A Proposed 

Risk Management 

Regulatory 

Framework

A Proposed Risk 

Management Regulatory 

Framework

A report to NRC Chairman Gregory B. Jaczko

from the Risk Management Task Force

Commissioner George Apostolakis, Head

Members

Christiana Lui, RMTF Executive Director

Mark Cunningham

George Pangburn

William Reckley

With contributions from

John Adams

Michel Call

Dennis Damon

Don Dube

Earl Easton

Timothy McCartin

Geary Mizuno

Joel Piper

April 2012 

Risk Management Task Force | iii

A Proposed Risk Management Regulatory Framework

ABSTRACT

At the request of Chairman Gregory B. Jaczko, a task force headed by Commissioner 

George Apostolakis prepared this report. The task force’s charter was to develop a strategic 

vision and options for adopting a more comprehensive, holistic, risk-informed, performancebased regulatory approach for reactors, materials, waste, fuel cycle, and transportation that 

would continue to ensure the safe and secure use of nuclear material. The proposed risk 

management regulatory framework builds upon well established practices, such as the NRC’s 

defense-in-depth philosophy and its policies to incorporate risk-informed and performancebased approaches into the agency’s regulation and oversight of byproduct, source, and special 

nuclear materials. Risk management is being adopted by many different organizations, 

including Federal agencies, and would seem to be the logical next step in the evolution of the 

NRC’s regulatory programs. The report describes a proposed risk management regulatory 

approach that could be used to improve consistency among the NRC’s various programs and 

discusses implementing such a framework for specific program areas. 

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A Proposed Risk Management Regulatory Framework

We Athenians, in our own persons, take our decisions on policy 

and submit them to proper discussions; for we do not think that 

there is an incompatibility between words and deeds; the worst 

thing is to rush into action before the consequences have been 

properly debated. And this is another point where we differ 

from other people. We are capable at the same time of taking 

risks and of estimating them beforehand. Others are brave out 

of ignorance; and, when they stop to think, they begin to fear. 

But the man who can most truly be accounted brave is he who 

best knows the meaning of what is sweet in life and what is 

terrible, and then goes out undeterred to meet what is to come.

 

Funeral Oration Delivered by Pericles circa 430 B.C. 

Thucydides, History of the Peloponnesian War, 

Book B, Paragraph 40

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CONTENTS

ABSTRACT.................................................................................................................. iii

A MODERN THINKER IN ANTIQUITY ......................................................................... v 

LIST OF FIGURES....................................................................................................... ix

LIST OF TABLES......................................................................................................... xi

EXECUTIVE SUMMARY............................................................................................ xiii

ACKNOWLEDGEMENT .......................................................................................... xxxi

1. INTRODUCTION

1.0 Background...........................................................................................1-1

1.1 The Risk Management Task Force Report ...........................................1-3

1.2 References............................................................................................1-4

2. A PROPOSED RISK MANAGEMENT REGULATORY FRAMEWORK

2.0 Background...........................................................................................2-1

2.1 Considerations and Findings ................................................................2-1

2.2 Recommendations................................................................................2-7

2.3 References..........................................................................................2-12

3. RISK MANAGEMENT DECISIONMAKING PROCESS

3.0 Background...........................................................................................3-1

3.1 Identify Issues.......................................................................................3-3

3.2 Identify Options.....................................................................................3-3

3.3 Analyze .................................................................................................3-4

3.4 Deliberate..............................................................................................3-5

3.5 Implement Decision ..............................................................................3-6

3.6 Monitor..................................................................................................3-7

3.7 Communications ...................................................................................3-7

3.8 References............................................................................................3-8

4. IMPLEMENTATION OPTIONS

4.1 Introduction......................................................................................4.1-1

4.1.1 References.........................................................................................4.1-3

4.2 Nuclear Reactors .............................................................................4.2-1

4.2.1 Power Reactors .................................................................................4.2-1

4.2.1a Background........................................................................................4.2-1

4.2.1b RMTF Findings and Recommendations for Power Reactors ............4.2-5

4.2.1.1 Implementation Options - Operating Power Reactors .....................4.2-14

4.2.1.2 Implementation Options - New Reactors ........................................4.2-16

4.2.1.3 Implementation Options - Generation IV Power Reactors...............4.2-19

4.2.2 Nonpower Reactors .........................................................................4.2-22

4.2.2a Background......................................................................................4.2-22

4.2.2b Findings and Recommendations for Nonpower Reactors ...............4.2-22

4.2.3 References.......................................................................................4.2-27

4.3 Material Uses....................................................................................4.3-1
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viii | Risk Management Task Force

4.3.1 Background........................................................................................4.3-1

4.3.2 RMTF Findings and Recommendations for Materials Uses ..............4.3-4

4.3.3 References.......................................................................................4.3-10

4.4 Waste ................................................................................................4.4-1

4.4.1 Low-Level Waste ...............................................................................4.4-1

4.4.1.1 Background........................................................................................4.4-1

4.4.1.2 RMTF Findings and Recommendations for Low-Level Waste...........4.4-4

4.4.2 High-Level Waste...............................................................................4.4-7

4.4.2.1 Regulatory Background and Framework on High-Level Waste Disposal..4.4-7

4.4.2.2 RMTF Findings and Recommendations for High-Level Waste..........4.4-9

4.4.2.3 Implementation Options for High-Level Waste ................................4.4-12

4.4.3 References.......................................................................................4.4-13

4.5 Uranium Recovery ...........................................................................4.5-1

4.5.1 Background........................................................................................4.5-1

4.5.2 RMTF Findings and Recommendations for Uranium Recovery ........4.5-3

4.5.3 Implementation Options for Risk Management in Uranium Recovery4.5-5

4.5.4 References ......................................................................................4.5-6

4.6 Fuel Cycle.........................................................................................4.6-1

4.6.1 Background........................................................................................4.6-1

4.6.2 RMTF Findings and Recommendations for Fuel Cycle .....................4.6-3

4.6.3 Implementation Options for Risk Management in Fuel Cycle Facilities 4.6-5

4.6.4 References.........................................................................................4.6-7

4.7 Interim Spent Fuel Storage .............................................................4.7-1

4.7.1 Background........................................................................................4.7-1

4.7.2 RMTF Findings and Recommendations for Spent Fuel Storage .......4.7-4

4.7.3 Implementation Options for Risk Management in Interim SNF Storage. 4.7-8

4.7.4 References.........................................................................................4.7-9

4.8 Transportation .................................................................................4.8-1

4.8.1 Regulatory Background and Framework ...........................................4.8-1

4.8.2 RMTF Findings and Recommendations ............................................4.8-3

4.8.3 Implementation Options for Risk Management in Transportation......4.8-5

4.8.4 References.........................................................................................4.8-6

5. CONCLUSIONS ..............................................................................................................5-1

APPENDIX A RISK MANAGEMENT SYSTEMS............................................................... A-1

APPENDIX B RISK MANAGEMENT – ANALYSES AND DELIBERATION .................... B-1

APPENDIX C DEFENSE-IN-DEPTH.................................................................................. C-1

APPENDIX D PERFORMANCE-BASED REGULATION.................................................. D-1

APPENDIX E EXPERIENCE WITH RISK-INFORMED REGULATION.............................E-1

APPENDIX F NUCLEAR POWER REACTORS – LICENSING-BASIS EVENTS............F-1

APPENDIX G SAFETY CLASSIFICATION OF SYSTEMS, STRUCTURES, 

AND COMPONENTS............................................................................... G-1

APPENDIX H NUCLEAR POWER REACTORS – CHAPTER 4 ALTERNATIVES.......... H-1

APPENDIX I STAKEHOLDER INTERACTIONS................................................................I-1

APPENDIX J FINDINGS AND RECOMMENDATIONS J-1
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LIST OF FIGURES

Figure 2-1 The Regulatory Decisionmaking Process 2-10

Figure 2-2 A Proposed Risk Management Regulatory Framework 2-11

Figure 3-1 The Regulatory Decisionmaking Process 3-1

Figure 3-2 Deliberations 3-6

Figure 4-1 Regulatory Framework for Nuclear Power Reactors 4.2-8

Figure 5-1 A Proposed Risk Management Regulatory Framework 5-1

Figure A-1 Framework for Environmental Health Risk Management

The Presidential/Congressional Commission on Risk

Assessment and Risk Management. Final Report 1997 A-4

Figure A-2 Understanding Risk: Informing Decisions in a Democratic

Society, National Resource Council, 1996 A-6

Figure A-3 International Risk Governance Council (IRGC)

Risk Governance Framework A-7

Figure A-4 GAO Risk Management Framework

(U.S. Government Accountability Office) A-8

Figure A-5 NASA Framework

 (U.S. National Aeronautics and Space Administration) A-9

Figure A-6 U.S. Coast Guard Framework A-10

Figure A-7 ISO Framework (International Organization for Standardization) A-11

Figure A-8 DHS Risk Management Process

 (U.S. Department of Homeland Security) A-11

Figure A-9 The Regulatory Decisionmaking Process A-15

Figure B-1 The Regulatory Decisionmaking Process B-1

Figure B-2 A Proposed Risk Management Regulatory Framework B-2

Figure B-3 Balancing Risk Assessments and Deterministic Techniques B-4

Figure B-4 Technical Analysis Techniques and Deliberation B-5

Figure B-5 Deliberations B-7

Figure B-6 RG 1.174 Integrated Decisionmaking Process B-10
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x | Risk Management Task Force

Figure B-7 RG 1.174 Process and Decision Criteria B-11

Figure B-8 ROP Significance Determination Guidance B-11

Figure B-9 F-C Curves from NUREG-1860 and NGNP B-12

Figure B-10 Generic F-C Curve, with ALARA Region B-12

Figure B-11 Risk-Informed Decisionmaking B-13

Figure B-12 Implementation of Decisions B-15

Figure E-1 Status of Risk-Informed Activities (2002) E-1

Figure F-1 Nuclear Event Analysis Margins (NEA, 2007) F-6

Figure F-2 Event Categories F-8

Figure F-3 Event Categories and IAEA Defense-in-depth F-9

Figure F-4 Licensing-basis Events, Documents and Programs F-9

Figure F-5 Regulatory Framework for Nuclear Power Reactors F-13

Figure G-1 10 CFR 50.69 Risk-Informed Safety Class Categories G-8

Figure G-2 Representations of Safety Classification Concepts G-17

Figure H-1 Regulatory Framework for Nuclear Power Reactors H-6

Figure H-2 Possible Alternatives for Nuclear Power Reactors H-7

Figure H-3 Alternative 1a (NRC identified, NRC approval) H-8

Figure H-4 Alternative 1b (NRC identified, performance-based) H-9

Figure H-5 Alternative 2, Risk Management (ALARA) H-13

Figure H-6 10 CFR 50.69 Risk-Informed Safety Class Categories H-18

Figure H-7 Event Categories and IAEA Defense-in-depth H-19
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LIST OF TABLES

Table F-1 Approaches to External Hazards F-4

Table F-2 Licensing Approach to Events (current approach) F-10

Table G-1 ANS/ASME Safety Classes G-3

Table G-2 IAEA Safety Categories G-12

Table H-1 Licensing Approach to Events (current approach) H-20

Table I-1 Summary of Comments from External Stakeholders I-10

Table J-1 RMTF Findings and Recommendations J-1

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A Proposed Risk Management Regulatory Framework

EXECUTIVE SUMMARY

Introduction

The Atomic Energy Act (the “Act”) and other applicable laws establish the fundamental basis 

by which the U.S. Nuclear Regulatory Commission (NRC) regulates the civilian uses of nuclear 

materials. The implementation of these laws has evolved considerably since the original Act, 

with concepts such as “defense in depth” and methods such as risk assessment emerging as 

important aspects of this evolution.

In early 2011, at the request of Chairman Gregory B. Jaczko, Commissioner George Apostolakis 

agreed to lead a Task Force to evaluate how the agency should be regulating 10 to 15 years 

in the future. More specifically, the Task Force was chartered “to develop a strategic vision 

and options for adopting a more comprehensive and holistic risk-informed, performance-based 

regulatory approach for reactors, materials, waste, fuel cycle, and transportation that would 

continue to ensure the safe and secure use of nuclear material.”

This report describes the findings and recommendations of this evaluation. The underlying 

analysis was performed by a team of NRC staff (named the Risk Management Task Force 

(RMTF)) under the direction of Commissioner Apostolakis, with contributions from additional 

NRC staff. The RMTF also benefitted from comments and suggestions provided by members of 

the public and other NRC staff.

The accident at the Fukushima Dai-ichi nuclear power plants in Japan occurred shortly after the 

RMTF was established. The team’s analysis has been influenced by the events at Fukushima 

and the subsequent studies, including the NRC Near-Term Task Force, and the continuing 

discussions on the accident’s implications for U.S. nuclear power plants.

The findings and recommendations are compiled into two groups, with the most important 

described below.1 The first group addresses agencywide, more strategic issues, describing a 

structure of goals and objectives that could be the framework for NRC regulatory activities 10 to 

15 years in the future. The second group addresses what changes would be needed in specific 

program areas (e.g., power reactors, materials) in the next several years to ensure that the 

framework is implemented. It is acknowledged that these recommendations, if adopted, would 

require actions and resources to fully address their implementation that go well beyond the 

charter of the RMTF.

Agencywide Findings and Recommendations

The RMTF found many positive attributes of the individual regulatory activities now performed 

by the NRC. As a result, the RMTF determined that its proposed strategic vision could and 

should be more evolutionary than revolutionary. The findings of the Task Force are listed below.

•	 Finding: Whether used explicitly, as for power reactors, or implicitly, as for materials 

programs, the concept of defense in depth has served the NRC and the regulated 

industries well and continues to be valuable today. However, it is not used consistently, 

and there is no guidance on how much defense-in-depth is sufficient.

1 The complete set of findings and recommendations is provided in Appendix J.
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xiv | Risk Management Task Force Executive Summary

•	 Finding: Risk assessments provide valuable and realistic insights into potential 

exposure scenarios. In combination with other technical analyses, risk assessments can 

inform decisions about appropriate defense-in-depth measures.

Considering these findings, the RMTF proposes a framework shown in Figure ES-1 and 

discussed below. 

The RMTF proposes that risk management should be stated as the NRC’s objective 

(“Objective” in Figure ES-1). Declaring that this is the agency’s objective is a natural next 

step in the evolution of the NRC’s regulatory practices. This term is widely encountered in the 

management literature and is gaining greater use in other Federal agencies. In addition, it 

explicitly recognizes that adequate protection of public health and safety is not synonymous with 

absolute safety and that the NRC’s role is to ensure that risks from the use of nuclear materials 

are well managed. Finally, establishing a common language of risk management across all 

NRC activities is consistent with the principles of good regulation. 

Figure ES-1 A Proposed Risk Management Regulatory Framework 
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A Proposed Risk Management Regulatory Framework

As used in the proposed framework, the concept of risk consists of answers to the three 

standard questions: What can go wrong? How likely is it? What are the consequences? It 

is acknowledged that some program areas, such as power reactors and high-level waste 

repositories, take a more explicit and quantitative approach to answering all three questions, 

whereas other program areas take more qualitative approaches. However, all NRC programs 

practice risk management. 

The RMTF recognizes the importance of translating the concept of risk management into more 

operational terms. The task force proposes doing this by integrating the traditional concept of 

defense in depth and the methods of risk assessment (“Risk Management Goal” in Figure

ES-1). Traditional analysis techniques are combined with risk assessment methods to define 

appropriate personnel training and qualifications, barriers and controls, including ensuring that 

the risks from the failure of these protections are acceptably low. This combination brings forth 

the systems analysis approach and provides a way to decide how much defense in depth is 

sufficient. The decision of what are acceptably low risks is not necessarily based on quantitative 

probabilistic metrics. For nuclear power reactors and waste repositories, the existing 

quantitative goals and requirements help define acceptably low levels of risk. Risk management 

for those program areas dealing with lesser amounts of radioactive materials is often achieved 

largely by standard radiation protection practices, which are established by all licensees in 

accordance with the NRC regulations defined in Title 10 of the Code of Federal Regulations 

(10 CFR) Part 20, “Standards for Protection Against Radiation.”

There is value in describing the basic steps to be performed in risk management 

decisionmaking (“Decisionmaking Process” in Figure ES-1). Although these steps have many 

similarities to current NRC decisionmaking processes (e.g., Regulatory Guide 1.174, “An 

Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-specific 

Changes to the Licensing-basis”), there are important benefits to standardizing these steps 

across all NRC program areas.

The proposed framework should be implemented for both safety and security-related issues. 

Although estimating the risk of security-initiated events is difficult, and methods are not well 

established, the NRC nonetheless should have as a goal managing the appropriate amount 

of security defense in depth and better integrating security vulnerability assessments and risk 

assessments for other safety issues.

As a first step in implementing the proposed Risk Management Regulatory Framework, the 

RMTF proposes the following recommendation:

Recommendation (Risk Management Regulatory Framework):

The NRC should formally adopt the proposed Risk Management Regulatory 

Framework through a Commission Policy Statement.

Consistent with normal agency practice, obtaining stakeholder input will be valuable in the 

development of the proposed policy statement. Recognizing the regulatory authority of 

Agreement States in the materials program area, they should have an early role in this process.
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xvi | Risk Management Task Force Executive Summary

This proposed framework includes several important benefits:

•	 Updated knowledge from contemporary studies, such as risk assessments, would be 

incorporated into the regulations and guidance, thereby improving their realism and 

technical basis.

•	 Implementation of a systematic approach would foster a consistent regulatory 

decisionmaking process throughout the agency and improve resource allocation.

•	 Consistency in language and communication would be improved across the agency 

and externally.

•	 Support of issue resolution would be achieved in a systematic, consistent, and 

efficient manner.

Implementation of the proposed framework would also pose challenges:

•	 A change would be required within the agency and externally to increase understanding 

of the value and use of risk concepts and risk management language.

•	 The proposed risk-informed and performance-based concept of defense in depth may 

require the development of additional decision metrics and numerical guidelines.

•	 The approach would likely require developing new or revised risk-assessment 

consensus codes and standards.

•	 A long-term commitment from the Commission and senior agency management would 

be required for implementation.

Program Area Findings and Recommendations

As noted above, the RMTF has developed findings and recommendations on what changes 

would be needed to ensure that the proposed risk management framework would be 

implemented in 10 to 15 years. The RMTF did not assess some NRC activities in detail 

(e.g., environmental reviews, decommissioning), but these activities could be addressed in a 

manner similar to those presented below. The program area findings reflect what the RMTF 

considers to be important gaps between how specific types of licensees are regulated today and 

how they would be regulated in the future using the proposed risk management framework. The 

recommendations suggest ways in which the gaps could be closed.

Power Reactors 

Power reactors in the United States have been licensed using 10 CFR Part 50, “Domestic 

Licensing of Production and Utilization Facilities.” Implementation of 10 CFR Part 50 has 

been achieved, for the most part, using deterministic methods and acceptance criteria. From 

a safety perspective, a set of licensing-basis events was established that was intended to 

ensure conservatism in design and protection from a wide spectrum of postulated events, up to 

and including design-basis accidents (DBAs). These postulated accidents are highly stylized 

and generally do not consider multiple failures of safety systems. Qualitative approaches 

for ensuring reliable safety systems, such as the single failure criterion, were implemented. 
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Testing plans and operational limits were established in technical specifications to ensure 

that safety systems would perform as intended if called upon. In addition to the failure or 

malfunction of plant equipment, the challenges to and protections of power reactor designs 

include consideration of external hazards, such as earthquakes, which may initiate a variety of 

plant transients while also challenging one or more of the defense-in-depth barriers provided to 

mitigate or contain potential releases of radioactive materials.

The NRC’s current regulatory approach for power reactors includes a set of events and 

accidents necessary for “adequate protection” and additional events and accidents licensees 

are required to address to provide an additional amount of safety. The former set consists 

mostly of the set of “design-basis” events and accidents (described above) established more 

than 30 years ago. The latter set has emerged in more recent times to address specific issues, 

such as station blackout accident risks.

The NRC’s power reactor regulatory program has been the subject of considerable work to 

increase the use of risk assessment methods and results. Requirements have been added, 

modified (including the development of risk-informed alternative rules), and deleted. Risk 

information is used in some licensing activities. The power reactor oversight program has 

important risk considerations included in resource allocation and the evaluation of 

inspection findings.

The RMTF assessed this regulatory structure in the context of the proposed risk management 

framework and developed a number of findings. The most important findings are listed below.

•	 Finding: The concept of design-basis events and accidents continues to be a sound 

licensing approach, but the set of design-basis events and accidents has not been 

updated to reflect insights from power reactor operating history and more modern 

methods, such as probabilistic risk assessment (PRA).

•	 Finding: Requirements for beyond-design-basis accident scenarios (e.g., station 

blackout) were established at different times and in different ways. Differences in 

implementation approaches have reduced the efficiency and consistency of the NRC’s 

regulatory and oversight activities.

•	 Finding: The extent to which licensee activities undertaken as part of voluntary industry 

initiatives can be credited has been a source of contention in the Reactor Oversight 

Process and has reduced the efficiency of that process.

•	 Finding: The process for establishing the external hazard design basis does not use 

consistent event frequency or magnitude methods.

•	 Finding: Differences in regulatory language and approaches between power reactor 

security and safety regulation may have reduced the efficiency and effectiveness of the 

NRC’s work.
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xviii | Risk Management Task Force Executive Summary

In light of these findings, the RMTF offers the following recommendations for power reactors.

Recommendations (Power Reactors):

•	 The set of design-basis events and accidents should be reviewed and 

revised, as appropriate, to integrate insights from the power reactor 

operating history and more modern methods, such as probabilistic 

risk assessment (PRA).

•	 The NRC should establish through rulemaking a design-enhancement 

category of regulatory treatment for beyond-design-basis accidents. 

This category should use risk as a safety measure, be performancebased (including the provision for periodic updates), include 

consideration of costs, and be implemented on a 

site-specific basis.

•	 The NRC should reassess methods used to estimate the frequency 

and magnitude of external hazards and implement a consistent 

process that includes both deterministic and PRA methods. 

Consideration of the risks from beyond-design-basis external hazards 

should be included in the proposed design-enhancement category.

•	 The NRC should develop and implement guidance for use in its 

security regulatory activities that uses a common language with 

safety activities and harmonizes methods with risk assessment 

and the proposed risk-informed and performance-based defense-indepth framework.

A significant change from current practice probably will be required for determining the design 

basis for Generation IV reactor designs. However, the NRC should be amenable to, and 

promote where practical, the use of a revised set of design-basis events and accidents for 

operating reactors and certified reactor designs.

In addition to the framework benefits and challenges noted above, the following observations 

are also relevant to the power reactor regulatory program:

•	 The proposed design-enhancement category would clarify the attributes of all requirements 

established as substantial safety (beyond-design-basis) improvements. This approach may 

contribute to the resolution of the “patchwork” issue identified by the Fukushima Near-Term 

Task Force.

•	 Consideration of cost in the proposed design-enhancement category would necessitate a 

reconsideration of the agency’s tools for performing cost-benefit analysis.
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A Proposed Risk Management Regulatory Framework

Nonpower Reactors 

Nonpower reactors (NPRs), generally known as research and test reactors, are nuclear reactors 

primarily used for research, training, development, and isotope production. They contribute 

to almost every field of science, including physics, chemistry, biology, medicine, geology, 

archeology, and environmental sciences.

NPRs have been licensed using a process that has many similarities to the process used 

to license power reactors. The concepts of defense in depth and design-basis events and 

accidents are used, as is the concept of a “maximum hypothetical accident.” However, NPRs 

have a substantially smaller radiological hazard than power reactors, because of their much 

smaller power rating and intermittent use.

The most important findings are:

•	 Finding: The analysis of design basis and the maximum hypothetical accidents based 

on conservative design limits, acceptance criteria, safety margins, and assumptions 

in conjunction with the application of a defense-in-depth philosophy continues to be a 

sound but highly conservative licensing approach to ensuring adequate safety of NPRs.

•	 Finding: While PRAs have been performed for NPRs by other organizations, modern 

risk assessment methods have not been used in NRC NPR licensing decisions.

In light of these findings, the RMTF’s recommendations for nonpower reactors are 

provided below.

Recommendations (Nonpower Reactors):

•	 The proposed defense-in-depth framework should be applied to the 

NPR licensing process to ensure that the current amount of defense in 

depth is appropriate given the relatively small radioactive hazard. This 

application should include safety and security licensing matters.

•	 The NRC should evaluate the utility of performing a pilot risk 

assessment, including consideration of external hazards, using modern 

risk assessment methods at an NPR. This evaluation would assess the 

value of the risk insights gained from the risk assessment on the basis 

of possible safety enhancements and possible contributions to a more 

efficient and effective risk-informed and performance-based regulatory 

framework for NPRs.

Materials

Reactor and accelerator-produced nuclear materials are used extensively throughout the 

United States for industrial applications, basic and applied research, manufacture of consumer 

products, academic studies, and medical diagnosis, treatment, and research. In addition, 

source materials are used in the production of processed uranium for nuclear fuel fabrication 
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xx | Risk Management Task Force Executive Summary

and a wide variety of other uses. The regulatory framework for use of these materials is 

contained in the Commission’s radiation protection standards in 10 CFR Part 20, in applicationspecific regulations in 10 CFR Part 30 through Part 39, and in related guidance and policies. 

The NRC’s materials regulatory programs are designed to ensure that licensees use these 

materials safely and securely so that they present no undue risk to public health and safety and 

the environment.

Under the provisions of Section 274 of the Atomic Energy Act, the NRC may enter into 

agreements with States under which the NRC discontinues its authority over certain radioactive 

materials and a State assumes that authority. The Agreement States, as they are known, now 

regulate more than 85 percent of the materials licensees in the United States. Presently, there 

are 37 Agreement States and, in concert with the NRC, they play a major role in the regulation 

of most materials uses.

The risks presented by the use of nuclear materials differ significantly from the risks presented 

by power reactors. The risk environment for reactors focuses predominantly on the prevention 

of low-frequency, high-consequence scenarios, whereas the risk environment for materials 

focuses primarily on higher-frequency, lower-consequence scenarios. Furthermore, in 

the materials area, risk assessments are largely qualitative; in the reactor program, such 

assessments are generally quantitative. Traditionally, the basis for the materials program 

largely has been a deterministic one, with rules and guidance developed over time and as a 

result of operational experience. Beginning in the 1990s, the NRC undertook a number of 

initiatives to better risk-inform and performance-base its nuclear materials (and other) regulatory 

programs. These initiatives led to fundamental changes in inspection frequency and approach, 

as well as licensing policies and practices, and regulations and guidance. Today, risk insights 

and performance considerations continue to be significant factors in materials program 

development and implementation. 

The RMTF assessed this regulatory program in the context of the proposed risk management 

framework and developed a number of findings. The most important of these findings are 

listed below.

•	 Finding: The terminology of defense in depth is not used consistently across the NRC’s 

materials regulatory programs.

•	 Finding: The materials program could benefit from a more structured application of the 

risk management process in resource allocation. This process would allow program 

managers to more systematically apply resources to those areas where the safety or 

security risk warrants it.

•	 Finding: Buy-in of the 37 Agreement States is essential to the success of the risk 

management process implementation.
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A Proposed Risk Management Regulatory Framework

In light of these findings, the RMTF’s recommendations for materials are provided below.

Recommendations (Materials):

•	 The NRC materials program should continue to apply risk insights 

and performance-based considerations, as appropriate, in 

rulemaking, guidance and policy development, and implementation 

in accordance with the proposed risk management framework. This 

consideration should include both safety and security licensing 

processes.

•	 The development and rollout of the recommended Risk Management 

Policy Statement should be closely coordinated with the leadership of 

the Agreement States.

Low-Level Waste

The NRC regulates the management and disposal of LLW through regulations, licensing, 

inspection, enforcement, guidance, and policy development. The primary regulations in 

this area are 10 CFR Part 20, 10 CFR Part 35, “Medical Use of Byproduct Material” (decayin-storage provisions), and 10 CFR Part 61, “Licensing Requirements for Land Disposal of 

Radioactive Waste.” The Agreement States also play a major role in LLW regulation since they 

regulate all the operating commercial LLW disposal sites, as well as the major LLW processors. 

The Low-Level Radioactive Waste Policy Act of 1980, and amendments to that act in 1985, 

established that LLW disposal was a State responsibility. The LLW Act encouraged States to 

enter into regional compacts that would develop common disposal facilities for use by member 

States of a compact. The Commission has directed the staff to expand its effort to bring a 

clearer risk-informed approach to 10 CFR Part 61 in a staff requirements memorandum issued 

in January 2012. 

The RMTF assessed this regulatory program in the context of the proposed risk management 

framework and developed a number of findings. The most important of these findings are 

listed below.

•	 Finding: The regulatory framework for LLW disposal has, for the most part, implicitly 

followed a risk-informed and performance-based approach. However, changes in the 

LLW environment and the maturing of the performance assessment method over the 

past 30 years have underscored the need to provide a stronger risk basis to 

the program. 

•	 Finding: Certain aspects of the LLW regulatory framework readily lend themselves 

better to the risk management approach, such as waste classification or concentration 

averaging. Applying the proposed risk management approach to comprehensive LLW 

licensing decisions, however, may be more challenging because it involves estimating 

facility performance due to events potentially far into the future.
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xxii | Risk Management Task Force Executive Summary

•	 Finding: The interlocking and reinforcing systems approach in 10 CFR Part 61 (site 

suitability, waste form and classification, intruder barrier, and institutional controls) 

represent an implicit consideration of defense-in-depth features, based on the risk posed 

by various classes of waste. 

In light of these findings, the RMTF recommendations for low-level waste are provided below.

Recommendations (Low-Level Waste):

•	 The NRC should adopt the concept of risk management to the 

LLW program, as well as any revisions proposed to 10 CFR

Part 61 (including performance assessment requirements) and 

related guidance documents.

•	 The NRC should develop an explicit characterization of how defense 

in depth, within the proposed risk management framework, applies to 

the LLW program and build this into current and future staff guidance 

documents and into training and development activities for the staff. 

•	 The NRC should include environmental reviews within the scope of its 

risk management framework.

High-Level Waste

U.S. policies governing the permanent disposal of high-level waste (HLW) are defined by the Nuclear 

Waste Policy Act of 1982, as amended (NWPA). This Act specifies that HLW will be disposed of 

underground, in a deep geologic repository, and that Yucca Mountain, NV, will be the single candidate 

site for characterization as a potential geologic repository. In light of the Federal Government’s 

decision to discontinue activities directed at ultimate disposal of HLW at the Yucca Mountain site, the 

future direction and options for the NRC’s regulation of HLW disposal are not certain at this time. The 

following paragraphs discuss how risk-informed and performance-based factors were considered in 

the development of the current regulations and guidance for HLW disposal and how the proposed risk 

management framework might interface with future regulatory development. 

Under the NWPA Act, the NRC is one of three Federal agencies that have a role in the disposal of 

spent nuclear fuel and HLW from the Nation’s nuclear weapons production activities:

•	 The U.S. Department of Energy (DOE) is responsible for designing, constructing, operating, 

and decommissioning a permanent disposal facility for HLW, under NRC licensing and 

regulation. 

•	 The U.S. Environmental Protection Agency (EPA) is responsible for developing site-specific 

environmental standards for use in evaluating the safety of a geologic repository. 

•	 The NRC is responsible for developing regulations to implement EPA’s safety standards and 

for licensing and overseeing the construction and operation of the repository. 
Executive Summary Risk Management Task Force | xxiii

A Proposed Risk Management Regulatory Framework

In the late 1970s, the NRC began developing regulations for HLW disposal. Since that time, risk 

information and risk concepts have been used increasingly to assist and guide the development 

of the HLW program. NRC regulations for geologic disposal are found in 10 CFR Part 60, 

“Disposal of High-Level Radioactive Wastes in Geologic Repositories,” (generic regulations 

for all sites other than Yucca Mountain, issued in 1983), and 10 CFR Part 63, “Disposal 

of High-Level Radioactive Wastes in a Geologic Repository at Yucca Mountain, Nevada” 

(specifically for Yucca Mountain, issued in 2001). Revisions to the standards and regulations 

for geological disposal to a potential repository at Yucca Mountain made significant use of risk 

information to support more effective and efficient standards and regulations. In particular, 

the NRC’s regulations in 10 CFR Part 63 reflected a risk-informed and performance-based 

approach that focused on regulatory compliance for the post-closure period on the dose to a 

reasonably maximally exposed individual (dose limit provided in the EPA standard for a potential 

repository at Yucca Mountain under Title 40, “Protection of Environment,” of the Code of 

Federal Regulations (40 CFR), Part 197, “Public Health and Environmental Radiation Protection 

Standards for Yucca Mountain, Nevada”). As experience with performance assessment 

methods and results (e.g., risk information) increased over time, so did the use of risk 

information to resolve technical issues and assist in the development of regulatory guidance.

The RMTF assessed this regulatory program in the context of the proposed risk management 

framework and developed the following findings:

•	 Finding: The development of regulations for geologic disposal of HLW at

Yucca Mountain was based significantly on risk information developed from 

performance assessments and closely followed the proposed Risk Management 

Regulatory Framework.

•	 Finding: The NRC’s regulatory philosophy of defense in depth is reflected in the 

multiple-barrier requirement for post-closure in 10 CFR Part 63. Compliance with the 

multiple barrier requirements is demonstrated through the performance assessment.

•	 Finding: As performance assessment capabilities and experience increased at the NRC 

during the past 30 years, so did the use of risk insights to help guide the HLW program. 

Risk insights and performance-assessment capabilities have been used to improve the 

efficiency and effectiveness of guidance documents, inform pre-licensing interactions 

with DOE, and help identify and direct data needs and experimental activities.

In light of these findings, the RMTF recommendation for high-level waste is provided below.

