Text

Presentation Slides - Periodic Advanced Reactor Stakeholder Meeting 06072023
	ML23157A018
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Issue date:	06/07/2023
From:	Katie Wagner; NRC/NRR/DANU/UARP
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Advanced Reactor Stakeholder Public Meeting June 7, 2023 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 856 640 616#

Time Agenda Speaker 10:00 am - 10:10 am Opening Remarks / Advanced Reactor Integrated Schedule NRC 10:10 am - 11:40 am Guidance for Reviewing Facility Training Programs NRC 11:40 am - 12:45 pm Lunch Break All 12:45 pm - 1:15 pm Regulatory Treatment of Potential High Temperature Fluid Releases in Advanced Reactor Designs Argonne National Laboratory 1:15 pm - 1:45 pm Regulatory Treatment of Non-Core Sources of Radioactivity Associated with Advanced Reactor Designs Argonne National Laboratory 1:45 pm - 2:00 pm Break All 2

Time Agenda (continued)

Speaker 2:00 pm - 2:30 pm Electronic Submittal of Advanced Reactor Applications NRC 2:30 pm - 3:00 pm Overview of the Advanced Reactor Construction Oversight Program (ARCOP)

Recently Issued SECY Paper NRC 3:00 pm - 3:15 pm Break All 3:15 pm - 4:35 pm Advanced Reactor Content of Application Project (ARCAP)/Technology Inclusive Content of Application Project (TICAP) Guidance Documents NRC 4:35 pm - 4:40 pm Future Meeting Planning and Concluding Remarks NRC 3

Advanced Reactor Integrated Schedule of Activities The updated Advanced Reactor Integrated Schedule is publicly available on NRC Advanced Reactors website at:

https://www.nrc.gov/reactors/new-reactors/advanced/integrated-review-schedule.html 4

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced/integrated-review-schedule.html 5

White Paper on Facility Training Programs Draft Review Guidance Jeff Correll NRR/DRO/IOLB June 7, 2023 6

Facility Training Program Guidance

This staff white paper has been prepared and is being released to support ongoing public discussions. This white paper uses a draft interim staff guidance (ISG) format because the staff is considering using this format to provide staff guidance in the near future to support the review of advanced reactor applications.

This paper has not been subject to NRC management and legal reviews and approvals, and its contents are subject to change and should not be interpreted as official agency positions.

The paper is publicly available at ADAMS ML23017A130 7

Facility Training Program Guidance

This white paper is intended to support both applications under the proposed Part 53 as well as near-term applications under Parts 50 and 52.

The guidance supports the NRC staff review of the portion of an application associated with the training program for plant personnel, including licensed operator initial and requalification training programs.

This guidance also facilitates the review of non-accredited training programs at commercial nuclear plants. This guidance may also be used to support training program inspection needs as currently specified in NUREG-1220.

This guidance covers:

- The 5 phases of the systems approach to training (SAT)

- Scope of facility training programs 8

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Background===

10 CFR Part 53 is currently with the Commission for review pending issuance as a proposed rule for public comment

- Guidance in this ISG is subject to change based on rulemaking

Key documents for Part 53 rulemaking can be found at Regulations.gov under Docket ID NRC-2019-0062 9

Goals

Establish reliable guidelines for training program developments based on current best practices from research and expertise on the Systems Approach to Training (SAT) Process 10

ISG Layout

Section A defines the five phases of SAT

- Evaluation criteria are provided for initial training program approval, and for ongoing training program inspections.

Section B outlines basic Training Program Guidance

- Defines the basic requirements that the staff would expect to see in a training program guide.

11

Section 1.0 Analysis Phase - Overview

Defines the three methods of Analysis:

Section 1.1 - Needs Analysis Section 1.2 - Job Analysis Section 1.3 - Task Analysis 12

Section 1.1 Analysis Phase - Needs Analysis 1.1 Conducting Training Needs Analysis 1.1.1 Needs Analysis is a process that includes training and line personnel.

1.1.2 Needs Analysis process is used to analyze internal and external factors.

.1 Initial Training Programs

.2 Existing Training Programs

.3 Needs Analysis process utilizes Job and Task Analysis process when applicable.

13

Section 1.1 Analysis Phase - Needs Analysis 1.1.3 Changes to the Task List and associated KSAs

.1 Changes to non-Commission approved training programs.

.2 Changes to Commission approved training programs

.3 Changes to the objectives and lesson plan material does not always require changes to the Task and KSA list 14

Section 1.1 Analysis Phase - Needs Analysis 1.1.4 Needs Analysis process maintains the initial and continuing training programs 1.1.5 Needs Analysis Process includes analyzing performance gaps 1.1.6 Training Exemptions are analyzed and documented 1.1.7 Needs Analysis Documentation 15

Section 1.2 Analysis Phase - Job Analysis 1.2 Conducting Job Analysis 1.2.1 Job Analysis is a process that includes training and line personnel 1.2.2 Job Analysis process groups tasks into Position/Role/Duty Areas 1.2.3 Job Analysis process produces a task list 1.2.3.1 Initial Training Job Analysis 1.2.3.1.1 Initial Job Analysis Considerations 1.2.3.2 Existing Training Job Analysis 16

Section 1.2 Analysis Phase - Job Analysis 1.2.4 Tasks are systematically selected for training 1.2.4.1 Licensed Operator Training includes items important to safe plant operation 1.2.4.2 Licensed Operator Retrain periodicity 1.2.5 Job Analysis Documentation 17

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Section 1.3 Analysis Phase - Task Analysis 1.3 Conducting Task Analysis 1.3.1 Task Analysis is an iterative process that includes training and line personnel 1.3.2 Task Analysis produces task characteristics for further training development 1.3.2.1 Operator Licensing Programs produce a comprehensive KSA list for Commission approval 1.3.3 Task Analysis Documentation 20

Licensed Operator Examinations (DRO-ISG-2023-01)

Job Analysis Task List Created Tasks are Selected for Training Licensed Operator Training includes Items important to safety Task Analysis Approved KSA List Develop Learning Objectives Develop Evaluation Items ANALYSIS DESIGN Licensed Operator KSA ranking process SAT Based Training Program:

Development

Implementation

Evaluation 21 SAT and Licensed Operator Examinations

Section 2.0 Design Phase - Overview

Defines:

- Target student population

- Objectives

- Evaluation instruments

- Instructional settings 22

Sections 2.1-2.2 Design Phase -Students and Learning Objectives 2.1 Define Target Student Population 2.2 Develop Learning Objectives (LO) 2.2.1 LOs contain Conditions, an Action, and Standards 2.2.2 LOs focus on desired results the trainee is expected to achieve 2.2.3 Lesson plans includes Terminal Objectives 2.2.4 Lesson Plans include enabling objectives to support the terminal objective goal.

