ML21042B889
| ML21042B889 | |
| Person / Time | |
|---|---|
| Site: | Nuclear Energy Institute |
| Issue date: | 02/11/2021 |
| From: | Nichol M Nuclear Energy Institute |
| To: | Coyne K Office of Nuclear Material Safety and Safeguards |
| SECY/RAS | |
| References | |
| 85FR71002, NRC-2019-0062, PR-53 | |
| Download: ML21042B889 (46) | |
Text
MARCUS R. NICHOL Senior Director, New Reactors 1201 F Street, NW, Suite 1100 Washington, DC 20004 P: 202.739.8131 mrn@nei.org nei.org February 11, 2021 Dr. Kevin Coyne Acting Director, Division of Rulemaking, Environmental, and Financial Support Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Industrys Concerns about NRC Proposed Approaches to Part 53, and Alternative Discussion Draft for the NRCs Rulemaking on, Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN-3150-AK31; NRC-2019-0062)
Project Number: 689
Dear Dr. Coyne:
The Nuclear Energy Institute (NEI)1 and its members appreciate the Nuclear Regulatory Commissions (NRC) staffs approach to develop preliminary rule language to facilitate discussion with stakeholders on the concepts for the Part 53 rule. To date, the staff has issued preliminary rule language for safety criteria, design and analysis, facility safety program and siting, which has been helpful in identifying and understanding potential approaches for Part 53 among all stakeholders. We also appreciate the NRCs public meetings, which have helped to foster stakeholder discussion on the potential approaches. We remain dedicated to providing detailed input on our perspectives in order to support the staffs efforts in a timely manner. However, we are concerned that what the staff has proposed is overly complex and not consistent with the goal to have a more efficient regulatory framework, and that the rulemaking could result in a Part 53 rule that would be an unattractive option to license advanced reactors.
The purpose of this letter is to provide the NRC with a discussion draft of a Part 53 rule that was developed with NEIs Part 53 Task Force (Attachment 1). This task force is comprised of members from advanced reactor companies, utilities and vendors. We believe that the attached draft would achieve the industrys vision, provided in our letter dated October 21, 2020, that licensing new reactors under the new Part 53 1 The Nuclear Energy Institute (NEI) is responsible for establishing unified policy on behalf of its members relating to matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEIs members include entities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect and engineering firms, fuel cycle facilities, nuclear materials licensees, and other organizations involved in the nuclear energy industry.
Dr. Kevin Coyne February 11, 2021 Page 2 rule will be the most efficient option for all new reactor applicants and will meet industry needs for schedule, cost and predictability.
We have several concerns, discussed in Attachment 2, with the NRCs proposed approaches and have provided alternatives that we believe would continue to achieve the NRCs high levels of safety for new reactors in a more modern, efficient, technology-inclusive, performance-based and risk-informed manner.
While the NRC staff has acknowledged receipt of the input we provided in our December 23, 2020 letter and in presentations in public meetings, the NRC staff has yet to provide substantive feedback on whether these approaches could be acceptable for the Part 53 rule. The lack of feedback makes it difficult for us to improve our proposed alternatives to address NRC concerns and to develop additional details for our ideas for Part 53. Furthermore, in the NRCs public meeting on February 4, 2021 the NRC staff appeared to not be receptive to alternative approaches to the rulemaking. For example, in response to stakeholder input on alternative approaches to address quantitative health objectives (QHO) and defense-in-depth (DID), the staffs response was that the NRC proposed approach is the only approach that could be taken in Part 53. It appears that the NRC staff has decided the approach to Part 53 before fully considering stakeholder input, and the NRCs position is not supported by the fact that the NRC proposed approaches to QHO and DID are different than what is done in Parts 50 and 52, and thus alternative approaches are obviously possible.
A successful Part 53 is very important to the industry and holds the promise of enabling future reactor licensing to be much more safety-focused and efficient.
The attached discussion draft provides the scope and approach to Part 53 requirements that we believe would 1) fulfill the NRCs statutory requirements for the regulation of new reactors, 2) implement the relevant provisions of the Nuclear Energy Innovation and Modernization Act (NEIMA), and 3) align with the vision and goals we identified in our October 21, 2020 letter and the success criteria in our December 23, 2020 letter. The discussion draft is arranged according to the following regulatory functions associated with the NRCs regulation of new reactors:
1.
Determine the standards for reasonable assurance of adequate protection, and extra-adequate protection, in terms of direct impacts on the of the public health and safety. (§53.2 and §53.3 in the discussion draft) 2.
Determine the facility characteristics that are needed to ensure the facility will meet the public protection criteria. (Roughly §53.4 in the discussion draft) 3.
Determine how to provide reasonable assurance that the facility is designed to achieve the facility safety characteristics. (Roughly §53.5 in the discussion draft) 4.
Determine how to provide reasonable assurance that the facility is constructed and will operate in accordance with the design. (Roughly §53.16 through §53.26 in the discussion draft) 5.
Determine the types of licenses, permits and design approvals that the NRC may issue, and the process for obtaining them. (Roughly §53.33 through §53.39 in the discussion draft)
Dr. Kevin Coyne February 11, 2021 Page 3 6.
Determine the information from an applicant or holder of a license, permit or design approval that needs to be reviewed and approved by the NRC to facilitate the regulation of the facility. (Roughly
§53.39 through §53.44 in the discussion draft)
Consistent with our earlier recommendation to follow a systematic approach to developing the Part 53 rule, this discussion draft incorporates our efforts to address Step 2 - Establish the scope of the rule, many of the elements of Step 3 - Create the safety paradigm, and some of the elements of Step 4 - Identify how to document the regulatory framework. The discussion draft represents our current thinking, not a final position, and we expect it to evolve based on NRC feedback and further stakeholder discussions.
The following key topics were identified through development of the discussion draft to have a profound influence on the outcome of the rule and which would benefit from more detailed discussions:
Safety criteria and safety paradigm - The paradigm for safety and security in Part 53 influences the structure and details of nearly all of the requirements. We approached Part 53 by following the NRCs safety and security paradigm in what is called the bow tie figure (§53.4), and incorporated the latest approaches to safety that have been identified in NEI 18-04 and other documents. There are a number of sub-topics in this area that should be included in this discussion, for example, the standards for adequate protection and extra-adequate protection, role of the quantitative health objectives, role of defense-in-depth, commercial quality standards, and ALARA.
Role of Probabilistic Risk Assessments (PRA) - Stakeholders have a diverse perspective on the role of PRA in Part 53, and we proposed a flexible requirement that allows a diverse set of approaches. We believe this offers significant advantages as compared to the NRCs proposal for a detailed and prescriptive set of specific PRA requirements. A topic for further discussion is whether guidance on a graded approach to risk evaluations would be helpful, and a better place to locate more detailed expectations for meeting the performance-based requirement.
Performance-based safety, security, siting and emergency preparedness - We approached Part 53 by approaching the safety, security, siting and emergency planning of the facility in a holistic manner, which enables significant opportunities to protect the public health and safety more efficiently. Pursuit of performance-based approaches requires rethinking the terms and constructs as they exist and innovating on a new paradigm to provide an equivalent level of protection of the public health and safety.
Organization of documentation and technical requirements - We approached Part 53 by separating the technical performance requirements from the documentation requirements to provide greater clarity in the rule language. We recognize that this is a departure from the approach in Part 50 and 52 to embed technical performance criteria in the application content requirements; however, the benefit of this separation is greater clarity in the regulations.
Level of detail in regulations and use of guidance - The requirements in the discussion draft are necessarily written at a high level in order to achieve a technology-inclusive, performance-based and risk-informed approach. These high-level requirements will place more emphasis on the use of guidance to define details historically included in regulations as they apply to specific technologies
Dr. Kevin Coyne February 11, 2021 Page 4 or licensing approaches. Stakeholder discussions may identify that, in some areas, a more detailed and prescriptive requirement to provide greater clarity and stability in the rule is preferred over a higher-level requirement that provides flexibility. These discussions are also expected to identify areas where guidance is needed to accompany the rule.
Relationship with Part 50 and 52 licensing processes - We approached Part 53 by referencing existing requirements where they are expected to align with the proposed technology-inclusive, performance-based and risk-informed approach, and including alternative requirements in Part 53 where modification is necessary. However, further discussion may determine more efficient approaches.
We believe that sharing our ideas on the complete set of Part 53 requirements early in the process will 1) help to better inform the NRCs proposed rule language, 2) facilitate discussion of key topics, and 3) allow more time to work on guidance to accompany the rule, and 4) better enable an efficient and useable rule in the timeframe established by the Commission. These ideas also provide feedback on the preliminary rule language released by the NRC to-date by proposing alternative language that we believe better achieves the rulemaking objectives.
We would appreciate the NRCs feedback on the attachments to this letter, so that we may understand where the NRC agrees and disagrees with the approaches proposed in the discussion draft, including the details of the rule text and key topics for further discussion. If you have questions concerning our input, please contact me at 202-739-8131 or mrn@nei.org.
Sincerely, Marcus Nichol Attachment c:
Ms. Andrea Veil, NRR, NRC Mr. Rob Taylor, NRR, NRC Mr. Mohamed K. Shams, NRR/DANU, NRC Mr. John Segala, NRR/DANU/UARP, NRC Mr. Robert H. Beall, NMSS/REFS/RRPB, NRC Mr. William D. Reckley, NRR/DANU/UARP, NRC Ms. Nanette Valliere, NRR/DANU/UARP, NRC Rulemaking.Comments@nrc.gov
February 11, 2021
- Industry Discussion Draft A of Rule Text for Part 53 Page 1 of 28 Table of Contents Relationship to Requirements in Other Parts of this Chapter..................................................... 3 Key Part 53 Topics: Industry Concerns with NRC Preliminary Rule Text for and Proposed Alternatives......................................................................................................................... 4
§ 5
3.1 Purpose and Scope
..................................................................................................... 5 Subpart A - Public Protection Paradigm..................................................................................... 5
§ 53.2 - Adequate Protection of Public Health and Safety......................................................... 5
§ 53.3 - Extra-Adequate Protection........................................................................................ 6
§ 53.4 - Facility Characteristics.............................................................................................. 7
§ 53.5 - Design and Analysis................................................................................................. 8
§ 53.6 through 53.10 - Reserved........................................................................................... 9
§ 53.11 - Quality Assurance.................................................................................................. 9
§ 53.12 through 53.15 - Reserved......................................................................................... 9 Subpart B - Construction and Operations................................................................................... 9
§ 53.16 - Construction Inspections........................................................................................ 9
§ 53.17 - Preoperational and Initial Operations Testing........................................................... 9
§ 53.18 through 53.20 - Reserved....................................................................................... 10
§ 53.21 - Conduct of Operations, Inspection, Maintenance..................................................... 10
§ 53.22 - Technical Specifications........................................................................................ 10
§ 53.23 - Emergency Plan................................................................................................... 11
§ 53.24 - Safeguards and Security Plans.............................................................................. 11
§ 53.25 - Radiation Protection............................................................................................. 12
§ 53.26 - Records, Reports and Notification.......................................................................... 12
§ 53.27 through 53.30 - Reserved....................................................................................... 12 Subpart C - Decommissioning................................................................................................. 12
§ 53.31 - Decommissioning Funding.................................................................................... 12
§ 53.32 - License Termination............................................................................................. 13 Subpart D - Licenses, Permits and Certifications....................................................................... 13
§ 53.33 Activities Requiring a License or Permit..................................................................... 13
§ 53.34 Specific Exemptions................................................................................................ 14
February 11, 2021
- Industry Discussion Draft A of Rule Text for Part 53 Page 2 of 28
§ 53.35 Issuance, Limitations, and Conditions of Licenses, Permits and Design Approvals.......... 14
§ 53.36 Licenses................................................................................................................ 14
§ 53.37 Permits.................................................................................................................. 15
§ 53.38 Certifications and Standard Design Approvals............................................................ 16
§ 53.39 Applications for Licenses, Permits and Certifications................................................... 16
§ 53.40 Limited Work Authorization..................................................................................... 19
§ 53.41 Licensee Controlled Changes to the Facility............................................................... 20
§ 53.42 Amendments to Licenses and Permits....................................................................... 21
§ 53.43 - Imposition of New Requirements and Protection Thereof (Backfitting)...................... 21
§ 53.44 - Revocation, Suspension, Modification and Emergency Operations by the Commission
{To be developed}............................................................................................................. 23
§ 53.45 - Applicability of a Part 50 or Part 52 Approval {To be developed}.............................. 23
§ 53.45 through 53.50........................................................................................................ 23 Subpart E - General Provisions and Administrative.................................................................... 23
§ 53.51 Definitions: {To Be Developed Later}....................................................................... 23
§ 53.52 Interpretations....................................................................................................... 23
§ 53.53 Communications..................................................................................................... 23
§ 53.54 Deliberate Misconduct............................................................................................. 25
§ 53.55 Employee Protection............................................................................................... 26
§ 53.56 Completeness and Accuracy.................................................................................... 28
§ 53.57 through 53.60 Reserved.......................................................................................... 28
February 11, 2021
- Industry Discussion Draft A of Rule Text for Part 53 Page 3 of 28 Relationship to Requirements in Other Parts of this Chapter
{The purpose of this section is to describe which requirements in other parts apply to Part 53 (e.g.,
Part 2, 20, 50, 51, 52, 55, 73, 74, 75, 100), and how Part 53 interfaces with the requirements. This is intended to provide clarity for the discussion draft rule text and is not proposed to be a Part 53 requirement. This list identifies the other parts that were considered in the discussion draft and recognizes that other part will need to be addressed in the future.}
a) Part 50 and 52 (Licensing Processes) - The discussion draft approach is to leverage the Part 50 and 52 licensing processes as much as possible. This will result in significant conforming changes to those parts. Further consideration should be given to improvements that can be made to the licensing processes based on the Part 53 safety paradigm and whether it is more efficient to include the licensing process details in Part 53 in lieu of reference Parts 50 and 52.
