitable for inclusion in Framework B, Subpart P
  • Importantly, the technical requirements in the areas of Concept of Operations, Functional Requirements Analysis, and Function Allocation provide a means of providing the necessary human-systems design insights needed to facilitate the use of similar flexibilities under either Framework
  • Subpart P adopts most requirements from Subpart F via cross-references or copying requirements with changes; structures of each are similar
  • Features of Subpart P that are different than what was presented under the first iteration of Subpart F will be discussed here, as well as differences between the two subparts in their current versions 16

Subpart P - Requirements for Operation Staffing, Training, Personnel Qualifications, and Human Factors (contd)

  • NEW: the staffing plan required in § 53.4226(f) must include a description of how engineering expertise will be available to support the on-shift operating personnel during all plant conditions
  • Engineering expertise requirement:

o Common to Subparts F and P o For all types of Part 53 plants, regardless of licensed operator category o To assist the on-shift crew with uncertainties - situations not covered by training or procedures o Must have qualifying degree/professional engineering license and familiarity with facility operation o Offers flexibility: could be met by someone filling a traditional shift technical advisor (STA) role or engineering expertise provided from offsite 17

Subpart P - Requirements for Operation Staffing, Training, Personnel Qualifications, and Human Factors (contd)

Several other differences exist between Subpart P provisions and those discussed in the first iteration of Subpart F; these include the changes summarized below that are now common within the current versions of Subparts F and P:

Load Following o Allowance for load following expanded to include process heat use such as for desalination, hydrogen production, and district heating

  • Change Control o Changes to approved programs permitted based on specific criteria
  • Simulator HFE testbed requirements o Removed; to be covered via guidance instead (current status quo) 18

Subpart P - Requirements for Operation Staffing, Training, Personnel Qualifications, and Human Factors (contd)

  • Primary difference between current versions of Subparts P and F stems from their differing provisions for operator licensing
  • Subpart F provisions for alternatives to using licensed ROs and SROs are not currently included in Subpart P
  • Staff continues to evaluate the suitability of including such staffing alternatives in Subpart P; this includes giving consideration to those plants applying the AERI process
  • Provisions for alternatives to using licensed ROs and SROs have been extensively revised within the second iteration of Subpart F and will be discussed in detail during the 6/24 session
  • Other changes to Subpart F that are consistent with requirements discussed for Subpart P today will be summarized during the 6/24 session 19

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 20

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.

21

  • 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 22

  • Technical content of application requirements consolidated in § 53.4730 o Reduces rule length o Minimizes the potential for requirements to diverge between application types Subpart R - Licenses, Certifications, and Approvals

§ 53.4730: General Technical Requirements 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

(37)

X X

X X

X X

X Application Type

§ 53.4730(a)

Requirement 23

(a)(7)

Combustible gas control.

(a)(12)

Post-accident radiation monitoring and protection.

(a)(19)

Organizational structure.

(a)(21)

Pre-operational testing and initial start-up.

(a)(22)

Normal operations and maintenance.

(a)(24)

Fitness-for-duty programs.

(a)(25)

Multi-unit sites.

(a)(26)

Technical qualifications.

(a)(33)

Minimization of contamination.

(a)(37)

Water-cooled reactor requirements.

Many § 53.4730 requirements derived directly from COL technical contents of application in § 52.79 Subpart R - Licenses, Certifications, and Approvals

§ 53.4730: General Technical Requirements 24

(a)(1)

Site safety analysis.

(a)(2)

Facility description.

(a)(4)

Design bases and principal design criteria.

(a)(21)

Pre-operational testing and initial start-up.

(a)(35)

Aircraft impact assessment.

Several § 53.4730 requirements modified from § 52.79 to ensure technology-inclusiveness Subpart R - Licenses, Certifications, and Approvals

§ 53.4730: General Technical Requirements 25

(a)(6)

Fire protection.

(a)(8)

Environmental qualification of electric equipment important to safety.

(a)(9)

Role of personnel.

(a)(10)

Maintenance rule.

(a)(11)

Effluent control.

(a)(15)

Emergency plans.

(a)(23)

Technical specifications.

(a)(27)

Training program.

(a)(28)

Physical security program.

(a)(31)

Radiation protection.