Recommendation (High-Level Waste):

•	 Any future revisions to the regulatory framework for geologic disposal 

of HLW should be done in accordance with the proposed risk 

management framework to ensure that risk information continues to 

be appropriately considered in the development of requirements and 

appropriately reflect any future HLW disposal paradigm. 
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xxiv | Risk Management Task Force Executive Summary

Uranium Recovery 

Uranium recovery is the first step in the nuclear fuel cycle. It involves extracting uranium 

from its parent ore and processing it into a physical and chemical form (yellowcake) that will 

allow additional processing and fabrication to become nuclear fuel. The NRC’s regulatory 

authority under the Atomic Energy Act (the Uranium Mill Tailings Radiation Control Act of 1978 

(UMTRCA)) and 10 CFR Part 20 and 10 CFR Part 40, “Domestic Licensing of Source Material,” 

does not extend to the mining of uranium-bearing ore by conventional methods. That activity 

is regulated primarily by States and other Federal agencies. Rather, the NRC regulates the 

processing of uranium ore to concentrate uranium and the disposal of the large amount of 

tailings resulting from that processing. 

Both NRC and EPA regulations in this area are largely deterministic and do not reflect riskinformed, performance-based considerations. The Commission considered a revision to the 

regulatory framework for uranium recovery in 2000 that would have updated and risk-informed 

the regulations, but the agency chose not to pursue it. A staff risk assessment in 2001 

concluded that in situ recovery (ISR) facilities are of inherently low risk. The NRC licensing 

guidance for ISR is currently being revised to reflect risk insights and licensing experience. ISR 

facility licenses also include a performance-based license condition that allows licensees to 

make certain changes without requesting NRC approval in the form of a license amendment.

The RMTF assessed this regulatory program in the context of the proposed risk management 

framework and developed the following findings:

•	 Finding: Uranium recovery facilities are of low radiological risk to workers and members 

of the public under normal operational conditions and most accident scenarios.

•	 Finding: Although the NRC staff has made inroads to risk-informed, performancebased licensing of uranium recovery facilities, the regulatory framework is largely a 

deterministic one. NRC regulations in 10 CFR Part 40 and Appendix A to 10 CFR Part 

40 reflect the requirements of UMTRCA and EPA’s rules in 40 CFR Part 192, which predate the Commission’s move to a risk-informed, performance-based framework in the 

mid-1990s. Similarly, program guidance, especially for conventional mills, could benefit 

from a greater risk basis. 

•	 Finding: The exact nature of the ISR rule under development by the EPA is not clear 

at this time and, therefore, presents an uncertainty to adoption of the proposed risk 

management regulatory framework.

•	 Finding: Consideration of environmental risks is a central part of the uranium recovery 

regulatory program.
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A Proposed Risk Management Regulatory Framework

In light of these findings, the RMTF recommendations for uranium recovery are provided below.

Recommendations (Uranium Recovery):

•	 Notwithstanding the current uncertainty associated with the EPA 

rulemaking, the NRC should adopt the proposed risk management 

regulatory framework to the uranium recovery program to provide 

greater efficiency, effectiveness, and predictability in policy 

development and regulatory decisionmaking.

•	 The NRC should work closely with the Agreement States and the 

regulated community to guide implementation of risk management in 

the uranium recovery program.

•	 The NRC should include environmental reviews within the scope of its 

risk management framework. 

Fuel Cycle 

The NRC regulates major fuel cycle facilities, including those involved in conversion of uranium 

ore to UF6, gaseous centrifuge and diffusion enrichment, reactor fuel fabrication, plutonium 

processing, and UF6 deconversion. Reactor fuel fabrication facilities include those that produce 

low-enriched uranium, high-enriched uranium, and mixed-oxide (Pu + U) fuels. The NRC 

also regulates possession of small amounts of special nuclear material (SNM), usually for 

research purposes. This regulatory program is primarily governed by regulations contained 

in 10 CFR Part 70, “Domestic Licensing of Special Nuclear Material,” and 10 CFR Part 76, 

“Certification of Gaseous Diffusion Plants,” as well as by 10 CFR Part 40.

The regulations applicable to licensing of fuel cycle facilities vary because of the differences 

in the nature and amounts of the materials licensed and the historical origins of the facilities. 

Regulation of risk to workers is a major focus for these facilities, since workers typically are 

in close proximity to the hazards. The regulations in 10 CFR Part 40 apply to conversion 

and deconversion facilities as licensees authorized to possess source material. The safety 

requirements for 10 CFR Part 40 uranium conversion licenses are general, not prescriptive, 

and do not require performance of any risk assessment. 10 CFR Part 70 applies to major lowenriched uranium fuel fabrication facilities, high-enriched uranium processing facilities, and new 

enrichment facilities (gaseous centrifuge). For these major fuel cycle facilities, 10 CFR Part 70, 

Subpart H, “Additional Requirements for Certain Licensees Authorized to Possess a Critical 

Mass of Special Nuclear Material,” requires licensees to perform an integrated safety analysis 

(ISA). ISAs share some elements in common with PRAs in that all significant accident 

scenarios must be evaluated, and consequences must be estimated. 

The RMTF assessed this regulatory program in the context of the proposed risk management 

framework and developed the following findings:
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xxvi | Risk Management Task Force Executive Summary

•	 Finding: The current fuel cycle regulatory approach incorporates several elements of 

the proposed risk management regulatory framework, such as the use of ISAs to identify 

safety significant items, and the implementation of a revised fuel cycle oversight program 

as directed by the Commission.

•	 Finding: The concept of defense in depth, as embedded in fuel cycle regulatory 

requirements and practices, is consistent with Commission guidance. Its implementation 

changes as the processes change at the fuel cycle facilities.

In light of these findings, the RMTF recommendation for fuel cycle is provided below.

Recommendations (Fuel Cycle):

The fuel cycle regulatory program should continue to evaluate the risk and the 

associated defense-in-depth protection by using insights gained from ISAs. 

ISAs should continue to evolve to support regulatory decisionmaking. 

Spent Nuclear Fuel Storage 

Spent nuclear fuel (SNF) dry storage systems are designed to be robust, passive systems. 

They are designed to withstand the effects of “worst-case” events or design-basis events 

and phenomena while still maintaining the capabilities to provide adequate shielding and 

confinement of radioactive contents and prevent nuclear criticality. The systems are designed 

to perform these functions while requiring minimal maintenance or repair. The regulations 

in 10 CFR Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear 

Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste,” 

include requirements based on risk to some degree. Both the NRC and the Electric Power 

Research Institute have conducted PRAs of dry cask storage systems and concluded that 

the risk associated with them is very low. The NRC has employed a conservative approach 

in its regulations, guidance, and licensing practices for independent spent fuel storage 

installations to minimize the likelihood of adverse consequences to public health and safety. 

The Commission has directed the NRC staff to revisit the paradigm for SNF storage and 

transportation to include evaluating the dry storage of SNF for periods significantly in excess of 

those previously envisioned.

The RMTF assessed this regulatory program in the context of the proposed risk management 

framework and developed the following findings:

•	 Finding: The regulatory approach for SNF storage is largely based on meeting 

applicable industry consensus standards and conservative guidance to ensure adequate 

safety margins in the facility and cask designs and operations. More recently, insights 

from a limited number of risk studies have been gradually factored into this regulatory 

approach. Furthermore, though qualitative, a systematic approach that parallels 

answering the risk triplet was used in the latest revision of the Standard Review Plan.
Executive Summary Risk Management Task Force | xxvii

A Proposed Risk Management Regulatory Framework

•	 Finding: The concept of defense in depth is not explicitly or consistently applied in the 

SNF storage regulatory program. 

In light of these findings, the RMTF recommendations for spent fuel storage are provided below.

Recommendation (Spent Fuel Storage):

•	 While elements of the proposed risk management approach have 

been used in the SNF storage regulatory approach to evaluate 

the acceptable level of risk and the sufficiency of defense in depth 

(physical barriers, controls or margins) more consistently, the NRC 

should develop the necessary risk information, the corresponding 

decision metrics, and numerical guidelines. This is important in 

guiding further changes to the existing SNF storage regulatory 

approach and the evaluation of strategies for extended SNF 

storage activities.

•	 As part of the implementation of the proposed risk management 

regulatory framework, the NRC should more consistently consider 

the concept of defense in depth explicitly and evaluate its proper 

use in the SNF storage regulatory program. The NRC should also 

improve appropriate parts of staff training to make this concept a 

central part of such training.

Transportation

The transportation of radioactive materials within the United States is regulated jointly by the 

U.S. Department of Transportation (DOT), the NRC, DOE, and State and local governments. 

The approval or certification of shipping package designs for radioactive materials is shared 

jointly by the NRC, DOT, and DOE. NRC and DOT responsibilities for the certification of 

shipping package designs are delineated in a 1979 Memorandum of Understanding between 

DOT and the NRC. The NRC’s primary “licensing” role in transportation safety is the review 

and certification of Type B and fissile material shipping package designs. The NRC inspects 

10 CFR Part 71, “Packaging and Transportation of Radioactive Material,” certificate holders and 

package component fabricators to ensure that transportation casks are fabricated and tested 

in accordance with the package specifications in the NRC certificate. The United States, which 

is a participating International Atomic Energy Agency (IAEA) member state, has also endorsed 

the concept that its domestic transportation regulations in Title 49, “Transportation,” of the Code 

of Federal Regulations (49 CFR) Part 173,”Shippers—General Requirements for Shipments 

and Packaging,” should be compatible with IAEA’s transportation regulations to the greatest 

extent practicable. The practical result is that the NRC and DOT periodically undertake a joint 

rulemaking effort to revise their respective regulations to be compatible with the latest revision 

to IAEA transportation regulations.

The basic physical tests implemented in the 1964 IAEA regulations remain as the primary tests 

used today for approving Type B and fissile material shipping packages. The continued use of 
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xxviii | Risk Management Task Force Executive Summary

these physical tests has been supported by numerous risk studies done in the United States, 

as well as in other IAEA member states. All of the studies have shown that the risk of shipping 

spent fuel is very low. 

The RMTF assessed this regulatory program in the context of the proposed risk management 

framework and developed the following findings:

•	 Finding: While the U.S. transportation regulatory approach is governed by the IAEA 

transportation regulations, the current NRC transportation regulatory approach uses 

several elements of the proposed risk management framework.

•	 Finding: Risk assessments have been conducted on the safety of transportation of 

spent fuel. However, there is a lack of risk information on the transportation of other 

radioactive materials.

In light of these findings, the RMTF recommendations for transportation are provided below.

Recommendations (Transportation):

•	 Considering the strong international regulatory basis for transportation 

and the need to conform U.S. standards to those of the IAEA and 

other member states, application of the proposed risk management 

framework should focus on implementation guidance.

•	 The risk management process should be used to influence the future 

outcome of IAEA deliberations on proposed changes in international 

transportation regulations.

•	 The NRC should explore the value of using risk insights to justify 

regulations different from the IAEA’s for domestic use only, such 

as regulations dealing with domestic storage and transportation of 

high burnup fuel. Risk information could be used to develop a more 

flexible approach toward implementing and making gradual changes 

to current transportation regulations.

Concluding Remarks

The NRC has made progress in its efforts to implement risk-informed and performance-based 

approaches into its regulation of the various uses of byproduct, source, and special nuclear 

materials. Nevertheless, it is necessary to reassess that progress and the underlying strategic 

vision from time to time.

The RMTF has found that the NRC’s programs do not require radical or revolutionary 

changes, but they could benefit from continuing the evolution that has been embraced 

throughout the agency’s history. To that end, a Risk Management Regulatory Framework 

is recommended as the next logical step for the NRC. This proposed framework uses a 

disciplined risk management process to identify and evaluate issues and make decisions on 
Executive Summary Risk Management Task Force | xxix

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appropriate defense-in-depth protections for various radiological hazards. The risk-informed 

and performance-based defense-in-depth protections provide sufficient barriers, controls, and 

personnel to prevent, contain, and mitigate the exposure of workers or the public to radioactive 

materials. The appropriate barriers, controls, and personnel are based on the hazards present, 

the relevant scenarios leading to possible exposures, and the associated uncertainties to ensure 

that the risks resulting from the failure of some or all of the established barriers are maintained 

acceptably low.

To a certain degree, some resistance to change is natural and perhaps even desirable for 

the NRC to maintain a clear and stable regulatory environment. In addition, the ongoing 

work activities within the agency and limited resources can impede the development and 

implementation of the proposed regulatory framework. However, a patchwork of regulatory 

requirements has been created as a result of addressing problems on a case-by-case basis 

for many years. The RMTF has concluded that cost-effective changes are possible and 

recommends that they be undertaken in a holistic manner across the NRC’s programs. With 

this in mind, implementation of the proposed Risk Management Regulatory Framework can 

be pursued in a planned and deliberate manner so that it does not disrupt the NRC’s mission, 

but ensures that the NRC continues to improve on how it protects public health and safety, 

promotes the common defense and security, and protects the environment.

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ACKNOWLEDGMENT

I thank Chairman Jaczko for proposing this work and the agency for providing resources to carry 

it out. The Risk Management Task Force (RMTF) consisted of Christiana Lui, RMTF Executive 

Director, Mark Cunningham, George Pangburn, and William Reckley. Their knowledge, 

innovative thinking, and dedication confirmed, once again, the high regard that I have for the 

NRC staff. Our sessions debating risk-informed and performance-based concepts will be 

among my fond memories of my tenure as a commissioner.

The RMTF benefited from contributions by the following staff members: John Adams, Michel 

Call, Dennis Damon, Don Dube, Earl Easton, Timothy McCartin, Geary Mizuno, and Joel Piper. 

Their contributions helped us cover the large range of regulatory activities of the agency, a 

nearly impossible task for any small group.

I also thank the many other individuals and organizations who provided comments and insights. 

In that regard, our meetings with NRC regional offices were particularly valuable.

George Apostolakis

Chapter 1: Introduction Risk Management Task Force | 1-1

A Proposed Risk Management Regulatory Framework

1. INTRODUCTION

1.0 Background

The wide-scale use of radioactive materials has evolved over only the last several decades 

of human history. During that time, many beneficial applications of radioactive materials and 

nuclear technology have been developed and put into use. Experiences during this period also 

have demonstrated the need to protect workers and the public from inadvertent exposures to 

radioactive materials. The importance of regulatory controls to govern the use or generation 

of radioactive materials is reflected in the congressional findings of the Atomic Energy Act (“the 

Act”), which states:

The processing and utilization of source, byproduct, and special 

nuclear material must be regulated in the national interest and 

in order to provide for the common defense and security and to 

protect the health and safety of the public.

The increasing use of radioactive materials during the decades that followed implementation 

of the Act is reflected in the various regulatory programs of the U.S. Nuclear Regulatory 

Commission (NRC). This includes a wide range of uses of radioactive materials in industrial, 

medical, and research applications; the use of nuclear reactors for electricity production and 

research; the transportation and storage of radioactive materials; and the disposal of radioactive 

wastes. The challenge undertaken by the NRC and its licensees is to enable the beneficial uses 

of radioactive materials while ensuring those activities pose no undue risk to the public health 

and safety. 

The NRC has established agencywide regulations and policies to help ensure that civilian 

uses of radioactive materials pose no undue risk. The Commission’s basic radiation protection 

regulations are contained in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, 

“Standards for Protection Against Radiation.” Regulations in 10 CFR Part 20 establish 

exposure limits and require licensees to make every reasonable effort to maintain exposures 

to radiation as low as is reasonably achievable. Other regulations, such as those contained in 

10 CFR Parts 30 through 70, describe additional basic safety requirements for NRC-licensed 

materials and facilities. Specific regulatory controls reflect the differences in the radiological 

hazards associated with different types of NRC-licensed materials, devices, and facilities.

In the 1970s, the NRC completed its first probabilistic risk assessment of two nuclear power 

reactors, which introduced a new way to measure nuclear safety and the effectiveness of the 

NRC’s regulations. The Commission subsequently established a policy on how risk assessment 

methods should be used to complement the NRC’s established regulations in all its regulatory 

programs. This policy, coupled with additional Commission guidance1 issued in 1999, has 

resulted in a variety of program-specific improvements.

While progress has been made, the NRC’s Strategic Plan and Principles of Good Regulation 

make it clear that improvements in efficiency, effectiveness, and reliability continue to be agency 

1 The NRC’s “White Paper on Risk-Informed and Performance-Based Regulation” (NRC,1999) contained 

descriptions of many terms, including defense-in-depth.
A Proposed Risk Management Regulatory Framework

1-2 | Risk Management Task Force Chapter 1: Introduction

goals. The NRC Strategic Plan (NRC, 2012) notes that the expanded use of risk-informed and 

performance-based insights and the use of state-of-the-art technologies are the means by which 

the agency enhances the effectiveness and realism of NRC actions. The Principles of Good 

Regulation reinforce these points, noting that regulatory activities should be consistent with the 

degree of risk reduction they achieve. Furthermore, regulations should be based on the best 

knowledge available from research and operational experience.

In a memorandum dated February 11, 2011, NRC Chairman Gregory Jaczko created a task 

force headed by Commissioner George Apostolakis to develop a strategic vision and options 

for adopting a more comprehensive and holistic risk informed, performance-based regulatory 

approach for reactors, materials, waste, fuel cycle, and transportation that would continue to 

ensure the safe and secure use of nuclear material (NRC, 2011). The task force was afforded 

the flexibility to provide options ranging from a complement to or alternative to the existing 

regulatory framework. The objectives of the task force were defined as follows:

Task Force Objectives:

The task force should identify the options and specific actions 

that the NRC could pursue to achieve a more comprehensive and 

holistic risk-informed, performance-based regulatory structure.

About 1 month after the creation of this task force, hereafter referred to as the Risk 

Management Task Force (RMTF), a significant accident occurred at the Fukushima Dai-ichi 

nuclear power facility in Japan. The Commission established a Near-Term Task Force (NTTF) 

“to conduct a systematic and methodical review of U.S. Nuclear Regulatory Commission 

processes and regulations to determine whether the agency should make additional 

improvements to its regulatory system and to make recommendations to the Commission for 

its policy direction, in light of the accident at the Fukushima Dai-ichi Nuclear Power Plant.” 

Recommendation 1 of the NTTF (NRC, 2011a) is listed below. 

NTTF Recommendation 1:

The Task Force recommends establishing a logical, systematic, 

and coherent regulatory framework for adequate protection that 

appropriately balances defense-in-depth and risk considerations.”

The RMTF has benefited from the discussion accompanying Recommendation 1 in the 

NTTF report. The RMTF report could contribute to the implementation of the NTTF’s 

recommendations.
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A Proposed Risk Management Regulatory Framework

1.1 The Risk Management Task Force Report

The report that follows is organized into five chapters. In Chapter 2, the RMTF describes the 

current regulatory approach and proposes a regulatory framework based on its findings. The 

proposed framework centers on the concept of risk management.

Chapter 3 provides a general discussion of risk management concepts, similar to those 

commonly used by other organizations, including Federal agencies. These concepts are 

then translated into the NRC-specific framework, should the Commission decide to pursue 

a risk management focus for the NRC’s regulatory programs. The chapter and related 

appendices discuss a common NRC goal related to risk-informed and performance-based 

defense-in-depth principles and techniques that can be used to evaluate risks, qualitatively 

or quantitatively, for different radiological hazards and applications. In addition, the chapter 

discusses possible factors used to determine the tolerability or acceptability of the risk 

estimations or characterizations resulting from the staff’s evaluations. The chapter details 

several approaches and possible actions that might be taken to develop and implement a risk 

management framework for the NRC’s regulatory programs. The RMTF offers its findings and 

recommendations in this chapter on the adoption of a risk management framework at the NRC.

Chapter 4 discusses how the NRC would implement a risk management framework in the 

regulation and oversight of licensees within various regulatory program areas. Licensees would 

ultimately manage risks to workers and the public based on the specific uses of byproduct, 

source, and special nuclear materials. Chapter 4 also discusses how the NRC would integrate 

a risk management framework into its processes used to regulate and oversee its licensees. 

The RMTF offers its findings and recommendations on specific regulatory programs and general 

or crosscutting topics or activities.

Chapter 5 summarizes the RMTF findings and discusses possible implementation strategies.

To maintain the report’s focus on the chartered tasks, many of the more detailed descriptions 

supporting the discussions in the chapters and related specific topics are presented in the 

following appendices:

•	 Appendix A: Risk Management Systems

•	 Appendix B: Risk Management—Analyses and Deliberation 

•	 Appendix C: Defense-in-depth 

•	 Appendix D: Performance-based Regulation 

•	 Appendix E: Nuclear Power Reactors—Risk-Informed Initiatives 

•	 Appendix F: Nuclear Power Reactors—Licensing-basis Events 

•	 Appendix G: Nuclear Power Reactors—Safety Classification 

•	 Appendix H: Nuclear Power Reactors— Chapter 4 Alternatives
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1-4 | Risk Management Task Force Chapter 1: Introduction

•	 Appendix I: Comment Summary 

•	 Appendix J: Findings and Recommendations

1.2 References

(NRC,1999) U.S. Nuclear Regulatory Commission, Staff Requirements 

Memorandum Regarding SECY-98-144, “White Paper on Riskinformed and Performance-Based Regulation,” March 1, 1999, 

Agencywide Documents Access and Management System (ADAMS) 

Accession No. ML003753601.

(NRC, 2011) U.S. Nuclear Regulatory Commission, “Assessment of Options for More 

Holistic Risk-Informed, Performance-based Regulatory Approach,” 

Memorandum dated February 11, 2011, to R.W. Borchardt and S.G. Burns 

from Chairman Gregory B. Jaczko, ADAMS Accession No. ML1104606111.

(NRC, 2011a) U.S. Nuclear Regulatory Commission, “Recommendations for Enhancing 

Reactor Safety in the 21st Century; The Near-Term Task Force Review of 

Insights from the Fukushima Dai-Ichi Accident,” July 2011, ADAMS Accession 

No. ML112510271.

(NRC, 2012) U.S. Nuclear Regulatory Commission, “Strategic Plan: Fiscal Years 

2008–2013,” NUREG–1614, Vol. 5, February 2012, ADAMS Accession 

No. ML12038A003.
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A Proposed Risk Management Regulatory Framework

2. A PROPOSED RISK MANAGEMENT REGULATORY FRAMEWORK

2.0 Background

In its efforts to propose options for a more holistic approach to the development and 

implementation of risk-informed and performance-based regulation of NRC licensees, the Risk 

Management Task Force (RMTF) identified a general framework based on established risk 

management concepts. The NRC mission is focused on protecting the public from possible 

hazards introduced by the use of radioactive materials. While some regulatory provisions, such 

as 10 CFR Part 20, cut across all NRC programs, the regulatory systems for reactors, materials, 

and other NRC program areas have been developed somewhat independently from each other. 

The first challenge for the RMTF was to develop a framework that could be used throughout the 

NRC’s regulatory programs. 

The value of soliciting insights from the NRC staff and external stakeholders regarding the 

RMTF activities was recognized from its formation. The RMTF prepared an internal survey 

and distributed it to a cross-section of NRC staff and managers in several key program offices, 

and issued a notice in the Federal Register (published on November 22, 2011 (76 FR 72220)) 

to solicit public input. A summary of the responses to the internal survey and Federal Register

notice is provided in Appendix I. The RMTF also benefited greatly from numerous informal 

discussions with NRC staff and managers, and external stakeholders. Furthermore, the RMTF 

visited all the NRC regional offices. These interactions with the regional staffs were especially 

useful because they provided the RMTF with an opportunity to hear firsthand the experiences 

and issues associated with the use of risk-informed and performance-based approaches in the 

NRC’s oversight process.

2.1 Considerations and Findings

The RMTF started with the NRC Mission Statement, which is derived from the Atomic Energy 

Act and is defined as follows in the NRC Strategic Plan (NRC, 2012):

NRC Mission

License and regulate the Nation’s civilian use of byproduct, 

source, and special nuclear materials to ensure adequate 

protection of public health and safety, promote the common 

defense and security, and protect the environment.

The Strategic Plan describes the NRC’s Principles of Good Regulation as follows:

Principles of Good Regulation

The safe and secure use of radioactive materials and nuclear 

fuels for beneficial civilian purposes is made possible by the 

agency’s adherence to the following principles of good regulation: 

independence, openness, efficiency, clarity, and reliability. In 

addition, regulatory actions are effective, realistic, and timely.
A Proposed Risk Management Regulatory Framework

2-2 | Risk Management Task Force Chapter 2: The Framework

The mission and principles of good regulation naturally translate into a high-level objective for 

the NRC to establish regulatory controls based on risks to the public that are introduced by 

licensed activities. The consideration of risks and tailoring regulations and oversight to manage 

those risks is inherent in current NRC programs. However, they are sometimes expressed 

using different terminology. This approach is consistent with several proposals for effective 

government, including Executive Order 13563, “Improving Regulation and Regulatory Review” 

(Executive Order, 2011). 

An example of the challenges related to terminology can be found in the often used terms 

“safety” and “risk” within the context of NRC programs and goals. The dictionary definitions 

of these two words highlight the relationship between and the difference in the terms. Such 

distinctions have resulted in occasional disagreements among knowledgeable practitioners 

about the terminology and related actions.

•	 Safety: the condition of being free from danger, 

 injury, or damage

•	 Risk: the possibility of damage, injury, or loss

Recognizing that it is not possible to achieve absolute safety or zero risk while benefitting from 

the use of radioactive materials, terms such as “adequate protection” and “no undue risk” are 

used in conclusions or decisions reached by the NRC. No matter an individual’s preference 

for “adequate protection” or “acceptable risk,” assessments and regulatory controls will involve 

asking the same basic questions, which are commonly referred to as the risk triplet (Kaplan and 

Garrick, 1981). These questions are listed below.

•	 What can go wrong?

•	 How likely is it? 

•	 What are the consequences?

While the RMTF observes that the NRC can and has treated “adequate safety” and “acceptable 

risk” as synonymous, it also recognizes that these concepts present challenges to the 

development and implementation of any generic framework. Related issues, such as the 

selection of analysis techniques, are discussed in later sections of this report.

The NRC has developed regulations that limit radiation exposure to workers and the public by 

requiring its licensees to limit the time persons are exposed to radiation, limit access to areas 

containing radioactive materials, and provide physical barriers between radioactive materials 

and individuals. These controls are associated with (1) requirements for radioactive materials to 

be contained within devices or facilities to prevent their inadvertent release, and (2) measures 

to address the possible degradation of barriers and controls intended to protect workers and 

the public. NRC rules and programs also address, to varying levels of specificity, the need 

for personnel to be trained and qualified for the tasks they perform. The totality of these 

requirements constitutes the way in which risk from the use of radioactive materials is managed.
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Requirements related to barriers, controls, and personnel to limit exposure to radioactive 

materials vary among NRC licensees, depending on the hazard present (e.g., the type and 

amount of radioactive materials), the relevant scenarios that might lead to exposures, and 

the associated uncertainties. The risks presented by most uses of byproduct materials differ 

significantly from the risks presented by nuclear power reactors. The risk environment for 

reactors focuses predominantly on the prevention and mitigation of low-frequency, highconsequence events whereas the risk environment for materials uses focuses primarily on highfrequency, low-consequence events. 

The potential for releases to the environment is a major focus of the reactor program. This 

potential is not as significant for the materials program because of the much smaller amounts 

of radioactive material possessed by most materials users and the lower risk posed by many of 

those specific uses.

The RMTF considered a number of approaches to accomplishing its charter. It considered the 

agency’s historic approach to consideration of risk, researched and reviewed literature across 

the field of risk, and received input from internal and external stakeholders to develop alternative 

approaches for a more comprehensive and holistic approach to its regulatory programs. These 

included: (1) no action, (2) a purely deterministic approach, (3) a single risk-based numerical 

criterion, and (4) an approach based on the concept of risk management. A more detailed 

discussion of these approaches is contained in Appendix A.

The U.S. Department of Homeland Security Risk Lexicon (DHS, 2010) defines risk management 

as follows:

Risk management is the process for identifying, analyzing, 

and communicating risk and accepting, avoiding, 

transferring, or controlling it to an acceptable level 

considering associated costs and benefits of any 

actions taken.

Risk management allows for various approaches to consider risks to the public in the NRC’s 

regulatory decisionmaking, including the use of both quantitative and qualitative tools. Such 

flexibility is essential given the broad range of NRC regulatory programs and related decisions. 

It represents a logical evolution from the risk-informed, performance-based philosophy that 

has governed NRC activities for many years. It may also provide program managers with a 

more systematic approach to resource allocation, whether in budget formulation, response to 

events, or licensing decisions. Risk management offers the potential for an improved regulatory 

framework. The sections that follow describe the general attributes of that framework. 

Whether the use of radioactive materials is in complex facilities, such as nuclear power reactors, 

or in simple sealed sources and devices, such as moisture density gauges, the NRC’s risk 

management approach has been (and still is) guided by the fundamental principle of defense-indepth. The NRC white paper on risk-informed and performance-based regulation (NRC, 1999) 

described many terms, including defense-in-depth. This description was modified in the 

statements of consideration for NRC’s final rule 10 CFR 50.69, “Risk-Informed Categorization 

and Treatment of Structures, Systems and Components for Nuclear Power Reactors,” 

(NRC, 2004), as follows:
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2-4 | Risk Management Task Force Chapter 2: The Framework

Defense-in-depth is an element of the NRC’s safety philosophy 

that employs successive measures to prevent accidents or 

mitigate damage if a malfunction, accident, or naturally caused 

event occurs at a nuclear facility. Defense-in-depth is a 

philosophy used by the NRC to provide redundancy as well as the 

philosophy of a multiple-barrier approach against fission product 

releases. The defense-in-depth philosophy ensures that safety 

will not be wholly dependent on any single element of the design, 

construction, maintenance, or operation of a nuclear facility. 

The net effect of incorporating defense-in-depth into design, 

construction, maintenance, and operation is that the facility tends 

to be more tolerant of failures and external challenges.

While this defense–in-depth definition comes from a power reactor document, it can be applied 

across all NRC regulatory areas by broadening “nuclear facility” to include other radioactive 

materials uses and by replacing “fission product releases” with “radioactive material releases” or 

“exposure to radioactive materials.” 

Although some NRC regulatory programs may not have used the defense-in-depth terminology, 

the general approach of identifying barriers to protect workers and the public from exposure 

to radioactive material is common to all NRC activities. For some radioactive materials, the 

radiological hazard may be minimal and the defense-in-depth determinations may result in 

simple containers and labeling systems to maintain the barrier. For complex reactor facilities, 

the defense-in-depth determinations may result in multiple layers, robust structures, and many 

support systems and procedures to maintain the basic barriers. Later sections of the report 

will discuss specific regulatory programs in more detail, but the basic approach being proposed 

by the RMTF is for the agency to adopt a common explanation of our risk management goal, 

which is defined in simple terms of which barriers, controls, and personnel are needed to protect 

individuals from exposure to radioactive materials. 

Even for the nuclear power reactor program, in which the defense-in-depth terminology is most 

ingrained, the concept is implemented through an elaborate system of regulations, guidance 

documents, and general industry practices. After decades of use, there is no clear definition 

or criteria on how to define adequate defense-in-depth protections. The traditional approach 

used by the NRC and industry to provide confidence in a reactor design’s defense-in-depth 

capabilities is based on analyzing stylized accident scenarios using approved conservative 

codes and criteria. The conservatisms added to design limits, acceptance criteria, and safety 

margins are intended to manage the uncertainties associated with accidents, including possible 

“unknown unknowns,” at the time a plant was designed. Safety margins are included in the 

analyses such that specific barriers are designed and constructed to ensure actual failures 

are not expected until key parameters well exceed the values assumed in the supporting 

engineering evaluations. Important limitations of this traditional regulatory approach are that 

(1) significant accident scenarios may not be identified or addressed by the defined barriers and 

controls, and (2) the stylized analyses and related barriers and controls may misdirect resources 

to address low-risk scenarios. 
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The RMTF review of NRC programs leads to the following finding regarding defense-in-depth. 

Finding 2.1: Whether used explicitly, as for power reactors, or implicitly, 

as for materials programs, the concept of defense-in-depth has served the 

NRC and the regulated industries well and continues to be valuable today. 

However, it is not used consistently, and there is no guidance on how much 

defense-in-depth is sufficient.

Risk-informed approaches have been developed over the last several decades to supplement 

the traditional regulatory approach by doing a more methodical assessment of the risk triplet 

questions (what can go wrong, how likely is it, and what are the consequences). The NRC 

white paper mentioned previously provides the following definition:

A “risk-informed” approach to regulatory Decisionmaking 

represents a philosophy whereby risk insights are considered 

together with other factors to establish requirements that better 

focus licensee and regulatory attention on design and operational 

issues commensurate with their importance to health and safety.

The term “risk insights” refers to the results and findings of risk assessments. A risk assessment 

is intended to be as realistic as possible, as opposed to the bounding, stylized approach of 

traditional methods. It explicitly and systematically assesses (quantitatively or qualitatively) the 

likelihood of a given scenario and considers failures of multiple barriers and controls.

The white paper also discusses the difference between prescriptive and performance-based 

regulatory requirements: 

A regulation can be either prescriptive or performance-based. 

A prescriptive requirement specifies particular features, actions, or 

programmatic elements to be included in the design or process, as 

the means for achieving a desired objective. A performance-based 

requirement relies upon measurable (or calculable) outcomes (i.e., 

performance results) to be met, but provides more flexibility to the 

licensee as to the means of meeting those outcomes. 

The benefits of using risk-informed and performance-based approaches throughout the 

NRC’s regulatory programs have been recognized and are reflected in various policy statements 

and initiatives. 

Within the nuclear power reactor program, there has been increased use of formal risk 

assessment techniques, such as probabilistic risk assessment (PRA).1 In contrast to the 

1 Internationally, this approach is known as probabilistic safety assessment (PSA). For the purposes of this 

report, there is no difference between PRA and PSA.
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2-6 | Risk Management Task Force Chapter 2: The Framework

traditional methods described above, a PRA models many credible accident sequences by 

considering the facility or operation as a “system of systems” consisting of structures, systems, 

components, and personnel. The risk assessments have resulted in the consideration of 

additional accident sequences during licensing (e.g., intersystem loss of coolant accident, 

station blackout), informed program areas, such as emergency planning, and have led to 

various initiatives to incorporate risk-informed, performance-based activities into the NRC’s 

regulation and oversight of nuclear power reactors.