b) Part 73 - Facilities licensed under 10 CFR Part 53 shall provide physical protection of plants and materials, in addition to the safety scope typically addressed in Parts 50 and 52. Part 53 establishes a technology-inclusive, risk-informed, and performance-based security approach to establish the requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage. Specifically, Part 53 requires, the establishment of a design features and/or human actions (§53.4(e)(2)) and a safeguards and security plan
((§53.24) to protect against the design basis threat (§53.4(d)(4)), such that the public protection criteria in §53.2 are met. This level of protection is equivalent to the technology-specific and prescriptive requirements in §73.55, such that duplicate requirements need not be applied to a Part 53 facility. Therefore, facilities licensed under Part 53 do not need to comply with §73.55. Additional security requirements in Part 73 (e.g., Access Authorization in §73.56) are applicable to Part 53 and are identified in §53.24.
c) Part 100 - Facilities licensed under 10 CFR Part 53 need to consider the siting of the reactor to provide reasonable assurance that each site of a nuclear power reactor will protect the public health and safety. Part 53 includes a technology-inclusive, performance-based and risk-informed approach to reactor site criteria. Specifically, §53.4(b) requires that the characteristics of the site that have a significant impact on the ability of the facility to meet the public protection criteria in §53.2 and §53.3 and the location of the site boundary be part of the facility characteristics. Part 53 also requires the establishment of design features and human actions (§53.4(e)) to protect against manmade hazards related to the site
(§53.4(d)(3)) such that the public protection criteria are met. Part 53 establishes the site boundary, in lieu of the low population zone and exclusion area boundary as the key boundary, in alignment with the on-going SMR Emergency Preparedness rulemaking, streamlining the licensing basis of the facility. Part 53 does not establish a requirement for distance from a population center, since it is not necessary, nor does it include specific requirements for the seismic and geologic criteria, since it is not needed for a technology-inclusive, performance-based and risk-informed approach. Therefore, facilities licensed under 10 CFR Part 53 do not need to comply with the reactor site criteria of Part 100.
February 11, 2021
- Industry Discussion Draft A of Rule Text for Part 53 Page 4 of 28 d) Part 20 - Radiation protection will be required for Part 53 (§53.25). The discussion draft approach is to leverage the Part 20 requirements as much as possible (Subparts C and D),
and recognizing that ALARA requirements are not necessary. Further consideration should be given to additional improvements and whether Part 53 should include RP requirements that better align with the safety paradigm in lieu of relying on Part 20.
e) Part 21 - Conforming changes to Part 21 will be needed so that it is applicable to Part 53.
The key term Basic Component will need to have a new definition as it applies to Part 53, since the safety paradigm differs from Part 50 and 52. Basic Component as it applies to Part 53 is: A component relied upon to meet the safety criteria in § 53.2.
f) Part 2 - It is recognized that Part 2 contains Agency Rules on Procedures and Practices, which are relied upon by Part 53. For example, the hearing process is relied upon for licensing through reference to licensing process requirements in Parts 50 and 52. There may be further opportunities to leverage Part 2 to reduce duplication, for example in the areas of communications, deliberate misconduct and employee protection, which are essentially identical between Part 50, 52 and 53. This approach may require conforming changes to Part 2.
Key Part 53 Topics: Industry Concerns with NRC Preliminary Rule Text for and Proposed Alternatives The following topics are discussed in more detail in Attachment 2.
Safety Objectives and Two-Tier Criteria As Low as Reasonably Achievable (ALARA)
Overall Safety Construct Occupational Exposures Quantitative Health Objectives Quantitative Frequencies Probabilistic Risk Assessments (PRA)
Defense-in-Depth Siting Facility Safety Program Addressing Uncertainties General Design Criteria Performance-Based Language
February 11, 2021
- Industry Discussion Draft A of Rule Text for Part 53 Page 5 of 28 Begin Discussion Draft Rule Text
§ 5
3.1 Purpose and Scope
The regulations in this part are promulgated by the Nuclear Regulatory Commission pursuant to the Atomic Energy Act of 1954, as amended (68 Stat. 919), and Title II of the Energy Reorganization Act of 1974 (88 Stat. 1242), to provide for the licensing of production and utilization facilities. This part provides an alternative to the provisions in Part 50 and Part 52 of this chapter. A plant may be licensed and regulated under Part 53, instead of Part 50 or Part 52 at the election of the applicant or licensee of a production or utilization facility. This Part also provides the provision to allow Part 50 or Part 52 licensees or applicants to convert their application or license to this Part.
This Part also gives notice to all persons who knowingly provide to any licensee, applicant, contractor, or subcontractor, components, equipment, materials, or other goods or services, that relate to a licensee's or applicant's activities subject to this part, that they may be individually subject to NRC enforcement action for violation of §53.54.
Subpart A - Public Protection Paradigm
{The purpose of this subpart is to provide the public protection paradigm for Part 53, based upon technology-inclusive, performance-based and risk-informed requirements for public protection standards, the facility characteristics necessary to meet those standards, and providing reasonable assurance that the design achieves the standards. This section includes security and siting considerations where their paradigm concepts are similar to the safety paradigm.}
§ 53.2 - Adequate Protection of Public Health and Safety
{Safety criteria in §53.2 and 53.3 were previously provided to the NRC in a letter dated December 23, 2020. Modifications to the safety criteria will be considered once NRC provides feedback.}
Each power reactor licensed pursuant to §53.35 must provide reasonable assurance of adequate protection of the public health and safety and the common defense and security. Adequate protection, which is focused on radiological risk, recognizes that some level of risk is expected when it comes to activities involving the use of a radioactive source, such that absolute protection is not required. The following criterion, when met by the power reactor, is necessary and sufficient to assure adequate protection of the public health and safety:
a) The contribution of total effective dose equivalent to an individual member of the public at the site boundary for infrequent event sequences, which may include one or more reactor modules, that {are not expected to occur in the life of a nuclear power plant} OR {have an expected frequency greater than once in 10,000 years} does not exceed:
- 1) 25 rem (250 mSv) for any 2-hour period following the radiological consequences of the event, and
- 2) 25 rem (250 mSv) from exposure to the radioactive cloud resulting from the radiological consequences of the event (during the entire period of its passage).
February 11, 2021
- Industry Discussion Draft A of Rule Text for Part 53 Page 6 of 28
§ 53.3 - Extra-Adequate Protection
{Safety criteria in §53.2 and 53.3 were previously provided to the NRC in a letter dated December 23, 2020. Considerations for the approach to quantitative health objectives were include in the presentation to the NRC at the meeting on January 7, 2021. Modifications to the safety criteria will be considered once NRC provides feedback.}
Each power reactor licensed pursuant to §53.35 must meet the additional criteria provided in paragraphs (a) and (b) of this section, which go above and beyond the requirements for adequate protection. Requirements above and beyond the requirements for adequate protection must have a reasonable nexus between the impacts being addressed and the statutory mission of protecting against radiological dangers. Requirements necessary to meet these criteria must provide a substantial safety improvement and be cost justified. Any such specific requirements must substantially improve the level of radiological safety as justified via a cost-benefit analysis that considers direct and indirect costs, and shall permit deviations from these requirements for licensees that demonstrate the application of these requirements to their specific design is not cost justified.
a) The contribution of total effective dose equivalent to an individual member of the public at the site boundary from normal operations, which may include one or more reactor modules, including events that are expected to occur one or more times during the life of a nuclear power plant does not exceed 0.1 rem (1 mSv) in a year and does not exceed 0.002 rem (0.02 millisievert) in any one hour in any unrestricted area.
b) {Each applicant or licensee shall develop, implement, and maintain mitigation strategies and guidance for rare event sequences that are not addressed in 53.2, which may include one or more reactor modules, and that [are not expected to occur in the life of a nuclear power plant]
OR [have an expected frequency greater than five times in 10,000,000 years] that are capable of being implemented site-wide and must include the following:
- 1) The capability to maintain or restore the required facility functions necessary to meet the criteria in 53.2.
- 2) The acquisition and use of offsite assistance and resources to support the functions required by paragraph (b)(1) of this section indefinitely, or until sufficient site functional capabilities can be maintained without the need for the mitigation strategies
- 3) Strategies and guidance to provide the capabilities in (b)(1) under the circumstances associated with loss of large areas of the plant impacted by the event, due to explosions or fire, to minimize radiological releases.}
OR b) {The cumulative plant risk to an average individual:
- 1) for early fatalities within 1 mile of the site boundary does not exceed 5 in 10,000,000 years, and
- 2) for latent cancer fatalities within 10 miles of the site boundary does not exceed 2 in 1,000,000 years.}
February 11, 2021
- Industry Discussion Draft A of Rule Text for Part 53 Page 7 of 28
§ 53.4 - Facility Characteristics
{The purpose of the requirement is to provide reasonable assurance that the facility, including credited human actions, is able to meet the public protection criteria. This is intended to align with the NRCs proposed bow tie diagram for facility safety. This approach addresses safety, security and siting in a performance-based manner and leverages the efficiencies of co-locating these topics that are based on the same conceptual paradigm.}
The facility achieves the public protection criteria through a combination of design features, human actions and/or programmatic controls that protect the public against the facilitys radiological hazard by achieving the public protection criteria in § 53.2 and 53.3 for relevant event sequences. The characteristics of the facility necessary to achieve the protection criteria shall consist of:
a) The radiological hazard of the facility, including the maximum power level, nature and inventory of radioactive materials, and the expected chemical and physical form during all phases of operation.
b) The characteristics of the site that have a significant impact on the ability of the facility to meet the public protection criteria, such as seismology, meteorology, geology, and hydrology, and the location of the site boundary within which the licensee has the authority to prevent access by the public.
c) The required facility functions that are necessary and sufficient to meet the public protection criteria in § 53.2 and 53.3 for the relevant event sequences.
d) The facility event sequences resulting from relevant initiating events of the following types:
- 1) Internal events, such as the failure of an SSC or human action to perform a required facility function, and internal plant hazards such as a fire or flood.