Several Subpart P programs referenced in § 53.4730 Subpart R - Licenses, Certifications, and Approvals

§ 53.4730: General Technical Requirements 26

  • Goal was to provide an equivalent level of safety by developing technology-inclusive analogs to applicable Part 50 and 52 requirements.
  • Leveraged previously developed language from the Part 5X effort.

o Part of the motivation was to provide an approach that better aligned with international regulatory paradigms, as appropriate and consistent with Commission policy.

  • Preliminary Proposed rule language maintains top-level acceptance criteria from Part 50 and 52.

Subpart R - Licenses, Certifications, and Approvals

§ 53.4730(a)(5): Accident Analyses and Initiating Events 27

(i)

Analysis and Evaluation: From § 52.79(a) with modifications to support technology-inclusiveness and Framework B event classifications.

(ii) Design Basis Accidents (DBAs): Technology-inclusive requirements for DBA analyses and SSC classification drawing from §§ 50.34(a)(4) and 50.46.

(iii) AOOs: Consistent with existing requirements including Part 20 acceptance criteria (iv) Beyond Design Basis Events (BDBEs): Technology-inclusive requirements for relevant BDBEs and analysis requirements for other BDBEs, drawn from Anticipated Transients Without Scram (ATWS)/Station Black Out(SBO) rulemakings; similar to international defense-in-depth requirements.

(v) Severe Accidents: Derived from § 52.79(a)(38), with modifications to support technology-inclusiveness (vi) Chemical Hazards: Derived from Framework A to address potential chemical hazards associated with licensed material Subpart R - Licenses, Certifications, and Approvals

§ 53.4730(a)(5): Accident Analyses and Initiating Events 28

Requirements split to acknowledge differences between non-LWR and LWR approaches to containment

  • For non-LWRs, § 53.4730(a)(36)(i) addresses:

o Set of barriers used to meet requirements for AOOs, DBAs, and siting criteria (functional containment) o Safety classification (i.e., safety-related) and qualification of SSCs making up functional containment barriers

o Meets the requirements of Part 50 Appendix J (also addressed in Subpart P) o Addresses any technically relevant requirements from LWR operating experience (containment isolation systems, penetrations, venting/purging)

Subpart R - Licenses, Certifications, and Approvals

§ 53.4730(a)(36): Functional Containment 29

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 alternate 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 30

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 31

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 32

Subpart U - Quality Assurance

  • Subpart U parallels structure and content of Framework A Subpart K
  • 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 33

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 34

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 35

Next Steps Advisory Committee on Reactor Safeguards

  • Full Committee:

July 6 - 9, 2022 Advanced Reactor Public Stakeholder Meeting:

June 30, 2022 Commission Meeting:

July 21, 2022 36

Final Discussion and Questions 37

A C R S S u b c o m m i t t e e M e e t i n g 6 / 2 4 / 2 2 Part 53, Framework B, Subpart R:

Alternative Evaluation for Risk Insights (AERI)

Agenda

  • Proposed AERI Entry Condition - Marty Stutzke
  • Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants - Mihaela Biro
  • Marty Stutzke - Technical Lead of the Graded PRA Working Group, Senior Technical Advisor for Probabilistic Risk Assessment, Division of Advanced Reactors and Non-power Production and Utilization Facilities (DANU), Office of Nuclear Reactor Regulation (NRR)
  • Mihaela Biro - Principal Author of Pre-decisional Draft Regulatory Guide DG-1413, Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants, Senior Reliability and Risk Analyst, Division of Risk Assessment (DRA), NRR
  • Alissa Neuhausen - Principal Author of Pre-decisional Draft Regulatory Guide DG-1414, Alternative Evaluation for Risk Insights (AERI) Framework, Reliability and Risk Analyst, DRA, NRR
  • Katie Wagner - Project Manager of the Graded PRA Working Group, Project Manager, DANU, NRR

Introductions

40

The Graded PRA Working Group Membership Project Manager

  • Marty Stutzke, NRR/DANU Working Group Members
  • Hosung Ahn, on rotation from NRR/Division of Engineering and External Hazard
  • Matt Humberstone, Office of Nuclear Regulatory Research (RES)/DRA
  • Robert Budnitz, consultant Management/Coordination
  • Candace de Messieres, NRR/DANU
  • Steve Lynch, NRR/DANU
  • Nathan Sanfilippo*
  • Former NRC employee 41
  • Evolved from the staffs graded PRA initiative starting in Spring 2021 o Began with intent to grade the technical content of the PRA o Moved to grading 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 (AERI) 42