Risk assessments and related insights also have been used in several NRC nonreactor 

programs. NUREG/CR-6642, “Risk Analysis and Evaluation of Regulatory Options for Nuclear 

Byproduct Materials Systems” (NRC, 2000), documented a programmatic quantitative analysis 

of tasks, hazards, barriers, and doses (to workers and members of the public) associated 

with a variety of byproduct materials and devices. These risk insights were used to inform the 

NRC inspection programs and were incorporated into licensing guidance in the NUREG-1556 

series, “Consolidated Guidance About Materials Licensees” (NRC, 1998). In another example, 

the NRC has required fuel cycle facilities to perform and document a type of risk assessment 

referred to as integrated safety assessments (ISAs). The high-level waste, spent fuel storage, 

and transportation programs also have all performed different types of risk assessments in their 

development of rules, guidance, or policy over the years. 

In addition to the tools and methods used to perform risk assessments, the NRC has developed 

criteria and guidance to help judge the acceptability of the risks identified by those analyses. 

The publication of the Commission’s quantitative safety goals (NRC, 1986) (and the adoption of 

the subsidiary objectives regarding core damage frequency and large, early release frequency) 

has allowed the use of PRA in helping to address a limitation of the traditional defense-in-depth 

approach for power reactors; namely, how much defense-in-depth is sufficient. The regulations 

in 10 CFR Part 63 established high-level waste repository performance objectives and a means 

to evaluate the adequacy of defense-in-depth in the context of those performance objectives. In 

the area of byproduct materials, NUREG-1556 provides risk-informed guidance on the barriers 

and controls that should be in place for meeting the dose criteria set forth in 10 CFR Part 20 for 

a wide variety of applications.

In summary, the introduction of risk assessments to supplement traditional approaches has 

provided systematic evaluation processes and more balanced considerations in several NRC 

regulatory programs. Some of these insights resulted in changes to NRC regulations and licensee 

programs that may have prevented serious challenges to or breaches of the barriers and controls 

put into place to prevent the release of radioactive materials. Other evaluations have shown that 

some existing regulatory requirements were not actually needed. These proactive changes are 

better than relying solely on events and operational experience to assess the effectiveness of 

current regulatory requirements and licensee practices. Therefore, the RMTF finds the following:

Finding 2.2: Risk assessments provide valuable and realistic insights 

into potential exposure scenarios. In combination with other technical 

analyses, risk assessments can inform decisions about appropriate defensein-depth measures.
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2.2 Recommendations

In addition to the NRC’s longstanding goal to move toward more risk-informed regulatory 

approaches, the benefits of performance-based approaches also have been recognized and 

encouraged. The NRC has recognized that purely deterministic and prescriptive approaches 

can limit the flexibility of both the regulated industries and the NRC to respond to lessons 

learned from operating experience and support the adoption of improved designs or processes. 

The RMTF recommends the following:

Recommendation 2.1: The goal to adopt risk-informed and performancebased approaches, where practical, should continue and should be 

incorporated into the revised regulatory framework.

As discussed above, the RMTF evaluated several options and determined that a framework 

based on risk management principles provided the best approach. Therefore, the RMTF offers 

the following recommendation: 

Recommendation 2.2: The general regulatory approach of the NRC should 

be defined in terms of “managing the risks” posed to workers and the public 

from the various uses of byproduct, source, and special nuclear materials.

To build on the mission and the Principles of Good Regulation from the NRC Strategic Plan and 

past agency performance, the RMTF adopted an “objective” for the agency. The objective is 

intended to help promote improved management by the NRC and its licensees of radiological 

hazards and associated risks. Key concepts to be captured by the agency’s objective include:

•	 Risk management as the vehicle to accomplish the NRC mission

•	 Applicability to all NRC programs (expressed in terms of byproduct, source, and special 

nuclear material)

•	 Maintaining the longstanding agency goal to use, whenever practical, performancebased approaches to regulation

•	 Applicability to NRC processes that establish requirements for or oversight of 

licensee programs 
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2-8 | Risk Management Task Force Chapter 2: The Framework

To support the above recommendation, the RMTF proposes the following statement as 

the objective for the NRC’s regulatory framework:

Objective

Manage the risks from the use of byproduct, source, and special 

nuclear materials through appropriate performance-based regulatory 

controls and oversight.

To a large degree, this objective has been part of the NRC’s existing practices. Therefore, it 

is not expected that refining the focus to risk management would require extensive changes in 

either the NRC’s philosophy or the regulatory requirements placed on its licensees. However, 

establishing a common language of risk management across all NRC activities would be 

beneficial. Consistent with Recommendation 2.1, the goal to use performance-based 

regulatory approaches whenever practical has been incorporated into the objective for the 

revised framework.

As stated in Finding 2.1, the concept of defense-in-depth has served the NRC and the 

industry well. This concept is a logical way to implement a risk management vision for the 

NRC, and the RMTF recommends that it be used across NRC programs and the life cycle of 

the related materials, devices, and facilities. The life cycle of regulatory activities generally 

can be described in terms of (1) design, fabrication, and construction, (2) operations, and 

(3) decommissioning. Defense-in-depth is used in this context to mean the barriers, controls, 

and personnel used to prevent, contain, and mitigate possible inadvertent exposure from 

radioactive materials. Although the principle of defense-in-depth can be applied to the NRC’s 

various regulatory programs, the nature of the hazards—in terms of the amount and form of 

radioactive materials—means that the actual barriers, controls, and requirements for personnel 

will vary considerably among the different types of NRC-licensed activities. 

Another important difference between NRC regulatory programs involves the amount of 

uncertainty associated with possible inadvertent exposure to radioactive materials. Some 

byproduct materials are used in small amounts, in stable forms, and within relatively simple 

devices. The radiological hazard in these cases is limited. There is a good understanding of 

release scenarios and possible pathways to the public for both normal operation and the loss 

of barriers and controls established for such devices. In contrast, power reactors are complex 

facilities with a large inventory of many different radioactive isotopes and possible scenarios that 

could interrupt safety functions. These functions include the control of reactivity, removal of heat 

from the reactor core, and confinement of fission products within the facility. These complexities 

also result in uncertainties associated with the response of plant equipment and personnel that 

comprise the barriers and controls to prevent the release of radioactive materials as a result 

of possible accidents. Risk assessment techniques have proven to be a valuable tool in the 

identification and resolution of uncertainties and in the evaluation of what are reasonable and 

sufficient levels of defense-in-depth for particular hazards. The RMTF, therefore, recommends 

the following:
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A Proposed Risk Management Regulatory Framework

Recommendation 2.3: In defining requirements for the protection of workers 

and the public, the NRC should recognize and address uncertainties 

associated with the hazards and the events, including human errors, which 

could challenge or degrade barriers and controls. A balanced approach 

that considers traditional and risk assessment techniques should be used to 

identify barriers and controls so that appropriate requirements are defined to 

prevent, contain, and mitigate exposures to radioactive materials.

To further define a revised regulatory framework, the RMTF defines a goal for the risk 

management objective that addresses Recommendation 2.3. Building on existing NRC 

practices, a risk-informed and performance-based defense-in-depth principle was identified as 

the best approach. Key concepts to be captured by the risk management goal include:

•	 Support implementation of the risk management framework

•	 Provide a logical way to implement the agency’s objective to manage the risks from the 

use of byproduct, source, and special nuclear materials

•	 Apply to all NRC programs (expressed in terms of hazard present)

•	 Incorporate performance-based objectives

•	 Address uncertainties associated with radiological hazards

•	 Address possible failure of some or all established barriers and controls

(including “cliff-edge” effects)

The RMTF proposes the following risk management goal to address Recommendation 2.3 and 

the attributes for a risk management goal:

Risk Management Goal

Provide risk-informed and performance-based defense-in-depth 

protections to:

•	 Ensure appropriate barriers, controls, and personnel to prevent, 

contain, and mitigate exposure to radioactive material according to 

the hazard present, the relevant scenarios, and the 

associated uncertainties.

•	 Ensure that the risks resulting from the failure of some or all of 

the established barriers and controls, including human errors, are 

maintained acceptably low.
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2-10 | Risk Management Task Force Chapter 2: The Framework

In program areas such as nuclear power reactors and waste repositories, the NRC has 

established quantitative goals to help define acceptably low levels of risk. Risk management 

for those program areas dealing with lesser amounts of radioactive materials often is achieved 

largely by standard radiation protection practices, which are established by all licensees 

in accordance with NRC regulations defined in 10 CFR Part 20. The use of the above 

risk management goal is addressed in Chapter 4 discussions of each regulatory program. 

However, if the agency decides to pursue the Risk Management Regulatory Framework, the 

NRC program offices would need to support its implementation by translating the above risk 

management goal into more practical criteria and guidance for specific devices and facilities. 

To ensure a consistent approach across the NRC’s regulatory programs, a common process 

should be used to identify and evaluate issues and then to make decisions and implement 

actions to manage the risks associated with the use of radioactive materials. Although not 

always presented in this form, the decisionmaking process shown in Figure 2-1, and described 

in more detail in Chapter 3, generally reflects the NRC’s way of regulating all of its licensees. 

This decisionmaking process is readily apparent in existing programs, such as regulatory 

analyses for rulemakings and the Reactor Oversight Process. 

There is a logical flow from the NRC’s mission to the proposed agency objective to manage 

risks, to the risk management goal provided by risk-informed and performance-based defensein-depth, and, finally, to a standard risk management decisionmaking cycle as shown above. 

This flow forms the proposed Risk Management Regulatory Framework shown in Figure 2-2.

Recommendation 2.4: The NRC should formally adopt the proposed 

Risk Management Regulatory Framework through a Commission 

Policy Statement.

A concerted effort will be needed to adopt the proposed Risk Management Regulatory 

Framework and improve its chances of success. This process would likely take years to 

develop and incorporate into NRC regulatory programs and procedures. An important first 

step in this process would be the formal adoption of the framework by the Commission using 

a vehicle such as a Policy Statement. This could then be followed by changes to guidance 

documents, such as management directives and office-level procedures, recognizing that many 

aspects of the proposed Risk Management Regulatory Framework are already in place, albeit 

using different terminology. For some regulatory programs, the implementation of the proposed 

Figure 2-1 The Regulatory Decisionmaking Process
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A Proposed Risk Management Regulatory Framework

framework could also involve changes to regulations, regulatory guides, and other processes 

that directly affect NRC licensees and other external stakeholders. Chapter 4 provides 

discussions of options for implementing the proposed framework for various NRC regulatory 

programs. This effort would also be an opportunity to update and consolidate previous policy 

statements, initiatives, and activities related to the adoption of risk-informed, performance-based 

approaches to improve the NRC’s regulation of source, byproduct, and special nuclear material.

Figure 2-2 A Proposed Risk Management Regulatory Framework 
A Proposed Risk Management Regulatory Framework

2-12 | Risk Management Task Force Chapter 2: The Framework

2.3 References

(DHS, 2010) U.S. Department of Homeland Security, “DHS Risk Lexicon,” 

September 2010.

(Executive Order, 2011) Executive Order 13563, “Improving Regulation and Regulatory 

Review,” January 18, 2011, published in the Federal Register on 

January 21, 2011 (76 FR 821).

(Kaplan and Garrick, 1981) S. Kaplan and B.J. Garrick, “On the Quantitative Definition of 

Risk,” Risk Analysis, 1:11–27, 1981.

(NRC, 1986) U.S. Nuclear Regulatory Commission, “Safety Goals for the 

Operations of Nuclear Power Plants; Policy Statement,” August 

1986, Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML011210381.

(NRC, 1998) U.S. Nuclear Regulatory Commission, “Consolidated 

Guidance About Materials Licensees,” NUREG-1556, 1998 

and later, multiple volumes.

(NRC, 1999) U.S. Nuclear Regulatory Commission, “White Paper on 

Risk-Informed and Performance-Based Regulation,” Staff 

Requirements Memorandum Regarding SECY-98-144, 

March 1, 1999, ADAMS Accession No. ML003753601.

(NRC, 2000) U.S. Nuclear Regulatory Commission, “Risk Analysis 

and Evaluation of Regulatory Options for Nuclear 

Byproduct Materials Systems,” NUREG/CR-6642, 

February 2000, ADAMS Accession No. ML003693052. 

(Not publicly available.).

(NRC, 2004) U.S. Nuclear Regulatory Commission, “Risk-Informed 

Categorization and Treatment of Structures, Systems, and 

Components for Nuclear Power Reactors, 10 CFR 50.69, 

published in the Federal Register on November 22, 2004, 

(67 FR 68008)

(NRC, 2012) U.S. Nuclear Regulatory Commission, “Strategic Plan: Fiscal 

Years 2008-2013,” NUREG-1614, Vol. 5, February 2012, 

ADAMS Accession No. ML12038A003.

.
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3. RISK MANAGEMENT DECISIONMAKING PROCESS

3.0 Background 

As discussed in Chapter 2, risk management provides the most useful framework for a 

comprehensive and holistic risk-informed, performance-based regulatory approach. Risk 

management concepts, process descriptions, and decisionmaking models are being widely 

used in various sectors, including Government agencies, financial institutions, and technology 

companies. There are several general descriptions of risk management processes and 

methodologies that share key points. These are summarized in Appendix A, “Risk Management 

Systems.” A basic structure with typical stages of a risk management process is included in the 

Risk Management Regulatory Framework (Figure 2-2) and is repeated below in Figure 3-1: 

In describing the options available to the NRC for improving its regulatory programs by adopting 

a risk management framework, it is important to recognize that the protection of workers and 

the public from radiological hazards is provided by the owners or operators of the materials, 

devices, and facilities licensed by the NRC. However, the use of radioactive materials 

introduces particular risks and public concerns; therefore, the NRC’s regulatory regime, adopted 

pursuant to the Atomic Energy Act and other statutes, provides the legal basis for ensuring that 

NRC licensees provide reasonable assurance of adequate protection of the health and safety of 

their workers and the public. 

The NRC has developed regulatory requirements for licensees to provide radiation protection 

programs, monitoring programs, and other controls to limit the radiation exposure to workers 

and the public to acceptably low levels. Managing the risks associated with a totally controlled 

radioactive material and population would be relatively straightforward; exposures could be 

limited using standard radiation protection programs (controlling time, distance, and shielding). 

However, many of the uses of radioactive materials introduce possible conditions or events, 

including human errors, which might lead to a breakdown of established barriers and controls. 

A key role of a Risk Management Regulatory Framework, as described in Chapter 2, is to 

ensure that no undue risk to the public from such unplanned events occurs. 

The Risk Management Task Force (RMTF) is recommending that the NRC meet its risk 

management goal by requiring its licensees to provide risk-informed defense-in-depth 

protections. Although the terminology may vary, defense-in-depth concepts have been used 

throughout the NRC’s history and across its regulatory programs. As mentioned in previous 

findings and recommendations (Chapter 2), defense-in-depth concepts have been instrumental 

in establishing effective regulatory controls and oversight for the civilian use of byproduct, 

Figure 3-1 The Regulatory Decisionmaking Process
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3-2 | Risk Management Task Force Chapter 3: Risk Management Process

source, and special nuclear materials. The NRC also has recognized and encouraged using 

performance-based approaches. Additional discussions of defense-in-depth are provided in 

Appendix C, “Defense-in-depth,” and other sections of this report. 

As discussed in the remainder of this report, appropriate evaluations and actions will vary 

depending on the type and amount of radioactive material and the complexity of the device or 

facility. For some materials and devices, concerns about an inadvertent exposure are small, 

and few controls beyond routine radiation protection programs are necessary. Facilities such 

as commercial nuclear power reactors, however, contain large amounts of radioactive materials 

that warrant multiple levels of protection to prevent, contain, and mitigate possible releases. 

The NRC uses processes and programs, such as rulemaking and guidance, licensing, 

environmental reviews, and oversight, to define and maintain confidence in the appropriate 

defense-in-depth measures to be provided by licensees. 

The risk management goal and related decisionmaking process are intended to be used 

universally for NRC actions on the regulation and oversight of licensed activities, including 

reactors, materials, waste, fuel cycle, and transportation. Therefore, if adopted, the proposed 

risk management framework would be used by major program offices and supporting offices 

whenever they make technical or other regulatory decisions. In many cases, the current NRC 

decisionmaking processes (e.g., integrated decisionmaking for nuclear reactors and riskinformed decisionmaking for materials licensees) are similar to the proposed risk management 

framework described in this report. Major actions, such as licensing of reactor or fuel cycle 

facilities, agency-level initiatives, and rulemakings, include project plans and processes that are 

often explained in terms of the six steps of the decisionmaking process (or in equivalent terms 

and steps). At first glance, the handling of some routine activities may appear simpler than the 

decisionmaking process, but this is probably because some steps—such as selecting technical 

analysis techniques and decisionmaking criteria—are already incorporated into procedures or 

guidance documents. Provided the existing guidance and processes ensure that appropriate 

barriers and controls are established and risks are maintained at an acceptably low level, the 

move to a risk management framework is unlikely to require significant or immediate changes to 

existing practices. 

Although it was beyond the immediate scope of the RMTF, general concepts of risk 

management and methodical decisionmaking are applicable to other agency activities. For 

example, a consistent risk management approach across NRC programs could support changes 

in management practices so that program managers would have increased flexibility to allocate 

resources to address risk or safety concerns. Insights from inspections, operating experience, 

scientific studies, or risk assessments could inform timely changes to programs and focus areas 

as part of the implementation of budgeted resources. The NRC also uses risk management 

processes in providing security to the agency’s information systems. 

Additional discussions on the six stages of regulatory decisionmaking are provided below and 

in Appendix A and Appendix B, “Risk Management─Analyses and Deliberations.” Insights on 

the implementation of the proposed framework for specific regulatory programs are provided in 

Chapter 4. 
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3.1 Identify Issues 

The regulatory decisionmaking process begins with an identified issue, a proposal from a 

licensee or other stakeholder, or some other problem that requires a regulatory decision by 

the NRC staff or management. It is important to clearly define the problem to be analyzed 

and resolved by the subsequent steps in the process. Framing or characterizing the issue 

or problem refers to defining the scope of the problem (e.g., applicable licensees or facilities) 

and its potential implications in terms of the risk management goal identified in Figure 2-2 and 

restated below: 

Risk Management Goal

Provide risk-informed and performance-based defense-in-depth 

protections to:

•	 Ensure appropriate barriers, controls, and personnel to 

prevent, contain, and mitigate exposure to radioactive 

material, according to the hazard present, the relevant 

scenarios, and the associated uncertainties.

•	 Ensure that the risks resulting from the failure of some 

or all of the established barriers and controls, including 

human errors, are maintained acceptably low.

Some problems, such as the determination of the appropriate response to the attacks of 

September 11, 2001, initially involved all NRC activities. Broadly applicable issues usually are 

divided into narrower problem statements for specific regulatory programs. Major licensing 

actions within NRC program areas often require an assessment of the overall risks associated 

with a device or facility, including how different measures or controls contribute to risk-informed 

and performance-based defense-in-depth. Most NRC decisions, however, relate to a problem 

or proposal that affects individual barriers or controls associated with specific devices or 

facilities already designed, licensed, and constructed. It is sometimes easy to forget that these 

decisions should also be made in the context of the overall risk-informed and performancebased defense-in-depth goal. 

3.2 Identify Options 

It is important to develop and consider choices in the resolution of issues or problems, or 

in the review of proposals from licensees. Some NRC processes, such as rulemaking and 

environmental reviews, include specific guidance for identifying and evaluating multiple options 

or alternatives, and they usually include a “no action” alternative. Other activities, such as the 

review of specific proposals from licensees, may not require the NRC staff to identify options 

beyond approving or denying an application, although discussion of alternatives with licensees 

is common even in these licensing-related deliberations. 

The development or consideration of alternatives is an opportunity for the NRC staff and 

licensees to propose a performance-based approach to meet regulatory requirements. As 
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3-4 | Risk Management Task Force Chapter 3: Risk Management Process

discussed in Appendix D, “Performance-Based Regulation,” performance-based approaches 

can improve the flexibility of both the regulated industries and the NRC to respond to lessons 

learned from operating experience and support the adoption of improved designs or processes. 

The use of performance-based approaches also can support more efficient regulation of 

licensees by reducing the number of licensing actions associated with the future operation of 

facilities (e.g., the technical specification improvement initiatives for nuclear power reactors). 

As such, the NRC staff should consider the use of performance-based approaches whenever 

practical and should identify and consider those approaches as alternatives as part of the risk 

management framework. This goal is reflected in the objective defined in Recommendation 2.1 

and Figure 2-2, which states that, where possible, the management of risks from the use of 

byproduct, source, and special nuclear material should be evaluated using performance-based 

regulatory controls and oversight. 

3.3 Analyze 

A major element of a risk management program is the evaluation of risks associated with the 

subject system or activity. A more detailed discussion of technical analyses and current NRC 

approaches is provided in Appendix B. In the context of this report, technical analysis refers 

to the general task of analyzing a problem or issue to support deliberations. A key concept 

in the development of a risk management process is that technical analyses can be done in 

many ways, including traditional mechanistic analyses (e.g., thermal-hydraulic calculations), 

probabilistic risk assessments (PRAs), and other techniques selected to support specific 

decisions related to particular issues and hazards. 

Within the proposed risk management framework, technical analyses will be used to support 

decisions on appropriate barriers and controls (including personnel) to prevent, contain, and 

mitigate exposures to radioactive materials. Technical analyses also will be used to ensure 

that the risks from events that degrade or challenge the barriers are maintained acceptably 

low. The appropriate barriers range from simple containers for some radioactive sources 

to complex structures for nuclear power plants. Likewise, the systems and actions taken to 

maintain barriers can range from labels and administrative controls to complex mitigation 

systems. Licensee organizations and individual workers are major contributors to the goal of 

preventing, mitigating, and containing the release of radioactive materials. Controls established 

for materials licensees can also address circumstances where the use of a device requires the 

temporary bypass of physical barriers (e.g., radiography and irradiation devices). The possible 

need for emergency preparedness requirements are based on the risk that barriers might be 

compromised in a way that the public could be exposed to radioactive materials and protective 

actions—such as sheltering or evacuation—might be warranted. 

Just as the ultimate determination on appropriate barriers, controls, and personnel depends 

on the radiological hazard and possible challenges to those protections, the selection of 

technical analysis techniques for a specific regulatory decision will consider how the issue 

relates to the hazards and challenges. In general, decisions on simpler devices and frequent 

events (e.g., activities with many licensees) can be supported by analyses based on traditional 

engineering approaches, operating experience, and qualitative risk assessments. Decisions 

related to more complex facilities and infrequent events (e.g., nuclear power reactors) can 

benefit from analyses that include more robust quantitative risk assessments. More complex 

facilities also can introduce additional uncertainties about how the facility will respond to relevant 

scenarios and the ability of barriers and controls to contain and mitigate possible releases of 

radioactive material. Such uncertainties should be addressed, as practical, within the technical 
Chapter 3: Risk Management Process Risk Management Task Force | 3-5

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analyses and be identified and accounted for as part of the deliberative process. Sensitivity 

studies can be a useful tool to evaluate uncertainties and otherwise inform the decisionmaking 

process. The technical analysis step of the process is further discussed in Appendix B. 

The key to selecting a risk assessment technique or combination of techniques is to ensure that 

the analysis will support the associated deliberation or decisionmaking process in an effective 

and efficient manner. Acceptance criteria or other rationale for decisionmaking should be 

established before using traditional methods, PRAs, or a combination of approaches. Many 

NRC decisions, in effect, are an evaluation of possible changes (e.g., license amendments, 

new understanding gained from operating experience) from an established baseline (e.g., 

initial licensing decision). In some cases, the evaluation (usually using traditional mechanistic 

techniques or engineering judgment) may conclude that the change does not have an adverse 

impact on a specific barrier or control and a broader assessment of the impact on overall 

risk may not be necessary. In other cases, a problem or proposal may involve a change to 

a barrier or control that requires consideration of the overall risk-informed and performancebased defense-in-depth protections. Previous NRC efforts to address appropriate balancing of 

traditional analysis methods and PRA methods include the integrated decisionmaking process in 

Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed 

Decisions on Plant-Specific Changes to the Licensing-basis” (NRC, 2011), for nuclear reactors 

and risk-informed decisionmaking for materials programs. The topic of acceptance criteria is 

discussed further in the following section. 

3.4 Deliberate 

The risk management framework is primarily an approach to provide structure and logic to the 

decisionmaking process. The technical analysis and identification of options are key inputs to 

the process, but they can only be successful if there is an equally disciplined approach to the 

actual decisionmaking. As stated in Chapter 2, the NRC’s decisions related to its core mission 

can be described in terms of the risk management goal, which is to provide risk-informed and 

performance-based defense-in-depth protections for the use of byproduct, source, and special 

nuclear materials. 

The report, “Understanding Risk: Informing Decisions in a Democratic Society” (National 

Research Council, 1996), defines deliberation as “any formal or informal process for 

communication and collective consideration of issues.” It further states that deliberations 

“formulate the decision problem, guide analysis to improve decision participants’ understanding, 

seek the meaning of analytic findings and uncertainties, and improve the ability of interested 

and affected parties to participate effectively in the risk decision process.” The concept of 

deliberation is called the “integrated decisionmaking process” in Regulatory Guide 1.174. A 

more detailed discussion of the deliberative processes to support a risk management framework 

is provided in Appendix B. 

The technical analysis is an important input into the deliberation process, but it is not the only 

factor influencing the final decision. As Regulatory Guide 1.174 states: 

“The results of the different elements of the engineering 

analyses ... must be considered in an integrated manner. None 

of the individual analyses is sufficient in and of itself.”
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3-6 | Risk Management Task Force Chapter 3: Risk Management Process

Decisionmakers need to consider the assumptions, uncertainties, and sensitivities associated 

with the technical analysis, as well as the analytical results and how they compare to decision 

criteria established for mechanistic approaches, PRAs, or a combination thereof. The 

deliberative process may also include consideration of resources and schedules for the agency 

and its licensees and the input received from various internal and external stakeholders. Other 

factors include legal requirements and desired consistency with other guidance documents, 

treaties, or standards. 

A representation of the deliberation process, including consideration of the different factors, is 

provided in Figure 3-2. 

Deliberation 

Stakeholder 

Input 

Assumptions, 

Uncertainties 

 and 

Sensitivities 

Technical

Analysis 

one or more 

techniques 

Decision 

Criteria 

Resource 

and 

 Schedule 

Constraints 

Other 

Factors 

Decision & 

Implementation 

Options 

Figure 3-2 Deliberations 

Guidance documents to help in the decisionmaking process include a variety of acceptance 

criteria for different regulatory programs, specific facilities, and for different types of events or 

operating conditions. Part of the long-term implementation of the proposed Risk Management 

Regulatory Framework would include updating guidance documents to reflect the goal of riskinformed and performance-based defense-in-depth. Given the variety of hazards, scenarios, 

and uncertainties associated with possible exposures to radioactive materials from different 

types of NRC-licensed devices and facilities, it would be necessary to update or develop such 

guidance within each regulatory program.

3.5 Implement Decision

Decisions resulting from deliberations on specific issues or proposals address whether the 

defense-in-depth protections associated with the radiological hazard are adequate, can be 

relaxed, or need to be strengthened. The NRC implements its decisions within regulatory 

processes that include the following: 

Figure 3-2 Deliberations
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•	 preparing regulations and guidance 

•	 reviewing proposed licensing actions 

•	 performing environmental reviews 

•	 executing its oversight programs

The actual defense-in-depth protections are implemented and maintained by licensees 

who process and control the radioactive materials or related facilities. Findings and 

recommendations for specific regulatory programs are discussed in Chapter 4. 

3.6 Monitor 

An important part of any decisionmaking model is monitoring and feedback to identify issues 

and to gauge the effectiveness of decisions and implementation actions. Evaluations provide 

important information about the effectiveness of the framework, the regulatory programs, and 

specific decisions. Evaluation tools include existing programs, such as operating experience 

programs, inspection and oversight programs, and interactions with various stakeholders. In 

some cases, the implementation of specific actions to address risks may include feedback 

provisions, such as reporting requirements or periodic assessments. In either case, evaluation 

plans should be built into the overall implementation plan to specify when evaluations will be 

conducted, who will conduct them, and what will be evaluated. New information may emerge 

during an evaluation that is sufficiently important to require repeating some parts of the risk 

management process. In such cases, the six-stage process would not be sequential; instead, it 

would require flexibility and iteration as important new information comes to light. 

An area of tension related to NRC processes and decisions is the need to balance goals for 

regulatory stability and predictability with the desire to identify and address new information, 

especially information related to risks associated with regulated activities. The RMTF 

recommendations on periodic risk assessments of the hazards and risks associated with specific 

uses of source, byproduct, and special nuclear materials are provided in Chapter 4. Such 

periodic assessments would then support the risk management framework to help determine if 

changes to regulatory requirements or oversight were in order. 

3.7 Communications 

Communications with internal and external stakeholders is vitally important during the 

decisionmaking process. It is notable that all of the risk management frameworks developed 

by various groups and outlined in Appendix A highlight the importance of communications, 

engagement of stakeholders, and collaboration as part of the deliberative process. 

The development and implementation of the proposed Risk Management Regulatory 

Framework will present a significant communication challenge within the NRC and with external 

stakeholders. Experiences associated with previous initiatives related to risk-informed, 

performance-based regulation demonstrated the importance of effective communication in 

explaining and promoting the changes being introduced. Some of these challenges and 

possible implementation strategies, including key communications, are discussed further in 

Chapter 5. 
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3-8 | Risk Management Task Force Chapter 3: Risk Management Process

3.8 References 

(National Research Council, 1996) “Understanding Risk: Informing Decisions in a Democratic 

Society,” National Academy Press, 1996. 

(NRC, 2011) U.S. Nuclear Regulatory Commission, “An Approach for 

Using Probabilistic Risk Assessment in Risk-Informed 

Decisions on Plant-Specific Changes to the Licensingbasis,” Regulatory Guide 1.174, Revision 2, May 2011, 

Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML100910006. 
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4. IMPLEMENTATION OPTIONS 

4.1 Introduction

This chapter discusses how the proposed Risk Management Regulatory Framework described 

in Chapters 2 and 3 could be implemented by the NRC in its licensing and oversight activities. 

Since the historical approach to licensing and oversight of particular facilities using nuclear 

materials has varied, the actions that would need to be taken to implement the framework would 

also vary.

For each major category of NRC-regulated activity, including reactors, materials, waste disposal, 

uranium recovery, fuel cycle, spent nuclear fuel storage, and transportation, the associated 

section provides the following:

•	 Background on regulatory context.

•	 Risk Management Task Force (RMTF) findings and associated recommendations on 

the potential implementation of the proposed framework (for both safety and security 

activities). The findings are related to what the RMTF considers to be important gaps 

between how NRC-licensed materials, devices, and facilities are regulated today and 

the recommendations are related to how they would be regulated in the future using the 

proposed framework.

•	 Recommendations on implementation approaches. 

In making its recommendations, the RMTF considered factors, such as who would be 

affected (the NRC, licensees, or both), the administrative mechanism that would be used for 

implementation (e.g., rulemaking), and the resources (time and cost) that would be required for 

implementation. Using these factors, the RMTF defined three implementation options:

Option A: Continue the current approach (“patchwork”)

This option would direct few, if any, additional resources to the development and implementation 

of a broad risk management approach across NRC programs. The efforts related to ongoing 

risk-informed and performance-based initiatives and activities related to the followup to the 

Fukushima accident would continue on their current courses. 

Option B: Implement the proposed Risk Management Regulatory Framework through selected 

guidance and rule changes

This option would emphasize specific rule and guidance changes to implement the proposed 

Risk Management Regulatory Framework. For specific rule changes, this option would maintain 

the basic structure of current regulations and oversight programs, but it would develop changes 

to or additions of specific regulations applicable to NRC licensees. An historical example of 

this approach is the development and issuance of 10 CFR 50.65, “Requirements for Monitoring 

the Effectiveness of Maintenance at Nuclear Power Plants” (NRC, 1991), which added a 

significant new requirement within the existing licensing and oversight programs. For guidance, 

changes would be largely aimed at changing how the NRC operates, but new requirements or 

expectations on licensees would not necessarily be imposed. An example of a successful effort 

developed and implemented through guidance is Regulatory Guide 1.174, “An Approach for 
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4.1-2 | Risk Management Task Force Section 4.1: Implementation Options

Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-specific Changes to 

the Licensing-basis” (NRC, 2011). Guidance-oriented activities likely would involve revisions 

or additions to Commission Policy Statements and guidance documents, such as regulatory 

guides, review plans, management directives, and office-level procedures. This option would 

require development of a plan describing actions to be taken and allocation of staff resources 

for the plan’s execution.

Option C: Implement the proposed Risk Management Regulatory Framework through broadscale regulatory framework changes

This option would involve a significant revision of the basic framework used for the licensing 

and oversight of a particular type of NRC-regulated activity. NUREG-1860, “Feasibility Study 

for a Risk-informed and Performance-based Regulatory Structure for Future Plant Licensing,” 

described such possible structural changes to the licensing approach for nuclear power 

reactors (NRC, 2007). Significant planning and resource allocations would be required to 

implement this option.

The following sections discuss the implementation of the proposed Risk Management 

Regulatory Framework for various NRC programs and are organized as follows:

•	 Power Reactors (Section 4.2.1)

o Currently operating reactors licensed under the provisions of 10 CFR Part 50, 

“Domestic Licensing of Production and Utilization Facilities, 

o “New” reactors licensed under the provisions of 10 CFR Part 52, “Licenses, 

Certifications, and Approvals for Nuclear Power Plants,”

o Generation IV reactors. 