- 2) External events are natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches.
- 3) Man-made events are the manrelated hazards (e.g., airports, dams, transportation routes, military and chemical facilities) specific to the site.
- 4) Security events are the design basis threat of radiological sabotage (DBT) as stated in §73.1(a)(1).
e) The design features, human actions and programmatic controls necessary and sufficient to perform the required facility functions during the relevant facility event sequences. The design features, human actions and programmatic controls should;
- 1) Utilize a combination of structures, systems and components to provide prevention, barriers, and mitigation.
- 2) SSCs, human actions, or a combination of both, to protect against security events in paragraph (d)(4) in this section shall be capable of detecting, assessing and communicating the threat to offsite law enforcement. For facilities that require human action to prevent the threat from causing damage that would exceed the public protection criteria in § 53.2, the human actions must be able to interdict and neutralize the threat.
February 11, 2021
- Industry Discussion Draft A of Rule Text for Part 53 Page 8 of 28
§ 53.5 - Design and Analysis
{The purpose of the requirement is to provide reasonable assurance that the design and analyses demonstrate that the facility is able to meet the public protection criteria in 53.2 and 53.3 within the facility characteristics in 53.4.}
The facility specified in § 53.4, except as it relates to 53.4(d)(4), shall be designed and analyzed to demonstrate that it meets the safety criteria in § 53.2 and 53.3. The design and analysis of the facility shall include:
a) A method for establishing the facility characteristics identified in § 53.4, including the radiological hazard, facility event sequences, required facility functions, design features, human actions and programmatic controls.
b) The principal design criteria of the facility, which establish the necessary design, fabrication, construction, testing, and performance requirements for the structures, systems, and components (SSCs).
c) The analysis and evaluation of the design and performance of structures, systems and components of the facility that establish the adequacy of SSC to perform the required facility functions and the margins for the facility during all relevant event sequences. Methods used for analysis must be appropriate for the conditions analyzed.
d) Design requirements and performance criteria for the SSCs required to meet the required facility functions. Codes and standards may be used for the design requirements and performance criteria where they are appropriate to the intended application.
e) A method of evaluating the risk to inform the design and analysis of the facility. The applicant may decide the role the risk evaluation serves in the design and analysis, and the role chosen may be based upon the maturity of the design. The risk evaluation may employ a probabilistic risk assessment (PRA) to be used as the primary or sole method to establish the required facility functions, event sequences, and design features. For applicants that employ deterministic methods for all, or some portions, of the facility evaluations, a qualitative risk assessment may be used to confirm the findings of the safety analyses.
f) Measures such as increased performance margins in the design of SSCs to address design and analytical uncertainties.
g) The methodology for placing facility SSCs into categories that will characterize their qualification requirements. The methodology may include previously accepted NRC categorization methods or established international standards. The required facility functions, event sequences, and design features, may be classified according to the safety criteria in § 53.2 and 53.3.
h) Qualification of the structures, systems and components by testing, analysis, or both to demonstrate the SSCs will perform their required facility function under the service conditions expected during the relevant facility event sequences.
i)
For design certification or standard design approvals a description, analysis, and evaluation of the interface requirements between the scope of the application and the balance of the facility.
February 11, 2021
- Industry Discussion Draft A of Rule Text for Part 53 Page 9 of 28
§ 53.6 through 53.10 - Reserved
§ 53.11 - Quality Assurance
{A high-level approach to quality assurance requirements will enable QA to be performed according to many acceptable approaches, including some commercial standards. It is expected that guidance will either endorse acceptable approaches, or present the criteria for acceptable approaches.}
A quality assurance program shall be used for the design, fabrication, construction and operation of the facility to provide reasonable assurance that the facility will perform its required facility functions during the facility event sequences. Quality assurance includes quality control, which comprises those quality assurance actions related to the physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, structure, component, or system to predetermined requirements. Applicants and holders of a permit, license, or design approval or certification shall have a quality assurance program that covers the scope of their regulated activities, such as designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying. The quality assurance program may rely on internationally accepted commercial standards and previously accepted NRC programs.
§ 53.12 through 53.15 - Reserved Subpart B - Construction and Operations
{The purpose of this subpart is to provide reasonable assurance that the facility is constructed and will operate in accordance with the design.}
§ 53.16 - Construction Inspections
{Purpose is to confirm that the plant has been constructed according to the design. For a COL, this would require ITAAC. More work is needed to determine if modifications to ITAAC for Part 53 are needed to align with the technology-inclusive, risk-informed and performance-based approach. No requirements are needed for an OL, since the reference to 50.57 in 53.36 incorporates 50.56 by reference, and provides the conditions for transferring from a CP to OL.}
For a combined license pursuant to 53.36(c), the facility shall meet the requirements for inspections during construction specified in § 52.99.
§ 53.17 - Preoperational and Initial Operations Testing Plans shall be established for preoperational testing and initial operations as part of a test program to demonstrate that SSCs will perform satisfactorily in service. The test program shall assure that all
February 11, 2021
- Industry Discussion Draft A of Rule Text for Part 53 Page 10 of 28 testing is identified and performed in accordance with written test procedures, which incorporate the requirements and acceptance limits contained in applicable design documents.
§ 53.18 through 53.20 - Reserved
§ 53.21 - Conduct of Operations, Inspection, Maintenance All operations, inspection and maintenance of the plant shall be conducted in a manner to provide reasonable assurance that the public protection criteria specified in §53.2 and §53.3 are satisfied.
The following shall be established and maintained for the facilitys conduct of operations:
a) An organization of facility personnel, which integrates the operational functions in this section with the emergency response functions in §53.23 and the security functions in §53.24, and is capable of performing the human actions in §53.4.
b) An operations program that identifies any functions that may only be performed by a licensed operator pursuant to Part 55 of this chapter, other functions to be performed by operators, and the training qualifications of operators.
c) An inspection and testing program shall be established and maintained to provide reasonable assurance that SSCs will perform their required facility function(s). The inspection and testing program shall identify the SSCs and associated required facility functions to be demonstrated, the intervals for inspection or testing of each SSC, and the conditions for the inspection or test.
d) A maintenance program shall be established and implemented in accordance with the provisions of §50.65(a). The scope of the maintenance program shall be for the SSCs in
§53.4 that perform a required facility function, and thus §50.65(b) is not applicable.
§ 53.22 - Technical Specifications Technical specifications shall be established for SSCs in §53.4 that perform required facility functions necessary to meet the public protection criteria in §53.2 and §53.3. The technical specifications shall provide reasonable assurance that the plant is operated within the design and performance characteristics identified in in §53.5. Technical specifications will include items in the following categories:
a) Adequate protection limits of the facility, that if exceeded could lead to exceeding the public protection criterion in §53.2.
b) Limiting safety settings for protective devices relating to required facility functions necessary to meet the public protection criterion in §53.2.
c) Operating conditions of the facility, and the applicability of the conditions during operation, that if exceeded could lead to a failure to perform a required facility function necessary to meet the public protection criterion in §53.2.
d) Action(s), including the completion time, to return the facility to compliance with the operating condition.
e) Technical bases for the operations of the facility.
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§ 53.23 - Emergency Plan
{This requirement will need updating to reflect the final SMR And ONT EP rule (Docket ID NRC-2015-0225) that establishes 50.160, and conforming changes to align the requirements of 50.47 to the Part 53 format of putting application content requirements in 53.39 and technical performance requirements for EP in 53.23.}
An emergency plan shall be established for the facility pursuant to §50.47 and 50.160.
§ 53.24 - Safeguards and Security Plans
{This requirement will need updating to reflect the final SMR And ONT Security rule (Docket ID NRC-2017-0227). Additional conforming changes may be needed so that Part 53 will provide a substitute for 73.55. A review of Part 74 and 75 may identify additional changes to this requirement, especially in consideration of micro-reactors.}
A safeguards and security plans shall be established to provide physical protection of the facility against theft, diversion and the design basis threat for radiological sabotage:
a) Physical security plan. The physical security plan must describe the SSCs identified in
§53.4(e)(2) that protects against the security events in §53.4(d)(4). The physical security plans shall include:
- 1) The achievable target sets, which are the combinations of SSCs that the security threat could destroy and would result in exceeding the criteria in §53.2, and that must be protected by human actions.
- 2) A description of the security organization.
- 3) A description of the human actions, if any, relied upon to prevent the threat from causing damage that would exceed the public protection criteria in § 53.2. Human actions are not required to perform the functions of detecting, assessing and communicating the threat to offsite law enforcement.
- 4) For facilities that require human action to interdict and neutralize the threat, a description of the security response plan and performance evaluation that provides reasonable assurance that the human actions will be accomplished during a security event.
- 5) A security plan established pursuant to Part 37 of this chapter to address physical protection of Category 1 and Category 2 radioactive sources, may be used to satisfy the requirements of paragraph (a) of this section.
b) Access authorization program in accordance with §73.56.
c) Cyber security program in accordance with §73.54.
d) Insider mitigation program to minimize the potential for an insider to adversely affect the ability to meet the criteria in §53.2.
e) Training and qualification plan pursuant to Appendix B,Section IV of Part 73.
f) Licensee safeguards contingency plan. The safeguards contingency plan must comply with the criteria set forth in Section II of Appendix C to Part 73. The implementing procedures
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g) The security plan and safeguards contingency plan shall protect the plans and other related Safeguards Information against unauthorized disclosure in accordance with the requirements of §73.21.
§ 53.25 - Radiation Protection
{The intent is to leverage Part 20 to the extent that it is applicable to Part 53. There may be more improvements to radiation protection for Part 53 beyond the topic of ALARA. Thus, there may be a need to include Part 53 specific radiation protection requirements in lieu of using Part 20.}
A radiation protection program shall be established that complies with the requirements in Part 20, with the following clarifications:
a) The dose limits applicable to the radiation protection program are specified in §53.3 and in Part 20, Subparts C and D.
b) The licensee is relieved from the requirements of §20.1101(b) to achieve occupational doses and doses to members of the public that are as low as reasonably achievable.
§ 53.26 - Records, Reports and Notification
{Requirements for the below functions are needed; however, these may be modified after further evaluation. Specifically, changes may be necessary to align these requirements with the technology-inclusive, risk-informed, and performance-based approach being developed for Part 53.}
Records, reports and notifications shall be performed in accordance with:
- 1) §50.71, Maintenance of records, making of reports
- 2) §50.72, Immediate notification requirements for operating nuclear power reactors and the amendment to §50.72
- 3) §50.73, Licensee event report system
- 4) §50.74, Notification of change in operator or senior operator status
- 5) §50.76, Licensees change of status; financial qualifications
§ 53.27 through 53.30 - Reserved Subpart C - Decommissioning
{Need to determine if decommissioning requirements will be established with the initial rulemaking or after 2024. The following requirements are needed for the licensing of a facility.}
§ 53.31 - Decommissioning Funding
{Need to evaluate whether conforming changes are needed here or in 50.75 to align with the technology-inclusive, risk-informed, and performance-based approach in Part 53.}
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§ 53.32 - License Termination
{Need to evaluate whether conforming changes are needed here or in 50.82 or 52.110 to align with the technology-inclusive, risk-informed, and performance-based approach in Part 53. This requirement is needed to provide a reference for 53.53(b)(5) and (6) for certification of permanent cessation of operations and certification of permanent fuel removal.}
Termination of the license, including certification of permanent fuel removal shall meet the applicable requirements in 50.82 or 52.110.