Uses of Probabilistic Risk Assessment

  • The Policy Statement on the Regulation of Advanced Reactors (73 FR 60612; October 14, 2008) references three PRA-related policy statements:
  • 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)
  • Severe Reactor Accidents Regarding Future Designs and Existing Plants (50 FR 32138; August 8, 1985)
  • The AERI approach and two pre-decisional draft regulatory guides (PDGs) have been developed to:
  • Clarify for potential applicants the logic and the expectations of the NRC staff
  • Provide sufficient risk information to inform licensing decisions
  • Address related ACRS recommendations Search for severe accident vulnerabilities Identify risk insights Meet the QHOs 43
  • October 7, 2019 - Letter concerning review of draft SECY paper, "Population - Related Siting Considerations for Advanced Reactors," ML19277H071:
  • Need to examine new designs with a clean sheet of paper.
  • Think carefully about the failures and combinations of failures that could occur.
  • Must remain vigilant and remember that nature provides surprises.
  • 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.
  • 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:
  • 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.
  • Include guidance that the initial search for initiating events and scenarios should be done without preconceptions or using existing lists.

ACRS Recommendations 44

The Original Approach to Identifying Licensing Events To avoid the very real and very great danger of an accidental release of radioactivity from a reactor, our committee established a simple procedure: We asked the planner of each reactor to imagine the worst possible accident and to design safety apparatus guaranteeing that it could not happen. The committee reviewed each reactor plan, trying to imagine an accident even greater than that conceived by the planner. If we could think of a plausible mishap worse than any discussed by the planner, his analysis of the potential dangers was considered inadequate.

Edward Teller*

First chair of the Atomic Energy Commission (AEC)

Reactor Safeguards Committee (1947-1949)

  • Teller, Edward with Allen Brown, The Legacy of Hiroshima, Double Day & Company, Garden City, NY, 1964.

45

Recognized Problems with the Original Approach to Identifying Licensing Events It is inherently impossible to give an objective definition or specification for "credible accidents" and thus the attempt to identify these for a given reactor entails some sense of futility and frustration, and, further, it is never entirely assured that all potential accidents have been examinedIt should be noted parenthetically, however, that this systematic search for credible accidents often contributes substantially to the safety of a facilityIn the plants finally approved for operation, there are no really credible potential accidents against which safeguards have not been provided to such extent that the calculated consequences to the public would be unacceptable.

Clifford K. Beck, AEC, 1959*

46

47 47 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 condition 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 47

Proposed AERI Entry Condition

§ 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 (EABs) 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 (CILCFR) = 2 x 10-6 per event o

Meet the QHO without developing a PRA to credit accident frequency in the risk estimate The AERI entry condition is not a safety or siting criterion.

48

Derivation of AERI Entry Condition (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.

49

Derivation of AERI Entry Condition (2 of 7)

There are two QHOs:

o Individual early fatality risk (IEFR) within 1 mile of the site o Individual latent cancer fatality risk (ILCFR) within 10 miles of the site The Safety Goal Policy Statement specifies the distances to be used.

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 studies indicate that IEFR is essentially zero for large LWRs.

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

50

Derivation of AERI Entry Condition (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

51

Derivation of AERI Entry Condition (4 of 7) 10 11 Apply the linear no-threshold (LNT) model, which relates cumulative radiation exposure to latent cancer 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.