•	 Nonpower Reactors (Section 4.2.2)

•	 Materials (Section 4.3)

•	 Waste Disposal (Section 4.4)

o Low-Level Waste

o High-Level Waste

•	 Uranium Recovery (Section 4.5)

•	 Fuel Cycle (Section 4.6)

•	 Spent Fuel Storage (Section 4.7)

•	 Transportation (Section 4.8)
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4.1.1 References

(NRC, 1991) U.S. Nuclear Regulatory Commission, “Monitoring the 

Effectiveness of Maintenance at Nuclear Power Plants,” Final 

Rule, published in the Federal Register on July 10, 1991 

(56 FR 31306).

(NRC, 2007) U.S. Nuclear Regulatory Commission, “Feasibility Study for a 

Risk-informed and Performance-based Regulatory Structure 

for Future Plant Licensing,” NUREG-1860, Vol. 1, December 

2007, Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML080440170.

(NRC, 2011) U.S. Nuclear Regulatory Commission, “An Approach for Using 

Probabilistic Risk Assessment in Risk-informed Decisions on 

Plant-Specific Changes to the Licensing-basis,” Regulatory 

Guide 1.174, Revision 2, May 2011, ADAMS Accession 

No. ML100910006.

Section 4.2: Nuclear Reactors Risk Management Task Force | 4.2-1

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4.2 Nuclear Reactors

4.2.1 Power Reactors

4.2.1a Background

The historical basis for power reactor licensing and oversight is described well elsewhere 

(Walker, 2010; Okrent, 1981) and will not be discussed extensively here. In brief, power 

reactors in the United States have been licensed using 10 CFR Part 50.1 Implementation 

of 10 CFR Part 50 has been achieved, for the most part, using deterministic methods and 

acceptance criteria. From a safety perspective, a set of licensing-basis events was established 

that was intended to ensure conservatism in design and protection from a wide spectrum of 

postulated events, up to and including design-basis accidents (DBAs). Appendix F provides 

additional information on this spectrum of postulated events. These postulated accidents are 

highly stylized and generally do not consider multiple failures of safety systems. Qualitative 

approaches for ensuring reliable safety systems, such as the single failure criterion, were 

implemented. Testing plans and operational limits were established in technical specifications 

to ensure that, if called upon, safety systems would perform as intended. In addition to the 

failure or malfunction of plant equipment, the challenges to and protections of nuclear power 

plants from various external hazards are described in plant licensing documents (e.g., final 

safety analysis reports (FSARs), Chapter 2). External hazards are events that could initiate a 

variety of plant transients while also challenging one or more of the barriers provided to mitigate 

or contain potential releases of radioactive materials.

One key concept that emerged early in the licensing history of nuclear power reactors was 

“defense in depth.” This concept was developed and applied to compensate for the recognized 

lack of complete knowledge of nuclear reactor operations and the consequences of potential 

accidents. Although both reactor operation experience and the knowledge base of potential 

(and actual) consequences have grown considerably since that time, the Risk Management 

Task Force (RMTF) concludes that the concept continues to be highly relevant today. That is, 

even with this increased experience, there continues to be knowledge gaps and uncertainties2

 that can be addressed by a defense-in-depth approach. Additional discussion of defense-indepth is provided in Appendix C.

A question that arises when considering changes to nuclear reactor regulations and the 

categorization of events or equipment is if requirements are needed for adequate protection of 

the health and safety of the public. The NRC establishes regulatory requirements for protecting 

public health and safety and common defense and security using a two-tier structure. This twotier structure is consistent with the statutory requirements in the Atomic Energy Act of 1954, as 

amended. The top tier consists of those requirements needed to ensure adequate protection of 

public health and safety, and to be in accordance with the common defense and security. The 

adequate protection standard is also described in terms of ensuring that licensed activities pose 

no undue risk to public health and safety or are not inimical to common defense and security. In 

a case brought before the Court of Appeals for the D.C. Circuit (Union of Concerned Scientists 

1 All currently operating power reactors have been licensed under 10 CFR Part 50. The newest currently 

operating reactor began operations in 1996. The operating license for Watts Bar Unit 2 (and potentially 

Bellefonte, Unit 1) also are being pursued using the provisions of 10 CFR Part 50.

2 The conservatism embedded in reactor design by the DBAs, external hazard analyses, and defense-indepth is also intended to protect the facility from accidents or conditions that the NRC has not thought of, the 

so called “unknown unknowns.”
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4.2-2 | Risk Management Task Force Section 4.2: Nuclear Reactors

v. NRC, 824 F.2d 108, 120 (D.C. Cir. 1987)), the Court recognized that adequate protection (no 

undue risk) does not equate to “zero risk.” The Court stated:

Similarly, under the adequate-protection standard of section 182(a), the NRC 

need ensure only an acceptable or adequate level of protection to public health 

and safety; the NRC need not demand that nuclear power plants present no risk 

of harm.

Measures taken to prevent, contain, and mitigate events or concerns in this tier historically 

correspond to the traditional design-basis accidents described in Appendix F and several 

programmatic requirements for dealing with beyond-design-basis events (e.g., emergency 

planning and loss of large areas due to fires or explosions). NRC requirements to address 

concerns of adequate protection are developed and imposed without consideration of cost. 

The second tier of the NRC safety structure was described by the Court as follows:

If it so desires, however, the Commission may impose safety measures on 

licensees or applicants over and above those required by section 182(a)’s 

adequate-protection standard. As we have noted, section 161 of the Act 

empowers the Commission to issue rules, regulations, or orders to “protect 

health or to minimize danger to life or property.” 42 U.S.C. Sec. 2201(b), (i). This 

section cannot be read simply to permit the Commission to provide adequate 

protection; another section of the Act requires the Commission to do that much. 

We therefore must view section 161 as a grant of authority to the Commission to 

provide a measure of safety above and beyond what is “adequate.” The exercise 

of this authority is entirely discretionary. If the Commission wishes to do so, it 

may order power plants already satisfying the standard of adequate protection to 

take additional safety precautions. When the Commission determines whether 

and to what extent to exercise this power, it may consider economic costs or 

any other factor. The Commission, after all, need not exercise the authority 

granted by section 161 at all; given this fact, the Commission certainly may use 

cost-benefit analysis to decide whether exercising the authority conferred by 

section 161 makes economic or policy sense.

In the development of its findings and recommendations, the RMTF considered how some 

requirements would be associated with the first-tier mission of the NRC to ensure adequate 

protection of public health and safety and the common defense and security, and how other 

requirements would address the second tier by going beyond adequate protection in an 

attempt to further reduce risk. As described by the Court of Appeals and in longstanding NRC 

regulations and guidance documents, the requirements established to reduce risk (beyond 

measures needed for adequate protection) will not attempt to eliminate all risk but will instead 

pursue reasonable reductions. An evaluation of the costs and benefits of proposals falling within 

the second tier has been used as part of the determination of what is a reasonable requirement 

to reduce risks to public health and safety and the common defense and security.

The RMTF considered the two-tier safety structure discussed above, along with the risk 

management goal described in Chapter 2, when preparing its findings and recommendations. 

The risk management goal is to:
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Provide risk-informed and performance-based defense-indepth protections to:

• Ensure appropriate barriers, controls, and personnel to 

prevent, contain, and mitigate exposure to radioactive 

material according to the hazard present, the relevant 

scenarios, and the associated uncertainties; and 

• Ensure that the risks resulting from the failure of some 

or all of the established barriers and controls, including 

human errors, are maintained acceptably low. 

Taken together, the above criteria for risk-informed and performance-based defense-in-depth 

address both tiers of the safety structure. However, the first criterion does not align with the 

first tier (adequate protection) by itself and the second criterion does not align with the second 

tier (additional protections) by itself. Instead, a specific design feature or operating control can 

be seen as addressing one or more of the factors in the risk management goal. The nature 

and importance of the protection provided will determine whether it serves to ensure adequate 

protection or provides additional protections to further reduce risk to life or property. 

The above observation on defense-in-depth is fully consistent with the general findings 

and recommendations in Chapter 2. However, as mentioned in earlier parts of this report, 

experience has also identified limitations with the traditional approaches, particularly in 

supporting specific decisions on what are reasonable and sufficient levels of defense-in-depth 

for particular hazards. As described in Recommendation 2.3, the RMTF recommends that 

a balanced approach, considering traditional and risk assessment techniques, is needed to 

identify appropriate barriers and controls.

Probabilistic risk assessment (PRA) was first used in the early 1970s for nuclear power reactors 

(NRC, 1975), after the designs of essentially all operating plants were fixed. The NRC uses risk 

assessment in a way that recognizes the particular strengths of this approach and complements 

the traditional, more deterministic, approach. That is, risk assessment provides:

•	 A systematic approach for assessing the wide variety of hazards that can challenge the 

safety of an NRC-licensed facility.

•	 A logical method for characterizing the capability of a facility’s design and operation to 

respond to the identified hazards.

•	 A method for estimating the consequences of combinations of hazards and unsuccessful 

responses to these hazards (thousands of realistic accident sequences are 

investigated, in contrast to the limited number of stylized accidents considered in the 

traditional approach).
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4.2-4 | Risk Management Task Force Section 4.2: Nuclear Reactors

•	 A model for assessing the occurrence frequencies of these hazards, the probabilities 

of unsuccessful response (with the associated design or operational failures), and the 

consequences of the unsuccessful responses. These frequencies and probabilities can 

be estimated quantitatively.

•	 Rankings of accident sequences and of systems, structures, and components (SSCs) 

according to their contributions to risk.

•	 Valuable input to risk management through the rankings of accident sequences and 

SSCs that allow the allocation of resources to be focused on what really matters to risk.

Proposed new power reactor designs are light-water reactors similar in many respects to 

operating power reactors. New reactor designs are being licensed using the various subparts 

of 10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” which 

in all technical areas essentially references 10 CFR Part 50. That is, new power reactors are 

licensed using the same set of licensing-basis events, including design-basis events that include 

assumptions such as the single failure criterion. Recognizing the value of PRA in providing 

a complementary view of safety, each new design is required to have a PRA, and any plant 

referencing that design is required to update the PRA by the time of fuel loading to reflect the 

final design and operational characteristics. As envisioned in the NRC’s Severe Accident Policy 

Statement (NRC, 1985), new reactors must satisfy a regulatory requirement in 10 CFR Part 52 

to include severe accident features in their designs. In addition, an analysis is required for new 

reactor designs to ensure that critical safety functions would be maintained following the impact 

of a large commercial aircraft. In general, the designs of these new power reactors are fixed and 

approved through an NRC-issued design certification, although none is in actual operation.

A subset of new reactors being pursued by the nuclear industry involves small modular 

designs that introduce some new features, compared to traditional large reactor designs. The 

small modular reactors (SMRs) currently being developed are integral pressurized-water 

reactors (iPWRs) with major components, such as steam generators, control rod drives, and a 

pressurizer contained within the reactor vessel, thereby eliminating the need for most reactor 

coolant piping. The power levels for SMRs generally are less than 300 megawatt-electric 

(MWe), and passive safety features are incorporated into the designs. Potential applicants 

and the NRC staff are planning on a licensing approach for iPWRs that is generally consistent 

with that used for larger new reactors with passive safety features, such as the Westinghouse 

AP1000. Some applicants are considering the two-phase 10 CFR Part 50 licensing process 

(i.e., a construction permit and operating license), but site-specific licensing would be linked to 

10 CFR Part 52 design certifications for subsequent deployments. Therefore, CFR 10

Part 52 provisions mentioned above would be addressed for SMRs. Some iPWR vendors have 

expressed an interest in performing a Level 3 PRA. However, it is not clear how these efforts 

would be incorporated into the overall licensing approach.

Generation IV power reactor designs, including, for example, the Next Generation Nuclear Plant 

(NGNP), are being considered by designers, although none has yet been submitted officially 

for NRC review. These designs vary considerably from the current set of operating reactors 

and the proposed new reactor designs, including the consideration of coolants such as helium, 

liquid sodium, and molten salt. Reflecting these design differences, discussions have been 

held with the NRC on using alternative regulatory approaches, such as using event frequency 

information, to define the set of design-basis accidents to be analyzed. The designs of these 

advanced reactors are still conceptual.
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4.2.1b RMTF Findings and Recommendations for Power Reactors

The RMTF has considered the history and process for licensing and oversight of power reactors 

in the context of the risk management framework and has developed a set of findings and 

recommendations. The findings relate to what the RMTF has identified to be important gaps 

between how power reactors are regulated today and how they would be regulated in the future 

using the proposed Risk Management Regulatory Framework. The RMTF assessed each 

power reactor class (operating, new and SMR, and Generation IV) individually and found many 

similarities with respect to the findings. Implementation recommendations for each class are 

provided in the following sections of this report. 

Design-basis Accident Licensing Approach

Methods used to certify new designs include the set of DBAs and qualitative reliability 

approaches developed more than 30 years ago, although more modern approaches exist and 

are used extensively in, for example, the oversight of operating reactors. Some licensees have 

used PRA to improve operational flexibility, even with the current set of DBAs and designs 

established long ago. Further, it appears that both licensees and the NRC benefit from the 

clarity that the DBA approach provides, making interactions between licensees and the NRC 

more efficient.

The RMTF examined power reactor regulatory requirements from a broad and strategic 

perspective. From this perspective, the RMTF observes a “within” design basis part of the 

power reactor regulatory framework that uses, among other things: qualitative approaches 

for ensuring system reliability (e.g., the single failure criterion) when more modern quantitative 

approaches exist; stylized considerations of human performance (e.g., operators are assumed to 

take no actions within, for example, 30 minutes of an accident’s initiation); and a set of accidents 

not reflective of operating experience and modern understanding (e.g., the now-understood 

downside risks of automatic containment spray actuation in pressurized-water reactors with 

large dry containments). The continued use of the DBAs established over 30 years ago runs 

counter to the NRC Principles of Good Regulation that regulations (and, by extension, DBAs) 

should be based on the best available knowledge from research and operational experience. 

The longstanding use of the same design-basis events has, however, helped maintain stability in 

regulatory processes, which is another aspect of the NRC’s Principles of Good Regulation. The 

RMTF, therefore, offers the following finding and recommendation for power reactors:

Finding PR-F-1: The concept of design-basis events and accidents continues to be 

a sound licensing approach, but the set of design-basis events and accidents has 

not been updated to reflect insights from power reactor operating history and more 

modern methods, such as PRA.

Recommendation PR-R-1: The set of design-basis events and accidents should 

be reviewed and revised, as appropriate, to integrate insights from power reactor 

operating history and more modern methods, such as PRA.
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4.2-6 | Risk Management Task Force Section 4.2: Nuclear Reactors

Beyond-design-basis Accidents

Several rules have been issued to address accidents identified in PRAs as having 

significant risks not recognized in or enveloped by the set of DBAs. An important example is 

10 CFR 50.63, the station blackout rule (NRC, 1988). However, these events were not added 

to the set of DBAs, and a consistent framework was not developed to address them. As a 

result, a “patchwork” of agency guidance was established to address beyond-design-basis 

accidents. The Fukushima Near-Term Task Force (NTTF) report (NRC, 2011a) included a 

number of observations on the results of this ad hoc approach and recommended that “a logical, 

systematic, and coherent regulatory framework for adequate protection that appropriately 

balances defense-in-depth and risk considerations” be established. 

One aspect of the NRC’s consideration of beyond-design-basis accidents has been the use of 

voluntary industry initiatives. For example, in the 1990s, the NRC and power reactor licensees 

developed and implemented accident management guidelines, the purpose of which was “to 

enhance the capabilities of the licensee’s emergency response organization to prevent and 

mitigate severe accidents and minimize any offsite releases” (NRC, 1996). As explained in 

that reference (SECY-96-088), the staff “accepted the industry commitment to implement 

A/M [accident management] at each NPP [nuclear power plant] pursuant to a formal industry 

position on A/M in lieu of pursuing other actions for obtaining improvements in industry A/M 

capabilities ...,” such as incorporating requirements into a regulation or operating licenses. 

In 2011, the Fukushima NTTF noted that “NRC inspection programs give less attention to 

beyond-design-basis requirements and little attention to industry voluntary initiatives since there 

are no requirements to inspect against.” During discussions between the RMTF and the NRC 

staff, the extent to which such voluntary initiatives could be credited in risk assessments (which 

are intended to be realistic) arose on several occasions. More specifically, the extent to which 

licensee activities undertaken in such initiatives can be credited has been a source of contention 

in the Significance Determination Process, which is part of the Reactor Oversight Process.

The Western European Nuclear Regulators’ Association (WENRA) prepared a document 

that included discussion of “design extension” analysis (WENRA, 2008).3 This analysis was 

intended to achieve the following:

Examine the performance of the plant in specified accidents beyond the 

design basis, including selected severe accidents, in order to minimize as far 

as reasonably practicable radioactive releases harmful to the public and the 

environment in cases of events with very low probability of occurrence. 

The document indicated that:

[B]eyond design basis events shall be selected (based on a combination of 

deterministic and probabilistic assessments as well as engineering judgment) 

and considered in the safety analysis to determine those sequences for which 

reasonable practicable preventive or mitigative measures can be identified and 

implemented, and that realistic assumptions and modified acceptance criteria 

may be used for the analysis of the beyond design basis events. 

3 WENRA is a network of Chief Regulators of European Union countries with nuclear power plants, and 

Switzerland, as well as of other interested European countries, that have been granted observer status. The 

main objectives of WENRA are to develop a common approach to nuclear safety, to provide an independent 

capability to examine nuclear safety in applicant countries, and to be a network of chief nuclear safety 

regulators in Europe exchanging experience and discussing significant safety issues.
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The document also listed a set of events that, at a minimum, needed to be considered, 

including station blackouts. WENRA and other organizations continue to promote the adoption 

of standardized approaches, including the “design extension” category, throughout the 

international nuclear community (IAEA, 2010).

As discussed in several appendices to this report, the inclusion of a design-enhancement4

category in the United States would result in the following framework for design-basis and 

beyond-design-basis events:

•	 Design-basis Events

o normal operation

o anticipated operational occurrences

o design-basis accidents

o design-basis external hazards

•	 Beyond-design-basis Events

o design-enhancement events 

−	 internal events

−	 external hazards

o residual risk scenarios

−	 internal events

−	 external hazards

Design-basis events traditionally have been associated with mechanistic analyses, reliance on 

and protection of safety-related equipment, and the establishment of technical specifications 

and other licensing-related operational controls that have been deemed necessary for adequate 

protection of public health and safety. Beyond-design-basis events have more often included 

the use of best-estimate type analyses, PRAs, and in establishing additional plant protections 

to further reduce risks to the public health and safety and common defense and security 

(e.g., additional protections for station blackout conditions and aircraft impacts).

Two key concepts related to the identification of relevant scenarios, categorization, and 

subsequent design features and operating limits are 1) the threshold to define when a scenario 

needs to be considered within a category and 2) the acceptance criteria to define when a design 

feature or operating limit provides the desired protection from the defined scenario(s). The 

RMTF considered the two-tier structure described by the U.S. Court of Appeals (i.e., adequate 

protection and additional protections), as well as the risk-informed and performance-based 

defense-in-depth concept used in the risk management goal and developed Figure 4.2-1, which 

represents a general regulatory framework for nuclear power reactors.

4 The RMTF has chosen to use the term “design-enhancement” rather than “design extension.”
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4.2-8 | Risk Management Task Force Section 4.2: Nuclear Reactors

As shown in Figure 4-2-1, the RMTF agrees with WENRA, the International Atomic Energy 

Agency (IAEA), and the Fukushima NTTF recommendation and proposes that a designenhancement category be established for beyond-design-basis accidents. The RMTF has 

defined a number of desirable characteristics of the design-enhancement category. These 

characteristics reflect consideration of current agency practices, including those associated with 

cost-beneficial safety enhancements. The desirable characteristics are listed below.

•	 Consistency of regulatory approach. The RMTF recommends that the designenhancement category be established by rulemaking to ensure appropriate regulatory 

controls, including, for example, change control and reporting requirements. Welldefined regulatory requirements would also clarify reliability and operational expectations 

and oversight and significance determination process considerations.

•	 Risk as safety measure. Where practical, PRA should be the tool used to measure the 

safety importance of accident sequences, SSCs, human actions, etc., in this category, 

and the potential safety improvements or relaxations associated with potential plant 

modifications not needed to ensure adequate protection. In the long term, these PRAs 

should be developed on a site-specific basis and be “full scope” and include all modes 

of operations and all initiating events (internal and external). They should be carried out 

to include offsite health consequences and should include uncertainty analyses. Toward 

this end, the approach defined in 10 CFR 50.71(h), which ties current PRA requirements 

to the availability of consensus standards, should be applied.

•	 Focus on performance. Regulations and other regulatory programs established in this 

category should be performance-based, using the guidance provided by the Commission 

in 1999 and discussed in Appendix D. Performance monitoring should include periodic 

reassessments of risk, considering new information from operating experience, research, 

and other sources. Toward this end, the approach defined in 10 CFR 50.71(h), which 

Figure 4-1 Regulatory Framework for Nuclear Power Reactors
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specifies that new reactors will perform a reassessment of risks (using PRAs meeting 

NRC-endorsed codes and standards) every 4 years, should be applied together with 

ad hoc reassessments with the development of new, potentially significant information 

(e.g., new seismic hazard information). Furthermore, a concept similar to the reliability 

assurance program (RAP) for new reactors should be developed to support the 

performance-based approach for SSCs identified as important in scenarios in this 

category. The SSCs included in RAP are identified as being risk-significant using a 

combination of deterministic, probabilistic, and other analysis techniques. Inclusion of 

SSCs in the RAP and the related controls and monitoring are intended to ensure that the 

subject SSCs are designed, constructed, and operated in a manner consistent with the 

supporting technical analyses (including reliability and availability).

•	 Consideration of costs. The Backfit Rule establishes the basis for the NRC’s 

consideration of costs in reactor regulatory decisionmaking. Implementation of 

regulations and other regulatory programs included in the design-enhancement category 

could reflect an “as low as is reasonably achievable” assessment with cost–benefit 

considerations similar to the Backfit Rule and other NRC regulatory analyses included in 

the determination of what is “reasonable.”

•	 Implementation on a site-specific basis. Regulations and other regulatory programs 

included in the design-enhancement category should be implemented, to the maximum 

extent practical, considering site-specific design and operations, as reflected in a sitespecific PRA.

Finding PR-F-2: Requirements for beyond-design-basis accident scenarios 

(e.g., station blackout) were established at different times and in different ways. 

Differences in implementation approaches have reduced the efficiency and 

consistency of the NRC’s regulatory and oversight activities.

Finding PR-F-3: The extent to which licensee activities undertaken as part of 

voluntary industry initiatives can be credited has been a source of contention in the 

Reactor Oversight Process and has reduced the efficiency of that process.

Recommendation PR-R-2: The NRC should establish through rulemaking a designenhancement category of regulatory treatment for beyond-design-basis accidents. 

This category should use risk as a safety measure, be performance-based 

(including the provision for periodic updates), include consideration of costs, and be 

implemented on a site-specific basis.
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4.2-10 | Risk Management Task Force Section 4.2: Nuclear Reactors

The addition of a design-enhancement category and related requirements for identifying 

events and developing measures to keep risks as low as is reasonably achievable or below 

an established level may change the landscape currently defined by design-basis events and 

several selected conditions (e.g., station blackout, ATWS, and aircraft impact for new reactors). 

This change may, in turn, enable changes to the current handling of design-basis events by 

providing a defined regulatory program for items that may not warrant the full requirements 

currently associated with design-basis events but that warrant some regulatory measures 

to manage the risks associated with off-normal scenarios. A review of the existing designbasis events and the overall combination of events within all categories could be performed 

to determine whether some elements of the current design-basis events could be better 

addressed within the design-enhancement category. Such a review and subsequent changes 

to NRC guidance documents could serve as a way to implement Recommendation PR-R-1. 

Any relocations from the design-basis categories to the design-enhancement category would 

likely support a more performance-based way to ensure risk-informed and performance-based 

defense-in-depth. 

External Hazards

NRC’s licensed facilities are subject to a wide variety of safety challenges from naturally 

occurring hazards. The traditional licensing approach for power reactors establishes what were 

perceived as conservative bounds on the magnitude of particular hazards using concepts such 

as design-basis floods and safe shutdown earthquakes. The extent of conservatism varies. 

The design basis for earthquakes might be associated with lower frequencies than, for example, 

external floods. Although new information on the frequencies and magnitudes of these hazards 

has been obtained, this information has not been routinely evaluated and communicated.

As stated above, the NRC’s approach to addressing the risk from naturally occurring hazards 

varies among the hazards. Methods to address earthquakes are the most advanced, perhaps 

because this hazard is judged to be of the highest risk significance. Information on earthquake 

frequencies is periodically updated. Even for this hazard, however, the risk assessment 

expertise within the NRC and the industry is very limited. In addition, studies have shown that 

there are serious limitations in attempting to address seismic hazards or other natural events 

(including related uncertainties) at extremely low frequencies (Johnson and Apostolakis, 2012).

Finding PR-F-4: The processes for establishing the external hazard design bases 

do not use consistent event frequency and magnitude methods.

Finding PR-F-5: New information that would provide the basis for external hazard 

frequency updates is not systematically collected, evaluated, and communicated.

Finding PR-F-6: PRA methods for assessing external hazard risks are available, but 

expertise in performing such studies is very limited. Uncertainty analyses and the 

recognition of the limitations of available scientific knowledge are a key element of 

these methods.
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Recommendation PR-R-3: The NRC should reassess methods used to estimate 

the frequency and magnitude of external hazards and implement a consistent 

process that includes both deterministic and PRA methods. Consideration of the 

risks from beyond-design-basis external hazards should be included in the designenhancement category described in Recommendation PR-R-2.

Recommendation PR-R-4: The NRC should establish a program to systematically 

collect, evaluate, and communicate external hazard information. 

Defense-in-depth

The term defense-in-depth traditionally has been used in the context of power reactor safety. As 

discussed in Appendix C, a number of different descriptions of this concept exist. For this and 

other reasons, the RMTF is recommending the risk-informed and performance-based defensein-depth characterization discussed in Chapters 2 and 3.

Methods for quantitatively assessing risk have been developed for power reactors to a 

much greater degree than for other NRC-regulated activities. Recognizing this, the RMTF 

has expanded upon the Chapter 2 risk-informed and performance-based defense-in-depth 

philosophy to create possible defense-in-depth guidance specific to power reactors. In addition, 

the RMTF has extended this philosophy to considerations of power reactor security. 

The more quantitative RMTF characterization of risk-informed and performance-based defensein-depth for power reactors is shown in italics below:

1. Establish appropriate barriers, controls, and personnel to prevent, contain, and mitigate 

exposure to radioactive material according to the hazard present, the relevant scenarios, 

and the associated uncertainties.

a. Each barrier is designed with sufficient safety margins to maintain its functionality 

for relevant scenarios and account for uncertainties.

b. Systems needed to ensure a barrier’s functionality are designed to ensure 

appropriate reliability for relevant scenarios.

c. Barriers and systems are subject to performance monitoring.

2. Ensure that the risks resulting from the failure of some or all of the established barriers 

and controls, including human error, are maintained acceptably low. 

Within the above construct of risk-informed and performance-based defense-in-depth for 

power reactors, safety margins refer to conservatisms added to ensure that plants and specific 

barriers are designed and constructed so that failures are not expected until key parameters 

well exceed the values assumed in the supporting engineering evaluations. Safety margins 

usually derive from the traditional approach to design-basis accidents, but they can be informed 

by risk assessment techniques. Measures to address the reliability of barriers and supporting 

systems have increasingly been introduced to the regulatory process for power reactors 
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4.2-12 | Risk Management Task Force Section 4.2: Nuclear Reactors

(e.g., maintenance rule and reliability assurance programs for new reactors), but additional 

improvements for establishing and monitoring reliability goals could be developed for some 

equipment considered important to safety (e.g., equipment used in response to the loss of large 

areas due to fires or explosions). The improvements related to reliability have resulted largely 

from risk assessments and their use in programs such as the Reactor Oversight Program. 

An important component of the risk-informed and performance-based defense-in-depth 

approach for power reactors is the consideration of relevant scenarios in which some or all of 

the established barriers and controls are challenged or fail. PRAs can be used to evaluate 

the risk profile of a plant to ensure that the design and operating practices satisfy the NRC’s 

safety goals for nuclear power reactors. In addition, PRAs and other evaluations can be used 

to identify potential cliff-edge effects in which failing a barrier or exceeding a design value 

(e.g., flood level) would lead directly to core damage and a release of radioactive material to 

the environment. Finally, if the risk from some sequences involving failures of defense-in-depth 

protections is significantly lower than what is “acceptably low,” the possibility of relaxing some of 

these protections could be considered.

Finding PR-F-7: The availability and broad-scale use of quantitative risk assessment 

methods (PRA) for power reactors provide an opportunity for a more quantitative 

characterization of defense-in-depth. 

Recommendation PR-R-5: The NRC should apply the risk-informed and 

performance-based defense-in-depth concept to power reactors in a more 

quantitative manner.

Security

The NRC’s security requirements are established to protect the same radioactive hazards 

as safety requirements. However, the regulatory approach used is somewhat different.5 In 

discussions with NRC staff, the RMTF found that differences in language and methods exist 

that reduce the efficiency of the NRC’s interactions with licensees. The proposed RMTF 

characterization of risk-informed and performance-based defense-in-depth, applied to security 

regulatory activities, could help to remedy this inefficiency.

With respect to language, the NRC security staff endeavors to blend the language of the 

traditional security concepts with the agency’s safety concepts, including defense-in-depth. 

While this approach has been used in recent licensing reviews, language differences may 

reduce efficiency.

With respect to methods, there are similarities between risk assessments and security 

vulnerability assessments. The latter must deal with a difficult challenge—estimating the 

5 The traditional approach taken for security is somewhat analogous to the deterministic approach to external 

hazards such as flooding. In the case of security, the equivalent to a maximum water level to protect against 

is the “design-basis threat,” which defines the number and capabilities of the attacking force.
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frequency of the threat (the “initiating event,” in safety terms) and the fact that the threat 

changes continuously and may actively seek to identify and exploit weaknesses in the 

established barriers. 

Risk assessment methods may provide opportunities for addressing security-related 

vulnerabilities, as early risk studies identified important accident scenarios (e.g., interfacing 

systems loss-of-coolant accidents) not fully captured by consideration of design-basis events 

and accidents.

In the last decade, considerable research has been performed on estimating security 

risks, much of it sponsored by the U.S. Department of Homeland Security. Tools have 

been developed such as ITRA (Integrated Chemical, Biological, Radiological, and Nuclear 

Terrorism Risk Assessment Threat Estimation), a program aimed at providing an integrated 

quantitative assessment of the relative risk associated with chemical, biological, radiological, 

and nuclear terrorism to the homeland, and RAPID (Risk Assessment Process for Informed 

Decisionmaking), a program aimed at developing a strategic-level process to gauge future risks 

across the full range of DHS responsibilities” (DHS, 2010). Some of these methods focus on 

the display of information to decisionmakers, including “break-even analysis,” a variant of cost–

benefit analysis that estimates the threshold value at which a policy alternative’s costs equal its 

benefits (DHS, 2010). While much remains to be accomplished in assessing security risks, the 

NRC should actively review and implement methods improvements for its own purposes and for 

ensuring that agency and licensee resources are focused on the most important issues.

Finding PR-F-8: Vulnerability assessments performed to assess security have 

important similarities in scope (e.g., facility equipment and radioactive hazards 

considered) and methods to risk assessments. 

Finding PR-F-9: Differences in regulatory language and approaches between 

power reactor security and safety regulation may have reduced the efficiency and 

effectiveness of the NRC’s work.

Finding PR-F-10: In the past decade, considerable research has been performed 

on estimating security risks, much of it sponsored by the U.S. Department of 

Homeland Security.

Recommendation PR-R-6: The NRC should develop and implement guidance for 

use in its security regulatory activities that uses a common language with safety 

activities and harmonizes methods with risk assessment and the proposed riskinformed and performance-based defense-in-depth framework.
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4.2-14 | Risk Management Task Force Section 4.2: Nuclear Reactors

4.2.1.1 Implementation Options—Operating Power Reactors

Background

The NRC’s current policy on the use of risk information (NRC, 1995) has been implemented 

to a greater extent for operating power reactors than for other facilities or devices regulated by 

the agency. Guidance on backfitting requirements for these reactors (NRC, 2004), allowing 

changes to their licensing-basis (NRC, 2011), and defining acceptable levels of risk from 

their operation (NRC, 1986), combined with focused attention on establishing sophisticated, 

standardized, quantitative risk assessment methods (ASME/ANS, 2009), has created an 

environment that should permit a more timely implementation of the risk management 

framework described in Chapters 2 and 3, including the risk-informed and performance-based 

defense-in-depth concept.

In practice, the RMTF recognizes that operating reactors are, for the most part, well into the 

operations phase of their life cycle, meaning that their fundamental design was established 

many years ago and their licensing-basis occasionally has been updated to include the 

“patchwork” of additional requirements, such as station blackout and anticipated transients 

without scram.

With this background, the RMTF has examined its power reactor findings and recommendations 

and has developed several alternatives for implementation for operating reactors. These 

alternatives are discussed further in Appendix H and within the specific recommendations 

that follow.

Design-basis Accident Licensing Approach

As discussed above, the RMTF has found that the continued use of design-basis accidents 

established more than 30 years ago runs counter to one aspect of the NRC’s Principles of Good 

Regulation—that regulations should be based on the best available knowledge from research 

and operational experience. However, this practice is consistent with another aspect of the 

NRC’s principles—the value of stability in nuclear operational and planning processes. For 

operating reactors, the RMTF concludes that stability currently outweighs the potential value of 

changes to the design-basis accidents. However, should a new set of design-basis accidents 

be established for new reactor designs or an approach is developed by licensees or owners 

groups, the NRC staff should be amenable to its use by operating reactors at some future time.