Subpart D - Licenses, Permits and Certifications
§ 53.33 Activities Requiring a License or Permit
{This section is a collection of general requirements. As the discussion draft evolves, it is expected that these may be relocated.}
a) Requirement for a license. Except as provided in § 50.11 of this chapter, no person within the United States shall transfer or receive in interstate commerce, manufacture, produce, transfer, acquire, possess, or use any production or utilization facility except as authorized by a license issued by the Commission.
b) Requirement to Begin Construction. No person may begin the construction of a production or utilization facility on a site on which the facility is to be operated until that person has been issued either a construction permit or a combined license under this part, an early site permit under § 53.37(b) of this part, or a limited work authorization under § 53.40 of this part.
c) Affirmative Safety Case. The Commission shall, to the maximum extent practicable, allow applicants to demonstrate reasonable assurance of adequate protection of the public health and safety and the common defense and security through the presentation of an affirmative safety case. An applicant may present an acceptable affirmative safety case by providing the information necessary and sufficient to provide reasonable assurance that the public protection criteria in §53.2 and 53.3 are met.
d) An applicant for a license to construct and operate a power reactor facility, or for an amendment to such license, is not required to provide design features or other measures for the specific purpose of protection against the effects of (a) attacks and destructive acts, including sabotage, directed against the facility by an enemy of the United States, whether by a foreign government or other person, or (b) use or deployment of weapons incident to U.S. defense activities.
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§ 53.34 Specific Exemptions The Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of this part. The Commissions consideration will be governed by § 50.12 of this chapter, unless other criteria are provided for in this part, in which case the Commissions consideration will be governed by the criteria in this part. Only if those criteria are not met will the Commissions consideration be governed by § 50.12 of this chapter. The Commissions consideration of requests for exemptions from requirements of the regulations of other parts in this chapter, which are applicable by virtue of this part, shall be governed by the exemption requirements of those parts.
§ 53.35 Issuance, Limitations, and Conditions of Licenses, Permits and Design Approvals Upon determination that an application for a license, permit or certification, pursuant to § 53.39 meets the applicable standards and requirements of the Act and regulations, and that notifications, if any, to other agencies or bodies have been duly made, the Commission will issue the license pursuant to § 53.36, permit pursuant to § 53.37, or certification pursuant to § 53.38, as applicable, in such form and containing such conditions and limitations as it deems appropriate and necessary.
§ 53.36 Licenses a) Part 53 licenses will be issued to named persons applying to the Commission and will be either class 104 or class 103, and will be issued to applicants that satisfy the requirements of
§50.21 for class 104 licenses or §50.22 for class 103 licenses, as applicable.
b) Operating License. The NRC may issue an operating license under this part based on the findings set forth in § 50.57. Compliance with the requirements of this part demonstrates that (i) there is reasonable assurance that the activities authorized by the operating license can be conducted without endangering the health and safety of the public; and (ii) such activities will be conducted in compliance with the regulations in this chapter; and (iii) issuance of the operating license will not be inimical to the common defense and security or to the health and safety of the public. Therefore, an applicants compliance with the requirements of this part fulfills the findings required in 10 CFR 50.57(a)(3) and 50.57(a)(6).
c) Combined Operating License The NRC may issue a combined license under this part based on the findings set forth in 10 CFR 52.97. Compliance with the requirements of this part demonstrates that (i) there is reasonable assurance that the facility will be constructed and will operate in conformity with the license, the provisions of the Act and the Commissions regulations; and (ii) issuance of the license will not be inimical to the common defense and security or to the health and safety of the public. Therefore, an applicants compliance with the requirements of this part fulfills the findings required in 10 CFR 52.97(a)(1)(iii) and 52.97(a)(1)(v).
d) Manufacturing License. The NRC may issue a manufacturing license under this part based on the findings set forth in 10 CFR 52.167. Compliance with the requirements of this part demonstrates that (i) there is reasonable assurance that reactor(s) will be manufactured and can be transported, incorporated into a nuclear power plant, and operated in conformity
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52.167(a)(3) and 52.167(a)(6).
e) {Prototype Licenses - A topic for resolution is how prototypes should be addressed in Part
- 53. One option is to have a Prototype License. Other options are for special requirements related to prototypes, address solely through guidance, or to not address it.}
f) The Commission may combine in a single license the activities which would otherwise be licensed separately.
§ 53.37 Permits a) A construction permit for the construction of a production or utilization facility will be issued before the issuance of an operating license if the application is otherwise acceptable, and will be converted upon completion of the facility and Commission action, into an operating license pursuant to §53.36(b). However, if a combined license for a nuclear power reactor is issued, the construction permit and operating license are deemed to be combined in a single license. A construction permit for the alteration of a production or utilization facility will be issued before the issuance of an amendment of a license, if the application for amendment is otherwise acceptable, as provided in § 53.42.
b) The NRC may issue an early site permit under this part based on the findings set forth in
§52.24. Compliance with the requirements of this part demonstrates that (i) there is reasonable assurance that the site is in conformity with the license, the provisions of the Act and the Commissions regulations; issuance of the permit will not be inimical to the common defense and security or to the health and safety of the public. Therefore, an applicants compliance with the requirements of this part fulfills the findings required in 10 CFR 52.24(a)(3)) and 52.24(a)(6).
c) An applicant may request, and the Commission may issue, a limited work authorization under § 53.40 of this part in conjunction with the early site permit. The application must include the information otherwise required by § 53.40(a). The Commission shall issue the limited work authorization if it determines that the requirements in § 53.40(b) have been satisfied.
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§ 53.38 Certifications and Standard Design Approvals a) Standard Design Certification. The NRC may issue a design certification under this part based on the findings set forth in 10 CFR 52.54. Compliance with the requirements of this part demonstrates that (i) there is reasonable assurance that the standard design conforms with the provisions of the Act and the Commissions regulations, and (ii) issuance of the standard design certification will not be inimical to the common defense and security or to the health and safety of the public. Therefore, an applicants compliance with the requirements of this part fulfills the findings required in 10 CFR 52.54(a)(3).
b) Standard Design Approval. The NRC may issue a design approval under this part based on the requirements in 10 CFR 52.131 through 52.147. The requirements in this part satisfy the contents of application and the standards for review and the requirements in 52.136, 52.137 and 52.139 are not applicable.
c) The requirements for communications in §53.53 will apply in lieu of the requirements in
§52.3.
§ 53.39 Applications for Licenses, Permits and Certifications
{This section focuses on the documentation requirements only, and references technical performance requirements as necessary. This approach to separate documentation and technical requirements differs significantly from Part 50 and 52, which tend to mix technical performance requirements in the requirements for application content. Note that significant conforming to changes to 2.101 are needed for Part 53}
Any person seeking a license, permit or design approval specified in § 53.36, 53.37 or 53.38 shall submit an application in accordance with 10 CFR 2.101. Applications filed under this part will be reviewed for compliance with the standards set forth in this part and its supporting appendices.
a) The applicant must:
- 1) Maintain the capability to generate additional copies of the application, and to update the application as directed by the NRC.
- 2) Make available, for audit or inspection by the NRC to the extent that it is required to make a safety determination, the analyses, and testing and inspection results for supporting licensing basis of the facility, but is not required to be included in section (c)(8) of this subpart.
- 3) (For permits and combined licenses) Update the application and serve the updated copies with an index of the updated application as directed by the Atomic Safety and Licensing Board appointed to conduct the public hearing required by the Atomic Energy Act.
b) The application and any amendments to the application must be:
- 1) Submitted to the U.S. Nuclear Regulatory Commission in accordance with § 53.53. The NRC will make available the application and other records pertinent to the matter which is the subject of the application for public inspection and copying.
- 2) Executed in a signed original by the applicant or duly authorized officer thereof under oath or affirmation.
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- 3) Prepared in such manner that all Restricted Data and other defense information, if any, are separated from the unclassified information; and such that Safeguards and Security-Related Information is protected against unauthorized disclosure in accordance with the requirements of § 73.21 of this chapter withheld from the public.
- 4) Describe the technical qualifications of the applicant to engage in the proposed activities.
- 5) Accompanied by the fee prescribed in part 170 of this chapter, unless otherwise exempted from part 170 of this chapter. No fee will be required to accompany an application for renewal, amendment, or termination of a permit, license or design approval except as provided in § 170.21 of this chapter.
c) The content of the application:
- 1) Shall state information about the applicant, including the name, address, business description, citizenship of the individual, partners, or directors and principle officers, the location where the business is registered, and any foreign affiliations. If the applicant is acting as agent or representative of another person in filing the application, identify the principal and furnish information required under this paragraph with respect to such principal.
- 2) May combine in its application several applications for different kinds of licenses under the regulations of this chapter. An applicant may incorporate by reference in its application information contained in previous applications, statements or reports filed with the Commission, provided, however, that such references are clear and specific.
- 3) The class of license applied for, the use to which the facility will be put, the period of time for which the license is sought, and a list of other licenses, except operator's licenses, issued or applied for in connection with the proposed facility.
- 4) Except for a utility applicant for a license under this part, information sufficient to demonstrate to the Commission the financial qualification of the applicant to carry out, in accordance with regulations in this chapter, the activities for which the permit or license is sought. As applicable, the applicant should demonstrate that they possess or have reasonable assurance of obtaining the funds necessary to cover estimated construction, fuel cycle and operation costs for the period of the license. The Commission may request an established entity or newly-formed entity to submit additional or more detailed information respecting its financial arrangements and status of funds if the Commission considers this information appropriate. This may include information regarding a licensees ability to continue the conduct of the activities authorized by the license and to decommission the facility.
- 5) (For regulated utilities under a class 103 license) A list of the names and addresses of such regulatory agencies as may have jurisdiction over the rates and services incident to the proposed activity, and a list of trade and news publications which circulate in the area where the proposed activity will be conducted and which are considered appropriate to give reasonable notice of the application to those municipalities, private utilities, public bodies, and cooperatives, which might have a potential interest in the facility.
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- 6) (For a license) Information in the form of a report, as described in § 53.32, indicating how reasonable assurance will be provided that funds will be available to decommission the facility.
- 7) General description of the facility, including a high-level overview of the structures and systems that do not perform a required facility function and thus are not part of the licensing basis.
- 8) The licensing basis of the facility comprised of:
- i.
A preliminary safety analysis report for permits and a final safety analysis report for licenses consisting of the information that describes the scope of the facility described in §53.4, with the exception of the scope related to §53.4(d)(4), as appropriate to the type of application. The safety analysis report shall include a description of the design and analysis of the facility pursuant to §53.5. The FSAR shall be updated pursuant to Sec. 50.71(e) of this part.
ii.
Quality Assurance program pursuant to §53.11, iii.
Plans for preoperational testing and initial operations pursuant to §53.17, iv.
Plan for the organization, training of personnel and conduct of operations, including maintenance, surveillance and periodic testing of structures, systems and components pursuant to §53.21
- v.
Technical specifications pursuant to §53.22, vi.
Emergency plans pursuant to §53.23, vii.
Physical security plan pursuant to § 53.24 and cyber security plan pursuant to § 73.54 of this chapter.
viii.
Radiation Protection program pursuant to §53.25,
- 9) Shall be accompanied by an Environmental Report required under subpart A of part 51 of this chapter where the construction or operation may be determined by the Commission to have a significant impact in the environment.
d) The NRC shall,
- 1) Docket the application within 60 days of receipt, or reject the application for not containing the content required in paragraph (c) of the section, and
- 2) At the time of docketing the application:
- i.
Issue to the applicant a review schedule for the review and issuance of the requested license, permit or design approval that does not exceed 2 years, except for special circumstances.
ii.
Refer a copy of the application to the ACRS, whom shall review and report on those portions of the application that directly impact safe operation of the facility.
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§ 53.40 Limited Work Authorization A limited work authorization may be granted to applicants for an early site permit, a construction permit, or a combined license under this part.
a) Request for limited work authorization. An application for a limited work authorization shall meet the applicable requirements of 53.39 and must include
- 1) A safety analysis report, as applicable, that demonstrates that activities conducted under the limited work authorization will be conducted in compliance with the technically-relevant Commission requirements.
