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{{#Wiki_filter:Regulatory Frameworks for Nuclear Safety Audience: The Slovenia Nuclear Innovation Seminar By: Michael I. Dudek, Chief of the New Reactor Licensing Branch Location: Virtual Presentation Date: Tuesday, October 19, 2021
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Objectives
* Provide an understanding of the NRCs licensing basis for U.S. nuclear power plants (AP1000)
* Provide an understanding of how the licensing basis may be changed by the licensee with NRC approval.
* Provide an understanding of the licensing basis for reviewing and approving SMRs and Advanced Reactor designs.
2
 
Vogtle Reactor Vessel 3
 
NRC Licensing Requirements and Documents
 
Purpose
* The purpose of this section is to provide an understanding of important NRC licensing requirements and documents associated with licensing nuclear power plants in the U.S.:
NRC Regulations Design certifications (DCs)
Combined licenses (COLs)
Technical specifications (TS)
Final safety analysis report (FSAR)
* The section also focuses on the license and the process for amending it 5
 
Regulations Governing Licensing of Nuclear Power Plants
* Title 10 Code of Federal Regulations (10 CFR)
Part 50, Domestic Licensing of Production and Utilization Facilities - Fundamental regulations under whose terms commercial nuclear power plants operate 6
 
Part 50 Licenses
* Operating license based on as-built plant ITAAC not necessary
* Operating license does not reference design certification License amendments do not require exemptions or rule changes 7
 
Regulations Governing Licensing of Nuclear Power Plants
* 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants - More recent regulations that provide an alternative path to licensing
* Plants licensed under Part 52 are also subject to much of Part 50
* Todays discussion uses an example of a plant licensed under Part 52 8
 
Part 50 vs. Part 52 Licensing 2-step licensing - Part 50                  Alternative licensing - Part 52 Preliminary Safety Analysis Report          Early Site Permit          Design Certification Construction Permit                              Site/design information Construction and further design                      Final Safety Analysis Report Final Safety          Construction                      Combined License Analysis Report          complete Construction Operating License Verification (ITAAC)
Operation*
                          *Includes Pre-operational testing  Operation*
9
 
Design Certification
* 10 CFR 52 Subpart B
* Allows NSSS* vendor or applicant to obtain pre-approval of design through rulemaking Reduces licensing uncertainty by resolving design issues early Facilitates standardization High degree of regulatory finality
  *Nuclear Steam Supply Systems 10
 
Design Certification
* A combined license applicant is not required to reference a certified design, though likely will do so
* An Early Site Permit (on a specific site) may be referenced by a COL applicant
* Certified design is an appendix to Part 52; i.e.
design information is a regulation - Hard to change or to challenge it 11
 
Design Certification
* The AP-1000 (subject of todays discussion) design is documented in Appendix D to Part 52, Design Certification Rule for the AP1000 Design
* Appendix D contains Design Descriptions and ITAAC (Inspections, Tests, Analyses, and Acceptance Criteria) 12
 
Combined License (COL)
* License to construct and operate a nuclear power plant; and to possess nuclear fuel
* Governed by 10 CFR 52 Subpart C, Combined Licenses
* Intended to reduce financial risk to applicants
* Findings of Reasonable Assurance:
The completed facility will comply with Atomic Energy Act requirements and NRC regulations The completed facility will not be inimical to common defense and security, or to health and safety of the public 13
 
Combined License (COL)
* License Conditions Startup testing ITAAC closure Site conditions Operational programs
* Technical Specifications in Appendix A to the COL
* Design Descriptions and ITAAC in Appendix C to the COL 14
 
Technical Specifications (TS)
* Part of the operating license
* Protect the health and safety of the public by imposing limits, operating conditions, and other similar requirements on the facility
* Define the limits of plant operation to ensure that the plant is operated within those boundaries established by the Safety Analysis
* Establish conditions or limitations that obviates the possibility of an abnormal situation or event
* Cannot be changed without prior NRC approval 15
 
Design Description and ITAAC (COL, Appendix C)
* Design Description means that portion of the design that is certified
* ITAAC are the inspections, tests and analyses which must be performed; and acceptance criteria which must be met; before a Part 52 COL plant can enter commercial operation 16
 