2.2.5 Enabling objectives are organized to facilitate student learning 2.2.6 Performance Objectives maximize the use of performance opportunity 2.2.7 Learning Objectives are reviewed and approved by Training and Line Supervision 23

Section 2.3 Design Phase - Evaluation Items 2.3 Develop Evaluation Items 1.Evaluation items evaluate the topic of the objective 2.Evaluation items are leveled to the objective 3.Test item conditions and standards match the learning objectives conditions and standards 4.Test item construction is appropriate method of evaluation for the objective 5.Pass/fail criteria 6.Evaluation items must be plausible 7.Performance Objectives written for individual trainee evaluation 8.Test item creation includes review and approval by Training and Line Supervision 24

Section 3.0 Development Phase - Overview

Defines the following:

Section 3.1 - Training Material Development Section 3.2 - Exam Development 25

Section 3.1 Development Phase - Training Material 3.1.1 Training Material Development Standards:

.1 Training material development is rooted in the plants SAT Analysis

.2 Training Material Content and Consistency

.3 Method of delivery ensures effective objective mastery.

3.1.2 Training Material Content:

.1 Lesson Plan Content 3.1.3 Training Material review, approval, and accuracy:

3.1.4 Curriculum Organization

.1 Delivery Timeframe

.2 Curriculum Sequencing 26

Section 3.2 Development Phase - Exams 3.2 Exam Development Standards 3.2.1 Cognitive Evaluations

.1 Objective Sampling

.2 Multiple exams are created with >40% differing questions.

.3 An exam question selection process exists.

.4 Clear pass/fail standards exist.

.5 Clear grading methods exist.

3.2.2 Performance Evaluations

.1 Individual performance and evaluation of performance objectives

.2 Clear pass/fail standards to allow consistent evaluation

.3 Guidance provides reproducible consistency between evaluators 3.2.3 Evaluation Item review, validation, and approval 27

Section 4.0 Implementation Phase - Overview

Defines the following:

Section 4.1 - Preparation and Scheduling Section 4.2 - Delivery of Training Section 4.3 - Exam Administration and Remediation Section 4.4 - Post Training Activities 28

Sections 4.1-4.2 Implementation Phase - Preparation and Delivery 4.1 Preparation and Scheduling:

4.1.1 Fixed vs Flexible Scheduling 4.1.2 Schedules approved by training and line 4.2 Delivery of Training:

4.2.1 Instructors are trained and qualified 4.2.2 Deliver effective training 29

Sections 4.3-4.4 Implementation Phase - Exams and Follow-up 4.3 Exam Administration and Remediation:

4.3.1 Exam Administration Standards - all formal training requires evaluation 4.3.2 Exam security standards 4.3.3 Exam administration 4.3.4 Exam process to include test review with trainees 4.3.5 Exam Remediation process 4.3.6 Exam Remediation standards 4.4 Post Training Activities:

4.4.1 Student feedback solicited post training for evaluation 4.4.2 Document the training occurrence

.1 Update training records and qualifications 30

Section 5.0 Evaluation Phase - Overview

Defines the following:

Section 5.1 - Evaluation Intake Section 5.2 - Assess Information Section 5.3 - Initiate Corrective Actions Section 5.4 - Conclusion 31

Section 5.1 Evaluation Phase - Intake 5.1. Evaluation Intake 5.1.1 Collect and Analyze Incumbent and Management Feedback

.1 Training Feedback Analysis

.2 Management Observations of Training

.3 Exam Item Analysis

.4 Post Training Performance Review 5.1.2 Facility Issues and Events 5.1.3 Inspection/Assessment/Evaluation reports 5.1.4 Facility modifications and procedure changes 5.1.5 Industry Regulatory and Operating Experience 32

Sections 5.2 - 5.4 Evaluation Phase - Assess and Initiate Actions 5.2 Assess Information 5.2.1 Assessing the approved training program effectiveness 5.2.2 Assessing the approved training program scope 5.3 Initiate Corrective Actions 5.3.1 Appropriate Actions are taken to improve the training program

.1 Actions that initiate Training Needs Analysis

.2 Actions that do not initiate a Training Needs Analysis 5.3.2 Performance Gaps produce Training Effectiveness Metrics 5.3.3 Training Evaluation documentation and Approval 5.4 Conclusion 33

Section B Facility Training Programs

Includes guidance for the following sections:

Section 1 - Program Description Section 2 - Program Eligibility Section 3 - Initial Training Programs Section 4 - Requalification Programs 34

Section 1 Program Description 1.1 General Requirements:

a. The purpose of the program

b. Job positions credited towards each role, as defined by the job and task analysis

c. Training organization teaching the course or supervising instruction of the course material.

d. The qualification requirements of the training staff personnel.

e. The course curriculum 35

Section 1 Program Description 1.2 Licensed Operator Programs:

a. The course curriculum and scheduling for each course required to achieve a license (RO and SRO), as identified in the SAT analysis.

b. A chart showing the proposed schedule for licensing personnel prior to criticality. The schedule should be relative to the expected fuel load date and should also display the preoperational test period.

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Section 2 Program Eligibility 2.1 General Requirements:

2.2 Licensed Operator Requirements:

2.2.1 - Procedures:

2.2.2 - Educational and Experience Requirements 37

Section 3 Initial Training Programs 3.1 General Requirements 3.2 Licensed Operator Requirements:

3.2.1 - Foundational theory of plant operations are included in the task list and KSA development for training program design 3.2.2 - Included in the timeline of the training program design should include:

- classroom training

- hands on training (OJT/TPE, simulator, or equivalent)

- proficiency training (under instruction watches)

- program exams 38

Section 4 Requalification Programs 4.1 General Requirements:

a.

Task list requiring retraining b.

Retraining schedule according to retrain periodicity requirements c.

Scope of required training 4.2 Licensed Operator Requirements:

4.2.1 Requal program must include training for performance and cognitive based tasks as identified in the job and task analysis for tasks selected for retrain.

4.2.2 Licensed Operator requalification training shall include the following:

A retrain periodicity not to exceed 24 months for specifically licensed operator training programs A process for review and maintenance of the program 39

Questions 40

Lunch Break Meeting will resume at 12:45 pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 856 640 616#

Advanced Reactor Stakeholder Public Meeting

REGULATORY CONSIDERATIONS REGARDING POTENTIAL HIGH-TEMPERATURE FLUID RELEASES Dave Grabaskas Manager, Licensing and Risk Assessments Group, Argonne National Laboratory Ben Chen, Matthew Bucknor, Mark Cunningham Argonne National Laboratory David Holcomb Oak Ridge/Idaho National Laboratory Richard Denning Consultant

DOEs Advanced Reactor Demonstration Program Demonstration 1 (X-Energy)

Demonstration 2 (Natrium)

Risk Reduction for Future Demonstration National Reactor Innovation Center Regulatory Development Advanced Reactor Safeguards Key Industry Interfaces and Inputs Includes:

Nuclear Energy Institute Industry Technology Working Groups Electric Power Research Institute Insights from other program National Technical Directors & DOE Federal Managers Participation in NRC public meeting interactions (incl. NRC Integrated Schedule) 43 DOE REGULATORY DEVELOPMENT EFFORTS Four elements within Regulatory Development:

MSR Regulatory Development R&D FR Regulatory Development R&D GCR Regulatory Development R&D Regulatory Framework Modernization

44 CURRENT PROGRAM EFFORTS IN THIS AREA Emergency Planning Under the LMP Approach Regulatory Treatment of Low Frequency External Events Regulatory Treatment of High Temperature Fluid Releases Regulatory Treatment of Non-Core Sources of Radioactivity Assessment of Fast Reactor Consensus Safety Standards Reviewed in current meeting Reviewed in upcoming NRC meetings Ongoing effort

45 REGULATORY CONSIDERATIONS REGARDING POTENTIAL HIGH-TEMPERATURE FLUID RELEASES

Motivation

New internal hazards The potential consequences associated with liquid metals, molten salt, and high-temperature gas releases require design and licensing consideration.