- 2) A description of the activities requested to be performed,
- 3) The design and construction information otherwise required by the Commission's rules and regulations to be submitted for a construction permit or combined license, but limited to those portions of the facility that are within the scope of the limited work authorization,
- 4) An environmental report in accordance with § 51.49 of this chapter; and
- 5) A plan for redress of activities performed under the limited work authorization, should limited work activities be terminated by the holder or the limited work authorization be revoked by the NRC, or upon effectiveness of the Commission's final decision denying the associated construction permit or combined license application, as applicable.
b) Issuance of a limited work authorization. The Director of the Office of Nuclear Reactor Regulation may issue a limited work authorization specifying the activities that the holder is authorized to perform only after:
- 1) The NRC staff issues an environmental assessment for the limited work authorization;
- 2) The presiding officer makes the finding in § 51.105(c) or § 51.107(d) of this chapter, as applicable;
- 3) The Director determines that the applicable standards and requirements of the Act, and the Commission's regulations applicable to the activities to be conducted under the limited work authorization, have been met. The applicant is technically qualified to engage in the activities authorized. Issuance of the limited work authorization will provide reasonable assurance of adequate protection to public health and safety and will not be inimical to the common defense and security; and
- 4) The presiding officer finds that there are no unresolved safety issues relating to the activities to be conducted under the limited work authorization that would constitute good cause for withholding the authorization.
c) Effect of limited work authorization. Any activities undertaken under a limited work authorization are entirely at the risk of the applicant and, except as to the matters determined under paragraph (b) of this section, the issuance of the limited work authorization has no bearing on the issuance of a construction permit or combined license with respect to the requirements of the Act, and rules, regulations, or orders issued under the Act. The environmental assessment or environmental impact statement for a construction permit or combined license application for which a limited work authorization was previously issued will not address, and the presiding officer will not consider, the sunk costs of the holder of limited
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d) Implementation of redress plan. If construction is terminated by the holder, the underlying application is withdrawn by the applicant or denied by the NRC, or the limited work authorization is revoked by the NRC, then the holder must begin implementation of the redress plan in a reasonable time. The holder must also complete the redress of the site no later than 18 months after termination of construction, revocation of the limited work authorization, or upon effectiveness of the Commission's final decision denying the associated construction permit application or the underlying combined license application, as applicable.
§ 53.41 Licensee Controlled Changes to the Facility
{This is equivalent to §50.59. It has been modified to align with the safety paradigm of Part 53, which could require new guidance to align with terms. Alternative approach would be to reference
§50.59; however, safety paradigm differences may lead to questions on how to implement NRC endorsed guidance.}
A holder of a license issued under this part may make changes in the facility and the procedures as described in the final safety analysis report (as updated), and conduct tests or experiments not described in the final safety analysis report (as updated) without obtaining a license amendment pursuant to §53.42 only if:
- a. A change to the technical specifications incorporated in the license is not required, and b) The change, test, or experiment applies to an SSC that is not required to meet §53.2, or applies to an SSC that is required to meet §53.2 but does not:
- 1) Result in more than a minimal increase in the frequency of occurrence of an event sequence previously evaluated in the final safety analysis report (as updated);
- 2) Result in more than a minimal increase in the consequences of an event sequence previously evaluated in the final safety analysis report (as updated);
- 3) Create a possibility for an event sequence of a different type than any previously evaluated in the final safety analysis report (as updated);
- 4) Result in a design basis limit for a fission product barrier as described in the FSAR (as updated) being exceeded or altered; or
- 5) Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.
c) In implementing this paragraph, the FSAR (as updated) is considered to include FSAR changes resulting from evaluations performed pursuant to this section and analyses performed pursuant to Sec. 53.42 since submittal of the last update of the final safety analysis report pursuant to Sec. 53.39(c)(8)(i) of this part.
d) Definitions for the purposes of this section:
- 1) Change means a modification or addition to, or removal from, the facility or procedures that affects a design function, method of performing or controlling the function, or an evaluation that demonstrates that intended functions will be accomplished.
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- 2) Departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses means:
- i.
Changing any of the elements of the method described in the FSAR (as updated) unless the results of the analysis are conservative or essentially the same; or ii.
Changing from a method described in the FSAR to another method unless that method has been approved by NRC for the intended application.
- 3) Facility as described in the final safety analysis report (as updated) means:
- i.
The structures, systems, and components (SSC) that are described in the final safety analysis report (FSAR) (as updated),
ii.
The design and performance requirements for such SSCs described in the FSAR (as updated), and iii.
The evaluations or methods of evaluation included in the FSAR (as updated) for such SSCs which demonstrate that their intended function(s) will be accomplished.
- 4) Tests or experiments not described in the final safety analysis report (as updated) means any activity where any structure, system, or component is utilized or controlled in a manner which is either:
- i.
Outside the reference bounds of the design bases as described in the final safety analysis report (as updated) or ii.
Inconsistent with the analyses or descriptions in the final safety analysis report (as updated).
e) The licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments, which include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment. The licensee shall submit in accordance with § 53.53(b)(4), as applicable, a report containing a brief description of any changes, tests, and experiments, including a summary of the evaluation of each. A report must be submitted at intervals not to exceed 24 months. For combined licenses, the report must be submitted at intervals not to exceed 6 months during the period from the date of application for a combined license to the date the Commission makes its findings under 10 CFR 52.103(g).
§ 53.42 Amendments to Licenses and Permits Amendments for licenses or permits shall be made pursuant to §50.90, §50.91 and §50.92. The requirements for communications in §53.53 will apply in lieu of the requirements in §50.4 or §52.3.
§ 53.43 - Imposition of New Requirements and Protection Thereof (Backfitting)
{This is equivalent to §50.109. It has been modified to align with the safety paradigm of Part 53.}
The Commission may seek to impose a modification of or addition to (1) systems, structures, components, or design of a facility; (2) the design approval or manufacturing license for a facility; or (3) the procedures or organization required to design, construct or operate a facility. The NRC may impose such a modification (i.e., backfit) only if the modification
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- 1) That modification or regulatory action is necessary to bring the facility into compliance with 53.2.
- 2) That a modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee, and
- i.
The Commission determines, based on the analysis described in paragraph (c) of this section, that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and ii.
that the direct and indirect costs of implementation for that facility are far outweighed by the benefits and are justified in view of this increased protection, and shall permit exception to these requirements by facilities for which the modification or regulatory action is not cost justified.
c) In reaching the finding in paragraph (b) of this section, the Commission will consider how the backfit should be scheduled in light of other ongoing regulatory activities at the facility and, in addition, will consider information available concerning any of the following factors as may be appropriate and any other information relevant and material to the proposed backfit:
- 1) Statement of the specific objectives and reasons for the modification and the basis for invoking the exception that the proposed backfit is designed to achieve;
- 2) General description of the activity that would be required by the licensee or applicant in order to complete the backfit;
- 3) Potential change in the risk to the public from the accidental off-site release of radioactive material;
- 4) Potential impact on radiological exposure of facility employees;
- 5) Installation and continuing costs associated with the backfit, including the cost of facility downtime or the cost of construction delay;
- 6) The potential safety impact of changes in plant or operational complexity, including the relationship to proposed and existing regulatory requirements;
- 7) The estimated resource burden on the NRC associated with the proposed backfit and the availability of such resources;
- 8) The potential impact of differences in facility type, design or age on the relevancy and practicality of the proposed backfit;
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- 9) Whether the proposed backfit is interim or final and, if interim, the justification for imposing the proposed backfit on an interim basis.
d) No licensing action will be withheld during the pendency of backfit analyses required by the Commission's rules.
e) The Executive Director for Operations shall be responsible for implementation of this section, and all analyses required by this section shall be approved by the Executive Director for Operations or his designee.
f) If there are two or more ways to achieve compliance with a license or the rules or orders of the Commission, or with written licensee commitments, or there are two or more ways to reach a level of protection which is adequate, then ordinarily the applicant or licensee is free to choose the way which best suits its purposes. However, should it be necessary or appropriate for the Commission to prescribe a specific way to comply with its requirements or to achieve adequate protection, then cost may be a factor in selecting the way, provided that the objective of compliance or adequate protection is met.
§ 53.44 - Revocation, Suspension, Modification and Emergency Operations by the Commission {To be developed}
{Need to determine Part 53 relationship with 50.100 through 50.103}
§ 53.45 - Applicability of a Part 50 or Part 52 Approval {To be developed}
{Need to determine how a Part 50 or 52 approval can be used in Part 53, for example a CP or ESP leading to a Part 53 license.}
§ 53.45 through 53.50 Subpart E - General Provisions and Administrative
§ 53.51 Definitions: {To Be Developed Later}
{This section will need to be updated as key terms needing clarification are identified.}
§ 53.52 Interpretations Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be recognized to be binding upon the Commission.
§ 53.53 Communications a) General Requirements. All correspondence, reports, applications, and other written communications from an applicant or holder of a license, permit or approved design to the Nuclear Regulatory Commission concerning the regulations in this part, or the terms and
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- Industry Discussion Draft A of Rule Text for Part 53 Page 24 of 28 conditions of a license, permit or approved design, must be sent either mail addressed: ATTN:
Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland, between the hours of 7:30 a.m. and 4:15 p.m. eastern time; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, e-mail, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information. If the communication is on paper, the signed original must be sent. If a submission due date falls on a Saturday, Sunday, or Federal holiday, the next Federal working day becomes the official due date.
b) Distribution requirements. Copies of all correspondence, reports, and other written communications concerning the regulations in this part or individual license conditions, or the terms and conditions of an early site permit or standard design approval, must be submitted to the persons listed in paragraph (b)(1) of this section (addresses for the NRC Regional Offices are listed in appendix D to part 20 of this chapter).
- 1) Applications for amendment of permits and licenses; reports; and other communications. All written communications (including responses to: generic letters, bulletins, information notices, regulatory information summaries, inspection reports, and miscellaneous requests for additional information) that are required of holders of licenses, permits, and design approvals issued under this part must be submitted as follows, except as otherwise specified in paragraphs (b)(2) through (b)(7) of this section: to the NRCs Document Control Desk (if on paper, the signed original), with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector, if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under subpart F of this part.
- 2) Applications and amendments to applications. Applications for licenses, permits, and design approvals and amendments to any of these types of applications must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector, if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under subpart F of this part, except as otherwise specified in paragraphs (b)(3) through (b)(7) of this section. If the application or amendment is on paper, the submission to the Document Control Desk must be the signed original.
- 3) Acceptance review application. Written communications required for an application for determination of suitability for docketing must be submitted to the NRC's Document Control
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- Industry Discussion Draft A of Rule Text for Part 53 Page 25 of 28 Desk, with a copy to the appropriate Regional Office. If the communication is on paper, the submission to the Document Control Desk must be the signed original.
- 4) Information related to the licensing basis of the facility, including any changes, pursuant to 53.39(c)(8).
- 5) Certification of permanent cessation of operations. The licensee's certification of permanent cessation of operations under § 53.32, must state the date on which operations have ceased or will cease, and must be submitted to the NRC's Document Control Desk. This submission must be under oath or affirmation.
- 6) Certification of permanent fuel removal. The licensee's certification of permanent fuel removal under § 53.32, must state the date on which the fuel was removed from the reactor vessel and the disposition of the fuel, and must be submitted to the NRC's Document Control Desk. This submission must be under oath or affirmation.
(c) Form of communications. All paper copies submitted to meet the requirements set forth in paragraph (b) of this section must be typewritten, printed or otherwise reproduced in permanent form on unglazed paper. Exceptions to these requirements imposed on paper submissions may be granted for the submission of micrographic, photographic, or similar forms.
(d) Regulation governing submission. Applicants, licensees, and holders of standard design approvals submitting correspondence, reports, and other written communications under the regulations of this part are requested but not required to cite whenever practical, in the upper right corner of the first page of the submission, the specific regulation or other basis requiring submission.