52

Derivation of AERI Entry Condition (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 53

Derivation of AERI Entry Condition (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 54

Derivation of AERI Entry Condition (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 55

Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants (PDG-1413)

  • Formatted like a regulatory guide; currently a pre-decisional draft regulatory guide
  • Section A: Applies to LWRs and non-LWRs licensed under Parts 50, 52, and 53 (Frameworks A and B)
  • Section B (Discussion):

o Identifies licensing events for each licensing framework o Provides historical perspectives (early licensing, development of the standard review plan (SRP))

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 (Staff Guidance) 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

o Reviews techniques for searching for initiating events and points the user to helpful references o Does not endorse or recommend any specific technique 56

Licensing Event Refers to the Designated Event Categories in 10 CFR Parts 50, 52, and 53 Licensing Basis Designated Licensing Event Categories 10 CFR Parts 50 and 52 Mix of initiating events and event sequences Based on a search of the regulations and associated regulatory guidance:

Design basis events (DBEs) (§ 50.49):

o AOOs o

DBAs o

External events o

Natural phenomena Non-DBA (§ 50.2, alternate ac source)

BDBEs ATWS SBO LMP as presented in NEI 18-04, Rev. 1 and endorsed in RG 1.233 Event sequences PRA used to identify licensing events Licensing events are collectively referred to as licensing basis events (LBEs), which include the following categories:

AOOs DBEs BDBEs DBAs 10 CFR Part 53, Framework A Event sequences PRA used to identify licensing events Licensing events are collectively referred to as LBEs, which include the following categories:

AOOs Unlikely event sequences Very unlikely event sequences DBAs 10 CFR Part 53, Framework B Mix of initiating events and partial event sequences AOOs DBEs DBAs BDBEs 57

Implications of Applicant Decisions on Licensing Event Identification Applicant Decisions Implications Licensing Basis Use of Risk Insights Reactor Design Exemptions Required Risk Evaluation Licensing Event Identificationa 10 CFR Parts 50 or 52 traditional LWR no PRA PDG-1413 non-LWR yes PRA PDG-1413 enhanced LWR or non-LWR yes PRA RG 1.233 Preliminary proposed 10 CFR Part 53 Framework A enhancedb LWR or non-LWR no PRA RG 1.233 Preliminary proposed 10 CFR Part 53 Framework B traditionalb LWR or non-LWR no PRA or AERIc PDG-1413 aLWRs must compare the plant design to the SRP regardless of how the licensing events are identified.

bDictated by the choice of licensing framework.

cVoluntary risk-informed applications require development of a PRA.

58

Staff Perspective The use of a blank sheet of paper approach helps to avoid pitfalls such as, but not limited to:

  • The unwitting or unquestioning carryover of assumptions about plant design or
behavior,
  • The tendency to focus on which predefined events apply (or do not apply) rather than which events are missing from the list,
  • The use of predefined lists that are dated and do not reflect contemporary commercial nuclear plant design or operating experience.

The identification of licensing events, conducted objectively and without preconceptions or over-reliance on predefined lists, helps to ensure that the final list of licensing events is comprehensive and, hence, that the plant design is appropriately analyzed and demonstrated to be safe based on the comprehensive set of licensing events.

59

Techniques for Identifying Initiating Events

  • The staff has reviewed many sources for identifying initiating events:

o NRC NUREGs o NRC and Canadian Nuclear Safety Commission, Joint Report on Terrestrial Energys Methodology for Developing a Postulated Initiating Events List for the Integral Molten Salt Reactor o IAEA-TECDOC-719 and IAEA SSG-3 o International Electrotechnical Commission, International Standard IEC 31010 o American Society of Mechanical Engineers and American Nuclear Society PRA standards o Center for Chemical Process Safety, American Institute of Chemical Engineers o Electric Power Research Institute technical reports o Open literature internet search

  • PDG-1413 does not endorse any specific standard or reference
  • PDG-1413 provides a guide to useful references; it is not intended to be a textbook 60

Examples of Techniques for Identifying Initiating Events Inductive Techniques

  • Double Failure Matrix
  • Failure Mode and Effects Analysis*
  • Failure Mode Effects and Criticality Analysis
  • Fault Hazard Analysis
  • Functional Hazard Analysis
  • Hazard and Operability Analysis*
  • Preliminary Hazard Analysis*

Deductive Techniques

  • Cause Consequence Analysis
  • Common Cause Failure Analysis
  • Fault Tree Analysis*
  • Markov Analysis
  • Master Logic Diagram*
  • Operating and Support Hazard Analysis
  • System Hazard Analysis
  • Most common and well-developed techniques PDG-1413 recommends the use of an inductive and a deductive technique when searching for initiating events to ensure the list of initiating events is comprehensive.

61

Identify application-specific factors (licensing framework, plant-specific design features, and site characteristics).