Design Description and ITAAC (COL, Appendix C)
* ITAAC provide assurance that plant is constructed per its design ITAAC derived from:
Design Certification Site specifics Licensee must close all ITAAC before fuel load 17
 
ITAAC and 10 CFR 52.103(g)
* For a Part 52 licensee, the NRC finding under 10 CFR 52.103(g) is required for the licensee to conduct pre-critical testing of the newly constructed reactor
* Finding that the acceptance criteria (ITAAC) in the combined license are met
* Once the Commission finds the ITAAC have been met, the licensee may proceed with the testing and power ascension program 18
 
Exemption Requirement
* All of the Design Descriptions, and most of the ITAAC in Appendix C of a Part 52 COL, are derived from the certified design
* Because this information is in Appendix D of Part 52, it is a federal regulation and referred to as Tier 1
* Therefore, any deviation from this information requires an exemption to Part 52
* Changes to Tier 2 information may not require an exemption 19
 
Final Safety Analysis Report
* Principal purpose of the final safety analysis report (FSAR) is to describe:
The nature of the plant The plans for its use Safety analyses performed to evaluate whether plant can be constructed and operated without undue risk to the health and safety of the public 20
 
Final Safety Analysis Report
* FSAR - Principal document for describing the licensing basis The basis for conclusions on protection of health and safety of the public The basis for issuance of the license
* Basic document used by NRC inspectors to determine whether the facility is being constructed and operated within the licensed conditions 21
 
FSAR Content
* 10 CFR 50.34(a) and 52.79 describe minimum information required in FSAR
* Scope and level of detail varied based on applicable guidance at time of application
* Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, provides standard guidance
* Regulatory Guide 1.206 provides similar guidance for reactors licensed under Part 52 22
 
Amending a License
* When a holder of a license, including a construction permit and operating license under 10 CFR Part 50, and an early site permit, combined license, and manufacturing license under 10 CFR part 52, desires to amend the license or permit, the licensee must submit a license amendment request in accordance with 10 CFR 50.90 23
 
10 CFR 50.59 and 10 CFR 50.90
* 10 CFR 50.59 requires submittal of a license amendment request (LAR) in accordance with 10 CFR 50.90 for any change to facilities or procedures described in the FSAR that:
Results in more than a minimal increase in the frequency or consequence of an accident or malfunction Creates a possibility for an accident or malfunction of a different type; or Results in exceeding or altering a design basis limit for a fission product Changes a method of evaluation used in establishing the design bases or in the safety analyses 24
 
Other Change Control Processes
* Various requirements pertain to changing other mandated safety-significant programs and documents; for example:
Security plan - 10 CFR 50.54(p)
Emergency plan - 10 CFR 50.54(q)
* Changes may or may not require prior NRC approval For example, licensee may change security plan without prior NRC approval as long as change does not decrease effectiveness as defined in guidance 25
 
Vogtle Construction Site 26
 
NRC Licensing of Small Modular Reactors (SMRs)
 
Purpose
* The purpose of this section is to provide an understanding of the differences that the NRC is encountering with licensing SMRs (small light-water reactor technology)
Modular designs Nuclear safety review areas Environmental review areas SMR review challenges SMR lessons-learned 28
 
Differences in Licensing SMRs
* In general, the same regulatory framework applies to SMRs as to large LWRs; however, gaps may exist
* Gaps in regulatory and guidance documents are a continually evolving challenge
* Gaps change according to changes in technologies and approaches
* Addressing gaps in a completely technology-neutral, meaningful way is difficult 29
 
Differences in Licensing SMRs
* Not all gaps warrant the same amount of NRC focus or resources Safety or security significance Likelihood of guidance in that area actually being needed Timing of the need
* The NRC is always scanning the horizon for future gaps and prioritizing them 30
 
Nuclear Safety Review Areas
* Site Characteristics and Site Parameters
* Systems, Structures, Components, and Equipment Design
* Reactor Internals
* Reactor Coolant and Connected Systems
* Engineered Safety Features
* Digital Instrumentation and Controls/Electrical Power
* Auxiliary Systems
* Steam and Power Conversion Systems
* Radioactive Waste Management and Radiation Protection
* Conduct of Operations
* Initial Test Program and ITAAC
* Transient and Accident Analysis
* Technical Specifications
* Quality Assurance Program
* Human Factors Engineering
* Severe Accidents 31
 