§ 53.54 Deliberate Misconduct a) Knowingly providing to any licensee, any applicant for a license, permit, standard design certification or standard design approval, or a contractor, or subcontractor of a person or entity subject to this section, any components, equipment, materials, or other goods or services that relate to a licensees or applicants activities is subject to enforcement action in accordance with the procedures in 10 CFR part 2, subpart B.
b) Deliberate misconduct is:
- 1) Engaging in deliberate misconduct that causes or would have caused, if not detected, a licensee, holder of a permit, standard design approval, or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license issued by the Commission, any standard design approval, or standard design certification; or
- 2) Deliberately submitting to the NRC; a licensee, an applicant for a license, permit, standard design certification or standard design approval; or a licensee's, standard design approval holder's, or applicant's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC.
c) This section applies to a:
- 1) Holder of a license, permit, design certification or standard design approval;
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- Industry Discussion Draft A of Rule Text for Part 53 Page 26 of 28
- 2) Applicant for a license, permit, design certification, or standard design approval;
- 3) Employee and any contractor (including a supplier or consultant), subcontractor, or employee of a contractor or subcontractor of an entity in paragraph (1) or (2).
§ 53.55 Employee Protection a) Discrimination by a holder or applicant for a Commission licensee, permit, or design approval, a contractor or subcontractor, against an employee for engaging in certain protected activities is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in Section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Atomic Energy Act or the Energy Reorganization Act.
- 1) The protected activities include but are not limited to:
- i. Providing the Commission or their employer information about alleged violations of either of the statutes named in the introductory text of paragraph (a) of this section or possible violations of requirements imposed under either of those statutes; ii. Refusing to engage in any practice made unlawful under either of the statutes named in the introductory text of paragraph (a) of this section or under these requirements if the employee has identified the alleged illegality to the employer; iii. Requesting the Commission to institute action against their employer for the administration or enforcement of these requirements; iv. Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in the introductory text of paragraph (a) of this section; and (v)
Assisting or participating in, or is about to assist or participate in, these activities.
- 2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee assistance or participation.
- 3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from their employer (or the employer's agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Atomic Energy Act of 1954, as amended.
b) Any employee who believes that he or she has been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Employment Standards
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- Industry Discussion Draft A of Rule Text for Part 53 Page 27 of 28 Administration, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.
c) A violation of paragraph (a), (e), or (f) of this section by a Commission licensee, a holder of a standard design approval, an applicant for a Commission license, standard design certification, or a standard design approval, or a contractor or subcontractor of a Commission licensee, holder of a standard design approval, or any applicant may be grounds for
- 1) Denial, revocation, or suspension of the license or standard design approval;
- 2) Withdrawal or revocation of a proposed or final standard design certification;
- 3) Imposition of a civil penalty on the licensee, holder of a standard design approval, or applicant (including an applicant for a standard design certification under this part following Commission adoption of final design certification rule) or a contractor or subcontractor of the licensee, holder of a standard design approval, or applicant.
- 4) Other enforcement action.
d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.
e)
- 1) Each holder or applicant for a licensee, permit or design approval, shall prominently post the revision of NRC Form 3, "Notice to Employees," referenced in 10 CFR 19.11(e). This form must be posted at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. Premises must be posted not later than thirty (30) days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license, permit, standard design certification, or standard design approval under 10 CFR part 53, and for 30 days following license termination or the expiration or termination of the standard design certification or standard design approval under 10 CFR part 53.
- 2) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in appendix D to part 20 of this chapter, via email to Forms.Resource@nrc.gov, or by visiting the NRC's online library at http://www.nrc.gov/reading-rm/doc-collections/forms/.
f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor under Section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision which would prohibit, restrict, or otherwise discourage an employee
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- Industry Discussion Draft A of Rule Text for Part 53 Page 28 of 28 from participating in protected activity as defined in paragraph (a)(1) of this section including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities.
g) Part 19 of this chapter sets forth requirements and regulatory provisions applicable to licensees, holders of a standard design approval, applicants for a license, standard design certification, or standard design approval, and contractors or subcontractors of a Commission licensee, or holder of a standard design approval, and are in addition to the requirements in this section.
§ 53.56 Completeness and Accuracy (a) Information provided to the Commission by a holder of a license, permit, design certification or standard design approval under this part, and an applicant for a license, permit, design certification or a standard design approval under this part, and information required by statute or by the Commission's regulations, orders, license conditions, or terms and conditions of a standard design approval to be maintained by the licensee, the holder of a standard design approval under this part, the applicant for a standard design certification under this part following Commission adoption of a final design certification rule, and an applicant for a license, permit, a standard design certification, or a standard design approval under this part shall be complete and accurate in all material respects.
(b) Each applicant or licensee, each holder of a standard design approval under this part, and each applicant for a standard design certification under this part following Commission adoption of a final design certification regulation, shall notify the Commission of information identified by the applicant or the licensee as having for the regulated activity a significant implication for public health and safety or common defense and security. An applicant, licensee, or holder violates this paragraph only if the applicant, licensee, or holder fails to notify the Commission of information that the applicant, licensee, or holder has been identified as having a significant implication for public health and safety or common defense and security. Notification shall be provided to the Administrator of the appropriate Regional Office within 2 working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or updating requirements.
§ 53.57 through 53.60 Reserved End Discussion Draft Rule Text
February 11, 2021
- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 1 of 12 Some of the following concerns and industry positions (e.g., quantitative health objectives and as-low-as-reasonably achievable) have been previously provided to the NRC in the NEI letter dated December 23, 2020 and in presentations in NRC public meetings on January 7, 2021 and February 4, 2021. The NRC has not yet provided meaningful input that will help us engage in the Part 53 rulemaking effort. Industry positions are also provided on additional topics (e.g., probabilistic risk assessments and defense-in-depth)
Contents Safety Objectives and Two-Tier Criteria..................................................................................... 2 ALARA.................................................................................................................................... 3 Overall Safety Construct........................................................................................................... 4 Occupational Exposures........................................................................................................... 4 Quantitative Health Objectives.................................................................................................. 4 Quantitative Frequencies.......................................................................................................... 7 Probabilistic Risk Assessments.................................................................................................. 7 Defense-in-Depth.................................................................................................................... 9 Siting................................................................................................................................... 10 Facility Safety Program.......................................................................................................... 10 Addressing Uncertainties........................................................................................................ 11 General Design Criteria.......................................................................................................... 11 Performance-Based Language................................................................................................. 12
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- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 2 of 12 Safety Objectives and Two-Tier Criteria The NRC proposes a new approach in Part 53 in terms of defining the safety objectives. Consistent with 10 CFR Parts 50 and 52, NRCs draft section 53.200 requires advanced nuclear plants to provide reasonable assurance of adequate protection of the public health and safety and the common defense and security. However, the NRC further requires such plants to take such additional measures to protect public health and minimize danger to life or property as may be reasonable when considering technology changes, economic costs, operating experience, or other factors. In its November 2020 Discussion Table, the NRC staff ascribes these two requirements to Atomic Energy Act (AEA) Section 182 (License Applications) and Section 161 (General Duties of the Commission), respectively. In NRC draft sections 53.220 and 53.230, the NRC staff proposes two tiers of safety criteria that apparently are intended to coincide with the two safety objectives.
The NRCs draft rule text is confusing and does not lend itself to easy implementation. As a threshold matter, we are concerned about the manner in which the staff distinguishes between adequate protection and minimize danger as two different safety objectives or standards to be applied in the initial licensing of an advanced reactor. As the NRCs Director of the Office of New Reactors noted in an August 29, 2018 memorandum (ML18240A410) prepared in consultation with OGC, [t]he legal standard for [NRC] licensing decisions is that [the agency] have reasonable assurance of adequate protection - not the elimination of all risk. We agree that reasonable assurance of adequate protection is the operative standard for purposes of new reactor licensing, but believe that any additional measures to protect public health and minimize danger to life would be above and beyond what is adequate. This principle is discussed in the D.C. Circuit decision cited by the staff in its Discussion Table. See Union of Concerned Scientists v. NRC, 824 F.2d 108, 118 (D.C. Cir. 1987) (Under section 161 of the Act, the Commission may order plants to provide extra-adequate protection; in deciding whether to establish or enforce such requirements, the Commission may take into account economic costs.). Importantly, as reflected in its backfitting regulations, the NRC has required such extra-adequate protection measures only when they have demonstrable safety benefits that are justified in light of their economic costs.
Accordingly, in our discussion draft in Attachment 1, we propose structural changes to the draft rule text to more closely tie the safety objectives to the safety criteria, essentially establishing two sets of public protection criteria/requirements: (1) those that are necessary to establish reasonable assurance of adequate protection without regard to economic costs (NEI § 53.2), and (2) those which provide substantial, additional protection (i.e., extra-adequate protection) that is justified on the basis of costs/benefits (NEI § 53.2). We further propose that the NRC provide greater clarity regarding the application of the adequate protection standard to reduce uncertainty and unintended regulatory expansion during application reviews. In particular, we suggest that the staff consider the guidance on adequate protection provided in the aforementioned August 29, 2018 NRO memorandum, and in the January 15, 2019 memorandum (ML19015A290) issued by the Director of
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- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 3 of 12 the Office of Nuclear Material Safety and Safeguards (NMSS). Finally, insofar as the staff opts to retain use of the minimize danger language in its draft regulations (which NEI does not believe is necessary), it should make clear that any associated requirements are intended to provide cost-justified extra-adequate protection.
We believe that this structure aligns with the NRCs intent to address the Atomic Energy Act provisions in a manner that is more clearly understood within the broader context of Part 53 requirements and is more aligned with efficient implementation in the preparation of applications and regulatory decision making.
ALARA The NRCs regulatory approach should be to establish the limits that set the standards for protecting the public health and safety and then regulating in a philosophy that safe is safe. We believe these safety limits are established in the discussion draft NEI § 53.2 and NEI § 53.3, which establishes 0.1 rem during normal operations as its most stringent limit. We believe that a regulatory philosophy of regulating to an undefined and limitless safer than safe standard increases regulatory burden without enhancing the protection of the public health or occupational workers and distracts the licensee from focusing on matter truly important to safety. Furthermore, regulations regarding ALARA drive costs for regulatory compliance without a commensurate safety benefit, and are therefore inconsistent with the development of more risk-informed, performance-based and efficient regulatory framework.
In reviewing the Atomic Energy Act, we note that there is no nexus between ALARA and statutory requirements for the NRCs regulation of nuclear reactors. As the Commission has noted, the ALARA concept is intended to be an operating principle rather than an absolute. 56 Fed. Reg. 23359, 23366 (May 21, 1991). The draft regulations, however, appear to treat ALARA as the latter, and in fact institute an expansion of the ALARA principle beyond what is currently in place for the operating reactors. The issue here is how ALARA applies to engineered features and the degree to which engineered features are required for ALARA purposes. We therefore recommend that the NRC avoid imposing ALARA requirements in Part 53 (e.g., NRC § 53.230(a) and NRC § 53.260(b) in NRC draft rule text.)
Achieving radiation exposure as low as reasonably achievable (ALARA) through the implementation of the licensees radiation protection program is a wise policy. The industry will implement ALARA practices regardless of the NRCs position on the topic. If there are concerns about the industrys commitment to ALARA, a potential solution would be to establish an NRC policy encouraging the industry to implement ALARA practices. Since ALARA requirements are included in 10 CFR Part 20 requirements for radiation protection, Part 53 will need to include an approach to incorporate
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- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 4 of 12 radiation protection requirements (e.g., by either referencing 10 CFR Part 20, except for ALARA requirements, or establishing radiation protection requirements in Part 53).