Conduct a systematic and comprehensive search for initiating events.

Use a systematic process to delineate a comprehensive set of event sequences.

Group initiating events and event sequences into designated licensing event categories according to the selected licensing framework.

Provide assurance that the set of licensing events is sufficient.

Overarching Principles 62

Assemble Multi-disciplinary Team Select Initiating Event Identification Methods

  • Inductive methods
  • Deductive methods
  • Human-induced events (Appendix provides discussion and references)

Define Initiating Event Grouping Strategy and Characteristics Collect information on plant design, plant operating states, and site characteristics 1

3 6

10 9

Identify Plant-specific Safety Functions

  • Systems needed to achieve safety functions
  • Operator actions needed to achieve safety functions
  • Success criteria X

Select the Licensing Framework 2

Collect Application-Specific Information Select Analysis Methods Select Analytical Methods for Event Sequences (e.g., Event Trees, Event Sequence Diagrams)

Define Analysis Scope and Level of Detail (PRA vs. AERI) 7 Define Plant-specific End States for Event Sequences 8

11 Identify Radiological Sources and Transport Barriers from The Source To The Environment 5

Identify Chemical Hazards 4

Technology-Inclusive Identification of Licensing Events (Sheet 1 of 3) to Sheet 2 63

Apply Selected Analytical Methods

  • Identify initiating event impact on safety functions
  • Identify the impact of front-line and support system dependencies on safety functions
  • Identify the impact of operator actions on safety functions Independent Review and QA List of Event Sequences 18 17 Account for Relevant Operating Experience and for Insights from Earlier Analyses of Relevant Designs 16 Independent Review and Quality Assurance (QA)

Account for Relevant Operating Experience and for Insights from Earlier Relevant Analyses List of Initiating Events 14 15 Initiating Event Analysis Event Sequence Selection Apply Initiating Event Identification Methods Apply Initiating Event Grouping Strategy 13 12 X

Y Technology-Inclusive Identification of Licensing Events (Sheet 2 of 3) from Sheet 1 to Sheet 3 64

Is a PRA being developed to support the application?

Define Licensing Event Grouping Strategy and Characteristics

  • Group by frequency o Qualitative o Quantitative
  • Group by type o Plant response following the initiating event (sequence of events, timing) o Similar challenge to safety functions o End state Apply Licensing Event Grouping Strategy Compare to Predefined Lists (e.g., SRP Chapter 15, previous CP, OL, DC, SDA, ML, or COL applications) and identify differences from SRP (only for LWRs)

Identify Limiting Cases (bounding or enveloping scenarios) for each group Independent Review and QA List of Licensing Events 22 23 24 25 26 Follow NEI 18-04, Rev. 1 as endorsed in RG 1.233 21 19 Identify Required Categories of Licensing Events for Licensing Framework Provide initiating events and event sequences to the PRA AERI or traditional use of PRA

  • Part 50 or 52 without LMP
  • Part 53 Framework B Enhanced use of PRA
  • Part 50 or 52 with LMP
  • Part 53 Framework A no (AERI) yes (PRA) 20 Y

Technology-Inclusive Identification of Licensing Events (Sheet 3 of 3) from Sheet 2 65

Alternative Evaluation for Risk Insights (AERI) Framework (PDG-1414)

  • Formatted like a regulatory guide; currently a pre-decisional draft regulatory guide
  • Section A (Introduction): Only applies to LWRs and non-LWRs licensed under Part 53 Framework B
  • Sections B (Discussion) & C (Staff Guidance): Components of the AERI approach:

o Identification and characterization of the bounding event 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 condition 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 66

67 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 AERI entry condition 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 Applicant decision PDG-1413, Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants PDG-1414, Alternative Evaluation for Risk Insights (AERI) Framework LMP guidance - NEI 18-04, Rev. 1, as endorsed in RG 1.233 67

Identification and Characterization of the Bounding Event(s)

Identify and analyze the bounding event AERI Q1 - Demonstrably conservative risk estimate Q3 - Risk Insights Q2 - Severe accident vulnerabilities Continue design and licensing activities O

no AERI entry condition met?

