ML20239A957

Technology Inclusive Content of Applicatin Project, Advanced Reactor Content of Application Project, and Construction Permit Guidance Public Meeting

ML20239A957
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Issue date:	08/27/2020
From:	Joseph Sebrosky NRC/NRR/DANU/UARP
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Sebrosky J
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Download: ML20239A957 (81)

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Technology Inclusive Content of Application Project, Advanced Reactor Content of Application Project, and Construction Permit Guidance Public Meeting August 27, 2020 Telephone Bridgeline: 877-779-7419 Passcode: 9782874 1 of 81

Agenda Time Topic Presenter 10: -10:10 am Introduction NRC 10:10 - 10:40 amUpdates on industry-led Technology Inclusive Content of Application Project Southern (TICAP) efforts including status of fundamental safety function mapping, status of table top exercises 10:40 - 11:20 am Discussion of the concept (including examples) of Principle Design Criteria, Southern and Complementary Design Criteria 11:20 - 11:45 am TICAP next steps Southern 11:45 - 12:00 pm Feedback on fundamental safety function mapping and other areas NRC 12:00 - 12:30 pm Stakeholder questions All 12:30 - 1:30 pm Break All 1:30 -1: 50 pm Updates on Construction Permit Guidance for light water small modular NRC reactors 1:50 - 2:20 pm Industry and other Stakeholder feedback Industry, All 2:20 - 3:00 pm Discussion of Advanced Reactor Content of Application Project Including NRC/Idaho Additional Thoughts on Use of Performance-Based Approach National Lab 3:00 - 3:25pm Industry and Other stakeholder feedback All 3:25- 3:30 pm Next Steps and Concluding Remarks All 2 of 81

Technology Inclusive Content of Application Project (TICAP) Presentations Steve Nesbit, LMNT Consulting Brandon Chisholm, Southern Company TICAP - Nuclear Regulatory Commission (NRC) Working Meeting August 27, 2020 3 of 81 1

Outline of Todays TICAP Presentations

• Introduction and Overview (Steve)

• Content of Application Guidance (Steve)

• Tabletop Exercises (Brandon)

• Principal Design Criteria and Complementary Design Criteria (Brandon)

Please note that we will be discussing work in progress, not vetted and finalized results. We request your indulgence and welcome your feedback.

4 of 81 2

TICAP Overview

• Product: Develop an endorsable Guidance Document that proposes an optional formulation of advanced reactor application content that

- Benefits from the insights and knowledge gained through licensing and safely operating the current US-based nuclear fleet for over 40 years to ensure adequacy of proposed content requirements.

- Is based on describing a technology-inclusive affirmative safety case that meets the underlying intent of the current requirements

>> To optimize application content (add where additional content is needed and reduce where current content requirements are not commensurate with the contribution to risk)

>> To provide the needed regulatory agility to accommodate review of spectrum of designs that are expected to submit licensing application,

- Is risk-informed, performance-based to right size the required information in an application (based on the complexity of the safety case) to increase efficiency of generating and reviewing an application

- Its scope is governed by the Licensing Modernization Project (LMP)-based safety case to facilitate a systematic, technically acceptable, and predictable process for developing a designs affirmative safety case

- Provides similar information as is currently required from a light water reactor (LWR) applicant 5 of 81 3

Background

LMP-Driven Application Content

• Projects Expected Outcomes:

- A standardized content structure that facilitates efficient

>> preparation by an applicant,

>> review by the regulator, and

>> maintenance by the licensee.

- A content formulation that, based on the complexity of a designs safety case, optimizes

>> the scope (the functions, the structures, systems, and components (SSCs), and the programmatic requirements that need to be discussed) based on what is relevant to the design specific safety case.

>> the type of information to be provided (e.g., licensing basis events (LBEs), Required Safety Functions (RSFs), Safety-Related SSCs, Defense-in-Depth (DiD), etc.),

>> level of detail to be provided

• based on the importance of the functions and SSCs to the safety case (risk-informed, performance-based details).

• based on the relevance to the safety determination 6 of 81 Creating Clarity, Predictability, and Transparency 4

Affirmative Safety Case LMP-Based Affirmative Safety Case Definition - A collection of scientific, technical, administrative and managerial evidence which documents the basis that the performance objectives of the technology inclusive fundamental safety functions (FSFs) are met by a design during design specific Anticipated Operational Occurrences (AOOs), Design Basis Events (DBEs), Beyond Design Basis Events (BDBEs), and Design Basis Accidents (DBAs) by

- Identifying design specific safety functions that are adequately performed by design specific SSCs and

- Establishing design specific features (programmatic (e.g., inspections) or physical (e.g., redundancy)) to provide reasonable assurance that credited SSC functions are reliably performed.

7 of 81 5

Foundation of the TICAP Affirmative Safety Case Description The underlying intent of the current application content (within TICAP scope) is met by providing the LMP-Based Safety Case, anchored around principal design criteria (PDC), on the basis that

>> The LMPs approach to meeting the radiological risk performance objectives provides evidence that the underlying safety objectives of the regulations for providing reasonable assurance of adequate protection . . . are met.