Overall Safety Construct The NRCs safety criteria include many elements that we view as part of the overall safety construct (e.g., safety functions in NRC § 53.210, licensing basis events in NRC § 53.240, and defense in depth in NRC § 53.250). In the discussion draft in Attachment 1, we propose a more technology-inclusive, performance-based and risk-informed alternative, which takes into account the lessons learned in regulation nuclear reactors for over 40 years and insights from more modern approaches to establishing the safety case for advanced reactors. The result is a more integrated approach to safety, security, siting and emergency preparedness. The proposed requirement in NEI § 53.4 includes all of the aforementioned concepts and integrates the consideration of the radiological hazard, site characteristics, and the design features, human actions and programmatic controls that are necessary to demonstrate that the safety criteria are met. The concept of this requirement aligns with the NRCs proposed bow tie diagram that has formed the core thinking about the Part 53 approach to safety. The requirement is also directly tied to the safety criteria in a manner that is more clearly understood within the broader context of Part 53 requirements and is more aligned with efficient implementation in the preparation of applications and regulatory decision making.
Occupational Exposures We agree with the NRC that occupational exposure limits need to be addressed, and we generally agree with the NRCs approach to reference Part 20. The NRCs approach to elevate occupational exposure to safety criteria is not consistent with the current regulatory framework and would effectively expand the regulation of the engineering design of the plant beyond what currently exists. Furthermore, occupational workers undergo extensive training on radiological exposures and protection. We recommend that occupational exposures therefore be addressed under requirements for radiation protection, rather than as engineering-based safety criteria. Requirements for occupational exposures should be located in the Operations subpart of the rule, which is included in the discussion draft in Attachment 1 as NEI § 53.25.
Quantitative Health Objectives The NRC has proposed including the quantitative health objectives (QHOs) in the rule text for 10 CFR Part 53 safety criteria. The NRCs proposed use of the QHOs is as a Tier 2 criteria, which is intended to be part of the set of criteria that would meet the Atomic Energy Acts (AEAs) Section 161 requirement to minimize danger. This would be the first time the NRC has included the QHOs
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- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 5 of 12 in the regulations. Currently, the NRCs Safety Goals Policy Statement1, consist of two qualitative safety goals (QSGs) (individual and societal risks) backed up by two QHOs. The QHOs are:
- 1. The risk to an average individual in the vicinity of a nuclear power plant of prompt fatalities that might result from reactor accidents should not exceed one-tenth of one percent (0.1 percent) of the sum of prompt fatality risks resulting from other accidents to which members of the U.S. population are generally exposed.
- 2. The risk to the population in the area near a nuclear power plant of cancer fatalities that might result from nuclear power plant operation should not exceed one-tenth of one percent (0.1 percent) of the sum of cancer fatality risks resulting from all other causes.
Industry is evaluating whether we should recommend that the QHOs should remain as an NRC policy statement, or that the QHOs should be codified in rule language. It is recognized that regardless of whether the QHOs are in the Safety Goal Policy or Rule Language, the design, analysis and licensing approach that would be taken by an applicant, and the NRC scope of review would be the same. Likewise, the risk-informed approach in NEI 18-04 would be implemented the same under both approaches. The difference is in the legal compliance with the requirements that exists for the license and the potential to eliminate other requirements, if the QHOs are in the rule language.
The following advantages (pros) and disadvantages (cons) are developed to assist the industry in establishing its position on how the QHOs should be addressed in Part 53.
PROS AND CONS FOR INCLUSION OF QUANTITATIVE HEALTH OBECTIVES IN 10 CFR PART 53 Pros Cons
- 1. Enhances regulatory stability by making it harder for the NRC to change the limits, or make arbitrary judgements.
- 1. Increases regulatory uncertainty by establishing requirements without specifying the consequence limits (i.e., dose for immediate fatalities and latent cancers).
- 2. Enhanced clarity by providing specific limits of acceptable risk to the public for beyond design basis events (BDBEs).
- 2. Reduces regulatory stability since changes to the consequence limits (i.e., dose for immediate fatalities and latent cancers) will now be regulatory limits instead of policy goals.
1 The Safety Goal Policy was first established in 1983 (NUREG-0880) as a qualitative safety goal that the risk from nuclear power plant operation should not be a significant contributor to a persons risk of accidental death or injury, and the QHOs first appears in the 1986 Safety Goal Policy Statement (51FR30028). NRC issued implementation guidance in SECY 89-102. Possible revisions to the Reactor Safety Goal Policy Statement were discussed with the Commission and guidance was provided in the SRMs on SECY-98-101, SECY-99-191, and SECY-00-0077. The SRM to SECY 01-009 disapproved proposed changes and asked the staff to pursue more significant revisions when further progress has been made on the agency's various risk-informed regulatory initiatives.
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- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 6 of 12
- 3. Ensures that regulations explicitly result in risk levels that comply with the QHO limits.
- 3. Is counter to Commissions intent that the QHOs are goals, and not limits.
- 4. The QHOs are more understandable to the public because they are expressed in terms of public health effects.
- 4. Not having consequence limits, and the complexity of demonstrating the QHOs are met, invites contentions by intervenors and increases licensing risk.
- 5. The QHOs are the maximum acceptable consequences, and therefore avoid more conservative surrogate requirements.
- 5. Changes to non-radiological risks (fatalities and cancers due to other causes) can result in changes to the requirements that can force changes to the facility design or operational programs to ensure continued compliance with the new limits. (Note that QHOs would apply to the life of the facility).
- 6. Potential to eliminate the need for some other requirements (e.g., mitigation of beyond design basis events).
- 6. Puts the burden of demonstrating compliance on the applicant (QHO as a Policy Statement puts burden on NRC staff). Analyses and calculations related to demonstrating the QHOs are met are now used to demonstrate legal compliance with the requirements.
- 7. Risks a revision to the QHOs. The NRC discontinued its efforts circa 2000 to update the safety goals so that improvements can be more significant and incorporate experience with risk-informed decision making.
While including the QHO in the rule text is a purer approach to meeting the Safety Goal Policy, doing so could also introduce unforeseen licensing complications, since this is the first time QHOs would be in the regulations. This is particularly true since the NRC proposed requirement for the QHOs does not include the dose limits associated with early fatalities or latent cancer fatalities. Since the NRCs inclusion of the QHOs appears to be used to address beyond design basis events, we believe a discussion of alternative requirements would be helpful. One option, consistent with the NRCs recent rulemaking to address Fukushima type accidents, would be to implement mitigation strategies for beyond design basis events. We have included both options in the discussion draft in to facilitate further discussion before making our recommendation on the preferred option. We further note that if the QHOs are included in the rule text, the language needs to be improved for clarity and consistency with the Safety Goal Policy.
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- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 7 of 12 Quantitative Frequencies The NRC draft rule text proposed to include quantitative frequencies (e.g., once per 10,000 years).
While this is consistent with a risk-informed rule, we also note that using qualitative frequencies in the rule text (e.g., not expected to occur in the life of a nuclear power plant) and including the quantitative frequencies in guidance is consistent with a risk-informed rule. The concern with including quantitative frequencies in the rule text, which to our knowledge would be the first-time quantitative frequencies are included in the rule text, is the possibility of creating unforeseen challenges to the review and approval of new reactor license applications. We are still evaluating which approach we believe should be pursued in Part 53 and have included both options in the discussion draft in Attachment 1, as shown in bracketed italics, such as {are not expected to occur in the life of a nuclear power plant} OR [{have an expected frequency greater than once in 10,000 years}, to facilitate further discussion before making our recommendation on the preferred option.
Finally, the upper bound frequencies should not be used as it introduces an additional level of complication into quantitative metrics and effectively creates an undefined cap based on statistical inferences or temporal changes.
Probabilistic Risk Assessments Industry members have a diverse perspective on the role that probabilistic risk assessments (PRA) should have in Part 53, and we believe Part 53 should accommodate a wide range of approaches to risk evaluation. Some potential applicants plan to use PRAs that align with the ASME/ANS Non-LWR PRA Standard that the NRC is planning to endorse later this year. Other potential applicants plan to use more qualitative approaches to incorporating risk insights, because they believe the burden of performing a PRA is not justified by the marginal benefits of a PRA (versus a qualitative approach) for a simple design with large safety margins. A common view among the industry is that the NRCs proposed requirements (NRC § 53.450(b)) for risk evaluations are overly prescriptive and focused on a single approach to risk evaluation.
In the discussion draft in Attachment 1, we propose a flexible requirement to risk evaluation in Part 53 that allows a wide variety of approaches, which is important to achieve a technology-inclusive and performance-based regulatory framework. We believe the discussion drafts flexible requirement (NEI § 53.5(e)) offers significant advantages as compared to the NRCs proposal. It recognizes that risk-informed is not just a key philosophy of the rule, but also an important consideration in the design of the facility. We recognize that the high-level performance-based approach can result in a lack of clarity about the details of NRC expectations. Thus, we recommend more discussion on whether guidance for a graded approach to risk evaluations would be helpful.
We recognize that Part 52 requires a PRA be included in the application for a license, that the PRA be updated and maintained for the life of operations, and that the NRC is in the process of changing
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- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 8 of 12 Part 50 to align with the Part 52 requirements on PRA. We also understand that the Part 52 requirements and the NRC position on the use of PRAs is based on the Commissions PRA Policy Statement (policy). However, we also believe that transformational improvements can only be achieved in Part 53 if we reassess the applicability of all Part 50/52 requirements, policies and long-held staff positions, to determine whether what was developed based on LLWR technology is applicable to a technology-inclusive approach. As it relates to the use of PRAs, we think a one-size-fits-all approach is not consistent with a technology-inclusive approach, and can result in increased regulatory burden for some designs.
The NRCs PRA Policy Statement encourages increased use of PRA in all regulatory matters where practical and to the extent supported by the state of the art in the PRA and data and in a manner that complements deterministic approaches. It is important to note that the policy clarifies that deterministic-based regulations have been successful in protecting the public health and safety.
The policy is consistent with the recognition that a PRA is not needed to provide reasonable assurance of adequate protection of the public health and safety. In fact, the advantages of using a PRA are stated in the policy in terms of the benefits in regulatory efficiency:
- 1. Focusing regulations on items most important to safety
- 2. Eliminate unnecessary conservatism
- 3. Meeting requirements defined in probabilistic terms
- 4. Estimating safety of systems with very large uncertainties
- 5. Support evaluation of additional regulatory requirements in the cost-benefit safety improvements for backfit analyses
- 6. Support defense-in-depth philosophy by identifying and addressing weaknesses or overly conservative requirements The intention of the NRCs PRA Policy Statement to increase the use of PRAs and achieve the expected benefits may not be realized for all designs regulated under the technology-inclusive Part
- 53. Specifically, simple designs with and/or those with large safety margins (an example of which may be micro-reactors) could see increased regulatory burden if required to use a PRA instead of being allowed to rely upon the use of risk insights from a qualitative risk evaluation. For these designs, the increased regulatory burden resulting from an inflexible PRA requirement in Part 53 would be in conflict with the expected benefits in the NRCs PRA Policy Statement related to increased regulatory efficiency. Further, the use of a PRA versus a qualitative risk for these designs may not even result in a substantial safety improvement.
Finally, we believe that the NRCs PRA Policy Statement would allow (in particular by the use of where practical as a conditional imposition of the policy) the use of qualitative risk evaluations in lieu of PRAs for designs where the burdens of using a PRA are outweighed by benefits. A qualitative risk evaluation, for certain designs, would achieve the underlying purpose of the policy by using risk insights to support the defense-in-depth philosophy, and the intended benefits of increasing
February 11, 2021
- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 9 of 12 regulatory efficiency. However, if the policy is not flexible enough to allow for a qualitative risk evaluation under some conditions, then a new PRA policy for Part 53 would be beneficial to ensure new rule enables regulatory efficiency. In either case, we believe guidance would be an appropriate place to document the criteria for which a design may utilize a qualitative risk evaluation in lieu of a PRA.