Recommendation OR-R-1: For operating reactors, the establishment of the designenhancement category can be followed by a review of design-basis events and 

accidents and related revisions to anticipated operational occurrences and DBAs 

to integrate insights from operating history and more modern methods. The NRC 

need not impose such a requirement, but it should be amenable to related industry 

initiatives should they pursue revisions to design-basis events based on the 

introduction of the design-enhancement category. 
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Beyond-design-basis Accidents

In Appendix H, the RMTF defined a number of alternatives that could be used to implement 

the creation of the design-enhancement category. These alternatives included 1) the NRC 

defining additional specific events or conditions (including acceptance criteria) to be addressed 

for power reactors, 2) the NRC requiring licensees to perform risk assessments to identify 

design-enhancement events exceeding a defined threshold and reducing risks to levels as 

low as is reasonably achievable (ALARA), and 3) the NRC requiring licensees to perform 

risk assessments to identify design-enhancement events exceeding a defined threshold and 

reducing risks to levels below defined acceptance risk criteria. The RMTF does not recommend 

any of these alternatives, but provides them to assist in staff implementation. 

Recommendation OR-R-2: For operating reactors, the RMTF recommends that 

the NRC should establish through rulemaking a design-enhancement category of 

regulatory treatment for beyond-design-basis accidents.

External Hazards

As discussed above, the RMTF has found that the NRC’s consideration of external hazards 

varies in the establishment of the design basis and the assessment of their risk. Although the 

Fukushima accident may not be directly relevant to all U.S. power reactors, it did highlight that 

the potential safety significance of external hazards needs to be addressed more systematically 

for U.S. reactors.

Recommendation OR-R-3: The NRC should reassess the methods used to estimate 

the frequency and magnitude of external hazards and implement a consistent 

process that includes both deterministic and PRA methods. For operating reactors, 

the RMTF recommends that the design-enhancement category rulemaking include 

consideration of external hazards.

Recommendation OR-R-4: For operating reactors, the RMTF recommends that 

the NRC develop and implement guidance for the collection and dissemination of 

external hazard information.

Defense-in-depth

As discussed above, operating power reactors have been the subject of sophisticated risk 

assessment analyses and currently have an infrastructure of methods and a large record 

of experience (both positive and negative). As such, an opportunity exists to expand the 

characterization of risk-informed and performance-based defense-in-depth to be more 

quantitative. This expansion could build upon and make use of existing performance monitoring 

approaches, such as the maintenance rule, reliability assurance programs, and the ROP’s 

Mitigating System Performance Indicator (MSPI).
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Recommendation OR-R-5: The NRC should apply the risk-informed and 

performance-based defense-in-depth concept to power reactors in a more 

quantitative manner. For operating reactors, the RMTF recommends that this 

recommendation be implemented in the form of guidance to the NRC staff and in 

future requirements established for operating reactor licensees.

Security

As discussed above, the efficiency of the NRC’s activities is reduced by differences in 

terminology and methods between security and safety assessments. Advancements being 

made by other organizations (DHS) may help remedy this inefficiency. Use of the RMTF 

characterization of risk-informed and performance-based defense-in-depth in both safety and 

security activities also could help. 

Recommendation OR-R-6: For operating reactors, the RMTF recommends that 

guidance be developed and implemented to better harmonize terminology and 

methods used for reactor safety and security.

4.2.1.2 Implementation Options—New Power Reactors

Background

The consideration of risk insights for new power reactors has built upon the requirements and 

initiatives associated with operating reactors. New reactors, for example, are required to have 

and update PRAs. Insights from these PRAs are used for activities such as the improvement 

of plant designs and the establishment of reliability assurance programs. New plants also 

incorporate severe accident mitigation features to provide some protection against scenarios 

involving significant core damage. The RMTF offers the following recommendations for new 

reactors, including both large and small light-water reactor designs.

Design-Basis Accident Licensing Approach

As with operating power reactors, the continued use of traditional DBAs requires consideration 

of both the use of the best available knowledge and regulatory stability. Given that the 

designs and regulatory reviews for new large light-water reactor designs are completed or 

well underway, the RMTF concludes that they should be treated similar to operating reactors. 

However, the NRC should be amenable to a more risk-informed approach to DBAs if it is 

proposed for a future design of a large light-water reactor. The iPWR designs being developed 

are currently at the conceptual stage and a certain degree of adjustment to DBAs will be 

necessary as a result of expected design features (e.g., no large reactor coolant piping). 

Designers and potential licensees want to move relatively quickly to complete the designs and 

begin the design and licensing reviews. Their current plans and schedules would be somewhat 

complicated by a significant change to the DBA requirements. The NRC should, however, 
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be amenable to, and even promote, a more risk-informed and performance-based approach 

for iPWR DBAs, to the degree it could be done without unduly increasing the schedules or 

resources needed for design development and review. 

Recommendation NR-R-1: For new reactors, the RMTF recommends that the NRC 

be amenable to and promote, where practical, the adoption of more risk-informed 

and performance-based approaches for the selection of more relevant scenarios for 

design-basis events. Changes pursued for operating reactors 

(OR-R-1) should also consider applicability to new reactors.

Beyond-design-basis Accidents

There has been some consideration of beyond-design-basis accidents in the licensing of new 

reactors in terms of assessing PRA results, defining the regulatory treatment of non-safety 

systems for passive reactor designs, and establishing reliability assurance programs. The 

treatment of established events, such as station blackout and anticipated transients without 

scram, are handled basically the same way as they are for operating reactors, except where 

actual design differences warrant specific treatment (e.g., reduced reliance on alternating 

current power systems by the passive reactor designs). The similarities between new and 

operating reactor treatment of beyond-design-basis accidents result in a 

similar recommendation:

Recommendation NR-R-2: Apply Recommendation PR-R-2 (design-enhancement 

category) to new reactors.

It should be recognized, however, that differences in plant designs and risk profiles are already 

reflected in the licensing and oversight processes for new reactors. For example, vendors and 

applicants have identified non-safety-related equipment to be included in reliability assurance 

programs based, in part, on PRA insights. Many new reactor designs are also subject to review 

by regulators in other countries, which may have defined design-enhancement requirements. 

For these reasons, as well as the existing requirements for new plant designs and facilities to 

have PRAs, the RMTF foresees easier implementation of this recommendation for new plants 

than might be anticipated for some operating plants. 

External Hazards

The licensing of new reactors has benefited from updates to the guidance related to specific 

external hazards. However, the same inconsistency in handling various external hazards that 

was discussed for operating reactors is also applicable to new reactors. This leads the RMTF to 

suggest the same recommendations for new reactors as those proposed for operating reactors.
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4.2-18 | Risk Management Task Force Section 4.2: Nuclear Reactors

Recommendation NR-R-3: Apply Recommendation PR-R-3 (include external events 

in design-enhancement category) to new reactors.

Recommendation NR-R-4: Apply Recommendation PR-R-4 (periodically evaluate 

new information regarding external hazards) to new reactors.

Defense-in-depth

The new reactor designs have benefited from various studies and discussions of defense-indepth. One example is efforts to incorporate design features to address the Commission’s 

guidance in the Advanced Reactor Policy Statement (NRC, 2008) and a general reduction in the 

estimated risks associated with new reactor designs compared to operating reactors. Although 

the new reactor designs are somewhat improved in considering the features of risk-informed 

and performance-based defense-in-depth, adoption of the relevant recommendation from 

operating reactors could provide further improvements in areas such as the identification and 

resolution of cliff-edge effects.

Recommendation NR-R-5: Apply Recommendation PR-R-5 (issue guidance to 

adopt risk-informed and performance-based defense-in-depth) to new reactors.

Security

New reactor designs and reviews have benefited from more recent requirements and guidance 

related to security. For example, although not incorporated into NRC regulations, the staff 

prepared guidance related to security assessments for new reactor designs and combined 

licensed applications that was used by both applicants and NRC reviewers. Applicants were 

able to incorporate features into reactor designs or site layouts to help address security 

concerns as well as new requirements to address aircraft impact and the loss of large areas 

due to fires and explosions. These activities have positioned new reactors to more easily 

incorporate risk insights into the security program. However, the RMTF recognizes that the 

adoption of common terminology and approaches for both safety and security areas would 

be beneficial.

Recommendation NR-R-6: Apply Recommendation PR-R-6 (develop guidance and 

consistent approach between safety and security) to new reactors.
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4.2.1.3 Implementation Options—Generation IV Power Reactors

Background

The primary distinctions between Generation IV and operating and new power reactors are: 

(1) they introduce significant changes in reactor technology, such as different fuel forms and 

reactor coolants, and (2) the designs and supporting analyses are not yet finalized. A high 

degree of uncertainty remains as to if, and, or when the NRC would be asked to review a 

Generation IV reactor design and what technology ultimately will be pursued. The NGNP 

program at the U.S. Department of Energy (DOE) continues to work on technical research 

and development and a licensing plan for a high-temperature gas-cooled reactor. DOE and 

private companies are also working on liquid-metal-cooled fast reactor designs and hightemperature fluoride salt-cooled designs. The NGNP activities and previous efforts related to 

Generation IV designs (e.g., PRISM and sodium advanced fast reactor (SAFR) preliminary 

designs, and modular high-temperature gas-cooled reactor (MHTGR) preliminary design) have 

incorporated risk assessments into the design and licensing processes to a larger degree 

than has been done for light-water reactors. This is due in large part to a recognition that 

existing NRC technical requirements in 10 CFR Part 50 were developed specifically for lightwater reactor designs and the identification and resolution of issues would benefit from risk 

assessment techniques. For its part, the NRC developed and issued NUREG-1860 as part of 

its activities related to a potential “technology neutral” regulatory framework for future reactor 

designs (NRC, 2007). The technology neutral concept is to develop regulatory requirements 

that could be applicable to any advanced reactor design (e.g., gas-cooled reactors, liquid-metal 

cooled fast reactors). An alternative approach would be to develop regulations specific to each 

reactor technology. For either approach, the RMTF offers the following recommendations for 

Generation IV power reactors.

Design-Basis Accident Licensing Approach

The technology differences between current light-water reactor designs and Generation IV 

reactors require a fundamental reassessment of DBAs related to plant malfunctions. While 

some of the critical safety functions such as controlling reactivity and maintaining core cooling 

are similar, the design features and operating practices used to accomplish them will change. In 

general, Generation IV reactors have inherent features, such as higher system heat capacities, 

which will provide slower rates of temperature rises within the reactor core than what exists 

for light-water reactors. The Generation IV reactor designs also take advantage of passive 

safety features to reduce the reliance on active equipment, electrical power, and dedicated heat 

sinks. Differences in the fuel form and reactor coolant also have required alternate acceptance 

criteria for design-basis events since existing safety limits and regulatory acceptance criteria 

are defined specifically for light-water reactors with uranium dioxide fuel pellets within zirconium 

alloy cladding. 

The NRC staff’s interactions with NGNP and fast-reactor communities indicate that DBAs are 

under consideration for Generation IV technologies. These DBAs would share some of the 

attributes of current DBAs, such as reliance on safety-related equipment only. Other aspects 

of current practices, including the analysis of anticipated operational occurrences and the use 

of the single failure criterion, are being evaluated by the Generation IV reactor designers. The 

Generation IV activities have not proposed to go as far as NUREG-1860, “Feasibility Study for 

a Risk-Informed and Performance-Based Regulatory Structure for Future Plant Licensing,” in 

terms of combining PRA and traditional methods to define special treatment requirements for 
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4.2-20 | Risk Management Task Force Section 4.2: Nuclear Reactors

specific SSCs. Based on information from NRC staff interactions with staff of Generation IV 

programs and general stakeholder feedback that DBAs are useful in establishing functional 

requirements and safety margins, the RMTF offers the following recommendation:

Recommendation GIV-R-1: For Generation IV reactors, the RMTF recommends 

that the concept of design-basis accidents be maintained, but the NRC should 

be amenable to and promote, where practical, the adoption of more risk-informed 

approaches for the selection of relevant scenarios (e.g., alternatives to the single 

failure criterion) for design-basis accidents.

Beyond design-basis-Accidents

The NRC’s experiences related to Generation IV reactor designs have included increased 

consideration of beyond-design-basis accidents. This increased focus is because future 

reactors have defined new events and placed them into categories using similar criteria as 

provided by NRC regulations and industry codes and standards developed for 

light-water reactors. 

However, light-water reactor assessments used more operating experience and engineering 

judgment, since the need to define design-basis events predates the wide use of PRAs. 

Given that the Generation IV design and licensing activities began after the development of 

risk assessment techniques, PRAs were used to support plant design activities and licensing 

strategies for each of the subject Generation IV designs. The evaluation of beyond-designbasis events was also included in the licensing strategies for Generation IV designs in support 

of assessing plant design features and operational programs, such as emergency planning. As 

with design-basis events, alternative acceptance criteria usually are defined for 

Generation IV reactors, since the common surrogate measures for the NRC’s safety goal (core 

damage frequency and large early release frequency) were developed for light-water reactor 

designs. NGNP and NUREG-1860 proposed acceptance criteria of dose at the site boundary, 

which is more directly related to the health objectives in the NRC safety goal. 

NGNP and other Generation IV reactor initiatives have largely adopted the identification and 

disposition of what has been referred to as design-enhancement events within this report. 

Generation IV activities largely have been an international effort; therefore, this concept, as well 

as other features of the IAEA standards and guides, has been incorporated into the design and 

licensing programs (although sometimes using different terminology). Establishing the proposed 

Risk Management Regulatory Framework in a timely way for Generation IV reactors will support 

the design and pre-application interaction between designers and the NRC. Given the general 

alignment of Generation IV activities with the recommendations in this report, the RMTF 

recommends the following:

Recommendation GIV-R-2: Apply Recommendation PR-R-2 (design-enhancement 

category) to Generation IV reactors.
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External Hazards

The licensing of new reactors has benefited from updates to guidance related to specific 

external hazards, and some external events are being revisited in response to the Fukushima 

nuclear accident. However, the same inconsistency in handling various external hazards that 

was discussed for operating reactors and new reactors will, without NRC actions, probably 

become applicable to Generation IV reactors. This leads the RMTF to suggest the same 

recommendations for Generation IV reactors as those proposed for operating and new reactors.

Recommendation GIV-R-3: Apply Recommendation PR-R-3 (include external 

events in design-enhancement category) to Generation IV reactors.

Recommendation GIV-R-4: Apply Recommendation PR-R-4 (periodically evaluate 

new information regarding external hazards) to Generation IV reactors.

Defense-in-depth

Generation IV reactor designs and especially recent activities related to NGNP have benefited 

from various studies and discussions of defense-in-depth. Examples include specific white 

papers on the incorporation of defense-in-depth into the design and licensing approaches 

for NGNP and the increased use of IAEA standards and guides within specific Generation IV 

reactor technology groups. Although Generation IV reactor designs have largely embraced 

the previous discussions of risk-informed and performance-based defense-in-depth, the RMTF 

recommends formalizing the definition and overall Risk Management Regulatory Framework to 

support future interactions and ensure the identification and resolution of cliff-edge effects for 

advanced reactors.

Recommendation GIV-R-5: Apply Recommendation PR-R-5 (issue guidance to 

adopt risk-informed and performance-based defense-in-depth) to Generation 

IV reactors.

Security

Generation IV reactor designs would be expected to benefit from NRC guidance related to 

security assessments for new reactor designs and combined licensed applications. Designers 

are considering how to better incorporate security features into reactor designs or site layouts 

to help address requirements, including analysis of aircraft impacts and the loss of large areas 

due to explosions or fires. As previously discussed for operating and new reactors, the RMTF 

recognizes that the adoption of common terminology and approaches for both safety and 

security areas would be beneficial for Generation IV reactor programs.

Recommendation GIV-R-6: Apply Recommendation PR-R-6 (develop guidance and 

consistent approach between safety and security) to Generation IV reactors.
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4.2-22 | Risk Management Task Force Section 4.2: Nuclear Reactors

4.2.2 Nonpower Reactors

4.2.2a Background

In addition to the 104 commercial nuclear power plants licensed to operate in the United States, 

the NRC is also responsible for licensing and oversight of the nation’s 42 Nonpower reactors 

(NPRs). NPRs, also called research and test reactors, are nuclear reactors primarily used for 

research, training, and isotope production. They contribute to almost every field of science, 

including physics, chemistry, biology, medicine, geology, archeology, and environmental 

sciences. Of the 42 NPRs, 31 have operating licenses and 11 have terminated their operating 

licenses and are either awaiting, or actively involved in, facility decommissioning. 

The NRC’s authority to license and regulate NPRs is provided in Sections 103 and 104 of the 

Atomic Energy Act (the Act) (NRC, 2011b). Section 103 of the Act pertains to the licensing 

of “industrial or commercial” reactors that may or may not be an NPR. Section 104 of the 

Act pertains to the licensing of NPRs for the purpose of “medical therapy and research and 

development.” All NPRs currently licensed by the NRC are licensed under Section 104 of the 

Act. Unique to this authority are the provisions contained in paragraphs 104b and 104c of 

the Act. These paragraphs require the Commission to impose the minimum amount of such 

regulation and terms of license that will permit the Commission to fulfill its obligation under this 

Act to promote the common defense and security and to protect the health and safety of the 

public with the intent to permit the conduct of widespread and diverse research 

and development. 

Nonpower reactors in the United States have been licensed using 10 CFR Part 50. 

Implementation of 10 CFR Part 50, as it applies to NPRs, has been achieved using only 

deterministic methods and acceptance criteria. Licensing decisions allowing construction and 

operation of NPRs have focused on assurance that worker and public doses are maintained 

within the limits contained in 10 CFR Part 20 for research reactors and 10 CFR Part 100, 

“Reactor Site Criteria,” for test reactors. As was the case with power reactors, a set of licensingbasis events was established that was intended to ensure conservatism in design and protection 

from a wide spectrum of postulated events, up to and including design-basis accidents. Those 

accidents are highly stylized and do not consider multiple failures of safety systems. Qualitative 

approaches for ensuring reliable safety systems, such as the single failure criterion, were 

implemented. Testing plans and operational limits were established in technical specifications 

to ensure that if called upon safety systems would perform. 

The licensing of NPRs includes an analysis of a maximum hypothetical accident (MHA). 

Analysis of the MHA is necessary because many NPRs are designed and operated so that an 

accident involving a radioactive release is not credible. The MHA assumes an incredible failure 

that results in consequences that bound all credible DBA consequences. The MHA assumes a 

radioactive release with radiological consequences that exceed those of any credible accident. 

Because the MHA is not expected to occur, only the potential consequences are analyzed and 

not the initiating event and scenario details.

4.2.2b Findings and Recommendations for Nonpower Reactors

The concept of defense-in-depth was developed early in NPR licensing and has been 

applied to compensate for the recognized limited knowledge of nuclear reactor operations 

and the consequences of potential accidents. A comprehensive defense-in-depth approach 
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forms the foundation for all licensed NPRs. Nonpower reactor license application guidance 

(NUREG-1537, Part 1) (NRC, 1996a) specifically states that license applicants should provide 

a discussion in the safety analysis report of the multiple design features that comprise the 

facility’s defense-in-depth. This guidance specifically identifies discussions of the restricted area 

surrounding the reactor to exclude and protect the public, confinement or containment designs 

for the control of radioactive releases, limitations on operation that will ensure thermal-hydraulic 

parameters remain well below the designed capabilities of the fuel and cladding, diversity and 

redundancy of instrumentation and control systems, and active or passive engineered safety 

features included to mitigate the consequences of accidents. Both experience with NPR 

operations and the knowledge base of potential consequences have grown over time. However, 

even with increased experience and knowledge, gaps in knowledge and uncertainties remain. 

These gaps and uncertainties continue to be effectively addressed by a defense-in-depth 

approach. As such, the concept of defense-in-depth remains relevant with NPRs.

Finding NPR-F-1: The concept of defense-in-depth for NPRs remains relevant. 

Knowledge gaps and uncertainties continue to be effectively addressed by the 

defense-in-depth approach.

The applications of a defense-in-depth approach at both power reactors and NPRs share the 

common goal of preventing the radiological release to the environment. However, they differ 

significantly on potential accident consequences. These differences result from significantly 

different operating characteristics, including maximum power level, duration of operation, fission 

product inventories, and accumulation of spent fuel. To put these differences into perspective, 

consider the following: 1) the maximum licensed power level of the power reactor can be two to 

nine orders of magnitude greater than the Nonpower reactor, 2) the duration of operations of the 

power reactor is nearly continuous versus the periodic operation of the majority of NPRs, and 

3) the onsite accumulation of significant quantities of spent fuel (20 years’ to 40 years’ worth) 

in fuel pools at power reactors as compared to little or no spent fuel at NPRs. Each of these 

factors contributes to a significantly smaller radioactive inventory and accident source term at 

the NPR when compared to the power reactor.

Emergency planning considerations for NPRs and power reactors are similar. The significantly 

smaller accident source term of the NPRs results in a reduction of potential consequences and 

emergency planning zones that are typically bounded well within ownercontrolled areas. The 

boundary of the room or building in which the NPR is housed often comprises the emergency 

planning zone. Additionally, of the four emergency classes defined in Appendix E, “Emergency 

Planning and Preparedness for Production and Utilization Facilities,” to 10 CFR Part 50, the 

classification of general emergency is not included for any of the currently licensed NPRs since 

the MHA results for each NPR do not demonstrate a significant radiological impact at substantial 

distances from the reactor.
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Finding NPR-F-2: The analysis of design basis and the maximum hypothetical 

accidents based on conservative design limits, acceptance criteria, safety margins, 

and assumptions in conjunction with the application of a defense-in-depth 

philosophy continues to be a sound but highly conservative licensing approach to 

ensuring adequate safety of NPRs.

While significant conservatism has contributed to the demonstrated safety of NPRs, it is 

reasonable to assume that conservative design beyond some point does not yield an equivalent 

safety benefit. The imposition of excessively conservative NPR design and licensing criteria 

could be viewed as inconsistent with Section 104c of the Act. As presented previously, 

Section 104c requires the Commission to impose the minimum amount of such regulation and 

terms of license that will permit the agency to fulfill its obligation under this Act to promote the 

common defense and security and to protect the health and safety of the public with the intent of 

permitting the conduct of widespread and diverse research and development. The imposition of 

more stringent design requirements once an adequate level of safety or an acceptable level of 

risk has been achieved could be viewed as exceeding the requirements of the Act. 

Recommendation NPR-R-1: The proposed defense-in-depth framework should be 

applied to the NPR licensing process to ensure that the current amount of defensein-depth is appropriate given the relatively small radioactive hazard. This application 

should include safety and security licensing matters.

Implementation of a Risk Management Approach 

The assessment of risk at NPRs has been qualitative and based on conservative deterministic 

assumptions, traditional engineering analyses, and operational experience. To date, the 

operators of NRC-licensed NPRs have not used modern risk assessment methods in support of 

licensing activities. No reactor safety goals and objectives that parallel the Commission’s Safety 

Goal Policy Statement (NRC, 1986) have been developed for NPRs. 

The Department of Energy has sponsored PRAs on some of its NPR facilities. Eight 

international NPRs are known to have Level I PRA models. No Level 2 and 3 models are 

known to exist for NPRs.

The development of NPR risk assessment through the use of modern risk assessment 

methods and techniques may provide additional safety insight and benefit, especially for NPRs 

with maximum licensed power levels greater than 2 megawatts. However, the potential for 

additional safety insight and benefit will likely diminish rapidly as the facility power decreases. 

The greatest benefit of the development of a modern risk assessment would be a qualitative 

or quantitative measure of risk that is based on realistic information. Such information could 

provide a better understanding of the importance of facility systems and components to the 

overall risk presented by the facility, support the comparison of facility risk to an agency 

common risk management goal, and define the necessary barriers and controls to establish 

an adequate level of safety for the protection of individuals from radioactive materials. As 
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discussed in Appendix B, the selection of appropriate analyses techniques depends on the 

radiological hazards, relevant scenarios, and associated uncertainties. The use or development 

of analytical tools less complicated than the detailed PRAs used for operating reactors may be 

appropriate given the simpler designs and lesser amounts of radioactive materials associated 

with NPRs. Recognizing the small staff sizes and operating budgets of NPRs, even less 

complicated risk assessments may require external financial support.

Excessive conservatism or the imposition of requirements that do not result in a proportional 

benefit to safety or only add minimally to safety beyond an already existing adequate level of 

safety can be contrary to an efficient and effective regulatory framework. The combination 

of the conservatisms introduced through the consideration of an incredible accident scenario 

(e.g., the MHA), the use of restrictive 10 CFR Part 20 standards for evaluation of the effects 

of a postulated accident at research reactors, and large safety margins associated with 

the traditional engineering analyses, may result in an overly conservative NPR regulatory 

framework. If that is the case, the expenditure of resources in the execution of licensing 

activities and oversight may not be providing a corresponding safety or security benefit. The 

performance of an NPR risk assessment using modern methods informed by the best and most 

realistic knowledge available would be valuable in identifying areas of risk previously unknown 

and not adequately addressed by the NPR licensing and oversight processes, or areas where 

licensing and oversight efforts are directed toward areas with little or no importance to facility 

risk. In either case, the risk assessment would be useful in providing information to a formal risk 

management process for a decision on appropriate changes to regulatory requirements.

Finding NPR-F-3: The application of modern risk assessment methods at NPRs 

could provide valuable insights into accident scenarios not previously identified by 

the earlier deterministic safety assessment and could be valuable in focusing the 

application of licensing and oversight resources on areas of risk importance. Risk 

assessment insights, in conjunction with a formal risk management decisionmaking 

process, could significantly contribute to the development of a more efficient and 

effective NPR regulatory framework. NPR PRA models developed by others 

could be used as a starting point for facility-specific PRA models at NRC-licensed 

NPRs. Even with this background, however, funding such assessments could be 

problematic for NPRs.

Recommendation NPR-R-2: The NRC should evaluate the utility of performing a 

pilot risk assessment, including consideration of external hazards, using modern 

risk assessment methods at an NPR. This evaluation would assess the value of 

the risk insights gained from the risk assessment on the basis of possible safety 

enhancements and possible contributions to a more efficient and effective riskinformed and performance-based regulatory framework for NPRs.
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4.2-26 | Risk Management Task Force Section 4.2: Nuclear Reactors

External Events

Nonpower reactors licensed by the NRC are required to demonstrate, in their design basis, 

reasonable assurance that external events would not preclude safe operation and shutdown 

of the reactor. They must also demonstrate that provisions are included to mitigate or prevent 

an uncontrolled release of radioactive material and the consequences of an external event are 

considered or bounded by analyzed accidents. The traditional licensing approach for NPRs 

considers documented historical averages and extremes, credible frequencies, and predictive 

potential for the specific external events. At a minimum, each research reactor facility is 

required to meet the local building codes for the specific type of event and test reactors are 

required to meet the requirements of 10 CFR Part 100.

The traditional NPR licensing approach attempted to establish conservative bounds on the 

magnitude of particular naturally occurring hazards using regional features. This approach 

is similar to that used for power reactors, and shares the same limitations. As such, power 

reactor recommendations PR-R-3 and PR-R-4 that call for a reassessment of the methods 

used to estimate the frequencies and magnitudes of external hazards, the implementation of 

a consistent process that includes both deterministic and PRA methods, the establishment of 

a program to systematically collect external hazard information, and the periodic evaluation of 

information for site-specific implications are also valid recommendations in the NPR case.

Finding NPR-F-4: The traditional NPR licensing approach shares the same 

limitations as the power reactor approach for methods used to estimate frequencies 

and magnitudes of external events.

Recommendation NPR-R-3: NPRs should be considered, to the extent 

practical, in the implementation of power reactor Recommendation PR-R-3 and 

Recommendation PR-R-4.

Security

Nonpower reactor security requirements are established to protect the same radioactive 

hazards as safety requirements. The physical protection measures at NPRs are established 

to address the current threat. As the threat changes significantly, appropriate changes are 

made to security requirements and physical protection measures. Security at NPRs uses a 

graded approach with a focus on the prevention of theft and diversion of materials for NPRs 

with a maximum licensed power level less than 2 megawatts and prevention of both radiological 

sabotage and theft and diversion of materials for NPRs 2 megawatts and greater. Current 

NPR security regulations are not risk-informed. The development of risk-informed security 

regulations or guidance would enhance efficiency by focusing resources on areas 

of importance.
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Finding NPR-F-5: NPR security requirements are not risk-informed beyond the use 

of a graded approach based on NPR power levels.

Finding NPR-F-6: The development of risk-informed security regulations or 

guidance would enhance efficiency by focusing resources on identified areas 

important to facility security. 

Recommendation NPR-R-4: If the NRC decides to develop and implement a riskinformed and performance-based defense-in-depth regulatory framework to ensure 

the safety of NPRs, then the agency should also develop guidance for use in its 

NPR security regulatory activities that uses a common language with safety activities 

and harmonizes methods with the risk-informed and performance-based defense-indepth framework.

4.2.3 References

(ASME/ANS, 2009) American Society of Mechanical Engineers/ American Nuclear 

Society, ASME/ANS RA-Sa-2009, “Standard for Level 1/Large 

Early Release Frequency Probabilistic Risk Assessment for 

Nuclear Power Plant Applications,” Addendum A to RA-S-2008, 

ASME, New York, NY; ANS, LaGrange Park, IL.

(DHS, 2010) U.S. Department of Homeland Security, “DHS Risk Lexicon, 

2010 Edition,” September 2010.

(IAEA, 2010) International Atomic Energy Agency, “Draft—Safety of Nuclear 

of Power Plants: Design,” Draft Safety Guide DS-414, 

September 2010.

(Johnson and B.C. Johnson and G.E. Apostolakis, “Seismic Risk Evaluation

Apostolakis, 2012) within the Technology Neutral Framework,” Nuclear Engineering 

 and Design, 242 (2012) 341– 352.

(NRC, 1975) U.S. Nuclear Regulatory Commission, “Reactor Safety Study— 

An Assessment of Accident Risks in U.S. Commercial Nuclear 

Power Plants,” NUREG-75-014 (WASH-1400), October 1975. 

(NRC, 1985) U.S. Nuclear Regulatory Commission, “Policy Statement on 

Severe Reactor Accidents Regarding Future Designs and Existing 

Plants,” August 1985, published in the Federal Register on 

August 8, 1985 (50 FR 32138)
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(NRC, 1986) U.S. Nuclear Regulatory Commission, “Safety Goals for the 

Operation of Nuclear Power Plants,” Final Policy Statement, 

published in the Federal Register on August 21, 1986 

(51 FR 30028). 

(NRC, 1988) U.S. Nuclear Regulatory Commission, “Loss of All Alternating 

Current Power,” Final Rulemaking, published in the Federal 

Register on June 21, 1988 (53 FR 23215).

(NRC, 1991) U.S. Nuclear Regulatory Commission, “Monitoring the 

Effectiveness of Maintenance at Nuclear Power Plants,” Final 

Rule, published in the Federal Register on July 10, 1991 

(56 FR 31306).

(NRC, 1995) U.S. Nuclear Regulatory Commission, “Use of Probabilisitic 

Risk Assessment Methods in Nuclear Regulatory Activities,” 

Final Policy Statement, published in the Federal Register on 

August 16, 1995 (60 FR 42622).

(NRC, 1996) U.S. Nuclear Regulatory Commission, “Status of the Integration 

Plan for Closure of Severe Accident Issues and the Status of 

Severe Accident Research,” Commission Paper SECY–96–088, 

April 29, 1996. 

(NRC, 1996a) U.S. Nuclear Regulatory Commission, “Guidelines for Preparing 

and Reviewing Applications for the Licensing of Nonpower 

Reactors: Format and Content,” NUREG-1537, Part 1, 

February 1996, Agencywide Documents Access and Management 

System (ADAMS) Accession No. ML042430055.

(NRC, 1996b) U.S. Nuclear Regulatory Commission, “Guidelines for Preparing 

and Reviewing Applications for the Licensing of Nonpower 

Reactors: Format and Content,” NUREG-1537, Part 2, 

February 1996, ADAMS Accession No. ML042430048.

(NRC, 2004) U.S. Nuclear Regulatory Commission, “Regulatory Analysis 

Guidelines of the U.S. Nuclear Regulatory Commission,” NUREG/

BR-0058, Revision 4, September 2004, ADAMS Accession 

No. ML042820192.

(NRC, 2007) U.S. Nuclear Regulatory Commission, “Feasibility Study for a 

Risk-Informed and Performance-Based Regulatory Structure for 

Future Plant Licensing,” NUREG-1860, Vol. 1, December 2007, 

ADAMS Accession No. ML080440170.

(NRC, 2008) U.S. Nuclear Regulatory Commission, “Policy Statement on 

the Regulation of Advanced Reactors,” Final Policy Statement, 

October 7, 2008, ADAMS Accession No. ML082750370.
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(NRC, 2011) U.S. Nuclear Regulatory Commission, “An Approach for Using 

Probabilistic Risk Assessment in Risk-Informed Decisions on 

Plant-Specific Changes to the Licensing-basis,” Regulatory 

Guide 1.174, Revision 2, May 2011, ADAMS Accession 

No. ML100910006.

(NRC, 2011a) U.S. Nuclear Regulatory Commission, “Recommendations for 

Enhancing Reactor Safety in the 21st Century; The Near-Term 

Task Force Review of Insights from the Fukushima Dai-Ichi 

Accident,” July 2011, ADAMS Accession No. ML112510271.

(NRC, 2011b) U.S. Nuclear Regulatory Commission, “Nuclear Regulatory 

Legislation” NUREG-0980, Vol. 1, No. 9, January 2011.

(Okrent, 1981) Okrent, David. “Nuclear Reactor Safety: On the History of the 

Regulatory Process,” Madison, Wisconsin: University of Wisconsin 

Press, 1981.

(Walker, 2010) Walker, S., “A Short History of Nuclear Regulation, 1946–2009,” 

NUREG/BR-0175, September 2010.

(WENRA, 2008) Western European Nuclear Regulators’ Association, “WENRA 

Reactor Safety Reference Levels,” Reactor Harmonization 

Working Group, January 2008.