>> The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public (introduction to 10 CFR 50 Appendix A) 8 of 81 6

Technology Inclusive Content of Application Project (TICAP)

Content of Application Guidance Steve Nesbit TICAP - NRC Working Meeting August 27, 2020 9 of 81 7

NEI Guidance Document Annotated Outline

• NEI Guidance Document

- Key product from TICAP

- Guidance for structure, scope, and level of detail for portions of an advanced reactor safety analysis report (SAR) related to the affirmative safety case developed in accordance with NEI 18-04

- To be submitted by the Nuclear Energy Institute to the NRC around September 2021

• Guidance organized around Safety Analysis Report (SAR) outline

- Development of guidance ongoing 10 of 81 8

SAR Organization 11 of 81 9

Something I Noticed on the Way to the Forum

• Traditionally SARs include a large amount of information, much of which does not directly relate to public radiological safety

• Example: recent Lee Nuclear Station combined license SAR for two AP1000 reactors

- Reference 205 is the website for the Thunder Road Marathon

>> Mentioned in 2.1.3.3.2 Transient Population Between 10 and 50 Miles

>> Once a year event 50 miles from the proposed reactor site

- Section 2.5.1 (Basic Geological Information) includes 97 pages of text and references, four tables, and 58 figures

- Section 2.3 (Meteorology) has 89 figures Things are the way they are because they got that way.

- Gerald Weinberg, American computer scientist 12 of 81 10

There May Be a Bit of a Problem TICAP Philosophy - Focus on Affirmative Safety Case 13 of 81 11

What Are the Goals for the SAR?

• Satisfy regulatory requirements

• Provide a basis for a licensing decision in a manner that is understandable to the applicant, the NRC, and other stakeholders

- Present information at the appropriate level of detail

- Focus on safety

• Capture the design basis of the facility in a manner that is straightforward to apply and maintain

• Not goals for the SAR

- Highlight things considered important

- Discuss features and programs that do not impact the safety case

- Compile all data related to a parameter of interest 14 of 81 12

The Dilemma of Technology-inclusive Guidance

• Detailed guidance is appealing

- Predictable

- Standard of acceptability - whatever was done last time

• Technology-inclusive guidance

- Wide variation in safety cases and number and type of

>> Licensing basis events

>> PRA Safety Functions

>> Safety Related SSCs and Non-Safety-Related with Special Treatment SSCs

- Past a certain point, detail in guidance is counterproductive

- Focus should be on formulation of content

- Tabletop exercises will play a key role in refining and validating guidance 15 of 81 13

TICAP and ARCAP - Drawing the Line

• TICAP - focused on the LMP Affirmative Safety Case

• ARCAP - more general guidance

• How big is the TICAP box?

- Is every input to the affirmative safety case covered by TICAP?

>> Example - Design Basis External Hazard Levels

- What about chapters which include material fundamental to TICAP plus other stuff?

>> Example - Chapter 1 (General Plant and Site Description and Overview of Safety Case) 16 of 81 14

Construction Permit (CP) Guidance

• TICAP baseline is 10 CFR Part 52 combined license

• Other licensing paths to be addressed

- Design Certification

- 10 CFR Part 50 CP followed by operating license (CP/OL)

• NRC CP decision recognizes the design is generally maturing

- Major concepts and high level performance requirements are established

- Detailed means for achieving the performance requirements may still be evolving

- Finality of the safety case is not required

• TICAP plans to assume a CP/OL applicant which seeks minimal finality at the CP stage 17 of 81 15

Technology Inclusive Content of Application Project (TICAP)

Tabletop Exercises Brandon Chisholm TICAP - NRC Working Meeting August 27, 2020 18 of 81 16

Overview of Tabletop Exercises

• Objectives

- Exercise the TICAP guidance for content of application so that the guidance can be validated and, where necessary, improved

- Provide examples of an affirmative safety case including the use of Principal Design Criteria (PDC) and Complementary Design Criteria (CDC)

- Refine understanding of the broad set of inputs required to produce an affirmative safety case

- Develop feedback for the TICAP team (e.g., information about how decisions were made and how analyses were verified) to assist in the refining of the Guidance Document

• Organization (includes socialization & NRC observation)

- Initiation Phase - Define project structure and organization

- Planning Phase - Define a set of goals for each exercise

- Preparation Phase - Prepare necessary foundation to execute each tabletop exercise

- Execution/Facilitation Phase - Execute each tabletop exercise and produce results

- Wrap-up and Documentation Phase - Refine 19 of 81 results for incorporation into TICAP Guidance Document and sharing 17

Tabletop Exercise Update and Path Forward

• Tabletop Charter Document complete

• Negotiations (e.g., scope and schedule) started with 5 vendors

• Tabletop reports (i.e., final deliverables) will be publicly available

• Vendors support NRC participation in tabletops as observers

- Due to intellectual property, the working meetings will not be public Vendor Design Tabletop Status GE Hitachi PRISM Contract Negotiations +

LMP Demonstration Westinghouse eVinci Contract Negotiations +

LMP Demonstration Kairos KP-FHR Contract Negotiations +

LMP Demonstration X-energy Xe-100 Contract Negotiations +

LMP Demonstration TerraPower Molten Chloride Fast Contract Negotiations 20 of 81 Reactor 18

Technology Inclusive Content of Application Project (TICAP)

Principal Design Criteria (PDC) and Complementary Design Criteria (CDC)

Brandon Chisholm TICAP - NRC Working Meeting August 27, 2020 21 of 81 19

TICAP FSF Chart Required When? Required Safety Functions (RSFs)

Functional Design Criteria SR SSC Design Criteria (SRDC)

How?