Defense-in-Depth The NRC proposed approach to defense-in-depth in NRC § 53.250 is an overly prescriptive approach this is likely to increase regulatory burden without enhancing safety. The proposed Part 53 requirement apply uniformly across the entire scope of the facility, which could result in the need to include systems, structures and components that are otherwise unnecessary to the safety case and to meet the public protection criteria. Furthermore, the NRC proposal contains the following statement, No single design or operational feature, no matter how robust, should be exclusively relied upon to meet the safety criteria of 10 CFR part 53, which is overly prescriptive and simply does not belong in a performance-based rule. As an example, consider a design for which the failure of a single design feature, relied upon exclusively to meet the public protection criteria, would result in a dose to the public that is less than 1 rem. In this case, the consequences of the failure are so far below the public protection limit, i.e., 25 rem, it should not require a redundant design feature or additional barrier. Such a prescriptive requirement would not allow a risk-informed approach to the design. In contrast, the approach in the regulatory framework for Parts 50 and 52 is to apply defense-in-depth as a design philosophy, with a few targeted requirements, for example in Fire Protection and Risk Categorization, and even in these cases the focus of the requirement is to maintain defense-in-depth, not to prescribe how it must be achieved.
We do agree that defense-in-depth is an important design philosophy that should be applied to new reactors licensed under Part 53. We also believe that an approach consistent with Parts 50 and 52, where there is no cross-cutting defense-in-depth requirement, but rather it is a design philosophy, is the only approach that can achieve a technology-inclusive, performance-based and risk-informed regulatory framework. This does not mean that requirements should be established without consideration or expectation that defense-in-depth will be part of the new reactor design. In the discussion draft in Attachment 1, we include two requirements related to defense-in-depth. In NEI
§ 53.4(e)(1), there is a requirement that the facility utilize a combination of structures, systems and components to provide prevention, barriers and mitigation. And in NEI § 53.5(f), there is a requirement that the design and analysis include, measures, such as increased performance margins in the design of SSCs to address design and analytical uncertainties. While we believe the latter requirement is more related to the purpose of NRC § 53.250, we believe that the combination of the discussion draft requirements provide a technology-inclusive, performance-based and risk-informed regulatory approach that provides reasonable assurance that new reactors licensed under Part 53 will adequately address defense-in-depth. However, if the NRC believes more clarity is
February 11, 2021
- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 10 of 12 needed on expectations related to implementing defense-in-depth philosophies in the design of new reactors, then guidance could be created to accompany the rule.
Siting Facilities licensed under 10 CFR Part 53 will need to consider the siting of the reactor to provide reasonable assurance that each site of a nuclear power reactor will protect the public health and safety. While we agree with the NRC approach to include siting requirements in Part 53, we are concerned that the NRC proposal does little more than relocate the siting requirements from one part to another. We believe an incremental approach to Part 53 is a missed opportunity to achieve transformational changes that result in a more efficient regulatory framework to protect the public health and safety. Part 53 should completely reevaluate the approach to siting by recognizing that it is largely the same as it was originally conceived in 1960s/1970s, and that Part 53 is being built upon more a more modern and flexible regulatory framework.
The alternative approach in the discussion draft in Attachment 1 is a technology-inclusive, performance-based and risk-informed approach to reactor site criteria that better aligns with the Part 53 public protection paradigm. Specifically, the alternative approach requires (NEI §53.4(b))
that the characteristics of the site that have a significant impact on the ability of the facility to meet the public protection criteria in NEI §53.2 and NEI §53.3 and the location of the site boundary (NEI
§53.4(b)) be part of the facility characteristics. Part 53 also requires the establishment of design features and human actions (NEI §53.4(e)) to protect against manmade hazards related to the site (NEI §53.4(d)(3)) such that the public protection criteria are met. Part 53 establishes the site boundary, in lieu of the low population zone and exclusion area boundary as the key boundary, in alignment with the on-going SMR Emergency Preparedness rulemaking, streamlining the licensing basis of the facility. Part 53 does not establish a requirement for distance from a population center, since it is not necessary, nor does it include specific requirements for the seismic and geologic criteria, since it is not needed for a technology-inclusive, performance-based and risk-informed approach. Therefore, facilities licensed under 10 CFR Part 53 do not need to comply with the reactor site criteria of Part 100.
Facility Safety Program During the NRC public meeting on January 7, 2021, the NRC proposed a requirement for a Facility Safety Program under Subpart F of Part 53. The proposed requirement has no regulatory precedent in the U.S. and would increase the regulatory burden that the NRC imposes on the industry. We are still evaluating our position on the proposed requirement, and while we see that this could have operational benefits, we would need to have confidence that the benefits to the industry outweigh the increased burden in order to support the idea. The NRC stated that the need for such a requirement is based upon an expectation that there will be a diverse set of technologies (e.g., light-
February 11, 2021
- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 11 of 12 water small modular, high temperature gas, molten salt and micro-reactors) to be regulated at a large number of facilities (resulting from the small size of many designs) requiring NRC oversight.
However, it is not clear that the expected future conditions would necessitate a different approach to addressing operating experience and generic issues, nor is it clear that a facility safety program would reduce NRC oversight costs, since the inherent safety and simplicity of advanced reactors is expected to result in fewer resources to conduct oversight. While the NRC offered a potential benefit to industry in terms of more efficient licensing, we are skeptical that the NRC staff would be able to adapt their review practices to achieve the envisioned benefits. If the NRC intends to pursue this requirement, then they should demonstrate that the benefits to industry clearly outweigh the burden. Such benefits could include the elimination of the NRCs licensee reporting requirements (e.g., 50.72 and 50.72) and elimination of the NRC generic issues program. The NRC should also provide examples that demonstrated how past operating experience would have been more efficiently addressed for licensee that met the proposed facility safety program requirements.
Addressing Uncertainties We agree with the ACRS that uncertainties associated with limited information for the first reviews of novel technologies will need to be addressed in a systematic manner. However, we believe that the ACRSs recommendation to return to the methods used in the earliest days of licensing nuclear reactors would be a step back to more deterministic methods. The approach cited by the ACRS would seem to have the applicant postulate the worst possible accident, and then the ACRS try to postulate an even more severe accident based on plausibility. This subjective, deterministic approach is inconsistent with the direction of NEIMA to pursue a risk-informed performance-based regulatory framework. This would also ignore the progress made in modern analysis methods capability to systematically identify initiating events, event sequences and address uncertainties, including the approach in NEI 18-04 Modernization of Technical Requirement for Licensing of Advanced Non-Light Water Reactors that was endorsed by the NRC in Regulatory Guide 1.233 and by the Commission in SRM-SECY 19-0117. The ACRS also recommends that the NRC include a pathway for licensing prototype facilities to address designs where there is a lack of operating experience or an inability to perform experiments with sufficient similitude to the planned full-scale design to reduce uncertainties to an acceptable level. While we agree that a pathway to licensing prototype facilities is one option to address these situations, we believe that it is not necessary to require a prototype license and that applicants should be allowed to utilize other pathways to address uncertainties related to limited operating experience or full-scale testing data. Our proposed approach to address uncertainties in the design and analysis of the plant is included in the discussion draft in Attachment 1 as NEI § 53.5(f).
General Design Criteria
February 11, 2021
- Industry Concerns with NRC Preliminary Rule Text for Key Part 53 Topics and Proposed Alternatives Page 12 of 12 The ACRS recommended that the concept of the general design criteria (GDC) specified in Appendix A to 10 CFR Part 50 be incorporated into the Part 53 framework. They note that the GDC improved the predictability and efficiency of NRC reviews of licensing applications. We agree that the use of design criteria to establish design-specific requirements for the design, fabrication, construction, testing and performance is useful in ensuring that needed structures, systems and components are capable of providing their credited functions. Part 50 is able to utilize the GDC because it was developed solely around large light-water reactors (LWRs). However, the inclusion of GDC in the Part 53 regulations is problematic for a technology-inclusive Part 53 rule, since the variation in reactor technologies is so large that it is not possible to develop a single set of general criteria. This is reflected in the NRCs development of advanced reactor design criteria, which resulted in several technology-specific design criteria, which also are expected to lead to numerous deviations at the design-specific level. We recommend that Part 53 not specify design criteria in the regulations, and we expect further details of how they are addressed to be discussed related to the overall safety construct topic (rather than during the safety criteria topic). Our proposed approach to address principal design criteria is included in the discussion draft in Attachment 1 as NEI § 53.5(b).
Performance-Based Language We recognize that the NRC proposed rule text is a draft and likely will be revised to be more performance-based and clear. However, the structure of the NRCs preliminary rule text is overly complicated and the detail is overly prescriptive, which results in an inflexible rule that is difficult to understand and implement. The rule text included the Attachment 1 discussion draft is developed to be structured and written with a simplification philosophy that we believe leads to more clarity and efficiency in the implementation of the rule through development of applications and regulatory decision making. To aid these revisions, we have identified that the use of certain terminology (e.g.,
high-confidence, mean frequency) are not needed in the regulation, since they are details more appropriately addressed in guidance, and they can lead to later complications. Similarly, the NRC repeatedly uses the term design features and programmatic controls in describing what the applicant must do to meet the performance-based requirement. Defining the specifics of what must be provided is overly prescriptive. Furthermore, requirements related to the design and programs are better addressed in the safety construct, i.e., what needs to be done to meet the safety criteria, to be developed later.
From:
NICHOL, Marcus To:
Coyne, Kevin Cc:
Veil, Andrea; Taylor, Robert; Shams, Mohamed; Segala, John; Beall, Bob; Reckley, William; Valliere, Nanette; RulemakingComments Resource
Subject:
[External_Sender] Industrys Concerns about NRC Proposed Approaches to Part 53, and Alternative Discussion Draft for the NRCs Rulemaking on, Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN-3150-AK31; NRC-2019-0062)
Date:
Thursday, February 11, 2021 12:12:45 PM Attachments:
02-11-2021_NRC_Part 53 Industry Concerns and Alternatives with attachments.pdf THE ATTACHMENT CONTAINS THE COMPLETE CONTENTS OF THE LETTER
February 11, 2021
Dr. Kevin Coyne Acting Director, Division of Rulemaking, Environmental, and Financial Support Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Industrys Concerns about NRC Proposed Approaches to Part 53, and Alternative Discussion Draft for the NRCs Rulemaking on, Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN-3150-AK31; NRC-2019-0062)
Project Number: 689
Dear Dr. Coyne:
The Nuclear Energy Institute (NEI) and its members appreciate the Nuclear Regulatory Commissions (NRC) staffs approach to develop preliminary rule language to facilitate discussion with stakeholders on the concepts for the Part 53 rule. To date, the staff has issued preliminary rule language for safety criteria, design and analysis, facility safety program and siting, which has been helpful in identifying and understanding potential approaches for Part 53 among all stakeholders. We also appreciate the NRCs public meetings, which have helped to foster stakeholder discussion on the potential approaches. We remain dedicated to providing detailed input on our perspectives in order to support the staffs efforts in a timely manner. However, we are concerned that what the staff has proposed is overly complex and not consistent with the goal to have a more efficient regulatory framework, and that the rulemaking could result in a Part 53 rule that would be an unattractive option to license advanced reactors.
The purpose of this letter is to provide the NRC with a discussion draft of a Part 53 rule that was developed with NEIs Part 53 Task Force (Attachment 1). This task force is comprised of members from advanced reactor companies, utilities and vendors. We believe that the attached draft would achieve the industrys vision, provided in our letter dated October 21, 2020, that licensing new reactors under the new Part 53 rule will be the most efficient option for all new reactor applicants and will meet industry needs for schedule, cost and predictability.
Marcus R. Nichol Senior Director New Reactors
Nuclear Energy Institute 1201 F St NW, Suite 1100
Washington, DC 20004 www.nei.org P: 202.739.8131 M: 202.316.4412 E: mrn@nei.org
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