P yes no Q

  • The analysis of the bounding event should be capable of estimating the doses and consequences used in the demonstrably conservative risk estimate that result from evaluating the limiting initiating event for the design, considering credit only for inherent safety features.
  • The process should:

o use the full set of licensing events o consider both core and non-core radiological sources associated with the reactor unit or multiple units

  • The bounding event:

o should be defined by parameters that include source term, meteorology, atmospheric transport, protective actions, dosimetry, health effects, economic factors, and consequence quantification o may combine features of several individual licensing events

  • Multiple bounding events should be considered when, for example, a design has more than one event with approximately similar annual likelihoods of occurrence and with similar overall radiological impacts, but with different radiological characteristics of the analyzed release (isotopic composition, chemistry, timing) 68

The bounding event(s) is used to both:

Risks of acute and long-term radiation exposures

  • Conduct the AERI o Develop a demonstrably conservative risk estimate Uses a consequence estimate based on the bounding event(s) for risks of prompt and latent cancer fatalities) o Search for severe accident vulnerabilities (all licensing events) o Identification of risk insights (all licensing events) o Assessment of defense-in-depth Identify and analyze the bounding event AERI Q1 - Demonstrably conservative risk estimate Q3 - Risk Insights Q2 - Severe accident vulnerabilities Continue design and licensing activities O

no AERI entry condition met?

P yes no Q

69

Confirmation that the AERI Entry Condition is Met

  • When electing to use the AERI approach, it may not be known at first whether the entry condition is met
  • Confirmation is based on a realistic dose estimate and realistic description of uncertainties (preferred)
  • Conservatisms are identified and addressed
  • For multiple bounding events, the entry condition should be met for each bounding event Identify and analyze the bounding event AERI Q1 - Demonstrably conservative risk estimate Q3 - Risk Insights Q2 - Severe accident vulnerabilities Continue design and licensing activities O

no AERI entry condition met?

P yes no Q

70

Development of a Demonstrably Conservative Risk Estimate

  • Demonstration that a reactor design meets the QHOs
  • Risks of prompt radiation-caused fatalities and latent cancer fatalities to offsite populations
  • No realistic estimate of annual frequency expected
  • Assumed frequency of 1/yr, represents the sum of event sequence frequencies and equal to the sum of initiating event frequencies; based on LWR statistics
  • Applicant may use a different frequency with justification and NRC staff will review on a case-by-case basis 71 71 71 Identify and analyze the bounding event AERI Q1 - Demonstrably conservative risk estimate Q3 - Risk Insights Q2 - Severe accident vulnerabilities Continue design and licensing activities O

no AERI entry condition met?

P yes no Q

71

Search for Severe Accident Vulnerabilities

  • Encompass entire set of licensing events and any additional severe accidents
  • Severe accident vulnerabilities should be eliminated through modifications to the design, operations, or maintenance
  • Justification may be provided for why a severe accident vulnerability is acceptable for the design Identify and analyze the bounding event AERI Q1 - Demonstrably conservative risk estimate Q3 - Risk Insights Q2 - Severe accident vulnerabilities Continue design and licensing activities O

no AERI entry condition met?

P yes no Q

72

Definition of Severe Accidents and Severe Accident Vulnerabilities

  • Severe accidents defined in 10 CFR 53.4730(a)(5)(v)(B) are those events that progress beyond the DBAs, in which substantial damage is done to the reactor core whether or not there are serious offsite consequences
  • In PDG-1414, severe accident vulnerabilities are those aspects of a facility design that represent an overreliance on a single design feature, whether for accident prevention or mitigation, and that could lead to a severe accident after accounting for SSC reliability, human actions, and defense-in-depth Identify and analyze the bounding event AERI Q1 - Demonstrably conservative risk estimate Q3 - Risk Insights Q2 - Severe accident vulnerabilities Continue design and licensing activities O

no AERI entry condition met?

P yes no Q

73

Identification of Risk Insights

  • Based on entire set of licensing events
  • Qualitative descriptions should be provided
  • Quantitative descriptions should be provided when available
  • Risk insights should provide an understanding of the hierarchy of event sequences ranked by frequency
  • Risk insights identified using AERI may be used to support other licensing decisions Identify and analyze the bounding event AERI Q1 - Demonstrably conservative risk estimate Q3 - Risk Insights Q2 - Severe accident vulnerabilities Continue design and licensing activities O

no AERI entry condition met?