(RFDC)

Functions Safety SR SSC SR SSC Special Provided in the Related Performance Treatment Design (SR) SSCs Targets Requirements LBEs from LMP Design Basis Design Basis PRA Safety (AOOs, DBEs, Accidents External Hazard Functions (PSFs) and BDBEs) (DBAs) Levels (DBEHLs)

Input to Design and Frequency- Other Risk Fundamental Consequence Significant Safety Functions and Cumulative Safety Functions Non-SR NSRST SSC (FSFs) NSRST SSC Risk Targets with ST Special Performance (NSRST) Treatment Content of Application Targets SSCs Requirements Other Safety Functions for Adequate DID What? Non-SR Other Safety With No ST Functions SSCs (NST) 22 of 81 How Well?

20

PDC and CDC are answers to How?

Principal Design Criteria (PDC)

Required Required Safety Functional SR SSC Design Functions (RSFs) Design Criteria Criteria (SRDC)

(RFDC)

Frequency-Fundamental Consequence Safety SR SSC SR SSC Special Safety Functions and Cumulative Related Performance Treatment (FSFs)

Risk Targets (SR) SSCs Targets Requirements LBEs from LMP Design Basis Design Basis PRA Safety (AOOs, DBEs, Accidents External Hazard Functions (PSFs) and BDBEs) (DBAs) Levels (DBEHLs)

Other Risk Input to Design and Significant Functions Safety Functions Provided in the Non-SR NSRST SSC NSRST SSC Design with ST Special Performance (NSRST) Treatment Content of Application Targets Other Safety SSCs Requirements Complementary Design Functions for Adequate DID Criteria (CDC)

Non-SR Other Safety With No ST Functions SSCs (NST) 23 of 81 21

Allocating Design Criteria to SR SSCs Safety Case Element Definition Reference Radionuclide (Rn) Starting point for defining the scope of the PRA ASME/ANS RA-S-1.4-2020 Source which includes all Rn sources with the potential (Non-LWR PRA Standard) for producing a risk significant event sequence Fundamental Safety Performance objectives related to the safety NEI 18-04 Function (FSF) functions that are common to all reactor IAEA-TECDOC-1570 Performance Objective technologies and designs (including control heat generation, control heat removal, and confinement of radioactive material)

PRA Safety Function Reactor design-specific SSC functions modeled ASME/ANS RA-S-1.4-2020 (PSF) in a PRA that serve to prevent and/or mitigate a (Non-LWR PRA Standard) release of radioactive material from a specified source or to protect one or more barriers to release Required Safety A PSF that is required to be fulfilled to maintain NEI 18-04 Function (RSF) the consequence of one or more DBEs or the frequency of one or more high-consequence PDC BDBEs inside the F-C Target Required Functional Reactor design-specific sub-functions and NEI 18-04 Design Criteria (RFDC) functional criteria that are necessary and sufficient to meet the RSFs Safety-Related Design Design criteria for SR SSCs (in performing their NEI 18-04 Criteria (SRDC) RSFs) that are necessary and sufficient to fulfill the RFDCs for those SSCs24selected of 81 to perform the RSFs 22

PDC vs. CDC Principal Design Criteria (PDC) - Define plant capabilities that:

• Support demonstration of the performance objectives for the FSFs

• Are credited to perform RSFs for Design Basis Accidents (DBAs)

• Are classified as Safety-Related (SR) with appropriate treatment requirements

• Establish the foundation for making the adequate protection determination Complementary Design Criteria (CDC) - Define plant capabilities that:

• Provide additional means to perform required safety functions

• Are not credited to perform RSFs for DBAs

• Are classified as non-Safety-Related with special treatment requirements appropriate to the functions performed

• Support plant functions related to risk significance or DID as defined in NEI 18-04 25 of 81 23

Introduction to MHTGR Safety Case The Modular High Temperature Gas-Cooled Reactor (MHTGR) safety case (Rev. 3, 1987) predates the LMP approach but can be used to demonstrate aspects of PDC and CDC Tasks related to PDC identification and information flowdown

- Identification of RSFs

- Description of RFDC (i.e., PDC)

- Selection of SR SSCs

- Identification of SR SSC Design Criteria (SRDC)

Tasks related to identification of CDC

- Identification of other PSFs (non-SR) that support SR SSC to perform RSFs (i.e., CDC)

- Description of SSCs performing these other PSFs 26 of 81 24

MHTGR RSFs and SR SSCs 27 of 81 25

MHTGR RSFs and SR SSCs 28 of 81 26

MHTGR PDC Examples Required Required Safety Safety Required Functional Design Criteria (RFDC)

Sub-Functions Function (RSF)

The reactor shall be designed, fabricated, and operated in such a manner that the inherent nuclear feedback characteristics will ensure that the reactor thermal power will not exceed acceptable values.

Additionally, the reactivity control system(s) shall be designed, fabricated, and operated in such a manner that during insertion of reactivity, the reactor thermal power will not exceed acceptable values.

Transfer Heat to A highly reliable, passive means of removing the heat generated in the Ultimate Heat Sink reactor core and radiated from the vessel wall shall be provided. The system shall remove heat at a rate which limits core and vessel temperatures to acceptable levels during a loss of forced circulation.

Conduct Heat from The reactor core shall be designed and configured in a manner that will ensure Core to Vessel Wall sufficient heat transfer by conduction, radiation, and convection to the reactor vessel wall to maintain fuel temperatures within acceptable limits following a loss of forced cooling. The materials which transfer the heat shall be chosen to Control Heat withstand the elevated temperatures experienced during this passive mode of heat Removal removal. This criterion shall be met with the primary coolant system both pressurized and depressurized.