Section 4.3: Material Uses Risk Management Task Force | 4.3-1

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4.3 Material Uses 

4.3.1 Background

Reactor and accelerator-produced nuclear materials are used extensively throughout the 

United States for industrial applications, basic and applied research, manufacture of consumer 

products, academic studies, and medical diagnosis, treatment, and research. In addition, 

source materials are used in the production of processed uranium for nuclear fuel fabrication 

and a wide variety of other uses. The NRC materials regulatory programs are designed to 

ensure that licensees use these materials safely and securely and to present no undue risk 

to public health and safety and the environment. The regulatory framework for use of these 

materials is contained in the Commission’s radiation protection standards in 10 CFR Part 20, 

in application-specific regulations in 10 CFR Part 30 through Part 39, and in related guidance 

and policies. 

Under the provisions of Section 274 of the Atomic Energy Act, the NRC may enter into 

agreements with States under which the NRC discontinues its authority over certain radioactive 

materials and a State assumes that authority. The Agreement States, as they are known, now 

regulate more than 85 percent of the materials licensees in the United States. Presently, there 

are 37 Agreement States and, in concert with the NRC, they play a major role in the regulation 

of most materials uses. 

The NRC materials regulatory program is the Agency’s overall program that ensures safe 

and secure use of these materials by approximately 2,900 specific licensees. In addition, the 

program provides guidance and oversight to the Agreement States. Each year the NRC and 

Agreement States issue thousands of licensing actions (new applications, renewals, and license 

amendments) and conduct thousands of routine and reactive inspections. The program also 

regulates general licenses, licenses and inspects decommissioning activities and uranium 

recovery operations, implements the low-level waste program and conducts materials-related 

rulemakings.

The risks presented by the use of nuclear materials differ significantly from the risks presented 

by power reactors. The potential for releases to the environment are a major focus of the 

reactor program; that potential is not a significant part of the materials program because of the 

much smaller amounts of radioactive material possessed by most materials users and the lower 

risk posed by many of those specific uses. Therefore, the risk analysis for reactors focuses 

predominantly on the prevention of low-frequency, high-consequence scenarios, whereas the 

risk analysis for materials uses focuses primarily on higher-frequency, lower-consequence 

scenarios. Materials scenarios or events may include, for example, the loss of a portable 

gauge from the back of a truck, a well-logging device that is lost beneath the earth’s surface, 

or a radiography camera whose source becomes stuck in an exposed position. These types 

of events are not unusual, but they generally do not pose substantial radiological hazard or 

risk. Furthermore, in the materials program, risk assessments are largely qualitative, based 

on operational experience; however, in the reactor program, such assessments are more often 

quantitative and probabilistic. 

The materials program differs from the reactor program in other ways as well. Most of the 

technologies involved in materials uses are mature, having been implemented for decades, 

and are not undergoing fundamental change. New device and source development and uses 

are relatively infrequent. Costs, including regulatory costs and the difficulty in finding disposal 
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4.3-2 | Risk Management Task Force Section 4.3: Material Uses

options for radioactive sources, have led many private sector firms to seek nonradioactive 

technologies for their needs. All these factors have led to a relatively static or even shrinking 

number of materials licensees nationwide. 

The primary concern in the materials program is radiation protection. Accordingly, the focus 

is on ensuring that worker and public doses are maintained within the limits contained in 

10 CFR Part 20. There are no parallels to reactor safety goals in the materials program and, 

for the most part, there are no parallels to the General Design Criteria of 10 CFR Part 50 or 

emergency planning considerations because most materials users do not possess enough 

material to result in a credible accident scenario that might exceed EPA’s Protective Action 

Guidelines (PAGs). Finally, the concept of defense-in-depth, which is a central part of reactor 

regulation, is more of an implicit rather than explicit part of the materials program. 

Traditionally, the basis for the materials program has been largely a deterministic one, with 

rules and guidance developed over time and as a result of operational experience to provide 

reasonable assurance of adequate protection of public health and safety. Beginning in the 

1990s, the NRC began a shift to a more risk-informed, performance-based approach for all of 

its regulatory programs. This shift was embodied in a series of Commission SECY papers, 

the “Commission’s Final Policy Statement on Use of Probabilistic Risk Assessment Methods in 

Nuclear Regulatory Activities” and “Commission Direction-Setting and Policymaking Activities” 

papers that resulted from the agency’s strategic assessment and rebaselining process. At that 

time, the NRC undertook a number of initiatives to better risk-inform and performance-base 

its nuclear materials (and other) regulatory programs. These led to fundamental changes in 

inspection frequency and approach as well as licensing policies and practices, and regulations 

and guidance. 

Risk insights and performance considerations continue to be significant factors in materials 

program development and implementation, including rulemaking, guidance development, 

inspection, enforcement, and licensing. For example, in 2011, a working group of experienced 

Agreement State and NRC managers and staff conducted a re-evaluation of the nuclear 

materials inspection program to ensure its continued efficiency, effectiveness, and focus on 

safety and security. The group looked at various aspects of the inspection program, including 

risk insights. It considered several operational measures of risk—operational data, enforcement 

data, and expert elicitation through interviews with experienced NRC and Agreement State 

inspectors and program managers—in reaching its conclusion that the program is meeting 

its objective of ensuring the safe and secure use of nuclear material in a risk-informed, 

performance-based manner. 

Medical Uses

The NRC and Agreement States issue licenses to hospitals and physicians for the use of 

radioactive materials in medical treatments. Medical uses fundamentally differ from all 

other activities regulated by NRC and Agreement States because it is the only area in which 

individuals are intentionally exposed to radioactive materials. Approximately one third of all 

NRC and Agreement State licenses in the United States are for medical uses. In addition, the 

NRC develops guidance and regulations for use by licensees and maintains a committee of 

medical experts (the Advisory Committee on Medical Uses of Isotopes, or ACMUI) to provide 

advice to the staff on the use of byproduct materials in medicine. The NRC regulations in 

10 CFR Part 35, “Medical Use of Byproduct Material,” require physicians and physicists to have 

special training and experience to practice nuclear medicine. 
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Industrial and Commercial Uses

The NRC and Agreement States license a wide variety of industrial and commercial 

applications, including industrial radiography, fixed and portable gauges, well logging, nuclear 

laundries, research and development, and manufacturing. These uses have varying risks 

associated with them, reflective of the type and amount of material involved, the environments 

in which they are used, and the operational experience of the users. For example, radiography 

uses high-activity gamma radiation sources to find structural defects in metallic materials and 

welds, in both fixed and field locations. Commercial irradiators are also relatively high-risk 

facilities that expose products, such as food, food containers, spices, medical supplies, and 

wood flooring, to radiation to eliminate harmful bacteria, germs, and insects, or for hardening 

or other purposes. The NRC and Agreement States license approximately 50 commercial 

irradiators nationwide that can contain upwards of 1 million curies of cobalt 60. 

Fixed and portable nuclear gauges are used as nondestructive devices (e.g., to determine the 

thickness of paper products, fluid levels in oil and chemical tanks, and the moisture and density 

of soils and material at construction sites). In contrast to the other commercial uses described 

above, both fixed and portable gauges are relatively low-risk materials uses. They are regulated 

under the general requirements of 10 CFR Part 30, “Rules of General Applicability to Domestic 

Licensing of Byproduct Material,” require little in the way of operator training, and are inspected 

at a 5-year frequency (the lowest inspection frequency for materials uses).

General Licensees

Generally licensed devices are devices that typically contain small amounts of radioactive 

material and are used to detect, measure, gauge, or control the thickness, density, level, or 

chemical composition of various items. Examples of such devices are gas chromatographs 

(detector cells), density gauges, filllevel gauges, tritium exit signs, and static elimination devices. 

Because of the small amounts of radioactive material contained in these devices, as well as 

their inherent robust design, an individual is not required to apply to the NRC or an Agreement 

State for a license. The devices may be received from any licensed vendor, and they carry 

basic accountability requirements. Most generally licensed devices are not subject to routine 

inspection oversight. 

Security in the Materials Program

Over the past several years, the NRC and the Agreement States have increased security 

requirements on radioactive materials. These requirements were developed and implemented 

in a risk-informed manner, taking into account the form, quantity, and other aspects of the most 

risk-significant radionuclides. The NRC worked with its international partners in developing 

the International Atomic Energy Agency (IAEA) Code of Conduct on the Safety and Security 

of Radioactive Sources. The Code was published in 2004 and it lists 26 radionuclides and 

3 activity thresholds for each (IAEA, 2004). Those activity levels are Category 1, 2, and 3, 

respectively, with Category 1 being the most risk significant. The NRC also issued orders, 

additional security measures, and increased controls to licensees that possessed Category 1 

and 2 levels of material. The Commission recently approved a final rule, 10 CFR Part 37, 

which would place materials security requirements in the regulations. Working with other 

Federal agencies, such as the U.S. Department of Homeland Security, the NRC has also 

implemented a voluntary program of additional security improvements for materials licensees. 

In January 2009, the NRC deployed its National Source Tracking System (NSTS), by which the 
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agency and the Agreement States track the manufacture, distribution, and ultimate disposal 

of Category 1 and 2 sources. Licensees use the NSTS, a secure Webbased system, to enter 

uptodate information on the receipt or transfer of tracked radioactive sources. In addition, the 

NRC is developing the Web-Based Licensing system and the License Verification System to 

track license information for materials licensees and plans to integrate these systems with the 

existing NSTS. Integration of the three systems will provide a Web-based solution to: enable 

accounting of the possession of the most risk-significant radioactive sources in the Nation, 

authenticate the validity of radioactive materials licenses, and modernize materials licensing. 

Together, these activities have made radioactive sources more secure and less vulnerable to 

potential terrorists.

Environmental Considerations in the Materials Program

The Commission’s environmental protection regulations for implementing the requirements of 

the National Environmental Policy Act in the NRC’s domestic licensing and regulatory functions 

are contained in 10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing 

and Related Regulatory Functions.” Section 51.22 lists those categories of actions that the 

Commission has determined are excluded from environmental review based on a determination 

by the Commission that they do not “...individually or cumulatively have a significant effect 

on the human environment.” Subsection 51.22 (c)(14) identifies issuance, amendment, or 

renewal of materials licenses as one such category. As a result, environmental reviews in the 

form of environmental assessments or impact statements are not a routine part of the materials 

program. Accordingly, the RMTF did not consider environmental risks to materials. 

4.3.2 RMTF Findings and Recommendations for Materials Uses 

As noted above, the risk environment for materials uses is different from reactors and other 

regulated uses. Nonetheless, risk-informed, performance-based considerations have been built 

into the program for regulation of materials uses. Some of the more successful 

examples include:

•	 Programmatic Risk Assessments. The PRA Policy Statement stated that PRA should 

be used in the NRC’s regulatory programs wherever appropriate. Staff in NMSS 

undertook a first step in doing that in NUREG/CR-6642, “Risk Analysis and Evaluation 

of Regulatory Options for Nuclear Byproduct Materials Systems,” which performed a 

quantitative analysis of tasks, hazards, barriers, and doses (to workers and members 

of the public) to determine relative risk and regulatory options of various materials 

uses and systems (NRC, 2000). For purposes of this study, risk was defined in terms 

of the likelihood of workers or members of the public receiving doses of radiation that 

exceeded regulatory limits from normal operations or accidents. The study looked at 

more than 40 material uses or systems, including diagnostic nuclear medicine, high 

dose rate afterloaders, field radiography, pool irradiators, and portable nuclear moisture 

density gauges. This analysis accomplished its stated goals of providing the staff with 

regulatory options for these uses and systems that were informed by risk considerations, 

but noted that the diversity of materials uses and conditions posed significant challenges 

in terms of the lack of data and making comparisons to the use of quantitative 

probabilistic tools for regulatory purposes. The risk insights from NUREG/CR-6642 were 

used in the development of the NUREG-1556 series and to inform rulemakings, but for 

the aforementioned reason, they have not been used as a basis for individual materials 

licensing decisions. 
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•	 Changes to Inspection approach and frequency. The Phase II Byproduct Material 

Review was completed by a team of NRC and Agreement State managers and staff in 

2001 (NRC, 2001). The team was charged with conducting a broad independent review 

of the nuclear byproduct materials program to, among other things, seek to provide a 

more rigorous risk basis to the program. The team applied risk insights from 

NUREG/CR-6642 as well as operational experience and enforcement data in making 

decisions on materials inspection approach and frequency. This work led to a 

fundamental restructuring of the inspection approach to focus inspection activities on risk 

significant aspects of licensed operations as opposed to only compliance and document 

review. It also led to significant changes in inspection frequency based on licensee 

performance and relative risk of licensed activities.

•	 Materials Licensing Guidance. The NUREG-1556 series, Volumes 1–21, “Consolidated 

Guidance About Materials Licensees,” was developed in the late 1990s to pull together 

guidance documents written over the years for the wide variety of materials licensees. 

These documents allow license applicants to find the applicable regulations, guidance, 

and acceptance criteria used in granting a materials license, and help streamline the 

application and staff review processes. Operational experience (performance) and 

risk insights guided the development of these documents; higher risk activities with 

significant performance challenges have more prescriptive regulations and guidance 

than lower-risk activities. Over time, guidance in NUREG-1556 has been revised to 

further incorporate performance considerations and a new revision to the series is under 

development to address security issues. 

The RMTF received a number of comments regarding materials users in response to the 

November 22, 2011 Federal Register notice, including the following: 

•	 Due to the wide variety of licensed materials uses, there is not a common understanding 

of the terms risk-informed, performance-based, and defense-in-depth within NRC or with 

these licensees. 

•	 The NRC’s graded approach to radiation protection—focusing on higher dose, higher 

risk activities—has been one of the primary successes in the materials program.

•	 Performance-based inspections work well. They provide opportunities to recognize 

exemplary licensee performance and can help both the NRC and Agreement States to 

better focus their resources on risk-significant activities.

•	 Certain deterministic activities or concepts are necessary and effective—setting 

dose limits and possession limits, for example—but broader reliance on deterministic 

approaches can lead to ineffective use of licensee and regulatory resources.

•	 The NRC understands and attempts to build upon the different levels of risk associated 

with the various materials users and should build upon these efforts to establish and 

maintain a flexible regulatory approach that allows for and reflects the relative risks of 

these licensed activities.

•	 Regulation of highest risk, highest exposure activities—particularly medical—would likely 

benefit the most from transition to a risk management framework.
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•	 The NRC should inform its deliberations on a holistic risk management structure by a 

thorough stakeholder engagement process in which specific input is solicited by the 

NRC from different categories of licensees, as was done in the development of the 

Safety Culture Policy statement and, in a more limited manner, in the development of the 

Commission’s approach to security for cesium chloride irradiators.

•	 Major challenges to implementation of the framework include the need to work closely 

with the Agreement States; assuring that NRC management and staff support and can 

articulate the basis for the framework, and prioritizing ongoing regulatory initiatives to 

ensure that resources are available for transition to the risk management framework.

•	 The transition to a more holistic, risk management framework will likely take 

5 to 10 years or more.

Implementation of the proposed risk management regulatory framework in the materials 

program does not pose a significant cultural challenge since many of the concepts are already 

in use. For example, NRC regional managers use risk management considerations as a 

routine part of the licensing and inspection oversight functions for materials licensees. The 

NRC Regions, as well as the Agreement States, issue licenses, conduct inspections to assure 

the safe and secure use of materials, issue violations, and take enforcement actions where 

necessary. Each of these areas provides for risk-informed decisionmaking, feedback on 

performance and appropriate adjustments to licensing, inspection and enforcement operations, 

and policy. Regional and Agreement State managers also routinely make risk management 

decisions in determining the appropriate level of followup for events involving radioactive 

materials. Based on health and safety considerations, as well as resources, events may 

warrant immediate dispatch of an inspector to the licensee’s site, a note to the file for the 

next routine inspection to look into the matter, or no action at all. Irrespective of the level 

of risk and performance information already incorporated into NRC and Agreement State 

materials programs, risk enhancement of those programs can improve regulatory and licensee 

performance and further efficiency and effectiveness, provided it is done in a measured, 

systematic manner. 

Finding M-F-1: The materials program has successfully developed and 

incorporated risk insights and performance considerations into its rulemaking, policy 

development, and routine licensing and inspection activities. 

Finding M-F-2: Deterministic approaches such as dose limits and possession limits 

are useful concepts across the wide range of materials uses that should be retained, 

but broader use of deterministic approaches can lead to ineffective use of limited 

licensee, NRC, and Agreement State resources.

Finding M-F-3: Buy-in of the 37 Agreement States is essential to the success of 

risk management process implementation given their role in regulating more than 

85 percent of the materials licensees in the United States.
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Recommendation M-R-1: The NRC materials program should continue to apply 

risk insights and performance-based considerations, as appropriate, in rulemaking, 

guidance and policy development, and implementation in accordance with the 

proposed risk management framework. 

Recommendation M-R-2: The development and rollout of the recommended 

Risk Management Policy Statement should be closely coordinated with the 

leadership of the Agreement States and a joint NRCAgreement State Working 

Group should be established to guide risk management implementation in the 

materials area.

The proposed risk management regulatory framework described in Chapters 2 and 3 could be 

helpful in better harmonizing approaches to safety and security risks in the materials program 

and improving efficiency and effectiveness. Presently, there are apparent disparities between 

these two areas, as exemplified by inspection frequency. Safety and security inspections 

are done separately, even though there may be benefits from doing them at the same time. 

The inspection frequency, for example, for self-shielded irradiators less than 10,000 curies is 

set at 5 years for the safety inspection and 3 years for the security inspection. A recent selfassessment of the materials inspection program recommended that “...safety and security 

inspections should be done together and, in the longer term, a revision to IMC [Inspector 

Manual Chapter] 2800 should be made to add/modify focus elements to fully integrate security 

and safety into one overall inspection.”

Finding M-F-4: Differences in regulatory language and approaches between safety 

and security in the materials area may have reduced the efficiency and effectiveness 

of the NRC and Agreement State regulatory programs.

Recommendation M-R-3: The NRC should apply common risk approaches to 

safety and security based on the proposed risk management and defense-in-depth 

regulatory framework. 

The proposed risk management regulatory framework described in Chapters 2 and 3 is very 

broad and represents an evolutionary, not revolutionary, approach to the agency’s mission of 

protecting public health, safety, and the environment. While the framework is predicated on 

a defense-in-depth philosophy, that term is not commonly used within the materials program. 

However, the defense-in-depth concepts of hazards and barriers described above are implicit 

in the materials program. Considering the three primary components of materials licensing—

specific licenses, general licenses, and exemptions—NRC and Agreement State regulations, 

licenses, and guidance provide for barriers to the hazard presented by radioactive material 

commensurate with the risk presented by the type and form of that material. 
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For example, licensing requirements for panoramic irradiators in 10 CFR Part 36, “Licenses and 

Radiation Safety Requirements for Irradiators,” are arguably the most detailed requirements 

in the materials programs. The rule includes a system of defense-in-depth considerations 

that include physical barriers, engineered safeguards, access controls, and administrative and 

procedural controls designed to protect workers and members of the public from potentially 

significant exposure.

The licensing requirements for less hazardous uses, types, and amounts of radioactive 

materials can be and are correspondingly less prescriptive and reflect a less robust 

consideration of defense-in-depth. For example, portable and fixed gauges use small 

radioactive sources that are double encapsulated and contained within a relatively robust 

housing. The gauges can be used by individuals with a modicum of training that can be 

taken online. 

Within 10 CFR Part 35 there are also defense-in-depth considerations to greater or lesser 

degrees based on the hazard or risk posed by the material or modality. For example, the 

requirements for therapeutic applications of byproduct material, particularly those involving 

highactivity sources, such as high-dose rate afterloaders or gamma stereotactic radiosurgery 

units, are more robust than those for diagnostic nuclear medicine and may include multiple 

physical barriers and administrative controls to protect workers, patients, and members of 

the public. 

Defense-in-depth considerations are built into the design and manufacture of generally licensed 

devices so that an individual can possess and use such a device with no formal training or 

experience and only minimal requirements for accountability. For certain devices, which contain 

a sufficient amount of radioactive material that could pose a greater hazard, the NRC has 

required individuals to be registered (but not licensed).

Finding M-F-5: The terminology of defense-in-depth is not used consistently across 

the agency’s materials regulatory programs. 

Recommendation M-R-4: As part of the implementation of the proposed risk 

management regulatory framework, the RMTF recommends that the materials 

program should more explicitly consider the defense-in-depth philosophy in 

rulemaking, guidance, and program implementation, and modify appropriate parts of 

staff training to make these concepts a central part of such training.

So while there are numerous implicit applications of defense-in-depth consideration in the 

materials program, what is missing is explicit consideration of that philosophy as part of program 

development, implementation, and oversight. 

As noted earlier, the risk management concept is largely embodied in the rulemaking, guidance, 

and inspection aspects of the materials program. It is lacking, however, in the budgeting 
Section 4.3: Material Uses Risk Management Task Force | 4.3-9

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process, particularly because of the amount of time between when budget formulation begins 

and when budget execution takes place, which is on the order of 18–24 months. During this 

time, operational factors that influenced the decisionmaking process of budget formulation can 

and often do change. Fact-of-life changes are considered during this interim period, but not in a 

systematic way. 

The agency could benefit from a more structured application of the risk management process 

during this period. This would allow program managers to better apply resources to those areas 

of safety and security where risks are determined to warrant it. The annual nuclear materials 

events database report, the Agency Action Review Meeting, along with other operational data, 

would serve as the basis for decisions on changes in allocation of resources within major budget 

areas to reflect where priorities need to be adjusted to better focus on emerging safety and 

security risks. For example, if operational data indicated that a performance by a given class of 

licensees warrants greater regulatory attention, program managers at the NRC Headquarters or 

Regions would shift resources to provide that attention. Transition to this approach would be a 

challenge and could take considerable time and effort to change the current budget approach.

Finding M-F-6: The materials program could benefit from a more structured 

application of the risk management process in resource allocation. This process 

would allow program managers to more systematically apply resources to those 

areas where the safety or security risk warrants it. 

Recommendation M-R-5: Headquarters and Region managers should undertake a 

more formal review of operational data and risk considerations to help determine if 

and where budgeted resources need to be adjusted. This option takes into account 

the fact that budgets are developed well in advance of budget implementation and 

that operational realities may change in the intervening time.

Implementation Options for Risk Management in Materials

In the introduction to Chapter 4, three options were identified for implementation of the proposed 

risk management regulatory framework for the NRC. Following are what those options might 

look like for the materials program. 

Option A: Continue the current approach (“patchwork”)

Under this option, the materials program would continue its current approach to risk-informed, 

performance-based programs. Insights from operational data and enforcement data would 

be considered in programmatic development, including rulemaking, guidance development, 

and inspection and enforcement. No significant new initiatives—including development of a 

Commission Policy Statement—would be undertaken based on risk considerations. Periodic 

self-assessments of specific program areas or lessons learned in response to events or other 
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factors would continue to be performed as necessary but not under the direction provided by an 

overall risk management framework. 

Option B: Implement the proposed risk management regulatory framework through selected 

guidance and rule changes

This option would emphasize specific rule and guidance changes to implement the proposed 

risk management regulatory framework contained in the proposed Commission Policy 

Statement. This would include early and substantive outreach to the Agreement States and 

formation of a Working Group to determine where risk management might offer the greatest 

opportunities. These might include changes to the materials inspection approach and frequency 

(including security), training on risk concepts, ongoing revisions to the NUREG-1556 series, and 

specific targeted rulemakings. Rulemakings might include revisions to the GL rule, changes to 

training and experience requirements in 10 CFR Part 35, systematic revision of 10 CFR Part 40, 

and the incorporation of ICRP Publication 103, “The 2007 Recommendations of the International 

Commission on Radiological Protection,” into a revised 10 CFR Part 20. Implementation of this 

option could take up to 5 years.

Option C: Implement the proposed risk management regulatory framework through broad-scale 

regulatory framework changes

For materials, this option would involve a fundamental revision of the basic framework used 

for licensing and oversight for a specific materials area or areas. Such an option might begin 

with a specific area, such as industrial radiography, and taking into account risk management 

considerations and defense-in-depth, how that area might be regulated differently, while 

maintaining reasonable assurance of adequate protection of public health, safety, and the 

environment. That is, how might risk management and defense-in-depth lead to different 

licensing and oversight approaches in the regulations, guidance, and programmatic 

implementation As noted in Option B, the Agreement States would play a substantive role in 

the development of this option. Implementation of this option would take more than 5 years and 

could be as long as 10–15 years. 

The RMTF recommends Option B. It provides a way to transition the agency to the risk 

management framework and process in a reasonable amount of time and reflects ongoing 

operational priorities and resource constraints. It will also ensure continuity and stability in the 

regulated community through the continued use of existing regulations and guidance while 

implementing the risk management framework and process. 

4.3.3 References

(IAEA, 2004) International Atomic Energy Agency, Code of Conduct on the 

Safety and Security of Radioactive Sources, January 2004.

(NRC, 2000) U.S. Nuclear Regulatory Commission, “Risk Analysis and 

Evaluation of Regulatory Options for Nuclear Byproduct Materials 

Systems,” NUREG/CR-6642, February 2000, Agencywide 

Documents Access and Management System (ADAMS) Accession 

No. ML003693052 (Not publically available).
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(NRC, 2001) U.S. Nuclear Regulatory Commission, Phase II Byproduct Material 

Review, August 2001, ADAMS Accession No. ML012270095.

(NRC, 2011) U.S. Nuclear Regulatory Commission, Materials Inspection 

Working Group Final Report, May 2011, ADAMS Accession 

No. ML112020347.

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4.4 Waste

4.4.1 Low-Level Waste

The NRC regulates the management and disposal of low-level radioactive waste (LLW) through 

regulations, licensing, inspection, enforcement, guidance, and policy development. The primary 

regulations in this regard are 10 CFR Part 20, Part 35, “Medical Use of Byproduct Material,” 

(decay-in-storage provisions), and 10 CFR Part 61, “Licensing Requirements for Land Disposal 

of Radioactive Waste.” The Agreement States also play a major role in LLW regulation since 

they regulate more than 85 percent of the materials licensees in the United States and the 

operating commercial LLW disposal sites. A comprehensive history of low-level radioactive 

waste management, regulation, and legislation is provided in a white paper prepared by the 

former Advisory Committee on Nuclear Waste (NRC, 2007).

4.4.1.1 Background

LLW may be generated by any licensee and includes a wide range of forms, activity, and 

associated hazard. LLW is as diverse as contaminated clothing, miscellaneous equipment, 

medical syringes, animal carcasses, ion-exchange resins, and decommissioning wastes. 

Similarly, the activity levels in LLW may range from material that is barely above background to 

material that could result in individual exposures well in excess of occupational limits and could 

pose operational challenges at the disposal site. 

Before the development of 10 CFR Part 61, land disposal of radioactive wastes was governed 

by the general requirements of 10 CFR Part 20, as there were no specific regulations in place 

for such disposal. In response to operational issues at the existing LLW disposal facilities 

and a 1976 General Accounting Office review of those issues and other problems, the NRC 

began development of 10 CFR Part 61 in 1978 and the rule was completed in 1982. The 

agency performed many technical analyses to support the rulemaking. It sought to take into 

account the wide geographic and geologic variation in potential disposal sites by identifying four 

broad performance objectives—protection of the public from radioactive releases, protection 

against the inadvertent intruder, protection of disposal site workers, and site stability after 

closure—and specific technical requirements that needed to be met for granting a license for 

a disposal facility. The rule also took into account the variety in waste forms and activities by 

developing a waste classification system, in which waste was labeled Class A, B, or C based 

on radiological hazard. As the hazard increased, the disposal requirements for that class of 

waste became more robust. Certain wastes—commonly referred to as Greater than Class C—

were identified as generally unacceptable for nearsurface disposal. Greater than Class C 

wastes are the responsibility of the U.S. Department of Energy (DOE) in terms of a long-term 

disposal alternative. In its final form, 10 CFR Part 61 contained a mix of prescriptive (technical 

requirements in Subpart D), risk-informed (waste classification requirements) and performancebased (performance objectives in Subpart C) requirements for the nearsurface disposal of 

radioactive waste. Although no quantitative risk analysis was performed, this rule was an early 

attempt to follow the risk-informed and performance-based approach supported by the best 

understanding of the LLW disposal scenarios and available methods and models at the time.

The only new LLW disposal sites licensed since the issuance of 10 CFR Part 61 have been in 

Agreement States—the Energy Solutions site in Clive, UT, and the Waste Control Specialists 

site in Andrews County, TX. In addition, the other LLW disposal sites—Barnwell, SC, and 

Richland, WA— and the major waste processors are also located in Agreement States. The 
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NRC’s role has focused on development of regulations, guidance, and policy development with 

some inspection activity related to special nuclear material disposal licenses. 

Since it was issued in the early 1980s, 10 CFR Part 61 has remained largely unchanged. 

Subsequently, however, there have been significant changes in the LLW environment. The 

Low-Level Radioactive Waste Policy Act of 1980 and amendments to that act in 1985 

established that LLW disposal was a State responsibility. States were encouraged to enter into 

regional compacts that would develop common disposal facilities for use by member States 

of a compact. The Act has not been successful in developing a nationwide system of regional 

or State disposal facilities, and one of the longstanding operational sites—the Barnwell LLW 

facility in South Carolina—was closed to out-of-compact waste in 2008. A number of operating 

reactors have entered decommissioning, which created large volumes of radioactive waste 

that needed disposal homes. Waste processors began considering blending different classes 

of LLW to achieve a lower waste classification. Volume reduction practices continued in an 

effort to reduce overall disposal volumes and, thus, costs. New waste streams not envisioned 

at the time when 10 CFR Part 61 was written, such as large volumes of depleted uranium from 

commercial uranium enrichment operations, have appeared, and many licensees have had to 

consider onsite extended storage of LLW because of the lack of access to disposal sites. 

The NRC staff has been proactive in recent years in responding to these events and identifying 

the changes needed to the national LLW program, as the following activities show:

•	 strategic assessment of the LLW program (SECY-07-0180)

•	 Commission paper on unique waste streams (SECY-08-0147)

•	 Commission paper on blending of LLW (SECY-10-0043) 

•	 Commission paper on risk-informing 10 CFR Part 61 (SECY-10-0165)

•	 risk-informing Branch Technical Position on concentration averaging

•	 development of a Commission Policy Statement on volume reduction (SECY-12-0003)

In addition, the Commission issued a staff requirements memorandum (SRM) on 

January 19, 2012, on the staff’s ongoing work on risk informing 10 CFR Part 61. The 

SRM directed the staff to expand its effort to bring a clearer risk-informed approach to 

10 CFR Part 61. Specifically, the staff is to provide to the Commission within 18 months an 

expanded proposed rule that would:

•	 Allow licensees the flexibility to use International Commission on Radiological Protection 

dose methodologies in a site-specific performance assessment.

•	 Include a two-tiered approach that establishes a compliance period that covers the 

reasonably foreseeable future and a longer period of performance.

•	 Provide flexibility for disposal facilities to establish site-specific waste acceptance criteria 

based on the site’s performance assessment and intruder assessment.
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•	 Establish Agreement State compatibility criteria that ensure alignment between the 

States and the Federal Government on safety fundamentals while providing the States 

with flexibility.

A key part of LLW disposal licensing is performance assessment. The NRC staff developed the 

following description for the agency’s Web site of the role of performance assessment in the 

regulation of waste disposal:

In the context of disposal of radioactive waste, a performance assessment is a quantitative 

evaluation of potential releases of radioactivity from a disposal facility into the environment, and 

assessment of the resultant radiological doses. The term performance assessment can refer to 

the process, model, or collection of models used to estimate future doses to human receptors. 

Typically, a performance assessment is conducted to demonstrate whether a disposal facility 

has met its performance objectives. In general, a performance assessment considers the 

following factors:

•	 Selected scenario (specific features and processes at the disposal facility and in the 

surrounding area, such as the location of the potential release, location and general 

characteristics of the receptors, and applicable transport pathways through which 

radionuclides might reach the environment and pose a threat to the selected 

receptor groups) 

•	 Performance of the cask or other engineered barrier system used to store low-level 

waste, limit the influx of water, and reduce the release of radionuclides 

•	 Release and migration of radionuclides through the engineered barrier system and 

geosphere (those deep-underground portions of the disposal facility where human 

contact is generally not assumed to occur) 

•	 Radiological dose(s) to the selected receptor group(s) 

Because it is not possible for computer models to precisely replicate all conditions of a 

realistic disposal facility, the staff of the U.S. Nuclear Regulatory Commission (NRC) uses 

abstraction to simplify the information to be considered in a performance assessment. The 

degree of abstraction normally reflects the need to improve reliability and reduce uncertainty, 

balanced with other practical considerations (such as making the model and its results 

easy for people to understand). Nonetheless, it is important for the model to be sufficiently 

detailed to ensure that it yields valid results for the performance assessment. Also, while 

traditional deterministic methods have been sufficient to ensure adequate safety, performance 

assessments can be more explicitly quantified through probabilistic approaches. In particular, 

a probabilistic performance assessment considers the risk triplet: “What can go wrong?” “How 

likely is it?” and “What are the consequences?” Use of performance assessment tools and 

methodologies aids the NRC in applying a risk-informed and performance-based approach to 

regulatory decisionmaking.

Because the existing 10 CFR Part 61 requirements were established when performance 

assessment was at an early stage of its development, a performance assessment is not 

required by 10 CFR Part 61 to demonstrate that the performance objectives have been 

met, nor is it discussed in the staff’s standard format and content guide for LLW applications 
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4.4-4 | Risk Management Task Force Section 4.4: Waste

(NUREG-1199, “Standard Format and Content of a License Application for a Low-Level 

Radioactive Waste Disposal Facility”) and Standard Review Plan (NUREG-1200). 

4.4.1.2 RMTF Findings and Recommendations for Low-Level Waste

The RMTF received responses from several commenters about LLW and risk management in 

response to the November 22, 2011, Federal Register notice. Among the comments provided 

were the following:

•	 The NRC should use caution in attempting to adopt a single approach to risk 

management in light of the diversity of activities that the agency regulates.