Two aspects:

o Assessment of hazard as part of licensing o Response if an event were to occur during operation Objectives Examine the applicability (or non-applicability) of new and existing regulatory requirements and guidance regarding internal fire and internal flood.

Identify key regulatory considerations during licensing and operation regarding high-temperature fluid release events to ensure an appropriate regulatory treatment.

46 HIGH-TEMPERATURE FLUID RELEASES

Project Report A report was developed that reviews potential high-temperature fluid releases for advanced reactor designs, including the following factors:

o Phenomena o Prevention/mitigation features o Past regulatory experience o Applicability of existing internal hazard guidance o Design criteria considerations Report uses:

o Reference document for regulatory interactions o Educational material for regulator and industry staff new to advanced reactor designs o Collection of key reference material 47 HIGH-TEMPERATURE FLUID RELEASES ANL/NSE-22/35 OSTI:1972285

Phenomena Liquid sodium at operating temperatures of an SFR will burn when exposed to air and react energetically when in contact with water.

Major Considerations:

1) Preservation of sodium inventory to prevent core uncovery and ensure a heat removal pathway.

2) Prevention and mitigation of the consequences of a sodium release, such as the impact on structure, system, and component (SSC) functionality, etc.

48 EXAMPLE: SODIUM

Sodium Release Prevention/Mitigation:

Objectives:

1)

Protect important SSCs from the resultant pressure, heat, aerosol, and chemical effects of a sodium release 2)

Prevent the release of retained radionuclides (design dependent) 3)

Protect workers and the public from radiological and chemical effects Strategies:

1)

Careful consideration of sodium piping routing, SSC co-location, room sizing, etc.

2)

Sodium leak prevention (double-wall piping, guard piping, leak detection, etc.)

3)

Sodium leak mitigation (catch pans, drain tanks, compartment inerting, concrete insulation, etc.)

49 EXAMPLE: SODIUM

Past Regulatory Experience:

Review of past regulatory interactions:

Review of the evolution of design strategies for the prevention and mitigation of sodium releases that occurred in response to lessons learned from operating experience and regulatory interactions 50 EXAMPLE: SODIUM

Lessons Learned (example):

The sodium fire mitigation strategy for the Fast Flux Test Facility (FFTF) utilized an active nitrogen injection system to isolate and deprive sodium fires of oxygen Although approved during reactor authorization, difficulties were encountered with the system in practical application Subsequent SFR designs, such as the Clinch River Breeder Reactor, utilized alternative strategies:

o Transition to completely passive systems o Novel catch pan and suppression deck designs 51 EXAMPLE: SODIUM Clinch River Breeder Reactor:

Cascading Catch Pan Design

Applicability of Current Guidance:

Many similarities with the deterministic and risk-informed internal fire protection strategies associated with Appendix R and NFPA 805.

Applicability of leak-before-break (LBB) methodology?

Applicability of guidance regarding hazardous substance releases.

Design Criteria Considerations:

Reviewed the SFR-DCs in RG 1.232 and highlighted criteria that have relevancy to potential sodium releases and included specific considerations:

o SFR-DCs 3 and 73: Fire protection and sodium leaks o SFR-DCs 14,15,30-33,71,78,79: Primary coolant boundary integrity o SFR-DCs 75-77: Intermediate coolant boundary integrity o SFR-DCs 23,34,35: Preservation of safety functions 52 EXAMPLE: SODIUM

Open Issues:

Applicability of LBB method:

o Previous SFR applications have utilized the LBB method to limit the size of potential sodium releases o These approaches differ from the current application of the LBB method for LWRs, as currently defined in SRP 3.6.3 Withdrawal of ANS 54.8: Liquid Metal Fire Protection in LMR Plants o Published in 1988 o Withdrawn in 2000 o Previous SFR applicants referred to the standard in the regulatory submittals o Parallel program effort currently in progress to further explore and identify necessary safety standard development 53 EXAMPLE: SODIUM

Other Fluid Types:

Molten salt release (both fuel-salt and coolant-salts)

High-temperature gases 54 HIGH-TEMPERATURE FLUID RELEASES MSRE Reactor Cell Layout MSRE Final Shutdown Valve Leak MHTGR Pressure Mitigation Pathway

55 QUESTIONS

REGULATORY TREATMENT OF NON-CORE SOURCES OF RADIOACTIVITY Dave Grabaskas Manager, Licensing and Risk Assessments Group, Argonne National Laboratory Ben Chen Argonne National Laboratory Scott Ferrara, Jason Christensen, Jason Andrus Idaho National Laboratory

Motivation The ASME/ANS Non-LWR PRA Standard permits the inclusion of any source of radioactivity within the plant PRA.

Using a risk-informed performance-based (RIPB) licensing strategy for non-core sources may have advantages:

o Uniform, consistent methodology for the entire plant o Application simplification o Flexibility in licensing decision-making o Use of risk information as part of plant oversight For certain advanced reactor designs, the distinction between core and non-core is less straightforward.

Certain advanced reactor designs may include onsite facilities with characteristics close to fuel cycle facilities.

57 NON-CORE SOURCES Examples:

Spent fuel storage Purification systems Fuel processing systems Fuel movement

Objectives:

Identify the potential non-core sources of radioactivity for advanced reactors designs.

Identify the category of regulated material for each non-core source of radioactivity and the associated regulatory requirements and guidance.

Compare the RIPB treatment of non-core sources of radioactivity to the current regulatory requirements and associated guidance.

Identify potential gaps/discrepancies between current regulation & guidance associated with the licensing of non-core sources of radioactivity and a RIPB approach.

Provide recommendations regarding avenues to address or resolve identified gaps or discrepancies.

58 NON-CORE SOURCES

Scope:

A first step, focusing on regulatory criteria and associated high-level guidance Focus is on advanced reactor designs with non-core sources and monolithic advanced reactor plant sites that may include fuel facilities or similar Study utilized Licensing Modernization Project (LMP) and the Technology Inclusive Content of Applications (TICAP) as the RIPB approach for comparison, although other RIPB approaches are possible, potentially including simplified risk-informed approaches Study did not examine requirements associated with safeguards, security, offsite transportation, and final disposable of radioactive material Study did not explore the feasibility of risk-informing the treatment of non-core sources in terms of adequacy of PRA technology (i.e., are analysis methods and supporting data available)

Study focused on using RIPB approaches to satisfy existing regulatory requirements for non-core sources and not to increase the expectations regarding the fidelity of supporting analyses for such sources 59 NON-CORE SOURCES

Regulatory Requirements Review:

Reviewed the safety-relevant portions of the following:

o Byproduct Material: Parts 30-35 o Source Material: Part 40 o Special Nuclear Material: Part 70 o Interim storage: Part 72 Reviewed associated guidance:

o NUREG-1557: Consolidated Guidance about Material Licenses o NUREG-1513: Integrated Safety Analysis Guidance o NUREG-1520: Standard Review Plan for Fuel Cycle Facilities o NUREG-2215: Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities o RG 3.67: Standard Format and Content for Emergency Plans for Fuel Cycle and Material Facilities o Others 61 NON-CORE SOURCES

Analysis Areas:

Assessment focused on main areas associated with LMP and TICAP 62 NON-CORE SOURCES Event Classification and Criteria SSC Classification Emergency Planning Content of Applications

Event Classification and Criteria 10 CFR Part 20 applicable to all.