P yes no Q

74

Assessment of Defense-in-Depth

  • Encompasses entire set of licensing events
  • Regulatory guidance position is adapted from RG 1.174
  • NEI 18-04 may also provide guidance for the AERI framework related to assessing defense-in-depth (adapted from IAEA SSR 2/1)
  • Assessment of defense-in-depth complements the search for severe accident vulnerabilities and search for risk insights 75

Path Forward PDG-1413 & PDG-1414

  • Make revisions in response to ACRS and stakeholder feedback
  • Monitor changes to preliminary proposed rule text PDG-1414
  • Develop guidance for AERI maintenance and upgrades 76

Upcoming Meetings Periodic Advanced Reactor Stakeholder Public Meeting:

June 30, 2022 Advisory Committee on Reactor Safeguards Full Committee Meeting:

July 6, 2022 Commission Meeting - Update on 10 CFR Part 53 Licensing and Regulation of Advanced Nuclear Reactors:

July 21, 2022 77

Discussion 78

A C R S S u b c o m m i t t e e M e e t i n g 6 / 2 4 / 2 2 10 CFR Part 53 Subpart F Staffing, Personnel Qualifications, Training, and Human Factors (2nd Iteration)

Agenda Overview of Key Changes to Subpart F under the 2nd Iteration o Addition of Engineering Expertise Requirement o Expansion of Load Following Allowances o Removal of Simulator HFE Testbed Requirement o Replacement of Certified Operator Framework Generally Licensed Reactor Operators Questions 80

Theresa Buchanan, Senior Reactor Engineer (Examiner)

Dr. David Desaulniers, Senior Technical Advisor for Human Factors and Human Performance Evaluation Dr. Brian Green, Senior Human Factors Engineer -Team Lead Dr. Niav Hughes Green, Human Factors Psychologist Dr. Stephanie Morrow, Human Factors Psychologist Lauren Nist, Branch Chief, Operator Licensing and Human Factors Branch Maurin Scheetz, Reactor Engineer (Examiner)

Jesse Seymour, Reactor Operations Engineer (Human Factors)

Overview of Primary Staff Contributors (NRR & RES) 81

Key Changes to Subpart F under the 2nd Iteration 2nd iteration of Subpart F retains majority of requirements developed for the 1st iteration Some requirements have been relocated to more appropriate spots (e.g., grouping technical requirements)

Most changes made for the second iteration of Subpart F were mirrored in the contents of Subpart P; for that reason, most significant differences from the first iteration of Subpart F were discussed in detail during the Framework B presentation A summary of those major changes will be provided here 82

  • Addition of engineering expertise requirement o Staffing plan requirements of § 53.730(f) modified to include providing engineering expertise to operators
  • Expansion of load following allowances o §§ 53.725(b) and 53.740(e) - (f) requirements modified to expand load following to include process heat usage
  • Removal of simulator HFE testbed requirement
  • Specific change management for approved programs
  • Certified operator provisions completely replaced with an all new generally licensed reactor operator framework Overview of Key Changes (continued) 83

Generally Licensed Reactor Operators (GLROs)

Under 2nd iteration of subpart F operator licenses now consist of general licenses and specific licenses.

A specific license is issued to a named person and is effective upon approval by the Commission of an application filed pursuant to the regulations in this part and issuance of licensing documents to the applicant. Specific licenses are issued to ROs and SROs.

A general license is effective without the filing of an application with the Commission or the issuance of licensing documents to a particular person. The general licensing of GLROs is addressed by the requirements of §§ 53.800 through 53.830.

84

  • What types of facilities would have GLROs?

o No operator action is needed to mitigate plant events and achieve acceptable accident performance o Defense-in-depth independent of operator action o Operators not significant factor in safety outcomes

  • What role would GLROs fulfil?

o Administrative functions historically done by an SRO; keeps facility in analyzed state within licensing basis o Conduct manual reactivity manipulations if needed o Supervise core alterations and refueling operations Generally Licensed Reactor Operators (continued) 85