Radiate Heat from The vessel shall be designed in a manner that will ensure that sufficient heat is Vessel Wall radiated to the surroundings to maintain fuel and vessel temperatures within acceptable limits. This criterion shall be met with the primary coolant system in both a pressurized and depressurized condition.

Maintain Geometry The design, fabrication, operation, and maintenance of the core support structure, for Conduction and graphite core and reflectors, core lateral restraint assembly, reactor vessel, reactor Radiation vessel support, and reactor building shall be in such a manner that their integrity is maintained during off-normal conditions so as to provide a geometry conducive to removal of heat from29the of 81reactor core to the ultimate heat sink and maintain fuel temperatures within acceptable limits. 27

Example of SR SSC Satisfying PDC

• SR SSC: Reactor Cavity Cooling System (RCCS)

- Passive reactor cavity cooling system relying on air natural convection to the environment to provide passive core heat removal and protect the vessel and supports

• SRDC for the RCCS

- The RCCS shall have the capability to remove sufficient decay heat from the reactor core to prevent overheating of the outer control rods, the reactor, vessel, and vessel internals.

- The RCCS shall have the capability of removing sufficient decay heat from the reactor core to maintain peak fuel temperatures below 1600°C (2900°F).

- The RCCS shall provide the required decay heat removal capability for the duration of the HTS and SCS shutdown whether the vessel is pressurized (with full primary coolant inventory) or depressurized.

- Offsite radionuclide releases are to be limited as necessary to meet the numerical dose guidelines of the Top-Level Regulatory Criteria.

- In the event of a loss of primary coolant pressure boundary integrity, the RCCS shall be capable of withstanding a 69 kPa (10 psi) differential pressure.

30 of 81 28

MHTGR Requirements Related to CDC

• The MHTGR did not use the NSRST class in SSC safety classification and also did not include an explicit evaluation of DID adequacy - hence the original safety case did not explicitly define any CDC.

• However, the MHTGR did include some requirements for non-safety related SSCs that could be viewed as a surrogate for example CDC.

• The Helium Purification System is not safety related, but is required to have a function to monitor radioactivity circulating in the primary system to confirm performance of the safety related fuel.

- CDC: Monitor radioactivity circulating in the primary system to confirm performance of the Safety Related fuel (PSF associated with DID)

- NSRST SSC: Helium Purification System 31 of 81 29

Introduction to PRISM Safety Case PRISM LMP Demonstration: ADAMS Accession ML19036A584 Tasks related to identification of PDC red = not included in PRISM LMP Demonstration Report

- Identification of RSFs

- Selection of SR SSCs

- Description of RFDC

- Identification of SR SSC design criteria (SRDC)

Tasks related to identification of CDC

- Analysis of other PSF risk significance

>> No SSCs classified at NSRST based upon SSC Risk Significance

- Preliminary evaluation of DID

>> As part of DID evaluation, SSCs identified as32candidates of 81 for NSRST classification 30

PRISM Required Safety Functions Based on four sensitivity cases, the following RSFs were proposed for the PRISM demonstration:

• Reactivity Control

• Heat Removal 33 of 81 31

PRISM Required Sub-functions and SR SSCs Four SSC cases were sufficient to determined that the Five studies were carried selected SR SSCs could be grouped into the following out to identify the following high-level categories:

required sub-functions:

• Digital I&C logic and load drivers (Reactor Protection System [RPS], Diverse Protection System [DPS],

Required Sub-RSF and Safety-Related Qualified Distributed Control and functions Information System [Q-DCIS])

Heat RVACS passive cooling Removal

• Control rods and drives and associated operator Tripping of Primary actions and Intermediate electromagnetic (EM)

• EM pump supply breakers and associated pumps operator actions Reactivity Inherent Reactivity

• 120-VAC equipment Control Feedback

• 125-VDC equipment Control Rod Insertion

• Reactor vessel and internals

• Reactor Vessel Auxiliary Cooling System (RVACS)

• Supporting structures 34 of 81 32

Example of PRISM PDC and Supporting Information

• RSF: Remove Core Heat

• Sub-function: EM pump trip ensures no pump heat is added to decay heat in the Primary Heat Transport System

• RFDC: When a reactor scram occurs, the primary EM pumps shall be tripped such that coolant and fuel temperatures remain within specified acceptable design limits

• SR SSC: EM Pump Supply Breakers

• SRDC: 3 of 4 Primary EM pump supply breakers are tripped open upon receiving a successful trip signal and the neutron flux below TBD%

35 of 81 33

Examples of PRISM CDC [1 of 2]

• Evaluation was conducted to understand how to minimize the frequency of SR SSC challenges to PRISM DID Layer 2 (i.e., control abnormal operation, detect failures, and prevent DBEs)

• Reactor Vessel Auxiliary Cooling System (RVACS) is only challenged after Intermediate Heat Transport System (IHTS) has failed to transport heat to the BOP or Alternate Cooling System (ACS) or when the BOP/ACS fail

• PRA Safety Function (PSF): Transfer heat to Steam Generator (SG)

Alternate Cooling System (ACS)

• NSRST SSCs: SG shell (not including feed water supply and steam supply to turbine), cooling fan, and dampers

- The SG ACS removes SG heat by successful opening the SG ACS inlet and outlet dampers and starting the cooling fan. Power is supplied to the ACS Fan.