Only Part 70 and 72 contain dose requirements for off-normal or accident scenarios.

o Recommended event classification available in guidance, but up to applicant to propose for NRC approval.

No QHOs for regulated material, although NRC-proposed Quantitative Health Guidelines (QHGs) are similar.

Part 70 definition of credible defers from LMP.

o LMP is more conservative Inclusion of worker dose in Part 70.

Inclusion of chemical hazard considerations.

63 FINDINGS AND RECOMMENDATIONS

Event Classification and Criteria

1) Preemptively seek NRC approval of LMP event classification for Part 70 and 72 event classification requirements.

2) Explore supplemental worker dose and chemical hazard considerations under the LMP framework.

o The application of the LMP approach for VTR included these factors using supplemental criteria. The strategy was approved by the authorization body (DOE).

64 FINDINGS AND RECOMMENDATIONS VTR Collocated Worker F-C Curve Part of VTR SSC Classification Requirements

SSC Classification Only Parts 70 and 72 have SSC classification requirements:

o Part 70: Items relied on for safety (IROFS) o Part 72: SSCs important to safety The LMP approach for SSC identification and establishment of design basis likely sufficient to meet these requirements.

Further detailed assessment needed to clarify expectations regarding special treatments for SSCs.

o The potential SSC special treatments outlined in NEI 18-04 likely beyond what is required under Part 70 and 72.

65 FINDINGS AND RECOMMENDATIONS

Content of Applications The LMP approach and TICAP guidance likely provide the necessary information to fulfill safety-relevant application requirements under Parts 30, 40, 70, and 72.

Part 70 requires an Integrated Safety Assessment (ISA).

o The ISA is an integrated analysis regarding potential radiological and chemical hazards, including possible accident sequences and consequences.

Only Part 72 requires a formal SAR.

A crosswalk of the required Part 70 ISA and Part 72 SAR content with TICAP guidance would be useful for expediting future licensing efforts.

66 FINDINGS AND RECOMMENDATIONS

Emergency Plan Parts 30, 40, 70 and 72 have nearly identical requirements regarding emergency plans LMP approach can provide the information necessary to fulfill most requirements In general, additional guidance needed regarding the use of LMP information in the formulation of emergency planning (including meeting Part 50 requirements)

Program effort currently underway to develop guidance for emergency plans based on the LMP approach and outputs 67 FINDINGS AND RECOMMENDATIONS Parts 30, 40, 70 and 72 Emergency Plan Elements

No Show-Stoppers Utilizing a RIPB pathway for the licensing of non-core sources associated with advanced reactors seems possible (from a safety perspective).

Recommended Next Steps

1) Pursuit of regulatory clarity o Alternative event category definitions under Part 70 and 72 o Utilization of supplemental worker dose and chemical hazard criteria as part of LMP o SSC special treatments under Part 70 and 72

2) Part 72 SAR crosswalk o Comparing TICAP content and necessary Part 72 SAR content

3) Guidance regarding emergency planning using the LMP approach (ongoing parallel effort) 68 OVERALL CONCLUSIONS ANL/NSE-22/62 OSTI: TBD

Observation Current draft 10 CFR Part 53 includes the QHOs as explicit regulatory requirements.

If an applicant is including non-core sources of radioactivity within their plant PRA, the impact of these events may be included when determining satisfaction of the QHOs.

Such an approach may introduce inconsistencies when compared to the current application of the QHOs to LWRs, when utilizing the surrogates of CDF and LERF*.

May be particularly important for external hazard scenarios that impact multiple sources of radioactivity at the plant simultaneously.

69 QUANTITATIVE HEALTH OBJECTIVES

Past NRC studies have examined the risk of spent fuel pools and comparison to the QHOs (NUREG-1738)

70 QUESTIONS

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Break Meeting will resume at 2:00 pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 856 640 616#

Advanced Reactor Stakeholder Public Meeting

Electronic Submittal of Advanced Reactor Applications NRR/DANU/UARP Advanced Reactor Stakeholder Meeting June 7, 2023 73

Electronic Submittals of Advanced Reactor Applications

Purpose:

To Provide Insights Regarding Submittals for Advanced Reactor Applications Outcome: Insights lead to efficient submittal and processing of advanced reactor applications 74

Electronic Submittals of Advanced Reactor Applications

History

- Because of past limitations for processing of large documents in ADAMS, NRC staff developed a process that included use of a packing slip

Previous presentation on use of packing slips available at: ML14071A344

Bellefonte combined license application had over 750 individual files

Current Guidance

- Electronic Submittal webpage has been updated to remove the mention of packing slips (see: https://www.nrc.gov/site-help/e-submittals.html) 75

Electronic Submittal of Advanced Reactor Applications

Current Guidance

- Electronic Submittal Guidance Document is available at: ML13031A056

Link to guidance document can also be found on NRC webpage:

https://www.nrc.gov/site-help/electronic-sub-ref-mat.html

Defines individual file requirements

- Current limitation on file size is provided in this document and it is substantially greater than the file size limitation in 2007 time frame of 25 megabytes

- Increase in allowed individual file size could lead to less individual files being submitted 76

Electronic Submittal of Advanced Reactor Applications

Current Guidance

Recent examples of large electronic submittals include:

-Vogtle 3 and 4 updated final safety analysis report, Tier 1, Technical Requirement Manual, and Technical Specifications Bases Submittal (see:

https://www.nrc.gov/docs/ML2217/ML22179A121.html

-North Anna Updated Final Safety Analysis Report (see:

ML22283A023) 77

Electronic Submittal of Advanced Reactor Applications

Current Guidance

- Treatment of Sensitive Information

Documents containing safeguards information may not be submitted via the electronic information exchange (EIE) process

- Documents with Safeguards Information (SGI) may be transmitted on optical storage media (OSM)

- The mailing package containing optical storage media with safeguards information must be processed, marked and transmitted in accordance with the requirements set forth in 10 C.F.R. § 73.22(e), (g), (h), and (f), as appropriate.

78

Electronic Submittal of Advanced Reactor Applications

Current Guidance

- Treatment of Sensitive Information

Sensitive Unclassified Non-Safeguards Information (SUNSI) may be electronically submitted through the EIE process or on OSM.

- Transmittal documents used to transmit one or more documents containing SUNSI must be marked to show SUNSI is contained in the documents being transmitted. A header marking must be placed on each page of the transmittal document showing the type of SUNSI (i.e., Security-Related InformationWithhold under 10 C.F.R. § 2.390, or Proprietary InformationWithhold under 10 C.F.R. § 2.390.)