  • Only a single operator license level exists within the GLRO framework (analogous to the SRO level)
  • Plants meeting the criteria for using GLROs would have to use GLROs in lieu of ROs and SROs for staffing
  • Like ROs and SROs, the GLRO training program must also be derived from a systems approach to training (SAT)
  • Prescriptive staffing and capabilities for GLROs:

o Continuous monitoring with continuity of responsibility o Monitor plant parameters, evaluate emergency conditions, initiate reactor shutdown, dispatch/direct ops & maintenance personnel, and implement emergency plan Generally Licensed Reactor Operators (continued) 86

§ 53.800 defines new class of plants using the design criteria previously developed for certified operator use o

Establishing new class done to conform with Atomic Energy Act

§ 53.805 establishes the responsibilities of facility licensees that use GLROs:

o Maintain GLROs qualifications for responsibilities o

Only GLROs may manipulate facility controls o

Develop/implement/maintain Commission approved programs for GLRO training, exams, & proficiency o

Ensure GLROs meet Part 26 & 73 requirements o

Report names of all GLROs to the NRC annually

§ 53.810 is the general license; it is granted provided that qualifications are established and maintained subject to restrictions. GLROs are subject to enforcement action.

Generally Licensed Reactor Operators (continued) 87

§ 53.815 covers GLRO training, retraining, and proficiency provisions.

o Training programs must be derived from SAT

Includes performing reactivity manipulations o Initial examination on knowledge and abilities o Continuing training and requalification exams o Requirements for use of simulation facilities o Records must be available for NRC inspection o Must establish a GLRO proficiency program o No specific medical requirements for GLROs

§ 53.830 covers expiration of the license for GLROs Generally Licensed Reactor Operators (continued) 88

Comparison of key aspects of the SRO/RO and GLRO frameworks:

Framework Aspect RO / SRO GLRO Licensed operators with responsibility for administrative requirements Yes Yes Licensed operators with role in event mitigation Yes No NRC has legal authority to suspend or revoke license for violations Yes Yes NRC approval of training & exam programs required?

Yes Yes NRC approval needed for exams, medical, simulator, renewals, terminations, and waivers?

Yes No Flexibility for requalification training & exam periodicity?

No Yes Specific Licensing (RO/SRO) versus General Licensing (GLRO) 89

1. STA Elimination - addressed via the new engineering expertise requirement
2. Use of non-licensed, certified operators - entirely replaced by a new licensed operator framework (GLRO)
3. Limited scope simulators - to be addressed via guidance and usage of existing standards
4. Training program reviews - new SAT program review guidance under development
5. Staffing plan provisions for first-of-a-kind builds, senior license holders, and organizational interfaces - to be addressed via guidance Summary of how ACRS letter concerns were addressed 90

Discussion 91

Final Discussion and Questions 92

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 93

ACRS Advisory Committee on Reactor Safeguards AEC Atomic Energy Commission AERI Alternative evaluation for risk insights AOO Anticipated operational occurrence ATWS Anticipated transient without scram BDBE Beyond design basis event BE Bounding event CFR Code of Federal Regulations CILCFR Conditional individual latent cancer fatality risk COL Combined license CP Construction permit DANU Division of Advanced Reactors and Non-Power Production and Utilization Facilities Acronyms EAB Exclusion area boundary DBA Design basis accident DBE Design basis event DC Design certification DG Draft regulatory guide DRA Division of Risk Assessment ESP Early site permit FR Federal Register GLRO Generally licensed reactor operator HFE Human factors engineering IAEA International Atomic Energy Agency IEFR Individual early fatality risk ILCFR Individual latent cancer fatality risk LBE Licensing basis event 94

LCO Limiting condition for operation LMP Licensing Modernization Project LNT Linear no-threshold LWR Light water reactor ML Manufacturing license NEI Nuclear Energy Institute NFPA National Fire Protection Association NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NUREG U.S. Nuclear Regulatory Commission technical report designation OL Operating license PDG Pre-decisional draft regulatory guide PRA Probabilistic risk assessment QA Quality assurance Acronyms RO Reactor operator QHO Quantitative health objective RES Office of Nuclear Regulatory Research RG Regulatory guide SAT Systems approach to training SBO Station black out SDA Standard design approval SRO Senior reactor operator SRP Standard review plan SSCs Structures, systems, and components STA Shift technical advisor TEDE Total effective dose equivalent TIRIMA Technology-inclusive, risk-informed maximum accident 95

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