Environmental Review Areas
* Seismology
* Geology
* Hydrology
* Meteorology
* Geography
* Demography (population distribution)
* Site Hazards Evaluation
* Radiological Effluent Releases
* Radiological Dose Consequences
* Emergency Preparedness (with FEMA)
* Security Plan Feasibility 32
 
SMR Review Challenges
* New first-of-a-kind designs Modular construction Completeness of the design Applicability of existing regulations Verification and validation of computer codes Applicability of prior operating experience
* Smaller applicant organizations in some cases Important that they have experience with nuclear licensing requirements Important that they are adequately staffed to address regulator questions in a timely manner during reviews 33
 
Modular Designs - AP1000 Modules CA01  Steam Generator / Refueling Channel CA02  In-Containment Refueling Water Storage Tank (IRWST) Interior Wall CA03  IRWST Perimeter Wall CA04  Reactor Vessel Cavity CA05  Access Tunnel / Passive Core Cooling and Volume Control System Equipment Room Wall CA20  Aux Building 34
 
SMR Lessons-Learned
* A rigorous safety basis and understanding is still essential for new designs
* If an organization is mostly familiar only with large light water reactors (LWRs), a mindset change may need to occur
* Sustained focus on the most safety-significant aspects of the design is important
* Complete understanding of the design as a whole assists the applicant and the regulator
* Successful completion of first-of-a-kind reviews will result in streamlined Nth-of-a-kind reviews
* Early and frequent communication with applicants is crucial, both prior to and during the regulators review 35
 
NRC Licensing of Advanced Reactors
 
Purpose
* The purpose of this section is to provide an understanding of the differences that the NRC is encountering with licensing Advanced Reactors (small non-light water reactor technology)
Advanced Reactor design landscape Ensuring readiness for licensing Advanced Reactor designs 37
 
New Non-Light Water Licensing Companies in active pre-application engagement:
X-Energy TerraPower & GE-Hitachi Oklo Aurora - first  Kairos Power advanced reactor Terrestrial Energy USA licensing application submitted March    General Atomics Energy 2020. Microreactor -  Westinghouse Electric Company 1.5 MWe 38
 
Advanced                        Reactor Broad Landscape of Advanced Reactor Designs Landscape Liquid Metal Cooled Fast    High-Temperature Gas-Cooled    Molten Salt Reactors              Micro Reactors                        Reactors                  (MSR)                  Reactors (LMFR)                          (HTGR)
TerraPower/GEH (Natrium)*                                                Kairos
* Westinghouse (eVinci)*
X-energy
* GEH PRISM (VTR)                                        Kairos (HermeslRTR)        BWX Technologies Framatome Advanced Reactor                                        Liquid Salt Cooled            X-energy Concepts                    StarCore Radiant l RTR Sodium-Cooled                    MIT Terrestrial
* Transportable Westinghouse                                              TerraPower TRISO Fuel                                            Ultra Safe Columbia Basin                                    Southern (TP MCFR) lRTR    University of Illinois RTR Hydromine            General Atomics (EM2)*            ACU lRTR
* Oklo Lead-Cooled              General Atomics                  Elysium Stationary ThorCon ARDP Awardees                                              Muons LEGEND Demo Reactors          In Licensing Review                  Flibe Risk Reduction
* Preapplication                    Alpha Tech RTR  Research/Test Reactor          Liquid Salt Fueled ARC-20 39
 
Ensuring Readiness for Licensing 40
 
Ensuring Readiness for Licensing
* Risk-Informed Performance-Based Licensing Approaches Draft Regulatory Guide (DG) 1353
* Technology Inclusive Policy Issues Functional Containment Security Emergency Preparedness
* Nuclear Energy Innovation and Modernization Act (NEIMA) 41
 
Questions?}}

Revision as of 19:32, 16 January 2022

Slovenia Nuclear Innovation Seminar, Regulatory Frameworks for Nuclear Safety- MID-FINAL
ML21293A041
Person / Time
Issue date: 10/19/2021
From: Michael Dudek
NRC/NRR/DNRL/NRLB
To:
Dudek M
References
Download: ML21293A041 (42)
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