- Note: PRISM LMP analysis has not yet been completed to determine if power is needed on a loss of offsite power for this PSF 36 of 81 34

Example of PRISM CDC [2 of 2]

• PRA Safety Function (PSF): Transfer heat to SG ACS

• Sub-function: Prevent a sodium-water reaction following a SG tube rupture from resulting in over-pressure of the Intermediate Heat Transport System (IHTS)

• NSRST SSC: Sodium-Water Reaction Protection System (SWRPS)

- The SWRPS detects a sodium-water interaction, actuates the integrated leak detection system, and actuates the steam isolation, feedwater isolation, water dump valves, steam relief valves and nitrogen purge valves.

>> Note: Not all components listed will be needed for PSF success and would therefore not all will be classified as NSRST

>> Note: Intermediate sodium loop pressure control is by passive means and not initiated by SWRPS 37 of 81 35

Additional Questions?

38 of 81 36

Light-Water Small Modular Reactor Construction Permit Review Guidance August 27, 2020 39 of 81

Light-Water SMR Construction Permit Guidance During the July 31 ARCAP meeting, the staff received the following feedback:

• The staffs guidance options (Interim Staff Guidance (ISG), Draft Strategy, and Office Instruction) are viable options that need further industry consideration.

• Of the three options, the ISG would likely be the most efficient and provide regulatory stability and durability.

• Challenges to developing guidance recognized:

- It may take 6-9 months to receive specific industry input needed in the guidance because of on-going activities for the DOE program.

- The first CP applications could be submitted by the end of 2021/early 2022. Draft guidance with application content or critical areas for a CP application by Spring 2021 could support this schedule.

• Prospective applicants have been encouraged to contact the NRC regarding the application process and the applicants schedule.

40 of 81

Light-Water SMR Construction Permit Guidance (continued)

• Specific public feedback:

- Based on a review of Regulatory Guide (RG) 1.70 and the Standard Review Plan (SRP):

• Clarity needed on Preliminary Safety Analysis Report (PSAR) content vs Final Safety Analysis Report (FSAR) content.

• Alignment needed regarding what specific level of detail is required for preliminary.

• Alignment needed on the applicability of specific [regulatory] requirements

[and staff guidance] identified in the SRP to advanced reactor design features.

- A more efficient approach may be the following:

• Developing a set of Regulatory Framework Documents as an integral part of the [applicants] Regulatory Engagement Plan.

• Using the Regulatory Engagement Plan to support early and often NRC preapplication engagement to gain alignment.

• Obtaining feedback from NRC management in writing that the applicant and NRC can refer to during the review.

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Light-Water SMR Construction Permit Guidance (continued)

The staff is considering developing the ISG to clarify the following:

Regulatory requirements and findings for issuing a CP.

Information needed and level of detail in an application to review and issue the CP.

Specific topics; e.g., siting.

The staff is interested in hearing feedback on:

The draft ISG details above.

Additional topics to consider.

42 of 81

Light-Water SMR Construction Permit Guidance (continued)

Next Steps:

• The staff plans to present additional initial considerations for the ISG during the monthly ARCAP meetings and is interested in hearing feedback on the considerations.

• The staff would like more information to better understand the guidance needs of prospective applicants.

• The staff encourages early engagement to better prepare and plan for a CP review.

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Construction Permit Application Guidance Marc Nichol Senior Director New Reactors August 27, 2020 44 of 81

©2019 Nuclear Energy Institute

Interest in Part 50 Construction Permits Part 50 Construction Permit: 13 of 20 (2 did not respond)*

By EOY 2021 2022-2025 2026 and later Number 2 7 4 Technologies Non-LWR LWR SMR & non- Non-LWR LWR Licensing basis LMP variations, LMP, and LMP, variations, and non-LMP variations and non-LMP Benefits of Part 50

• Earlier start to licensing

• Flexibility for changes during construction 45 of 81

©2019 Nuclear Energy Institute 2

• Member survey results may not include all industry application interest; results here do not show Part 52 interest

Goals for Guidance Minimum scope/level of detail needed for NRC to approve CP Cross-cutting generic issues to clarify acceptable level of details Interim Staff Guidance (likely 15 to 30 pages)

Uses for guidance

• Develop applications

• Predictable acceptance review

• Scope of audits and RAIs 46 of 81

©2019 Nuclear Energy Institute 3

Topics for Guidance (preliminary)

Commitments Analyses Design/SSC descriptions Programs Role of PRA Quality assurance Relationship with Topical Reports Relationship to operating license application Finality Non-applicability or exemption to requirements 47 of 81 Part 50 lessons learned rulemaking topics ©2019 Nuclear Energy Institute 4

Path Forward NEI paper - target December 2020 NRC ISG - target Spring 2021 Applications - potentially before end of year 2021 48 of 81

©2019 Nuclear Energy Institute 5

QUESTIONS?

49 of 81 By Third Way, GENSLER

U.S. Nuclear Industry Council Comments regarding Construction Permit at NRC Public Meeting Cyril Draffin Senior Fellow, Advanced Nuclear U.S. Nuclear Industry Council 27 August 2020 -

50 of 81

Construction Permit: NRC considered prior input

• USNIC appreciates NRC considering Stakeholder comments made at 31 July 2020 meeting regarding Light-Water SMR Construction Permit:

o Interim Staff Guidance (ISG) would likely be the most efficient and provide regulatory stability and durability o First CP applications could be submitted within the next year. Draft guidance with application content or critical areas for a CP application by Spring 2021 could support this schedule.

o Developing a set of Regulatory Framework Documents as an integral part of the applicants Regulatory Engagement Plan, and using the Regulatory Engagement Plan to support early and NRC preapplication engagement to gain alignment.

o Obtaining feedback from NRC management in writing that the applicant and NRC can refer to during the application review.