•	 10 CFR Part 61 is one area of the regulations that is ripe for full conversion to a risk 

management regulatory approach.

•	 There is not a common understanding and usage of the terms risk-informed, 

performance-based, and defense-in-depth within the NRC, as well as outside the NRC.

•	 Deterministic requirements are inherently straightforward and provide for robust facility 

designs. But, in some cases, those requirements provide no commensurate benefit to 

public health and safety.

•	 Integration of deterministic and risk management approaches is both inevitable and 

appropriate given the time and resource constraints on the NRC.

•	 Challenges to accomplish a holistic risk management structure include resources 

(the NRC’s as well as Agreement States’), difficulty in establishing performancebased measures of success, consistency in interpretation and approach; treatment of 

qualitative versus quantitative analyses; and availability of sufficient data. 

•	 The NRC should engage stakeholders through public meetings specific to individual 

program areas on the risk management approach. The outreach approach that the 

agency took in developing the Safety Culture Policy Statement was offered as a model 

for consideration.

The RMTF considered these comments in light of the findings and recommendations in 

Chapter 2 on adoption of the Risk Management Regulatory Framework and concluded that 

there is no inherent conflict between the framework and the above views offered. The proposed 

risk management framework is particularly well suited to the regulation of LLW given its mix of 

prescriptive, risk-informed, and performance-based components. 

The concept of defense-in-depth is implicit in the requirements and structure of 10 CFR Part 61, 

although the term itself is not explicitly used. The rule provides for a series of barriers or 

controls to assure that the performance objectives are met and that the public and the 

environment are adequately protected. For example, 10 CFR Part 61 requires that an applicant 

for a LLW disposal facility license to design disposal unit covers to minimize water intrusion into 

the disposal units. If water intrudes into the disposal units, other requirements in the rule on 

waste form, packaging, and placement serve as additional barriers or controls to minimize water 

coming into contact with the waste and serving as a transport mechanism for radionuclides. If 

somehow those radionuclides leach out of the waste, the rule requires additional barriers or 
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controls in the form of a buffer zone between the disposal units and the disposal site boundary, 

which must be of sufficient size to allow mitigation measures to be taken. 

As noted in Section 4.3, over the past several years, the NRC and the Agreement States have 

increased the security requirements on the most sensitive radioactive materials. The NRC 

issued orders, additional security measures, and increased controls to licensees that possessed 

Category 1 and 2 levels of material. LLW facilities were included in the group of licensees 

required to implement increased controls because they may possess Category 1 or 2 sources 

for disposal. 

Finding LLW-F-1: The regulatory framework for LLW disposal has, for the most part, 

implicitly followed a risk-informed and performance-based approach. However, 

changes in the LLW environment and the maturing of the performance assessment 

method over the past 30 years have underscored the need to provide a stronger 

risk basis to the program. The staff is currently implementing Commission direction 

to make it more so and will be sending various products to the Commission over 

the next year or two. However, as noted in earlier sections of this report, the terms 

risk-informed and performance-based are not well understood and have not been 

consistently applied across NRC regulatory programs. 

Finding LLW-F-2: Certain aspects of the LLW regulatory framework readily lend 

themselves better to the risk management approach, such as waste classification 

or concentration averaging. Applying the proposed risk management approach 

to comprehensive LLW licensing decisions, however, may be more challenging 

because it involves estimating facility performance due to events potentially far into 

the future.

Recommendation LLW-R-1: The NRC should adopt the concept of risk 

management to the LLW program, including any revisions proposed to 

10 CFR Part 61 (including performance assessment requirements) and related 

guidance documents.

Finding LLW-F-3: The interlocking and reinforcing systems approach in 10 CFR Part 

61 (site suitability, waste form and classification, intruder barrier, institutional 

controls, etc.) represents an implicit consideration of defense-in-depth features, 

based on the risk posed by various classes of waste. 

Finding LLW-F-4: The NRC has not developed an explicit characterization of 

defense-in-depth considerations for the LLW program. 
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Recommendation LLW-R-2: The NRC should develop an explicit characterization 

of how defense-in-depth within the proposed risk management framework applies 

to the LLW program and build this into current and future staff guidance documents 

and into training and development activities for the staff. 

The NRC staff is required by Subsection 51.20 (b)(11) of the Commission’s regulations in 

10 CFR Part 51, “Environmental Protection Regulations for Domestic Licensing and Related 

Regulatory Functions,” to prepare an environmental impact statement (EIS) or supplement to 

an EIS for issuance of a new license for land disposal of radioactive waste received from other 

persons. In addition, Subsection 51.20 (b)(12) requires an EIS or supplemental EIS to be 

prepared for an amendment to a LLW disposal license that authorizes: closure of a disposal 

site, transfer of the license for purposes of institutional control, or termination of the disposal 

site license. EIS preparation involves a consideration of impacts or risks to the environment 

against benefits from a proposed action or alternatives to that action. Renewal or significant 

amendments to the license other than those mentioned above may require an environmental 

assessment that can lead to preparation of an EIS. The requirements in 10 CFR Part 51 

are Federal requirements that apply to the NRC. The Agreement States have varying 

environmental review requirements, but are not bound to those in 10 CFR Part 51.

Finding LLW-F-5: Consideration of environmental risks as well as safety risks is a 

central part of the LLW regulatory program.

Recommendation LLW-R-3: The NRC should include environmental reviews within 

the scope of its risk management framework.

Implementation Options for Risk Management in LLW 

In the introduction to Chapter 4, three options were identified for implementation of the risk 

management framework for the NRC. Following are what those options might look like for the 

LLW program. 

Option A: Continue the current approach 

Under this option, the LLW program would continue its current approach to risk-informed, 

performance-based programs. No significant new initiatives, including development of a 

Commission Policy Statement, would be undertaken based on risk management or defensein-depth considerations. Under this option, the agency would take the position that the riskinformed regulation and defense-in-depth concepts already in place in the LLW regulatory 

program are sufficient and no further action is needed to comport with the framework proposed 

by the Task Force. The LLW regulatory framework, including 10 CFR Part 61 and related 
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guidance documents, would be revised in accordance with ongoing risk-informed, performancebased initiatives described earlier in this section. Periodic self-assessments or lessons learned 

in response to events or other factors would continue to be performed as necessary, but not 

under the direction provided by an overall risk management framework. 

Option B: Implement the risk management framework through selected guidance and 

rule changes 

Under this option, the agency would adopt RMTF Recommendation LLW-R-1, LLW-R-2, and 

LLW-R-3 to incorporate risk management and defense-in-depth into the basic philosophy of 

the LLW regulatory program. This option would emphasize specific rule and guidance changes 

to implement the risk management framework contained in the proposed Commission Policy 

Statement. NRC staff would undertake early and substantive outreach to the Agreement States 

(since all existing LLW disposal sites are in Agreement States) and form a working group to 

determine where risk management might offer the greatest opportunities. It would also help 

ensure that current and future regulatory initiatives had a stronger, more well-defined risk basis 

and would reduce deterministic, prescriptive regulation. Staff initiatives in progress, such as a 

revision to 10 CFR Part 61, consistent with the Commission’s January 19, 2012 SRM, would 

be guided by the risk management approach, including any proposed revisions to the waste 

classification, requirements for performance assessment, periods of compliance, considerations 

of uncertainty in performance assessment, etc. Implementation of this option could take up to 

5 years.

Option C: Implement the risk management framework through broad-scale regulatory 

framework changes

This option would involve redesigning the basic framework used for licensing and oversight 

for LLW disposal. In this case, that would mean a complete revision of 10 CFR Part 61 and 

its associated guidance documents to explicitly implement the Risk Management Regulatory 

Framework. As noted in Option B, the Agreement States would play a substantive role in 

developing this option. Implementation of this option would take more than 5 years and could be 

as long as 10 years to 15 years. 

The RMTF recommends Option B. It provides a way to transition the agency to the risk 

management framework and process in a reasonable amount of time and reflects ongoing 

operational priorities and resource constraints. It would also ensure continuity and stability in 

the regulated community through the continued use of existing regulations and guidance while 

implementing the risk management framework and process. 

4.4.2 High-Level Waste

4.4.2.1 Regulatory Background and Framework on High-Level Waste Disposal

U.S. policies governing the permanent disposal of high-level waste (HLW) are defined by the 

Nuclear Waste Policy Act of 1982, as amended (NWPA). This Act specifies that HLW will be 

disposed of underground, in a deep geologic repository, and that Yucca Mountain, NV, will be 

the single candidate site for characterization as a potential geologic repository. In light of the 

Federal Government’s decision to discontinue activities directed at ultimate disposal of HLW 

at the Yucca Mountain site, the future direction and options for the NRC’s regulation of HLW 

disposal are not certain at this time. The following paragraphs discuss how risk-informed and 
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performance-based factors were considered in the development of the current regulations and 

guidance for HLW disposal and how the proposed risk management framework might interact 

with future regulatory development. 

Under the NWPA Act, the NRC is one of three Federal agencies that has a role in the disposal 

of spent nuclear fuel and HLW from the Nation’s nuclear weapons production activities:

•	 DOE is responsible for designing, constructing, operating, and decommissioning a 

permanent disposal facility for HLW, under NRC licensing and regulation. 

•	 The U.S. Environmental Protection Agency (EPA) is responsible for developing sitespecific environmental standards for use in evaluating the safety of a geologic repository. 

•	 The NRC is responsible for developing regulations to implement EPA’s safety standards 

and for licensing and overseeing the construction and operation of the repository. 

In the late 1970s, the NRC began developing regulations for HLW disposal. Since that 

time, risk information and risk concepts have been used increasingly to assist and guide the 

development of the HLW program. Safety of an HLW geologic repository typically is divided 

into what is termed the “pre-closure” or “operational” period (the time when waste is emplaced 

in the repository and before the facility is permanently sealed) and the “post-closure” period 

(the time after the repository is permanently closed). Pre-closure safety is sufficiently similar to 

waste handling operations in other regulated activities (e.g., pool storage and spent fuel storage 

facilities); therefore, regulatory development in the pre-closure area has been guided principally 

by improvements implemented for similar activities for other operating facilities (e.g., security, 

emergency planning, worker safety and procedures, and operation of heavy-lift cranes). 

Conversely, the methods and approaches for evaluating post-closure safety were the principal 

subject of NRC development and improvement because of the unique aspect of evaluating 

the safety of a geologic repository thousands of years and longer into the future. Thus, this 

section primarily describes how post-closure performance assessment methods provided risk 

information that was used to 1) evaluate and improve the regulations for geologic disposal, 

2) assist the resolution of technical issues and development of regulatory guidance, 3) address 

defense-in-depth requirements for the post-closure period, and 4) assist regulatory reviews of 

geological disposal.

Development of Risk-informed and Performance-based Regulations

Existing NRC regulations for geologic disposal are found in 10 CFR Part 60, “Disposal of 

HighLevel Radioactive Wastes in Geologic Repositories,” (generic regulations for all sites 

other than Yucca Mountain, issued in 1983), and 10 CFR Part 63, “Disposal of HighLevel 

Radioactive Wastes in a Geologic Repository at Yucca Mountain, Nevada” (specifically for 

Yucca Mountain, issued in 2001). In the two decades between development of 10 CFR Part 60 

and 10 CFR Part 63, a variety of technical assessment improvements and insights have taken 

place. The Commission stated that 10 CFR Part 63 allows more effective and efficient methods 

of analysis for evaluating conditions at Yucca Mountain than the NRC’s existing generic criteria. 

These new methods were not envisioned when 10 CFR Part 60 criteria were established. 

It was thought that their implementation for Yucca Mountain would avoid the imposition of 

unnecessary, ambiguous, or potentially conflicting criteria that could result from the application 

of some of the Commission’s generic requirements of 10 CFR Part 60 (NRC, 1999).
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Perhaps the most significant change to the NRC regulations was the approach to defense-indepth during the post-closure period of a geologic repository (i.e., implementation of the multiple 

barrier requirements). A longstanding principle of geologic disposal has been a reliance on 

multiple barriers to limit the release and transport of radionuclides. Engineered barriers (such 

as waste packages and waste forms) should complement and work with the geological or 

natural barriers so that safety does not depend solely on a single barrier or phenomenon.

During the late 1970s and early 1980s, when the NRC was first developing generic repository 

criteria for 10 CFR Part 60, quantitative techniques for assessing repository performance were 

in their infancy. This lack of experience with and confidence in the quantitative methods for 

addressing uncertainty was a key factor in selecting an approach for implementing the multiple 

barrier requirement. Therefore, for 10 CFR Part 60, the NRC elected to prescribe minimum 

performance standards for each major system (as they were envisioned at the time) and 

require the overall system to comply with the EPA standards. The standards for each major 

system, which became known as “subsystem requirements,” were: 1) the length of time the 

waste package remains intact (300 to 1,000 years), 2) the rate of subsequent releases from the 

engineered system (release rate of 10-5 per year), and 3) the pre-emplacement ground water 

travel time to the accessible environment (1,000 years). Most of the bases for the NRC staff’s 

recommendations of the numerical values for the subsystem criteria were generic judgments of 

what was thought to be protective of the public health and safety and feasible for 

consistent evaluation.

Some of the earliest applications of performance assessment methods for examining regulatory 

criteria was the examination of the relationship between the NRC’s subsystem criteria 

(i.e., defense-in-depth or multiple barrier requirement) and compliance with the EPA’s overall 

system standard (limits regarding allowable releases from the repository integrated over 

10,000 years). These early examinations revealed that compliance with the subsystem criteria 

alone was not sufficient to ensure compliance with the EPA’s overall release limit (64 FR 8648; 

February 22, 1999). Additionally, questions were being asked about the merits of the EPA 

releases limits for carbon-14 relative to overall safety of the repository, based, in part, on 

performance assessment calculations for a potential repository at Yucca Mountain (NRC, 1990; 

EPA, 1993; NAS, 1995). 

In the 1980s and 1990s, work continued on the development of performance assessment 

methods for evaluating geologic disposal. This work significantly increased confidence in 

the technical ability to assess overall repository performance and to address and quantify the 

corresponding uncertainty (see for example Bonano, et al., 1989; Electric Power Research 

Institute [EPRI], 1990, 1992, and 1996; NRC 1992, 1995, and 1997; Sandia National 

Laboratories, 1992; Pacific Northwest Laboratory 1993; and DOE, 1995). Revisions to the 

standards and regulations for a potential repository at Yucca Mountain made significant use of 

risk information to support more effective and efficient standards and regulations. In particular, 

10 CFR Part 63 reflected a risk-informed, performance-based approach that, for the postclosure period, focused on meeting a dose limit to a reasonably maximally exposed individual 

(i.e., dose limit provided in the EPA standard at 40 CFR 197). 

4.4.2.2 RMTF Findings and Recommendations for High-Level Waste

Regulations in 10 CFR Part 63 impose criteria for performance assessment that must be used 

for demonstrating compliance with the overall safety standard (i.e., dose limit). As a result, the 

NRC found no need for added quantitative subsystem requirements. The quantitative criteria 
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for repository subsystems limited the flexibility of the repository developer, DOE, in achieving 

compliance with the overall performance goal. Instead, the NRC adopted an approach to 

regulation that provided DOE with the flexibility to select and defend the barriers it decides 

to rely on for demonstrating the safety of the repository. The 10 CFR Part 63 approach for 

implementing the multiple barrier approach for Yucca Mountain is tied to the performance 

assessment used to demonstrate compliance with the overall performance standard. The NRC 

stated that when it finalized its regulations for Yucca Mountain, it was “...confident that evidence 

for the resilience, or lack of resilience, of a multiple barrier system will be found by examining 

a comprehensive and properly documented performance assessment of the behavior of the 

overall repository system” (NRC, 2001a). Such an approach allows “...DOE to use its available 

resources effectively to achieve the safest repository without unnecessary constraints imposed 

by separate, additional subsystem performance requirements” (NRC, 2001a). 

Similar to the approach used for demonstrating compliance with the post-closure performance 

objectives (i.e., the use of a performance assessment), the NRC also prescribed a system 

analysis for evaluating compliance with the pre-closure performance objectives (i.e., dose 

limits applicable during the operational period when waste is being emplaced in the repository). 

Regulations in 10 CFR Part 63 identify the need for, and general scope of, the Pre-Closure 

Safety Analysis (PCSA) to be done to demonstrate compliance with the performance 

requirements for the operational phase of the repository. The PCSA was intended to refer to 

a broad category of analyses to be used by DOE in its evaluation of repository operations and 

design in meeting the pre-closure performance objectives. The regulations also prescribe the 

need for certain design measures to ensure the availability of safety systems (e.g., means to 

prevent criticality, radiation alarms, means to provide reliable emergency power, explosion and 

fire detection systems, and appropriate suppression systems). This approach is consistent with 

ensuring safety at an operating facility where actions could be taken to prevent or minimize 

accidents and their consequences. This is distinct from the post-closure period, where it is 

assumed actions will not be taken to ensure safety due to thousands of years and longer 

covered by the post-closure period.

As experience with performance assessment methods and results increased over time, so 

did the use of risk information to resolve technical issues and to assist the development of 

regulatory guidance. In particular, the NRC and DOE conducted regular technical exchanges 

on performance assessments for a potential repository at Yucca Mountain. The technical 

exchanges considered the data and science supporting the abstracted models used in 

performance assessment. They also considered the analysis of the results and interpretation 

of the risks insights derived from the analysis. These technical exchanges integrated 

technical work (e.g., data collection, confirmatory research) with results from the performance 

assessment that assisted prioritization of the technical work and improved performance 

assessment models. For example, geosphere flow and transport models in the performance 

assessment were appropriately simplified (i.e., risk-significant attributes of flow and transport 

could be sufficiently represented with simple models); further detail was added to previous 

models for waste package degradation and release of radionuclides from the repository; 

significantly more detail was added to models for disruptive events (i.e., seismic and volcanic 

events) specific to the Yucca Mountain site; detailed analyses and data collection on rainfall 

and infiltration were conducted to provide an infiltration rate in the performance assessment; 

and field investigations were initiated to better understand redistribution and mass loading of 

volcanic ash.
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The increased understanding of system performance and the increased development of 

scientific support and information for risk-significant aspects of repository performance were 

used to resolve technical issues and to develop the Yucca Mountain Review Plan (YMRP) 

(NRC, 2003). The post-closure portion of the YMRP was developed based on the attributes 

needed to develop a credible performance assessment. 

Finding HLW-F-1: The development of regulations for geologic disposal of HLW 

at Yucca Mountain was based significantly on risk information developed from 

performance assessments and closely followed the proposed Risk Management 

Regulatory Framework.

Finding HLW-F-2: The NRC’s regulatory philosophy of defense-in-depth is 

reflected in the multiple-barrier requirement for post-closure in 10 CFR Part 63. 

Compliance with the multiple barrier requirements is demonstrated through the 

performance assessment. 

Finding HLW-F-3: As performance assessment capabilities and experience 

increased at the NRC during the past 30 years, so did the use of risk insights to help 

guide the HLW program. Risk insights and performance assessment capabilities 

have been used to improve the efficiency and effectiveness of guidance documents; 

inform pre-licensing interactions with DOE; and help identify and direct data needs 

and experimental activities. 

Recommendation-HLW-R-1: Any future revisions to the regulatory framework for 

geologic disposal of HLW should be done in accordance with the proposed risk 

management framework to ensure that risk information continues to be appropriately 

considered in the development of requirements and appropriately reflect any future 

HLW disposal paradigm. 

Preparations for Regulatory Reviews

A fundamental aspect of the risk-informed, performance-based approach in 10 CFR Part 63 was 

that staff would focus its technical and regulatory reviews consistent with safety significance. 

During the staff’s technical review of DOE’s license application for Yucca Mountain, the NRC 

encountered some inefficiencies in applying its risk-informed approach partly because of the 

approach used to develop the YMRP and partly because of the somewhat more “open-ended” 

nature of the risk-informed, performance-based approach adopted in 10 CFR Part 63. 

First, the YMRP was developed well before DOE submitted its license application as a way to 

provide DOE and other stakeholders with information on how the NRC would review DOE’s 

license application. The NRC provided significant detail in the YMRP about how the staff would 

evaluate all the topics that “could” be relevant during the review. However, the NRC staff always 

understood and explained in public meetings on the YMRP that although the YMRP covered 
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4.4-12 | Risk Management Task Force Section 4.4: Waste

the full spectrum of topics, the regulatory review would focus on aspects that were significant 

to safety (i.e., a risk-informed approach). Despite the intended risk-informed approach, there 

was a strong desire (based on legal and adjudication concerns) to address “all” of the topics 

mentioned in the YMRP, simply because they appeared in the YMRP. Thus, the details 

included in the YMRP increased the topics that had to be addressed in the staff review 

regardless of their safety significance. It is appropriate to document how certain topics 

are determined to be not significant to safety; however, the selection of such topics to be 

discussed in technical evaluation reports should be based on the technical merits instead of the 

appearance of the topic in the YMRP. In hindsight, staff development of guidance (i.e., YMRP) 

that attempted to provide an identification of ‘all’ topics that might be significant unintentionally 

increased the burden of the regulatory review and potentially limited the effectiveness of the 

risk-informed approach. 

Secondly, the risk-informed, performance-based approach in 10 CFR Part 63 was intended 

to focus technical reviews on those aspects of the geological repository important to safety. 

As such, 10 CFR Part 63 prescribed in broad terms how the performance assessment of 

overall safety was to be performed (e.g., include features, events, and processes that have 

a significant effect on the overall dose estimate; consider events with an annual probability 

greater than 10-8), which provided the applicant flexibility to develop its safety case and design 

the facility as it deemed appropriate. This approach, however, also leaves open a wide range 

of topics that could potentially need to be adjudicated in the licensing hearing. For example, 

contentions can raise a topic related to performance; however, any detailed discussion on 

the quantitative impact on safety (i.e., risk significance) is generally considered a discussion 

addressed in the hearing. Therefore, a risk-informed, performance-based approach that 

provides flexibility to applicants in developing a safety case for a geological repository may also 

provide a potential for increased topics to be adjudicated in the licensing hearing. Although this 

may be an inevitable consequence of a risk-informed, performance-based approach for geologic 

disposal regulations, further development of regulations and guidance should consider how it 

affects the efficiency and effectiveness of the NRC’s licensing process.

Finding HLW-F-4: The risk-informed, performance-based approach for regulating 

geologic disposal (YMRP and regulations) was intended to provide flexibility to the 

applicant in developing the best approach in meeting the regulatory requirements, 

and allow the use the risk information to focus regulatory review on safetysignificant issues.

4.4.2.3 Implementation Options for High-Level Waste

The nature and future direction of a national program for HLW disposal within the United 

States remains to be determined. The recent Blue Ribbon Commission report was the 

first step in determining that direction in light of the cessation of activities for the Yucca 

Mountain site. The NRC has implemented an orderly closeout of its review of DOE’s license 

application for the Yucca Mountain site. However, the NRC must be prepared to exercise its 

regulatory responsibility for whatever decision is made on the national HLW program. The 

recommendations outlined above are intended to assist the agency in its future development of 

regulations and guidance, but because the timing and direction of the national HLW program are 

uncertain, the RMTF is not identifying any implementation options.
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4.4.3 References

(Bonano, E.J., et al. 1989) E.J. Bonano, et al., “Demonstration of a Performance Assessment 

Methodology for HighLevel Waste Disposal in Basalt Formation,” 

NUREG/CR-4759, U.S. Nuclear Regulatory Commission, 

Washington, DC, 1989.

(DOE, 1995) U.S. Department of Energy, “Total System Performance 

Assessment—1995: An Evaluation of the Potential Yucca 

Mountain Repository,” prepared by TRW Environmental Safety 

Systems, Inc., for DOE, Office of Civilian Radioactive Waste 

Management, B-01717-2200-00136, Revision 01, 1985.

(EPRI, 1990) Electric Power Research Institute, “Demonstration of a Risk-Based 

Approach to HighLevel Waste Repository Evaluation,” EPRI NP7057, Palo Alto, CA, 1990.

(EPRI, 1992) Electric Power Research Institute, “Demonstration of a Risk-Based 

Approach to HighLevel Waste Repository Evaluation: Phase 2,” 

EPRI TR-100384, Palo Alto, CA, 1992.

(EPRI, 1996) Electric Power Research Institute, “Yucca Mountain Total System 

Performance Assessment, Phase 3,” EPRI TR-107191, Palo Alto, 

CA, 1996.

(EPA, 2008) U.S. Environmental Protection Agency, “Public Health and 

Environmental Radiation Protection Standards for Yucca 

Mountain, Nevada; Final Rule; 40 CFR 197,” published in the 

Federal Register, on October 15, 2008, (73 FR 61256).

(IAEA, 2006) International Atomic Energy Agency, “Geological Disposal of 

Radioactive Waste,” Safety Requirements No. WS-R-4, 2006.

(NRC, 1983) U.S. Nuclear Regulatory Commission, “Disposal of HighLevel 

Radioactive Wastes in Geologic Repositories: Technical Criteria; 

Final Rule; 10 CFR 60,” published in the Federal Register on 

June 21, 1983, (48 FR 28194).

(NRC, 1992) U.S. Nuclear Regulatory Commission, “Initial Demonstration of 

the NRC’s Capability to Conduct a Performance Assessment 

for a HighLevel Waste Repository,” NUREG-1327, May 1992, 

Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML012980272.

(NRC, 1995) U.S. Nuclear Regulatory Commission, “NRC Iterative Performance 

Assessment Phase 2—Development of Capabilities for 

Review of a Performance Assessment for a HighLevel Waste 

Repository,” NUREG-1464, October 1995, ADAMS Accession 

No. ML040790450.
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(NRC, 1997) U.S. Nuclear Regulatory Commission, “NRC High-Level 

Radioactive Waste Program Annual Progress Report: Fiscal 

Year 1996,” prepared for NRC by the Center for Nuclear Waste 

Regulatory Analyses, NUREG/CR-6513, January 1997, ADAMS 

Accession No. ML012750518.

(NRC, 1999) U.S. Nuclear Regulatory Commission, “Disposal of HighLevel 

Radioactive Wastes in a Proposed Geologic Repository at Yucca 

Mountain, Nevada,” Proposed Rule; 10 CFR 63, published in the 

Federal Register on February 22, 1999, (64 FR 8640).

(NRC, 2001a) U.S. Nuclear Regulatory Commission, “Disposal of HighLevel 

Radioactive Wastes in a Proposed Geologic Repository at Yucca 

Mountain, Nevada; Final Rule; 10 CFR 63, published in the 

Federal Register on November 2, 2001, (66 FR 55732).

(NRC, 2001b) U.S. Nuclear Regulatory Commission, “Preliminary Performancebased Analyses Relevant to Dose-Based Performance 

Measures for a Proposed Geologic Repository at Yucca 

Mountain,” NUREG-1538, October 2001, ADAMS Accession 

No. ML020100312.

(NRC, 2003) U.S. Nuclear Regulatory Commission, “Yucca Mountain Review 

Plan; Final Report,” NUREG-1804, Revision 2, July 2003, ADAMS 

Accession No. ML032030389.

(NRC, 2007) U.S. Nuclear Regulatory Commission, “History and Framework 

of Commercial Low-level Radioactive Waste Management 

in the United States: ACNW White Paper,” NUREG-1853, 

January 2007, ADAMS Accession No. ML070600684.

(NRC, 2009) U.S. Nuclear Regulatory Commission, “Implementation of a Dose 

Standard After 10,000 Years; Final Rule; 10 CFR 63,” published in 

the Federal Register on March 13, 2009, (74 FR 10811).

(NRC, 2011) U.S. Nuclear Regulatory Commission, “Technical Evaluation 

Report on the Content of the U.S. Department of Energy’s Yucca 

Mountain Repository License Application, Post-closure Volume: 

Repository Safety after Permanent Closure,” NUREG-2107, 

August 2011, ADAMS Accession No. ML11223A273.

(NAS, 1983) National Academy of Science, “A Study of the Isolation System 

for Geologic Disposal of Radioactive Wastes,” Waste Isolation 

Systems Panel, Board on Radioactive Waste Management, 

National Academy Press, 1983.

(NAS, 1995) National Academy of Science, “Technical Basis for Yucca 

Mountain Standards,” Committee on Technical Basis for Yucca 

Mountain Standards, National Academy Press, 1995.
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(PNL, 1993) Pacific Northwest Laboratory, “Preliminary Total-System Analysis 

of a Potential High-Level Nuclear Waste Repository at Yucca 

Mountain,” PNL-8444, January 1993.

(SNL 1992) Sandia National Laboratory, “TSPA 1991: An Initial Total-System 

Performance Assessment for Yucca Mountain,” SAND-91-2795, 

July 1992.

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4.5 Uranium Recovery

4.5.1 Background

Uranium recovery is the first step in the nuclear fuel cycle. It involves extracting uranium from 

its parent ore and processing it into a physical and chemical form (yellowcake) that will allow 

additional processing and fabrication to become nuclear fuel. Although natural uranium is 

found throughout the United States, economically recoverable uranium ore reserves are found 

primarily in the western United States. The concentration of uranium in the parent ore is highly 

variable based on a number of geological and geographic factors. In general, the uranium 

concentrations in U.S. ore deposits are much lower (approximately 0.1 percent by weight) than 

those found in ore deposits in Canada and Australia (as high as 20 percent by weight). 

Because of this low concentration and variability, a variety of recovery methods and 

technologies have been employed in the United States. These include conventional mining and 

milling; insitu recovery (ISR), and heap leaching. 

Conventional uranium recovery involves mining of the uranium bearing ore, either by surface or 

underground mining methods, and milling (crushing, grinding, solvent extraction, concentration, 

drying, and product packaging and shipment). The U.S. Nuclear Regulatory Commission’s 

(NRC’s) regulatory authority under the Atomic Energy Act, the Uranium Mill Tailings Radiation 

Control Act of 1978 (UMTRCA), Title 10 of the Code of Federal Regulations (10 CFR) Part 20, 

“Standards for Protection Against Radiation,” and 10 CFR Part 40, “Domestic Licensing of 

Source Material,” do not extend to the mining of uraniumbearing ore by conventional methods. 

That activity is regulated primarily by States and other Federal agencies. Rather, the NRC 

regulates the processing of uranium ore to concentrate uranium and the disposal of the 

large amounts of tailings resulting from that processing. Conventional methods dominated 

U.S. uranium production through the mid1980s. 

As uranium prices dropped, reflecting a lack of demand beginning in the mid-1980s, other 

processes came into greater use that were less expensive and made lower quality or 

geologically deeper ore bodies economically attractive. The primary one is ISR. Unlike 

conventional mining and milling, ISR does not involve removal of ore from its place in nature. 

Rather, ISR is based on injection of a chemical solution—lixiviant—into the ore body, which 

releases uranium and the resultant loaded solution is brought to the surface through a recovery 

well. Uranium is removed from the solution through an ion exchange process and is then 

concentrated, purified, and dried at a processing plant in a manner similar to that used at a 

conventional mill. 

Heap leach is a third method used to extract uranium. This method involves placing ore 

over a lined pad that is overlain with perforated piping. An acid solution is then applied to the 

ore to release uranium, which percolates through the heap into the piping beneath it. The 

uraniumbearing solution is then collected and, as with ISR facilities, the uranium is removed 

from solution for processing into yellowcake. 

Uranium prices began to rebound beginning around 2003 and continued to increase into 2007, 

peaking at nearly $140 per pound, reflecting increased worldwide interest in nuclear power 

and other factors. This economic incentive has led to a resurgence of interest in new uranium 

recovery facilities, primarily ISRs, although the NRC has also received letters of intent for 

conventional mills and heap leach facilities. Presently, the NRC has received five applications 
A Proposed Risk Management Regulatory Framework

4.5-2 | Risk Management Task Force Section 4.5: Uranium Recovery

for new facilities and expects more than 20 additional applications for new, expanded, or 

restarted uranium recovery facilities. Uranium prices have declined since their peak in 2007 to 

less than $60 per pound, which has led to deferral or cancellation of a number of applications, 

but overall interest in new facilities remains high. 

As mentioned above, the NRC’s regulatory authority over uranium recovery is limited to the 

concentration of uranium as a result of processing and the disposal of tailings produced from 

that process. Before 1978, the NRC did not have specific authority over the disposal of tailings. 

Congress passed UMTRCA to ensure that there would be a comprehensive Federal and State 

program to deal with the environmental problems mill tailings had created. UMTRCA gave the 

U.S. Department of Energy (DOE) authority and responsibility for the cleanup of abandoned 

uranium mill processing sites and tailings under Title I of the Act. Title II gave the NRC authority 

and responsibility for regulating tailings at licensed processing sites and the U.S. Environmental 

Protection Agency (EPA) was given authority and responsibility for setting generally applicable 

environmental protection standards for uranium mills and tailings sites. NRC standards 

to implement the agency’s responsibilities under UMTRCA are required to conform to 

standards set by the EPA. In this regard, EPA’s uranium recovery standards are contained 

in Title 40, “Protection of Environment,” CFR Part 192, “Health and Environmental Protection 

Standards for Uranium and Thorium Mill Tailings.”

Agreement States may also regulate uranium recovery activities. Authority to regulate uranium 

recovery in Agreement States is based on an explicit request by a State to have 11(e)(2) 

byproduct material authority as part of their Agreement. Presently, the States of Colorado, 

Texas, Utah, and Washington have such authority and have active licensing and regulatory 

programs in place. 

The NRC’s regulatory framework for uranium recovery is contained in 10 CFR Parts 20 and 40, 

and in 10 CFR 40 Appendix A, “Criteria Relating to the Operation of Uranium Mills and the 

Disposition of Tailings or Wastes Produced by the Extraction or Concentration of Source 

Material from Ores Processed Primarily for Their Source Material Content.” Appendix A is the 

vehicle that NRC uses to conform its uranium recovery standards—primarily as they apply 

to the management of conventional uranium mills and mill tailings—to those of the EPA in 

40 CFR Part 192. These regulations (both those of the NRC and EPA) are largely deterministic 

and do not reflect risk-informed, performance-based considerations. The Commission 

considered a revision to the regulatory framework for uranium recovery in 2000 that would have 

updated and risk-informed the regulations but chose not to pursue it. 