- Preferred approach is to have a redacted publicly available portion of the application and an unredacted non-publicly available portion of the application 79

Electronic Submittal of Advanced Reactor Applications

Current Guidance Options

- Option 1 - application documents can be broken into manageable number of files (around 25 files) and total file size for the combined files is also manageable (around 1 gigabyte or less)

>> Electronic Information Exchange could be used

- Option 2 - application documents can be broken into manageable number of files (around 25 files) and total file size for the combined files is large (greater than 1 gigabyte)

>> Recommend that the submittal be provided to the document control desk via optical storage media such as compact disc (CD)/digital visual disc (DVD) or thumb drive

Concern is transfer speed capabilities via (EIE)

Could take a lengthy amount of time during which potential interruptions during the transfer could cause issues

Both above options assume major portions of the application will be submitted initially and in subsequent revisions (i.e., subsequent revisions will not be on a page replacement basis).

In addition, as discussed above applicants are cautioned to ensure appropriate steps are taken to protect SUNSI and that SGI is not to be processed in ADAMS.

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Electronic Submittal of Advanced Reactor Applications Questions?

81

Advanced Reactor Construction Oversight Program (ARCOP)

PENDING INFORMATION SECY (ARCOP INFO SECY)

Jon Greives Deputy Director, DANU/NRR 82

Topics Purpose of the ARCOP INFO SECY ARCOP Focus Monitoring Quality Enforcing Noncompliances Assessing Results Next Steps 83

84 ARCOP INFO SECY Vision for the NRCs Advanced Reactor Construction Oversight Program (ARCOP)

ARCOP Focus Quality provides confidence that SSCs will perform satisfactorily in service Manufacturers Establish reasonable assurance that facilities are built and will operate in accordance with their approved design and licensing bases Suppliers On-Site Construction 85

ARCOP Cornerstones of Safety 86 Quality of Construction Quality of Offsite Manufactured and Assembled Items and Services Quality of Onsite Construction Security &

Safeguards Programs Security &

Safeguards Operational Programs Operational Readiness Quality of Procured Items and Services Five Cornerstones of Safety Three Strategic Performance Areas

Areas of NRC Oversight

Performance Monitoring (Quality Monitoring)

Enforcing Noncompliance (including significance determination)

Assessing Results 87

Performance Monitoring Inspecting Quality 88 Facility-specific baseline inspection plans based on the potential of different construction activities to impact Fundamental Safety Functions Site-specific inspection plans for licensees Design-specific baseline inspection plans for manufacturers and vendors

Enforcing Noncompliance /

Significance Determination 89

- Significance determination based on impact to fundamental safety functions.

- Additional minor violation criteria including self-identification and correction credit

- Results in significance of noncompliance being appropriately aligned with potential risk of construction activities.

Assessment Overall Quality Assessment for OL issuance or 103(g)

Monitor On-Site Construction Quality through Licensee Performance Monitor Reactor/Module Quality through Manufacturer Performance COL/CP Holder Assessment Manufacturer Assessment COL, CP, LWA, or ESP Holder Manufacturer Licensee Inspection Plan Manufacturer Inspection Plan Vendor Vendor Inspection Plan

1. BIP Adjustment

2. Suppl. Inspections

3. Program Feedback

1. BIP Adjustment

2. Suppl. Inspections

3. Program Feedback 90

Next steps 91 IMC DRAFT COMPLETION New IMCs are being developed for ARCOP.

INTERNAL TABLETOPS Evaluate IMCs with data and experience EXTERNAL WORKSHOPS Engage AR stakeholders INTERNATIONAL COOPERATION Explore expanded cooperation for construction oversight INSPECTION TECHNOLOGY Explore use of new technologies to further improve inspection efficiency

TIMELINE 92 SUMMER 2023 FALL 2023 2024 2025 Draft and tabletop IMCs Conduct external workshops Finalize ARCOP procedures AR construction oversight program ready for use

questions Follow-up questions can be addressed to:

Jon Greives Deputy Director, NRR/DANU jonathan.greives@nrc.gov OR Phil OBryan Senior Reactor Operations Engineer and ARCOP Project Lead phil.obryan@nrc.gov 93

Break Meeting will resume at 3:15 pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 856 640 616#

Advanced Reactor Stakeholder Public Meeting

Updates on Advanced Reactor Content of Application Project (ARCAP) Interim Staff Guidance (ISG) Documents and Technology Inclusive Content of Application Project (TICAP) Draft Guide

Purpose 96

Facilitate stakeholder understanding of the ARCAP interim staff guidance (ISG) documents and the TICAP draft regulatory guide (DG) 1404

Provide guidance on how to submit written comments. Comments on the documents will not be taken in todays meeting. Please submit your comments in accordance with the instruction in the Federal Register notices.

Comment period ends on July 10, 2023 Note: ARCAP = Advanced Reactor Content of Application Project TICAP = Technology Inclusive Content of Application Project

ARCAP/TICAP Background 97

Guidance for developing and reviewing technology-inclusive, risk-informed, and performance-based non-light water (non-LWR) applications

Being developed to support 10 CFR Part 50 and 10 CFR Part 52 applications

Needed to support expected near-term non-LWR Part 50/52 applications using the licensing modernization project (LMP) process in NEI 18-04, Revision 1

The NRC staff intends to revise the guidance per the final Part 53 rulemaking language

TICAP Background 98

TICAP scope is governed by the LMP-based process

LMP uses risk-informed, performance-based approach to select licensing basis events, develop structures, systems, and components (SSC) categorization, and ensure that defense-in-depth is considered

Industry developed key portions of TICAP guidance See NEI 21-07, Revision 1, Technology Inclusive Guidance for Non-Light Water Reactors Safety Analysis Report Content for Applicants Utilizing NEI 18-04 Methodology, (ADAMS Accession No. ML22060A190)

DG 1404 proposes to endorse NEI 21-07, Revision 1, with clarifications and additions

ARCAP Background 99

Broad in nature and intended to cover guidance for non-LWR applications for:

combined licenses construction permits operating licenses design certifications standard design approvals manufacturing licenses

Encompasses TICAP

TICAP is guidance for off-normal reactor states only. ARCAP encompasses everything needed for a license application.

ARCAP and TICAP - Nexus 100 Outline Safety Analysis Report (SAR) -

Based on TICAP Guidance 1.

General Plant Information, Site Description, and Overview 2.

Methodologies and Analyses and Site Information*

3. Licensing Basis Event (LBE) Analysis

4. Integrated Evaluations

5. Safety Functions, Design Criteria, and SSC Safety Classification

6. Safety Related SSC Criteria and Capabilities

7. Non-safety related with special treatment SSC Criteria and Capabilities

8. Plant Programs Additional Portions of Application

Technical Specifications

Technical Requirements Manual

Quality Assurance Plan (design)

Fire Protection Program (design)

Quality Assurance Plan (construction and operations)

Emergency Plan

Security Plan

Cyber Security Plan

SNM physical protection program

SNM material control and accounting

Fire Protection Program (operational)

Radiation Protection Program

Offsite Dose Calculation Manual

Inservice inspection/Inservice testing (ISI/IST) Program

Environmental Report and Site Redress Plan

Financial Qualification and Insurance and Liability

Fitness for Duty Program

Aircraft Impact Assessment

Performance Demonstration Requirements

Nuclear Waste Policy Act

Operational Programs

Exemptions, Departures, and Variances )

Audit/inspection of Applicant Records Calculations Analyses P&IDs System Descriptions Design Drawings Design Specs Procurement Specs Probabilistic Risk Assessment

SAR Chapter 2 derived from TICAP guidance as supplemented by ARCAP interim staff guidance Chapter 2, Site Information Safety Analysis Report (SAR) structure based on clean sheet approach Additional contents of application may exist only in the SAR, may be in a separate document incorporated into the SAR, or may exist only outside the SAR.