1 l U.S. Nuclear Industry Council Aug 2020 - Construction Permit 51 of 81

Construction Permit: Interim Staff Guidance o Regulatory requirements and findings for issuing a CP o Information needed and level of detail in an application to allow timely review and issuance of CP 2 l U.S. Nuclear Industry Council Aug 2020 - Construction Permit 52 of 81

Construction Permit - Considerations

• Construction Permit (CP) guidance that appropriately considers elements of LMP/TICAP/ARCAP for vendors who may use LMP

• NRC has noted that non-LWR applicants not using TICAP should engage NRC early regarding safety classification and Defense in Depth; Part 50 and Part 52 both viable regulatory pathways

• NRC needs to ensure that Structures, Systems, and Components (SSCs) are selected appropriately; consultation with NRC staff on classification approach is advised

• Construction Permit guidance should consider stakeholder input, be timely, and have stability and predictability

• NRC should not sweep non-safety issues into applications 3 l U.S. Nuclear Industry Council Aug 2020 - Construction Permit 53 of 81

Construction Permit: Timing & Planning Considerations

• DOE selection of demo winners may influence and accelerate siting and construction permit application timing

• Need NRC guidance as soon as possible before applications:

• draft NRC guidance on critical areas by end of 2020

• draft guidance by spring 2021

• NRC should prepare Interim Staff Guidance that includes both LWR SMRs and non-LWRs

• NRC should indicate when and what kind of input they would like from industry on ISG 4 l U.S. Nuclear Industry Council Aug 2020 - Construction Permit 54 of 81

Proposal for Advanced Reactor Content of Application Project (ARCAP) Guidance Document 55 of 81

Background

• ARCAP Proposed Guidance document would provide a roadmap for developing an application

• Roadmap would leverage existing guidance or guidance that is under development

• Examples include:

• Technology Inclusive Content of Application Project (TICAP) developing portions of the application associated with the Licensing Modernization Project (LMP)

• Emergency planning and security rulemaking will provide insights to this portion of the application

• Never the intention of the ARCAP guidance document to attempt to replicate the Standard Review Plan for Light Water Reactors (NUREG-0800) 56 of 81

Background

• Figure provides an overview of some of the more important efforts underway to develop advanced reactor guidance

• TICAP will use the LMP (upper left of figure) to develop portions of the application 57 of 81

Proposal for ARCAP Guidance Document

• High level ARCAP proposal found in document referenced in meeting notice (ADAMS Accession No. ML20231A563)

• Proposed guidance includes table providing roadmap

• Table based on Idaho National Laboratory (INL) developed annotated structure for final safety analysis report (FSAR) portion of the application

• INL developed outline discussed in previous ARCAP meetings and can be found at ADAMS Accession No. ML20107J565

• Recognized that the TICAP FSAR proposed structure is different than INL-developed structure

• Table will be updated based on final version of TICAP structure 58 of 81

Proposal for ARCAP Guidance Document

• High level ARCAP proposal found in document referenced in meeting notice (ADAMS Accession No. ML20231A563) (continued) o First 14 items in table associated with FSAR, the rest of the items in the table associated with other portions of the application o Table color coded to note where proposed guidance would :

point to guidance that is being developed as part of another advanced reactor activity (e.g., TICAP)

Note where new ARCAP guidance is being developed Note where a combination of new ARCAP guidance is being developed and provide pointers to guidance that is being developed as part of another advanced reactor activity (e.g., TICAP, rulemaking etc.)

59 of 81

Proposal for ARCAP Guidance Document 60 of 81

Proposal for ARCAP Guidance Document 61 of 81

Proposal for ARCAP Guidance Document 62 of 81

ARCAP Chapters Under Consideration for a Performance-Based (PB) Approach (i.e., Approach 3) 63 of 81

Background

• In the July 31, 2020 ARCAP meeting, NRC provided additional details on a potential PB approach (Approach 3) for ARCAP Chapter 8, Control of Routine Plant Radioactive Effluents and Solid Waste. At the same meeting, industry suggested siting, EP and security as candidates for a PB approach.

• At the present time, the following ARCAP chapters are under consideration for a more PB approach:

- Chapter 2, Site Information

- Chapter 8, Section 8.3, Solid Waste

- Chapter 9, Control of Occupational Dose

• Since there are rulemaking activities underway for EP and security that may incorporate PB approaches, work on these topics will be dependent upon the outcome of the rulemakings.

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ARCAP Chapter 2, Site Information

• 10 CFR 100, Subpart B, requires that site characteristics be determined in order to establish (1) the external hazards (man-made and natural) the plant must be designed for, (2) the hydrological radionuclide transport properties, (3) if the site poses a significant impediment to EP and (4) that the individual and societal risk of potential accidents is low.

• Much of the above information is contained in Chapter 2 of the SAR, with the result that the chapter becomes very large. For example, the SARs contain information on historical records of the site (such as floods, temperatures, seismic events, etc.) as well as the results of recent site characterization work (e.g. meteorology, core samples).

