Gap 5: Safety and 

Risk Assessment Methodology

 NRC’s regulations require licensed facilities to demonstrate adequate 

assurances of safety, limiting risk to acceptable levels

 Existing regulations in 10 CFR Parts 50 and 70 do not adequately 

address the potential hazards, consequences, and risks of 

reprocessing facilities

• Potential source terms (e.g., fission products and actinides), 

number of scenarios, and sequence consequences are greater 

than existing fuel cycle facilities

• Given these potential risks, the existing requirements solely 

based on the integrated safety analysis (ISA) approach of 

10 CFR Part 70 may not be adequate
 Incorporate quantitative risk assessment in safety evaluations, through 

hybrid approach that includes ISA and probabilistic risk assessment 

methodologies

• Use ISA to identify and categorize accident sequences 

• Use probabilistic risk assessment methodologies (to the extent 

practicable) to evaluate high consequence and very high 

consequence events

• Apply safety controls (items relied on for safety) to reduce total risk

• Reduce risks as needed to meet the appropriate NRC risk 

limits/criteria

• Minimize total risks to receptors by applying As Low As 

Reasonably Achievable (ALARA)/As Low As Reasonably 

Practicable (ALARP)

• Identify general design criteria/controls to meet 

ALARA/ALARP

Gap 5: NRC Staff Position
 The staff also anticipates

• Recommending requirements analogous to 10 CFR 70.23(a)(3) 

and (a)(4) for high consequence and very high consequence 

events

• Developing guidance to support application quantitative risk 

assessment methodologies

Gap 5: NRC Staff Position (cont’d.)
Gap 5: NRC Staff Position (cont’d.)

( /yr)
 The Advisory Committee on Reactor Safeguards (ACRS) 

White Paper (NUREG-1909)

• In favor of using probabilistic risk assessment over ISA

• Recommends formulating ALARA requirements that 

establish design objectives

 NEI Paper, “Integrated Safety Analysis: Why It Is Appropriate 

For Fuel Recycling Facilities,” September 2010 

• In favor of using ISA with a quantitative assessment of 

fission product releases to members of the public for high 

consequence events, based on availability of data

 General Electric-Hitachi Industry Paper, “Applying Nuclear 

PRA to a Nuclear Fuel Facility Integrated Safety Analysis” 

• Discusses use of probabilistic risk assessment 

methodologies based on related nuclear industry data

Gap 5: Previous Stakeholder Input
 The Atomic Energy Act of 1954, as amended, requires 

production facilities to have licensed operators (42 USC 

2137)

 The current NRC regulations in 10 CFR Part 55—Operator’s 

Licenses are not applicable, in whole, to operators of 

reprocessing facilities

 The NRC staff is developing a regulatory framework to 

establish regulations for licensed operators and criteria for 

testing and licensing operators

Gap 7: Licensed Operators and 

Criteria for Testing and Licensing 

Operators
 Apply a risk-informed and performance-based approach to 

determine which personnel (including senior operators) need 

to be licensed and the requirements for their licensure

 Personnel whose actions are clearly related to safety, such as 

preventing or mitigating very high consequence events, will 

be licensed by NRC

 Include requirements similar to 10 CFR Part 55 approach for 

a systems approach to training, tests, simulation facilities, 

requalification, and roles and responsibilities

Gap 7: NRC Staff Position
 NEI White Paper, “Regulatory Framework for Recycling 

Nuclear Fuel,” addresses operator licensing

• Recommends operators be certified by the facility 

licensee according to an NRC-approved program

• Defines certified operators for possible fission product 

releases to members of the public for high consequence 

events

• No senior operators

Gap 7: Previous Stakeholder Input
 NRC establishes minimum requirements for proposed 

facilities or applications of licensed radioactive materials 

provide assurance that 

• Important to safety structures, systems, and components 

will have the ability and reliability to perform their 

intended safety functions 

• Uncertainties and errors, from design and analysis and 

from unknowns, are adequately addressed

• There is adequate defense in depth, redundancy, and 

diversity

• Balance of plant and unanalyzed situations do not 

negatively impact safety

 General design criteria currently do not exist for 

reprocessing facilities

Gap 9: General Design Criteria
 NRC staff proposes 10 categories for general design criteria:

Overall Radiological Protection 

Confinement and Containment Physical Security

Process Safety Material Control & Accounting

Criticality Safety Fuel and Radioactive Waste,

Siting Decommissioning 

 The NRC staff identified 78 potential general design criteria within 

these ten categories

Gap 9: NRC Staff Position
 NEI proposed 28 General Design Criteria (2008 White Paper)

• Largely from 10 CFR Part 50 for nuclear power plants

• Some additions from 10 CFR Parts 70 and 72

• Proposed thresholds for applicability of some general 

design criteria 

• Applicant to explain how general design criteria are 

addressed to achieve performance requirements

• Some general design criteria may not be necessary based 

on ISA results 

 Other stakeholders input includes

• Need to minimize regulatory uncertainty on general design 

criteria

• Establish general design criteria by regulation, not 

guidance

Gap 9: Previous Stakeholder Input
 Currently, there are no regulations for one-step licensing, 

combine license (COL) of a reprocessing facility

• 10 CFR Part 52 does not apply to spent nuclear fuel 

reprocessing facilities

• Requirements for approval of applications for domestic 

licensing of special nuclear material in 10 CFR 70.23 do 

not address reprocessing facilities

Gap 10: One-Step Licensing and 

Inspection Testing and Acceptance 

Criteria (ITAAC)
 General license authority similar to nuclear power plant 

requirements of 10 CFR Part 52

 One-step licensing process for reprocessing facilities will 

include inspection process (ITAAC or ITAAC-like) to confirm 

the facility meets the design, construction, and licensing 

requirements

 The proposed regulation will provide requirements and 

procedures for NRC to issue an early site permit that could 

be referenced in COL

 Reserve areas in the regulation for technology-specific 

requirements for particular technologies, such as aqueous 

and electro-chemical processes

Gap 10: NRC Staff Position
 NEI White Paper, “Regulatory Framework for Recycling 

Nuclear Fuel,” includes a framework that could allow an 

applicant the flexibility for either a two- or one-step licensing 

process 

 Industry has expressed interest in a one-step approach with 

the flexibility for a two-step licensing process

Gap 10: Previous Stakeholder Input
 The Atomic Energy Act of 1954, as amended, requires 

technical specifications for production facilities (42 USC 

2137)

 10 CFR Part 70 does not require technical specifications

 The IROFS to be addressed by technical specifications need 

to be established for new 10 CFR Part 7X regulation

Gap 11: Technical Specifications
 Define safety limits and limiting control settings, limiting 

conditions of operation, surveillance requirements, and 

design requirements for IROFS that address very high 

consequence accident sequences

 Additional technical specifications may be needed to ensure 

safe operation within the bounds of the safety analysis (e.g., 

spent fuel burn-up limits and effluents), as well as 

administrative/programmatic technical specifications 

Gap 11: NRC Staff Position
 NEI White Paper, “Regulatory Framework for Recycling 

Nuclear Fuel,” recommended technical specifications of 

items relied on for safety for high consequence events 

involving fission product releases to an individual located 

outside the controlled area

Gap 11: Previous Stakeholder Input
 Gap questions provided in handout material