The above list is for illustration purposes only.

Additional SAR Content -Outside the Scope of TICAP

9. Control of Routine Plant Radioactive Effluents, Plant Contamination, and Solid Waste

10. Control of Occupational Doses

11. Organization and Human-System Considerations

12. Post-construction Inspection, Testing and Analysis Programs

ARCAP/TICAP Background 101

Ten draft documents were publicly released in 2021 to encourage early stakeholder feedback

Nine ARCAP draft ISGs (released as White Papers)

One TICAP (DG 1404)

Draft Document Subject ADAMS Accession No.

Date Most Recent Version Released Review of Risk-Informed, Technology-Inclusive Advanced Reactor Applications - Roadmap ML21336A702 12/2/21 Chapter 2, Site Information ML21189A031 7/6/21 Chapter 9, Control of Routine Plant Radioactive Effluents, Plant Contamination and Solid Waste ML21189A033 7/6/21 Chapter 10, Control of Occupational Doses ML21189A035 7/6/21 Chapter 11, Organization and Human-System Consideration ML21309A020 11/5/21 Chapter 12, Post Construction Inspection, Testing and Analysis Program ML21294A266 10/21/21 Risk-Informed ISI/IST Programs ML21216A051 8/4/21 Licensing Modernization Project-based Approach for Developing Technical Specifications ML21133A490 5/10/21 Risk-Informed, Performance-Based Fire Protection Program (for Operations)

ML21294A266 10/21/21 Draft Regulatory Guide 1404, Guidance for a Technology Inclusive Content of Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Advanced Reactors ML21336A697 12/2/21

Revised ARCAP/TICAP Documents 102 All ten documents were reissued in May of 2023 (ADAMS Package No. ML23044A038). The 45-day formal public comment period started on May 25, 2023, and ends on July 10, 2023.

The NRCs Documents for Comment website https://www.nrc.gov/public-involve/doc-comment.html provides guidance on how to submit comments and provides links to the regulations.gov docket IDs.

ARCAP ISG Title ADAMS Accession #

Federal Register Regulations.gov Docket ID No.

DANU-ISG-2022-01, Review of Risk-Informed, Technology-Inclusive Advanced Reactor Applications - Roadmap ML22048B546 88 FR 33924 NRC-2022-0074 DANU-ISG-2022-02, Chapter 2, Site Information ML22048B541 88 FR 33940 NRC-2022-0075 DANU-ISG-2022-03, Chapter 9, Control of Routine Plant Radioactive Effluents, Plant Contamination and Solid Waste ML22048B543 88 FR 33930 NRC-2022-0076 DANU-ISG-2022-04, Chapter 10, Control of Occupational Doses ML22048B544 88 FR 33936 NRC-2022-0077 DANU-ISG-2022-05, Chapter 11, Organization and Human-System Consideration ML22048B542 88 FR 33928 NRC-2022-0078 DANU-ISG-2022-06, Chapter 12, Post Construction Inspection, Testing and Analysis Program ML22048B545 88 FR 33920 NRC-2022-0079 DANU-ISG-2022-07, Risk-Informed ISI/IST Programs ML22048B549 88 FR 33938 NRC-2022-0080 DANU-ISG-2022-08, Licensing Modernization Project-based Approach for Developing Technical Specifications ML22048B548 88 FR 33926 NRC-2022-0081 DANU-ISG-2022-09, Risk-Informed, Performance-Based Fire Protection Program (for Operations)

ML22048B547 88 FR 33922 NRC-2022-0082 Draft Regulatory Guide 1404, Guidance for a Technology Inclusive Content of Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Advanced Reactors ML22076A003 88 FR 33846 NRC-2022-0073

Common Changes for all ISGs and DG 1404 103

Applicability is now only for non-LWRs Recommends that light-water reactor applicants wanting to use ARCAP/TICAP guidance engage in pre-application discussions

All ISGs provide applicant guidance and NRC staff review guidance in separate sections

Removed references that did not have complete NRC staff review Appendices added to several ISGs to list in-development guidance documents that could affect future revision of those ISGs

Common ISG sections now contain uniform text and added two short sections to account for the Paperwork Reduction Act

Roadmap ISG Changes 104

Removed the definition of advanced reactor

Provides a listing of references that are associated with the ARCAP roadmap ISG guidance and includes references from other ISGs and TICAP DG

Purpose is to aid applicants and the NRC staff in the development and review of the application

Added guidance regarding Design of Structures, Components, Equipment, and Systems

Expanded discussion regarding site evaluation guidance

Roadmap ISG Changes (continued) 105

Added guidance regarding principal design criteria (PDC) development for those portions of a design outside the scope of TICAP

Revised guidance regarding the Technical Requirements Manual

Added guidance regarding special nuclear material (SNM) control and accountability

Added cyber security guidance

106

Added guidance regarding Financial Qualification and Insurance and Liability

Added Fitness for Duty section

Including referencing requirements for construction permits

Added guidance regarding Aircraft Impact Assessment

Added guidance for performance demonstration requirements in accordance with 10 CFR 50.43(e) requirements Roadmap ISG Changes (continued)

107

Added discussion of Nuclear Waste Policy Act contractual requirements

Added discussion of operational programs required by regulations

Minor changes made to Appendix A, Pre-Application Engagement Guidance

Added a discussion related to material qualification Roadmap ISG Changes (continued)

108

Added Appendix B, Analysis of Applicability of NRC Regulations to Non-Light-Water Power Reactors

Appendix B corrects errors in previous versions and clarifies certain matters

Noted that aircraft impact assessments are applicable for Construction Permits (CPs)

Added guidance in footnotes regarding the use of 10 CFR 50.69 and mitigation of beyond-design-basis-events for Part 52

Revised CP guidance in Appendix C

Many miscellaneous changes

Updated portions of the guidance copied from the final issuance of the LWR construction permit ISG (see: November 14, 2022 LWR CP ISG Federal Register Notice)

Added Appendix D, Draft ARCAP guidance documents under development as of May 2023 Roadmap ISG Changes (continued)

Chapter 2 ISG Site Information Changes 109

Added reference to the Commissions July 13, 2022, Staff Requirements Memorandum on SECY-00-0045 regarding a revision of Regulatory Guide 4.7 incorporating a risk informed assessment of population related issues in determining site suitability

Added Section 2.7.1 on volcanic hazards and Section 2.7.2 on a screening approach to identify other external hazards beyond design basis hazard levels (DBHLs)

Added reference to NUREG-0800, Section 3.5.1.6, Aircraft Hazards, to Section 2.3.2, review Guidance-Acceptance Criteria