• It is recognized that TICAP Chapter 1 will address siting, however, until TICAP Chapter 1 is provided, weve considered work on ARCAP Chapter 2 to identify areas where the amount of information that is required to be in the SAR might be reduced.

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ARCAP Chapter 2 - Changes Being Considered

• What is being considered is using the guidelines in NEI 98-03 Guidelines for Updating FSARs (endorsed by RG 1.181), developed to identify areas where information can be removed from FSARs, as the starting point for determining if it was needed in the first place. Examples include:

- Historical information (floods, storms, etc.)

- Information not expected to change with time (geological data, seismic data, etc.)

- Redundant information

- Excessive detail

• The intent is to limit the amount of material in SAR Chapter 2 to that which is necessary for establishing safety significant design parameters and performing the safety analysis, along with its supporting bases.

• If necessary, any additional supporting information (e. g. historical records, geological data) could be documented in a separate report available for audit.

• Note: Site population density considerations are dependent on Commission action on SECY-20-045, Population-Related Siting Considerations for Advanced Reactors and are not 66included of 81 in the ARCAP Chapter 2 work at this time.

ARCAP Section 8.3 and Chapter 9

- Overview

• Continue to develop performance-based guidance for additional non-TICAP safety analysis report chapters

- Section 8.3, Solid Waste

- Chapter 9, Control of Occupational Dose

• Related to the two performance-based content areas above, address continued applicability of NEI developed FSAR content templates:

- NEI 07-10A, Generic FSAR Template Guidance for Process Control Program (PCP)

- NEI 07-08A, Generic FSAR Template Guidance for Ensuring that Occupational Radiation Exposures are as Low as is Reasonably Achievable (ALARA) 67 of 81

ARCAP Section 8.3, Solid Waste

• Develop using same approach as Sections 8.1 and 8.2

• Reference applicable requirements for performance-based criteria, such as:

- 10 CFR 20.1301(a) regarding the allowable annual dose and allowable hourly dose to members of the public from routine operation

- 10 CFR 20.1301(e) regarding compliance with EPA's generally applicable environmental radiation standards in 40 CFR part 190

- 10 CFR Part 61 as it relates to requirements for classifying, processing, and disposing of dry solid and wet wastes

- 10 CFR 20.2006 and Appendix G to 10 CFR Part 20, as they relate to the requirements for transferring and manifesting radioactive materials shipments to authorized facilities (e.g., disposal sites, waste processors)

- 10 CFR 20.2007, as it relates to compliance with other applicable Federal, State, and local regulations governing any other toxic or hazardous properties of radioactive wastes, such as mixed wastes

- 10 CFR Part 71 and 49 CFR Parts 171-180, as they relate to the use of approved containers and packaging methods for the shipment of radioactive materials

- 49 CFR 173.443, as it relates to methods and procedures used to monitor for the presence of removable contamination on shipping containers, and 49 CFR 173.441, as it relates to methods and procedures used to monitor external radiation levels for shipping containers and vehicles 68 of 81

ARCAP Section 8.3, Solid Waste (cont.)

• Develop Acceptance Criteria - System Design, such as:

• Provide a high-level description of the solid waste management system (SWMS)

• Describe expected sources of waste

• Describe equipment design capacities for expected waste volumes and radioactivity inventories of Class A, B and C waste

• Describe design provisions to control and collect any solid waste spillage from equipment malfunction or puncture of waste containers 69 of 81

ARCAP Section 8.3, Solid Waste (cont.)

• Develop Acceptance Criteria - Operational Controls, such as:

- Provide a description of operational controls for waste processing and surveillance requirements which assure that:

• Allowable doses to members of the public remain within required levels

• The final waste product meets the requirements of applicable Federal, State and disposal site waste form requirements for burial at a 10 CFR 61 licensed Low-Level Waste (LLW) disposal site

- As an option, applicant may refer to NEI 07-10A, Generic FSAR Template Guidance for Process Control Program (PCP)

• If an applicant chooses to reference this template to address the above acceptance criteria no need to replicate text in the FSAR; may need to update/revise template to reflect operation of specific non-LWR 70 of 81

ARCAP Chapter 9, Control of Occupational Dose

• Develop using same approach as Chapter 8

• Address applicability to:

- Part 50 operating license and construction permit applications

- Part 52 design certification and combined license applications

- Non-LWRs and small modular LWRs

• Reference applicable requirements for performance-based criteria, such as:

- 10 CFR 19.12, as it relates to keeping workers informed who receive occupational radiation exposure (ORE)

- 10 CFR 20, Subpart C, Occupational Dose Limits (20.1201 -

20.1208)

- 10 CFR 20.1101 and the definition of ALARA in 10 CFR 20.1003, as they relate to those measures that ensure that radiation exposures resulting from licensed activities are below specified limits and ALARA 71 of 81

ARCAP Chapter 9, Control of Occupational Dose (cont.)

• Develop Acceptance Criteria - System Design, such as:

- Describe important equipment and facility design features used to ensure that occupational radiation exposures are ALARA such as, shielding, ventilation, area radiation and airborne radioactivity monitoring instrumentation and dose assessment

- Describe the design features provided to control access to radiologically restricted areas (including potentially very high radiation areas) and describe each very high radiation area and indicate physical access controls and radiation monitor locations for each of these areas

- Describe those features that reduce the need for maintenance and other operations in radiation fields, reduce radiation sources in areas where operations may be performed, allow quick entry and easy access, provide remote operation capability, or reduce the time spent working in radiation fields, as well as any other features that reduce radiation exposure of personnel

- Describe methods for reducing the production, distribution, and retention of activation products through design, material selection, water chemistry, decontamination procedures, and so forth 72 of 81

ARCAP Chapter 9, Control of Occupational Dose (cont.)