Chapter 9 ISG Routine Effluents Changes 110

Clarified the use of an alternative performance monitoring system

Clarified the use of exemptions

Added guidance and conditions for situations where a design does not generate any normal radioactive effluent releases

Added that a summary of the estimated doses to the public should be included in an application

Chapter 10 ISG Occupational Dose Changes 111

Added guidance for Standard Design Approvals and CPs

Added guidance regarding Combined License (COL) action items

Chapter 11 ISG Organization and Human-System Considerations Changes 112

Clarified guidance for applicants wanting approval of a licensed operator staffing plan that does not meet 10 CFR 50.54

Clarified guidance regarding situations where control room design details are not complete at the time of a COL application

Removed reference to remote operations white paper

Chapter 12 ISG Post-construction Inspection, Testing, and Analysis Program Changes 113

Clarified guidance regarding the scope of PITAP and the regulatory basis for post construction inspection requirements

Clarified the applicability of 10 CFR Part 50, Appendix B, Criterion III, Design Control, as a basis for PITAP Phase I requirements

Removed references to ASME Section VIII because Section III is the governing Code for SSCs addressed by this ISG

Added staff guidance regarding test description content

Technical Specifications (TS) ISG Changes 114

Clarified that design certifications (DCs) should have proposed generic TSs

Added that Manufacturing Licenses (MLs) must have final operational information

Added that applications referencing a DC or ML should replace bracketed information with site specific information.

Technical Specifications ISG Changes (continued) 115

Removed reference to the possible need for exemptions

In the Federal Register Notice for this ISG, the NRC is requesting comments on whether the correlation between the 10 CFR 50.36 text and the LMP process outputs require the NRC to consider an exemption

Revised the Safety Limit definition corresponding to NEI 18-04 output

Added an administrative control regarding reactor coolant system temperature and pressure limits report, if applicable to the specific design

116

Added staff recommendation that applicants use standard TS NUREGs for TS format

Clarified guidance regarding situations where applicants propose to use risk criteria that are different from the NEI 18-04 Frequency-Consequence (F-C) Target

Clarified text regarding the guidance for TS "Use and Application" information

Added Acceptance Criteria that the requirements of 10 CFR 50.36a are met Technical Specifications ISG Changes (continued)

ISI/IST ISG Changes 117

Added that the NRC staff assumes applicants will design and qualify their equipment to the latest NRC-accepted ASME Codes

Clarified guidance on components that perform active safety functions, such as moving fluid or transferring heat without mechanically interacting with the fluid

Identified that if components are safety significant and unique to a new design, additional examinations/testing may be necessary and sampling may not be sufficient

Fire Protection (Operations) ISG Changes 118

Added that all analyses related to the program should be available for audit

Clarified text regarding applicability of 10 CFR 50.48(c) and NFPA 805

TICAP DG Changes 119

Removed affirmative safety case, safety case, and licensing case terminology that comes from NEI 21-07

Added additional guidance for site evaluations with respect to external hazards and DBHLs and associated design requirements for SSCs

Added guidance on options using LMP based approach to address 10 CFR 50.34(a)(1)(ii)(D) and 52.79(a)(1)(vi) dose criteria (see staff position C.3.c)

Added guidance for the development of PDCs

TICAP DG Changes (continued) 120

Added the application must describe the safety features and components that require research and development (R&D).

Notes the applicant will conduct an R&D program to resolve any associated safety questions

Construction Permit Guidance Changes:

CP guidance moved from an appendix to main document

Added that the CP application should provide the necessary commitments to establish defense-in-depth adequacy

Added guidance regarding the scope of preliminary safety analysis descriptions in CP applications

TICAP DG Changes (continued) 121 Construction Permit Guidance Changes (continued)

Added that the CP application should include commitments to confirm reliability and capability targets consistent with the probabilistic risk assessment (PRA) for SR [and NSRST] SSCs in the operating license application, if not provided in the CP application

In addition, the application should describe any planned testing, validation, and special treatment to be applied to the SSCs to confirm their performance

Staff positions C.3.d (addition), C.3.e (clarification), and C.3.f (clarification) provide construction permit probabilistic risk assessment guidance

122

Added Appendix A, Draft ARCAP Guidance Documents Under Development as of May 2023, that could affect DG-1404

Includes four item related to probabilistic risk assessment

Includes an item related to materials compatibility

Potential future addition: List of generic safety issues that non-LWR applicants should address TICAP DG Changes (continued)

ARCAP/TICAP Title Regulations.gov Docket ID No.

DANU-ISG-2022-01, Review of Risk-Informed, Technology-Inclusive Advanced Reactor Applications -

Roadmap NRC-2022-0074 DANU-ISG-2022-02, Chapter 2, Site Information NRC-2022-0075 DANU-ISG-2022-03, Chapter 9, Control of Routine Plant Radioactive Effluents, Plant Contamination and Solid Waste NRC-2022-0076 DANU-ISG-2022-04, Chapter 10, Control of Occupational Doses NRC-2022-0077 DANU-ISG-2022-05, Chapter 11, Organization and Human-System Consideration NRC-2022-0078 DANU-ISG-2022-06, Chapter 12, Post Construction Inspection, Testing and Analysis Program NRC-2022-0079 DANU-ISG-2022-07, Risk-Informed ISI/IST Programs NRC-2022-0080 DANU-ISG-2022-08, Licensing Modernization Project-based Approach for Developing Technical Specifications NRC-2022-0081 DANU-ISG-2022-09, Risk-Informed, Performance-Based Fire Protection Program (for Operations)

NRC-2022-0082 Draft Regulatory Guide 1404, Guidance for a Technology Inclusive Content of Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Advanced Reactors NRC-2022-0073 123

1. Federal Rulemaking Website: Go to https://www.regulations.gov/ and search for the following Docket IDs. Comments must be received by July 10, 2023.

Two Ways to Submit Comments

124 Two Ways to Submit Comments (continued)

1. Federal Rulemaking Website: Go to https://www.regulations.gov/ using Docket IDs (continued)

Address questions about Docket IDs in Regulations.gov to Stacy Schumann; telephone: 301-415-0624; email: Stacy.Schumann@nrc.gov For technical questions, contact: Joseph Sebrosky, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: 301-415-1132, email: Joseph.Sebrosky@nrc.gov; or Michael Orenak, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: 301-415-3229, email:

Michael.Orenak@nrc.gov

2. Mail comments to: Office of Administration, Mail Stop:

TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Program Management, Announcements and Editing Staff.