• Develop Acceptance Criteria - Operational Controls, such as:

- Provide commitments to develop comprehensive worker protection programs, organizational structure, training and monitoring to ensure 10 CFR 19 and 10 CFR 20 requirements are met. Include commitments to any relevant regulatory guides, NEI templates, or standards

- As an option, applicant may refer to NEI 07-08A, Generic FSAR Template Guidance for Ensuring that Occupational Radiation Exposures are as Low as is Reasonably Achievable (ALARA)

• If an applicant chooses to reference this template to address the above acceptance criteria no need to replicate text in the FSAR; may need to update/revise template to reflect operation of specific non-LWR

- These criteria for operational controls could also be addressed in the Radiation Protection Program with a reference in the FSAR 73 of 81

U.S. Nuclear Industry Council Comments regarding ARCAP at NRC Public Meeting Cyril Draffin Senior Fellow, Advanced Nuclear U.S. Nuclear Industry Council 27 August 2020 74 of 81

ARCAP has better direction

• USNIC appreciates NRC considering industry ARCAP comments:

• NRC providing acceptance criteria, eliminating unnecessary material, adding flexibility for application depending on design and technology, making ARCAP technology-inclusive, and indicating how to consider exemptions

• NRC not requiring duplication of information--applicant can reference other documents

• Because there will be more reliance on industry actions then application details, important to consider what NRC oversight & inspection is needed so monitoring does not become too arduous, 1 l U.S. Nuclear Industry Council Aug 2020 - Advanced Reactors Content of Applications (ARCAP) 75 of 81

ARCAP NRC input to NEI on 17 August 2020

• NRC seems to be taking appropriate steps based in industry feedback from the 31 July 2020 meeting:

o Limited set of new guidance will be developed as part of ARCAP. ARCAP new guidance will be technology inclusive, to the maximum extent possible, so a light water or non-light water reactor applicant can use the guidance if they desire.

o ARCAP was never intended to be a comprehensive replacement or reiteration for all regulatory guidance for large light water reactors (e.g., NUREG-0800, other regulatory guidance) o Developed a roadmap document describing that ARCAP will provide high-level guidance that will provide pointers to advanced reactor guidance that is under development (e.g., TICAP, security and emergency planning rulemaking) and provide additional guidance (including in appendices) for areas that are not being addressed under an advanced reactor activity.

o Table provided listing of portions of the application and how the ARCAP would address the guidance seems reasonable.

2 l U.S. Nuclear Industry Council Aug 2020 Advanced Reactors Content of Applications (ARCAP) 76 of 81

Re-iteration of input on ARCAP

• Process should be clear, risk informed and consistent with NRC safety goals

• Reduce unnecessary burden, particularly where there is no nexus to safety--

focus on areas for elimination in a risk-informed review is appropriate

• Prompt elevation and expedited resolution of policy issues is needed

• Commissioners need to be fully engaged, recognizing a license application is being reviewed by the NRC and multiple license applications will be forthcoming in 2021-22

• If NRC will not have separate guidance on microreactors make sure approach considers their characteristics

• Outcome must be transformative

• ARCAP needs to provide a clear benefit to near and long term applicants 3 l U.S. Nuclear Industry Council Aug 2020 - Advanced Reactors Content of Applications (ARCAP) 77 of 81

Relation of ARCAP to Part 53

• ARCAP should not be the default basis for Part 53-- but ARCAP could provide elements that could be used in Part 53

• ARCAP is a bridge not a destination

• In near term could focus on key issues in ARCAP (perhaps with ISG), and that work could used in Part 53

• Making Part 53 more applicable to a wider variety of technologies will benefit staff and industry

• Must have clear high level requirements, fewer exemptions, and less iteration on existing rules 4 l U.S. Nuclear Industry Council Aug 2020 - Advanced Reactors Content of Applications (ARCAP) 78 of 81

Other input

• Be technology-inclusive

• Streamline near term application reviews to define appropriate scope and level of detail

• For ARCAP, NRC staff could indicate what they will do for siting (ARCAP Chapter 2) and physical security (e.g. role of security rulemaking)

• NRC need to define how and why requested information is used to make a regulatory decision

• USNIC looks forward to ongoing constructive engagement with Commission and Staff to develop regulatory review tools for advanced reactor deployment 5 l U.S. Nuclear Industry Council Aug 2020 - Advanced Reactors Content of Applications (ARCAP) 79 of 81

U.S. Nuclear Industry Council Contacts For questions contact Cyril W. Draffin, Jr. Jerey S. Merri"eld Senior Fellow, Advanced Nuclear, Chairman, US Nuclear Industry Council U.S. Nuclear Industry Council Advanced Reactors Task Force Cyril.Draffin@usnic.org U.S. NRC Commissioner (1998-2007)

Je.Merri"eld@pillsburylaw.com 6 l U.S. Nuclear Industry Council Aug 2020 80 of 81

Future Meetings 2020 Tentative Schedule for Technology Inclusive Content of Application Project Public Meetings September 24 October 22 November 19 81 of 81