125 QUESTIONS?

(Note: The NRC is not taking comments on the ISGs or the TICAP DG in this Q&A session)

Initial Observations on Draft NRC Reg Guide DG-1404

NEI 21-07 is intended to be part of a streamlined and predictable licensing pathway to deployment for advanced reactors under 10 CFR Part 50 or 52

NEI 21-07 applies to designs that follow the methodology of NEI 18-04

The NRC made draft Regulatory Guide DG-1404 on NEI 21-07 available on ADAMS on May 18, 2023, in conjunction with a number of interim staff guidance documents (ISGs) related to advanced reactors

Comments on DG-1404 and all of the ISGs are due July 10, 2023 By our count, DG-1404 contains

9 staff positions

2 combined clarifications and additions

8 clarifications

17 additions Introduction and Overview

©2023 Nuclear Energy Institute 128 Requests for clarification of some information in DG-1404 is provided in this presentation, supported by initial feedback on some items The TICAP team intends to formally submit comments on DG-1404 If an addition or clarification is not addressed in this presentation, that does not mean the TICAP team does not have a comment The TICAP team believes that a number of elements in DG-1404 would benefit from further discussions between the team and the NRC This presentation is focused on items in DG-1404 for which clarification is needed Initial General Feedback

Carryover of LWR business as usual where it is not beneficial but adds burden New NRC requirements that were not raised during the extensive discussions held between the NRC and the TICAP team in 2020 through 2022 Language and guidance in DG-1404 that is redundant to guidance already in NEI 21-07 The use of DG-1404 to provide guidance on matters outside the scope of NEI 18-04 and NEI 21-07, rather than putting that information in ARCAP where it would be more appropriate General Concerns (1 of 2)

©2023 Nuclear Energy Institute 130 Having a regulatory guide with a large number of additions and clarifications is not desirable from the standpoint of providing clear, understandable guidance for applicants and regulators A one and done Federal Register comment process on the draft Reg Guide may not result in the quality of guidance needed to enable an efficient and effective regulatory review of advanced non-LWRs The TICAP team desires to work with the NRC to develop the best possible guidance General Concerns (2 of 2)

NRC objects to the term LMP-based affirmative safety case and other variations of it NRCs specific concern is unclear NEI 21-07 defines the term LMP-based affirmative safety case so the meaning is not ambiguous Request for Clarifications on DG-1404

The NRC requests a summary table of regulatory guides directly applicable to the design What is the definition of directly applicable to the design Should it not be regulatory guides used in the design?

Regulatory guides are not mandatory, so their use is up to the applicant Request for Clarifications on DG-1404

The NRC requests additional information on radiological doses, referencing 10 CFR 50.34(a)(1)(ii)(D) or 10 CFR 52.79(a)(1)(vi)

How are these requirements related to NEI 18-04?

When would an exemption be needed for Option 1?

What are the meanings of the terms bounding DBA and bounding DBE in Option 2?

Request for Clarifications on DG-1404

The NRC wants applicants to identify and describe in Chapter 2 the non-PRA analysis and calculation methodologies used to establish their licensing bases What is the definition of analysis and calculation methodologies?

This seems rather open-ended; many analyses and calculations do not appear in SARs for current light water reactors Can the NRC provide examples?

Does NRC object to using Chapter 3 for documenting methodologies?

NEI 21-07 was written with the clear intent to allow applicants the option of documenting DBA analyses in Chapter 2 (if used in multiple applications) or in Chapter 3 with the associated DBA (if used only in that instance)

Request for Clarifications on DG-1404

Addition C.4.a:

The discussion of AOOs, DBEs, DBAs, and BDBEs in Chapter 3 of the SAR should include a description of the models, site characteristics, and supporting data associated with the calculation of the mechanistic source terms and radiological consequences (to the extent that such information does not appear in the discussions of methodologies and analyses in Chapter 2, the descriptions of systems and functions in Chapters 5-7, or other sections of the SAR).

Clarification C.4.b:

Section C.2.1.1 of NEI 21-07, Revision 1, contains adequate guidance on the level of detail in the SAR to describe non-DBA LBEs.

Taken together, the TICAP team interprets this information to mean that the information specified in 2.1.1 should be adequate to address the NRCs desire for information on AOO, DBE, and BDBE dose calculations Is that the correct interpretation?

Request for Clarifications on DG-1404

The CP application should provide a discussion in the SAR to establish DID adequacy. A discussion in the SAR to implement the DID adequacy assessment processes in RG 1.233 is considered acceptable for this purpose. Alternatively, the applicant should ensure that its DID process involves incorporating DID into design features, operating and emergency procedures, and other programmatic elements to ensure that performance requirements are maintained throughout the life of the plant. An applicant that chooses not to use the approach endorsed in RG 1.233 will need to explain its approach to DID and describe how it addresses DID in the application.

NEI 21-07 addresses DID needs for a CP - what is the purpose of the addition?

Why is there a carve out for applicants not following RG 1.233 here (for DID), but not in other areas covered by RG 1.233 (e.g., LBE identification)?

Did the NRC mean to state The CP application should provide a discussion in the SAR of the approach to establishing DID adequacy.?

Request for Clarifications on DG-1404

For each of the three plant performance metrics discussed above in Section 5 of this document (section C.4 of the application guidance), in addition to the results and margins, the SAR Chapter 4 should address the following (where different from the analysis performed for Chapter 3):

What is meant by where different from the analysis performed for Chapter 3?

The Chapter 3 AOO, DBE, and BDBE analyses and the quantitative integrated evaluations in Chapter 4 are performed by the same tool - the plant PRA Request for Clarifications on DG-1404

The NRC states that applicants should address human factors considerations in Chapter 4 (Integrated Evaluations).

Given that the NRC established Chapter 11 of the SAR to deal with human factors, what is the rational for requiring human factors information in Chapter 4?

Substantial previous discussions with the NRC on human factors occurred prior to the submittal of NEI 21-07.

NRC never suggested Chapter 4 should include human factors information Request for Clarifications on DG-1404

The NRC is asking for information on change control Why is this request not addressed by the Technology Inclusive Risk Informed Change Evaluation (TIRICE) project, which is developing guidance on change control?

What is desired in SAR Chapter 4 above and beyond the TIRICE guidance?

Request for Clarifications on DG-1404

The NRC asks for an extensive discussion of fuel qualification in Chapter 5 of the SAR This is a new requirement that was not raised during the extensive discussions on NEI 21-07 Why is Chapter 5 - Safety Functions, Design Criteria, and SSC Safety Classification - the appropriate location for such detailed information?

Isnt the information requested in Section (1) (a discussion of the role of the fuel in the safety analysis) already covered in Chapter 3?

Why is the guidance in (1) and (2) written like instructions for an NRC reviewers findings rather than for an applicant?

Request for Clarifications on DG-1404

The NRC asks for additional information beyond what is in NEI 21-07, but only for instrumentation and control (I&C) SSCs, not all SSCs Why is this expectation confined to I&C systems?

Request for Clarifications on DG-1404

©2023 Nuclear Energy Institute 142 The TICAP team believes that it is important to minimize duplicative discussions between NEI 21-07 and the associated NRC guidance document This will produce clarity and enhance usability and efficiency In a number of instances NRC clarifications and additions are unwarranted because NEI 21-07 already adequately addresses the issue Duplicative Clarifications and Additions Clarifications C.2.b C.3.e C.4.a(2)

C.3.f C.4.a(3)

C.3.a C.4.a(4)

Additions C.2.c C.3.g C.7.b(1)

C.8.a(1)

Clarification and Addition C.7.a

Thank You for Your Time and Attention

Future Meeting Planning

The next periodic stakeholder meetings are scheduled for July 20, 2023, and September 14, 2023.

If you have suggested topics, please reach out to Steve Lynch at Steven.Lynch@nrc.gov 144

How Did We Do?

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https://www.nrc.gov/pmns/mtg?do=details&Code=20230269

Then, click link to NRC public feedback form:

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