The NRC does not have specific regulations governing the licensing of ISRs. Licensing 

decisions are made based on staff guidance contained in NUREG-1569, “Standard Review Plan 

for In-Situ Uranium Recovery Licensing” (SRP) and implemented through license conditions 

(NRC, 2003). The SRP is in the process of being revised to reflect risk insights and licensing 

experience. Licenses for ISR facilities also include a performance-based license condition that 

allows licensees to make certain changes without requesting NRC approval in the form of a 

license amendment. This licensing approach recognizes the unique process, geological, and 

hydrological characteristics at ISR sites. The EPA has been working on an ISR rule for several 

years and the NRC staff expects that it could be issued as a proposed rule in the second half 

of 2012. The NRC would then need to conduct a conforming rulemaking to make its regulations 

track those of EPA. In the absence of these regulations, the staff plans to continue its current 

licensing approach of using the SRP and performance-based license conditions. 
Section 4.5: Uranium Recovery Risk Management Task Force | 4.5-3

A Proposed Risk Management Regulatory Framework

Security Considerations in the Uranium Recovery Program

Uranium is not one of the radionuclides contained within the International Atomic Energy 

Agency’s “Code of Conduct on the Safety and Security of Radioactive Sources.” (IAEA, 2004). 

Accordingly, uranium recovery licensees were not included as part of the group of licensees that 

received Increased Controls Orders by the NRC. Security of licensed material at these facilities 

is governed under the requirements of 10 CFR 20.1801 and 10 CFR 20.1802. 

4.5.2 RMTF Findings and Recommendations for Uranium Recovery

The risk environment for uranium recovery is different from power and Nonpower reactors 

as well as many materials uses. The environment was characterized in a risk assessment 

prepared by NRC staff and contractors in NUREG/CR-6733, “A Baseline Risk-informed, 

Performance-Based Approach for In Situ Leach Uranium Extraction Licensees,” issued in 

September 2001, which influenced development of the Standard Review Plan (NRC, 2001). 

The risk assessment concluded that ISR facilities are of inherently low risk, noting that they 

“...contain no operating reactors, no fission products and no high radiation areas requiring 

extensive shielding; and they have operating records that confirm low exposures to workers 

and members of the public.” The only significant operational radiological risk is posed by the 

yellowcake dryer which, under certain accident scenarios, could lead to occupational exposures 

in excess of NRC limits in 10 CFR Part 20. Radiological risk to members of the public is low. 

Uranium recovery facilities are not required to have offsite emergency plans since there are no 

credible accident scenarios that might exceed EPA’s Protective Action Guidelines. The risks 

associated with uranium recovery have largely been environmental ones, in terms of potential or 

actual ground water contamination, long-term tailings management, and radon emissions.

Finding UR-F-1: Uranium recovery facilities are of low radiological risk to workers 

and members of the public under normal operational conditions and most 

accident scenarios. 

Finding UR-F-2: Although the NRC staff has made inroads to risk-informed, 

performance-based licensing of uranium recovery facilities, the regulatory 

framework is largely a deterministic one. NRC regulations in 10 CFR Part 40 and 

Appendix A to 10 CFR Part 40 reflect the requirements of UMTRCA and EPA’s rules 

in 40 CFR Part 192, which pre-date the Commission’s move to a risk-informed, 

performance-based framework in the mid-1990s. Similarly, program guidance, 

especially for conventional mills, could benefit from a greater risk basis. 

Finding-UR-F-3: The exact nature of the ISR rule under development by the EPA is 

not clear at this time; therefore, presents an uncertainty to adoption of the proposed 

risk management regulatory framework.
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4.5-4 | Risk Management Task Force Section 4.5: Uranium Recovery

Recommendation UR-R-1: Notwithstanding the current uncertainty associated 

with the EPA rulemaking, the NRC should adopt the proposed risk management 

regulatory framework to the uranium recovery program to provide greater efficiency, 

effectiveness, and predictability in policy development and 

regulatory decisionmaking.

Recommendation UR-R-2: The NRC should work closely with the Agreement States 

and the regulated community to guide implementation of risk management in the 

uranium recovery program. 

Earlier sections of this report have discussed the concept of defense-in-depth as it applies to 

different program areas within the NRC. Similar to the finding in Section 4.3 for the materials 

program, the concept of defense-in-depth is not commonly used as an explicit consideration 

in the NRC’s regulation of uranium recovery. In large measure, this reflects the fact that 

uranium recovery is a relatively low-risk activity. There are instances, including design features 

and regulatory review of mill tailings impoundments, as well as the arrangement of injection, 

recovery and monitoring wells at ISR facilities that reflect defensein-depth considerations. 

The NRC staff is required by 10 CFR 51.20 (b)(8) to prepare an environmental impact 

statement (EIS) for issuance of new source material license authorizing uranium recovery. EIS 

preparation involves a consideration of impacts or risks to the environment against benefits 

from a proposed action or alternatives to that action. Renewal or significant amendments to a 

uranium recovery may require an environmental assessment that can lead to an EIS. These 

requirements and practices reflect the current and future potential risks posed by uranium 

recovery in terms of potential ground water contamination, long-term tailings management, 

radon emissions, and other factors. The NRC staff has been forward-looking in addressing 

these risks from increased interest in ISR licenses by preparing a “Generic Environmental 

Impact Statement [GEIS] for In-Situ Leach Uranium Milling Facilities (NUREG-1910)” 

(NRC, 2009). The GEIS assesses environmental impacts of common environmental issues 

associated with the construction, operation, and decommissioning of ISL facilities, as well as the 

ground water restoration at such facilities, if they are located in particular regions of the western 

United States. In addressing environmental issues common to the ISL process, the NRC 

staff will use the GEIS as a starting point for its site-specific environmental review of license 

applications for new ISL facilities and applications to renew or amend existing ISL licenses.

Finding UR-F-4: Consideration of environmental risks is a central part of the 

uranium recovery regulatory program. 

Recommendation UR-R-3: The NRC should include environmental reviews within 

the scope of its risk management framework.
Section 4.5: Uranium Recovery Risk Management Task Force | 4.5-5

A Proposed Risk Management Regulatory Framework

The RMTF received input from several commenters about uranium recovery and risk 

management in response to the November 22, 2011 Federal Register notice. Among the 

comments provided were the following:

•	 Section 84 of the Atomic Energy Act mandates that the Commission consider risk to 

public health, safety, and the environment in its management of 11(e)(ii) byproduct 

material, to the extent practicable.

•	 Risk-informed, performance-based approaches are good public policy because they 

promote efficient use of already-limited agency, licensee, and other 

stakeholder resources.

•	 Uranium recovery is by its nature a low-risk activity, as noted in NUREG/CR-6733, 

the GEIS, and other studies. Moreover, the location of uranium recovery facilities in 

remote, low-population areas in the western United States also contributes to low risk to 

the public. Risks of physical injury to workers at uranium facilities are far greater than 

radiological risks from uranium.

•	 There is a common understanding of the terms risk-informed and performance-based in 

the uranium recovery industry.

•	 Deterministic approaches limit the flexibility of both the NRC and the industry to respond 

to lessons learned from operational experience and should be eliminated to the greatest 

extent possible.

•	 A key characteristic of a holistic risk management regulatory structure is a ranking 

of radiological risks from uranium recovery along with the risks associated with other 

NRClicensed activities so that it is clear where they fit into the overall risk picture.

•	 There are many opportunities to identify and prioritize activities that are amenable to 

a risk management approach: regulations, licensing actions, policy development, and 

inspection and enforcement prioritization. Specific examples from the uranium recovery 

perspective include remediation or restoration of UR facilities, applying timeliness in 

decommissioning to uranium recovery wellfields, and contaminated soil cleanup at 

uranium recovery facilities.

The RMTF considered these comments in light of the findings and recommendations in 

Chapter 2 on adoption of the proposed risk management regulatory framework and concluded 

that there is no inherent conflict between the framework and the views offered by commenters. 

4.5.3 Implementation Options for Risk Management in Uranium Recovery

In the introduction to Chapter 4, three options were identified for implementation of the proposed 

risk management regulatory framework for the NRC. Following are what those options might 

look like for the uranium recovery program. 

Option A: Continue the current approach 

Under this option, the uranium recovery program would continue its current approach to riskinformed, performance-based programs. Insights from licensing reviews, environmental 
A Proposed Risk Management Regulatory Framework

4.5-6 | Risk Management Task Force Section 4.5: Uranium Recovery

reviews, and inspection findings would be considered in programmatic development, including 

rulemaking, guidance development, inspection, and enforcement as resources permit. 

No significant new initiatives, including development of a Commission Policy Statement, would 

be undertaken based on risk considerations. Assuming that EPA issues an ISR rule, the NRC 

would conform its regulations to the rule. Periodic self-assessments of specific program areas 

or lessons learned in response to events or other factors would continue to be performed as 

necessary but not under the direction provided by an overall risk management framework. 

Option B: Implement the proposed risk management regulatory framework through selected 

guidance and rule changes

This option would emphasize specific rule and guidance changes to implement the proposed 

risk management regulatory framework contained in the proposed Commission Policy 

Statement. This would include, assuming that EPA issues an ISR rule, making conforming 

changes based on a risk management framework. Development of updated guidance for ISR 

facilities, such as the revision to NUREG-1569, also would be based on risk management 

considerations, as would any guidance updates for conventional milling operations. It would 

also involve early and substantive outreach to the Agreement States and formation of a 

working group to determine where risk management might offer the greatest opportunities. 

Implementation of this option could take 5 years or more.

Option C: Implement the proposed risk management regulatory framework through broad-scale 

regulatory framework changes

The NRC’s ability to make broad scale regulatory framework changes in the uranium recovery 

program is somewhat limited because of the requirement that NRC rules must conform to those 

generally applicable environmental standards developed by EPA. However, the agency could 

consider how the proposed risk management regulatory framework might lead to different 

licensing and oversight approaches within the constraints of the EPA rule. As noted in Option B, 

the Agreement States would play a substantive role in developing this option. Implementation of 

this option would take more than 5 years and could be as long as 10 to 15 years. 

The RMTF recommends Option B. It provides a way to transition the agency to the proposed 

risk management regulatory framework and process in a reasonable amount of time and reflects 

ongoing operational priorities and resource constraints. It will also assure continuity and stability 

in the regulated community through the continued use of existing regulations and guidance 

while implementing the risk management framework and process. 

4.5.4 References 

(IAEA, 2004) International Atomic Energy Agency, Code of Conduct on the 

Safety and Security of Radioactive Sources, January 2004.

(NRC, 2001) U.S. Nuclear Regulatory Commission, “A Baseline Risk-informed, 

Performance-Based Approach for In Situ Leach Uranium 

Extraction Licensees,” NUREG/CR-6733, September 2001, 

Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML012840152.
Section 4.5: Uranium Recovery Risk Management Task Force | 4.5-7

A Proposed Risk Management Regulatory Framework

(NRC, 2003) U.S. Nuclear Regulatory Commission, “Standard Review Plan for 

In-Situ Uranium Recovery Licensing,” NUREG-1569, June 2003, 

ADAMS Accession No. ML031550272.

(NRC, 2009) U.S. Nuclear Regulatory Commission, “Generic Environmental 

Impact Statement for In-Situ Leach Uranium Milling Facilities,” 

NUREG-1910, May 2009, ADAMS Accession No. ML091480244. 

Section 4.6: Fuel Cycle Risk Management Task Force | 4.6-1

A Proposed Risk Management Regulatory Framework

4.6 Fuel Cycle

4.6.1 Background

The NRC regulates major fuel cycle facilities, including those involved in conversion of uranium 

ore to uranium hexafluoride (UF6), gaseous centrifuge and diffusion enrichment, reactor fuel 

fabrication, plutonium processing, and UF6 deconversion. Reactor fuel fabrication facilities 

include those that produce lowenriched uranium, highenriched uranium, and mixedoxide (Pu+U) 

fuels. The NRC also regulates possession of small amounts of special nuclear material (SNM), 

usually for research purposes. 

The regulations applicable to licensing of fuel cycle facilities vary because of differences in the 

nature and amounts of the materials licensed and the historical origins of the facilities. Hazards 

at these facilities include those from radioactive materials, toxic chemicals, and inadvertent 

nuclear criticality accidents. Regulation of risk to workers is a major focus for these facilities 

since these workers typically are in close proximity to the hazards. 

Regulations under 10 CFR Part 70, “Domestic Licensing of Special Nuclear Fuel,” apply to 

licensees authorized to possess SNM, which consists of specified fissile materials. In terms 

of fuel cycle facilities, 10 CFR Part 70 applies to major low-enriched uranium fuel fabrication 

facilities, high-enriched uranium processing facilities, and new enrichment facilities (gaseous 

centrifuge). For these major fuel cycle facilities, Subpart H, “Additional Requirements for 

Certain Licensees Authorized To Possess a Critical Mass of Special Nuclear Material,” to 

10 CFR Part 70, requires licensees to perform an integrated safety analysis (ISA). Regulations 

under 10 CFR Part 70 also apply to licensees possessing amounts of SNM less than that 

required to perform ISAs. Including the ISA requirement, 10 CFR Part 70 also applies to the 

Mixed-Oxide Fuel Fabrication Facility (MOX FFF) as a plutonium processing facility. 

As part of an ISA, Subpart H to 10 CFR Part 70 requires licensees or applicants to identify all 

accident sequences that could lead to “high” or “intermediate consequences” (as defined in the 

rule) to workers or the public. The rule also requires items relied on for safety (IROFS) to be 

applied to credible sequences sufficient to make high-consequence sequences “highly unlikely,” 

and intermediate-consequence sequences “unlikely”.1 The rule allows considerable flexibility to 

licensees for ISA methods and definitions of terms. For this reason, ISAs vary in methods and 

acceptance criteria. Some use considerable quantitative analysis and quantitative criteria. It 

is worth noting that for the purposes of the rule, ISAs may be conservative rather than realistic 

estimates. Unlike a Probabilistic Risk Assessment (PRA), no ISA sums risk from multiple 

sequences to individuals.

Regulations under 10 CFR Part 76, “Certification of Gaseous Diffusion Plants,” apply to 

certification of pre-existing DOE gaseous diffusion enrichment plants. The regulation does not 

require an ISA or PRA, but it does require technical specifications. There is just one operating 

10 CFR Part 76 facility at this time. 

The regulations in 10 CFR Part 40 address domestic licensing of source material (uranium and 

thorium ores). In terms of fuel cycle facilities, 10 CFR Part 40 applies to conversion 

1 Discussions and guidelines on which accident sequences are considered “highly unlikely” and “unlikely” are 

provided in NUREG-1520, “Standard Review Plan for the Review of a License Application for Fuel Cycle 

Facility” (NRC, 2010). In addition to the qualitative guideline, the quantitative guideline for “unlikely” is less 

than 10-4 per year and less than 10-5 per year for “highly unlikely.”
A Proposed Risk Management Regulatory Framework

4.6-2 | Risk Management Task Force Section 4.6: Fuel Cycle

and deconversion facilities as licensees authorized to possess source material. The safety 

requirements for 10 CFR Part 40 uranium conversion licenses are general, not prescriptive, 

and do not require performance of any risk assessment. However, the staff is working on a 

proposed amendment to 10 CFR Part 40 that would require licensees authorized to possess 

more than 2,000 kilograms (kg) of UF6 to perform ISAs.

Prior Risk-Informed/Performance-Based Efforts for Fuel Cycle

As directed in the staff requirements memorandum (SRM) to SECY-99-100, the NRC 

established a Risk Task Group (RTG) to define a framework for risk-informing certain nonreactor 

activities, including fuel cycle facilities (NRC, 1999). The RTG developed methods and criteria 

for risk-informing these activities, which were published in “Risk-informed Decisionmaking for 

Nuclear Material and Waste Applications” (NRC, 2008). This guidance, as directed, included 

guidelines on how safe is safe enough, analogous to the reactor safety goals, as well as 

decision algorithms for certain situations.2

As applicable to fuel cycle facilities, Subpart H to 10 CFR Part 70 is a risk-informed, 

performance-based rule that addresses all three elements of the risk triplet: What can go 

wrong? What are the consequences? What is the likelihood? The regulation includes 

qualitative performance requirements (10 CFR 70.61, “Performance Requirements”) 

formulated in terms of consequences and likelihoods of accident sequences. For example, 

“highconsequence” events are required to have IROFS to be applied until they are “highly 

unlikely”. This rule became final in 2000. As such, the industry and staff have more than 

10 years of experience in implementing this ISA rule. As a result, risk concepts are now more 

widely used in this industry. Lessons based on this experience were recently incorporated into 

Revision 1 of the fuel cycle Standard Review Plan (NRC, 2010). 

ISAs are required for new fuel cycle facilities, including the recent gaseous centrifuge plants 

(GCP). Information on consequences and riskindex likelihood results from two of the GCP ISAs 

were used to prioritize IROFS for the operational readiness reviews of these plants.

Ongoing Risk-Informed/Performance-Based Activities for Fuel Cycle

Issues relating to implementation of the ISA rule (Subpart H to 10 CFR Part 70) continue 

to be addressed by developing additional guidance. These issues include defining what 

plant features must be IROFS and clarifying reporting requirements. These definitions and 

clarifications relate to whether a given process design is acceptable for compliance with 

10 CFR 70.61 performance requirements; therefore, they relate to controlling risk. ISA analyses 

are also being systematically revisited by certain licensees in response to specific findings. 

The NRC inspection program and licensee corrective action programs inherently result in 

reevaluation, on a continuing basis, of parts of the ISAs in response to identified deficiencies 

or potential improvements. Revision of the fuel cycle oversight process (F-COP) to be more 

risk-informed, transparent, and predictable was proposed to the Commission in SECY-11-0140, 

“Enhancements to the Fuel Cycle Oversight Process,” and approved with added direction in a 

Staff Requirements Memorandum (NRC, 2011, and 2012). The proposed approach includes 

2 Detailed algorithms were provided for decisions related to new or revised safety requirements. These 

algorithms, in part, made use of pre-existing NRC guidance for regulatory analysis of rulemaking and backfit. 

However, detailed algorithms for other regulatory applications, such as license reviews and inspections, 

were not provided in this document. 
Section 4.6: Fuel Cycle Risk Management Task Force | 4.6-3

A Proposed Risk Management Regulatory Framework

a significance determination process, cornerstones, and an action matrix; all to be developed 

within 5 years. 

4.6.2 RMTF Findings and Recommendations for Fuel Cycle

Most fuel cycle facilities consist of a large number of separate process nodes. For example, 

about 60 to 80 process steps are required to produce a light-water reactor fuel bundle, given 

enriched UF6 as input. The safety controls and designs of processes are very diverse. Safety 

controls typically are not complex automatic controls requiring power to complete their safety 

mission. Safety or process controls usually are manual and governed by procedure and 

training. Process interactions may occur because of output and input relationships, or violent 

events, but processes are otherwise independent. Reliance on power or other motive forces 

for safety functions is limited. Confinement by ventilation is one exception for the MOX FFF. 

Chemical and nuclear criticality hazards vary substantially in magnitude between processes and 

licensees. Radiological hazards are not a dominant hazard except for the MOX FFF. 

ISAs share some elements in common with PRAs; namely, all significant accident scenarios 

must be evaluated and the consequences estimated. Consequences of fuel facility chemical 

and criticality accidents are acute effects, such as death or irreversible health effects. These 

effects have thresholds; therefore, the relevant risk metric is the probability of the discrete health 

effect instead of an expected value of dose. While most, but not all, ISAs have developed some 

form of quantitative or semiquantitative evaluation of accident sequence consequences and 

frequencies, these quantitative evaluations may be highly conservative and may not have the 

same consequence end point. As such, and as noted above, none of the ISAs sums accident 

sequence frequencies to obtain a metric of risk to specific individuals or of collective risk. 

Quantification of overall risk is further hampered by the fact that technologies for assessing 

quantitative risk to individuals onsite or offsite for the collection of processes in a fuel cycle 

facility have not been fully developed.3

The RMTF received the following comments regarding fuel cycle facilities in response to the 

November 22, 2011 Federal Register notice:

•	 Due to the wide variety of fuel cycle and other licensed operations, there is not a 

common understanding of the terms risk-informed, performance-based, and defense-indepth within NRC or with these licensees. 

•	 The NRC understands and attempts to build upon the different levels of risk associated 

with the various fuel cycle and other materials users and should build upon these efforts 

to establish and maintain a flexible regulatory approach that allows for and reflects the 

relative risks of such licensed activities.

•	 The NRC could and should improve its inspection of fuel cycle facilities to be more 

risk-informed and performance-based using the results of the facility-specific ISAs and 

required annual updates. Inspectors should spend less time evaluating strict compliance 

and focus more on evaluating the actual safety significance of facility operations.

3 For example, an analytical code for calculating potential chemical or radiological risk to an individual 

worker because of potential accidents at multiple locations in a plant, for a single accident scenario, or sum 

across all accident scenarios, does not exist. A computer code similar to MACCS, a probabilistic offsite 

consequence analysis code for potential power plant accidents, with a heavy gas model applicable to UF6 

releases also would be needed to estimate risk to individuals outside the controlled area to account for 

location and meteorology variations.
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4.6-4 | Risk Management Task Force Section 4.6: Fuel Cycle

•	 The licensing process, particularly amendments and renewals, could be improved by 

better ensuring that requests for additional information reflect a true safety benefit or 

need as opposed to nonsafety-significant matters.

•	 The NRC should inform its deliberations on a holistic risk management structure by a 

thorough stakeholder engagement process in which specific input is solicited by the 

NRC from various categories of licensees, as was done in the development of the Safety 

Culture Policy statement.

•	 The NRC should consider a more risk-informed approach to the annual license fee rule, 

given that license fees for a Category I fuel cycle facility are higher than for a reactor 

when the overall risk from fuel facilities is orders of magnitude lower than reactors. 

•	 As a class of licensees, fuel cycle facilities might benefit most from a more risk-informed, 

performance-based approach since they are primarily chemical processing plants that 

safely and securely handle licensed radioactive materials. 

The proposed risk management regulatory framework described in Chapters 2 and 3 applies to 

managing risk through the NRC fuel cycle regulatory program. Similar general concepts of riskinforming decisions in the nonreactor areas of NRC, including fuel cycle facilities, were laid out 

in the document “Risk-informed Decisionmaking for Nuclear Material and Waste Applications” 

(NRC, 2008).

Events with adverse consequences to persons can occur that involve fuel cycle facilities 

regulated by the NRC. Decisions need to be made concerning long-term policies, requirements, 

practices, and resource allocations to manage these potential adverse events. This is true if the 

adverse events result from routine plant operations, accidents, or malevolent acts. Hence, a 

common risk management situation confronts those who must make decisions about radiation 

health, safety, and security. NRC regulatory practices for both safety and security apply 

structured reasoning concerning consequences, likelihood, and other stable aspects about 

potential adverse events to set long-term policies, regulatory requirements, and allocation of 

resources. Therefore, it would be useful if both disciplines recognized the commonalities of their 

risk management frameworks and facilitated communication by a joint refinement of their risk 

management approach and terminology. 

The requirement for and definition of defense-in-depth in safety of fuel cycle facility 

processes is explicit in 10 CFR 70.64(b). That definition is identical to the one contained in 

SECY-98-144, “White Paper on Risk-Informed and Performance-Based Regulation,” which 

defined “risk-informed,” “defense-in-depth,” and related concepts (NRC, 1998). In addition, the 

double contingency principle has been an industry standard in the nuclear criticality safety field 

for decades and is also mandated by 10 CFR 70.64(a)(9). Thus, defense-in-depth is applied in 

regulation of fuel cycle facilities consistent with Commission guidance. However, unlike power 

reactors, where more permanent barriers and controls such as a containment are built into the 

design and operation, defense-in-depth for each fuel cycle unit process is different. As new 

processes are added or existing ones are changed, the design and maintenance of defensein-depth at these facilities are based on the characteristics of the most current operations. 

Therefore, defense-in-depth is a continuing process at fuel cycle facilities, not one permanently 

established by the initial design. 
Section 4.6: Fuel Cycle Risk Management Task Force | 4.6-5

A Proposed Risk Management Regulatory Framework

Finding F-F-1: The current fuel cycle regulatory approach incorporates several 

elements of the proposed risk management regulatory framework, such as the use 

of ISAs to identify safety significant items, and the implementation of a revised fuel 

cycle oversight program as directed by the Commission.

Finding F-F-2: The concept of defense-in-depth, as embedded in fuel cycle 

regulatory requirements and practices, is consistent with Commission guidance. Its 

implementation changes as the processes change at the fuel cycle facilities.

Recommendation F-R-1: The fuel cycle regulatory program should continue to 

evaluate the risk and the associated defense-in-depth protection by using insights 

gained from ISAs. ISAs should continue to evolve to support 

regulatory decisionmaking.

4.6.3 Implementation Options for Risk Management in Fuel Cycle Facilities 

In the introduction to Chapter 4, three options were identified for implementation of the proposed 

Risk Management Regulatory Framework for the NRC. Listed below are what those options 

might look like for fuel cycle programs. 

Option A: Continue the current approach 

Under this option, the fuel cycle programs would continue current approaches to risk-informed, 

performance-based programs. Insights from operational data and enforcement data would 

be considered in programmatic development, including rulemaking, guidance development, 

and inspection and enforcement. This approach would include the development of a more 

risk-informing performance-based fuel cycle oversight process, as authorized by the SRM to 

SECY-11-0140. No other major new initiatives, including development of a Commission Policy 

Statement, would be undertaken based on risk considerations. Periodic self-assessments 

of specific program areas or lessons learned in response to events or other factors would 

continue to be performed as necessary, but not under the direction provided by an overall risk 

management framework. 

Option B: Implement the proposed risk management regulatory framework through selected 

guidance and rule changes

This option would emphasize specific rule and guidance changes over time to implement the 

proposed risk management regulatory framework as contained in the proposed Commission 

policy statement. This would include outreach to fuel cycle stakeholders and formation of an 

NRC Fuel Cycle Working Group to determine where risk management might offer the greatest 

opportunities. These new opportunities might include guidance and tools development to 

support risk-informing for various regulatory activities, including licensing, inspection, and policy 

development, with particular emphasis on revision of the F-COP and evolution of ISAs. Other 
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4.6-6 | Risk Management Task Force Section 4.6: Fuel Cycle

new actions may include enhancement of riskrelated staffing and training, particularly in support 

of a risk-informed F-COP, development and implementation of methods to support prioritizing 

regulatory activities, and development of risk assessment tools applicable to fuel cycle facilities. 

Specific risk-informed changes to licensee safety programs may be identified and implemented 

as a result of the proposed revision of the F-COP. 

Practical Considerations

The staff would need to formulate objectives; then develop, test, and implement guidance. The 

NRC’s approach to involving stakeholders in the development of specific changes to regulatory 

practices is especially important for this program area because of the large number and variety 

of fuel cycle unit operations. Often, even at one facility, detailed knowledge about their risk 

is dispersed among licensee staff members instead of a few NRC experts. Therefore, these 

stakeholders have knowledge needed in the development of the risk management process. 

Likewise, developing applicable risk assessment tools would require a certain amount of 

external technical support because of the limited availability of qualified staff. Training would 

need to focus on practical applications of risk methods, in particular, to support a more riskinformed and performance-based F-COP with a significance determination process. 

This option, as in the case of the fuel cycle oversight process revision, faces the technical 

challenge of risk-informing regulatory activities without expending large resources. The above 

developments in support of an overall risk management approach could be implemented in a 

phased manner over a period of years. Rulemaking or backfit justification would be necessary if 

a particular change in regulatory practice so required. 

Option C: Implement the proposed risk management regulatory framework through broad-scale 

regulatory framework changes

This option would involve a revision of the basic framework and regulations used for licensing 

and oversight for fuel cycle facilities. Current regulations do not require quantitative risk 

assessment of fuel cycle facilities; therefore, rulemaking would be needed.

Practical Considerations

This option would require comprehensive development and application of quantitative risk 

assessment technology for all fuel cycle facilities. This would first require development and 

testing of technical PRA tools for such facilities. The rulemaking would have to follow the 

process and schedule required. 

The RMTF recommends Option B. It provides a way to transition the agency to the proposed 

risk management regulatory framework in a reasonable amount of time and reflects ongoing 

operational priorities, such as revision of the F-COP. It will also ensure continuity and stability in 

the regulated community through the continued use of existing regulations and guidance while 

implementing the proposed risk management regulatory framework.
Section 4.6: Fuel Cycle Risk Management Task Force | 4.6-7

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4.6.4 References

(NRC, 1998) U.S. Nuclear Regulatory Commission, SECY-98-144, “White 

Paper on Risk-Informed and Performance-Based Regulation,” 

June 22, 1998, Agencywide Documents Access and Management 

System (ADAMS) Accession No. ML992880068.

(NRC, 1999) U.S. Nuclear Regulatory Commission, Staff Requirements 

Memorandum Regarding SECY-99-100, “Framework for 

Risk-informed Regulation in the Office of Nuclear Material 

Safety and Safeguards,” June 28, 1999, ADAMS Accession 

No. ML003751901.

(NRC, 2000) U.S. Nuclear Regulatory Commission, Letter from B.J. Garrick and 

D.A. Powers to R.A. Meserve, “Use of Defense-in-depth in Risk 

Informing NMSS Activities,” ACRS Letter, May 25, 2000, ADAMS 

Accession No. ML003718610. 

(NRC, 2001) U.S. Nuclear Regulatory Commission, “Integrated Safety Analysis 

Guidance Document,” NUREG-1513, May 2001, ADAMS 

Accession No. ML011440260. 

(NRC, 2008) U.S. Nuclear Regulatory Commission, “Risk-Informed 

Decisionmaking for Nuclear Material and Waste Applications, 

Revison 1,” February 2008, ADAMS Accession No. ML080720238.

(NRC, 2010) U.S. Nuclear Regulatory Commission, “Standard Review 

Plan for the Review of a License Application for a Fuel Cycle 

Facility,” NUREG-1520, Rev. 1, May 2010, ADAMS Accession 

No. ML101390110. 

(NRC, 2011) U.S. Nuclear Regulatory Commission, SECY-11-0140, 

“Enhancements to the Fuel Cycle Oversight Process,” 

October 7, 2011, ADAMS Accession No. ML111180708.

(NRC, 2012) U.S. Nuclear Regulatory Commission, Staff Requirements 

Memorandum Regarding SECY-11-0140, “Enhancements to 

the Fuel Cycle Oversight Process,” January 5, 2012, ADAMS 

Accession No. ML120050322.

Section 4.7: Interim Spent Fuel Storage Risk Management Task Force | 4.7-1

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4.7 Interim Spent Fuel Storage

4.7.1 Background

The NRC has regulatory responsibility for the storage of spent nuclear fuel (SNF) and reactorrelated greater than Class C (GTCC) waste in an independent spent fuel storage installation 

(ISFSI); and the storage of SNF, solid highlevel radioactive waste (HLW) and reactor-related 

GTCC waste (in solid form) in a monitored retrievable storage installation (MRS). The NRC 

carries out its responsibilities in this area through rulemaking, licensing, and inspection.

Regulatory background and framework

The need for storage of commercial power reactor spent fuel stems from the Federal 

Government’s 1977 decision to forego reprocessing. While power reactor facilities include pools 

for storing SNF, they were not designed to hold all of the SNF generated by reactors during their 

operating life. Dry storage in ISFSIs was selected as a way to expand a reactor facility’s SNF 

storage capacity. ISFSIs were licensed under the agency’s general regulation 10 CFR

Part 70, “Domestic Licensing of Special Nuclear Material,” until late November 1980, when a 

new regulation specifically for ISFSIs was added as 10 CFR Part 72, “Licensing Requirements 

for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation.” This rule 

covers both wet and dry storage and licenses individual ISFSIs, the specific-license process.

The regulation has changed over time to address needs arising from policy and legislative 

requirements and those identified through petitions for rulemaking. To meet the requirements 

set out in the Nuclear Waste Policy Act (NWPA) of 1982, as amended, 10 CFR Part 72 was 

twice modified to introduce a licensing mechanism for MRSs and a mechanism for licensing 

SNF dry storage technologies, a general-license option coupled with approval of dry storage 

system designs. In addition to storage of SNF and solid HLW (MRS only), 10 CFR Part 72 now 

includes ways to license the storage of solid reactor-related GTCC waste at ISFSIs and MRSs.

Other relevant regulations for storage include the security requirements for licensees 

in 10 CFR Part 73, “Physical Protection of Plants and Materials,” as well as the 

Commission’s environmental protection requirements in 10 CFR Part 51, “Environmental 

Protection Regulations for Domestic Licensing and Related Regulatory Functions.” Also, 

U.S. Environmental Protection Agency (EPA) regulations contain public dose limits with which 

licensees must comply (e.g., 40 CFR 191.03(a)).

Licensing

The NRC reviews applications for specific licenses and specific-license amendments for ISFSIs 

and MRSs and certificates of compliance (CoC) and CoC amendments for SNF dry storage 

system designs. These reviews are conducted using guidance in the applicable Standard 

Review Plans (SRPs), NUREG-1567, “Standard Review Plan for Spent Fuel Dry Storage 

Facilities” (NRC, 2000), and NUREG-1536, Revision 1, “Standard Review Plan for Spent Fuel 

Dry Storage Systems at a General License Facility” (NRC, 2010). Applicable interim staff 

guidance (ISG) documents are also used, in addition to regulatory guides (RGs) and industry 

standards, to the extent appropriate. To address license and CoC renewal, the NRC recently 

published NUREG-1927, “Standard Review Plan for Renewal of Spent Fuel Dry Cask Storage 

System Licenses and Certificates of Compliance” (NRC, 2011). For obligations under the 
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4.7-2 | Risk Management Task Force Section 4.7: Interim Spent Fuel Storage

National Environmental Policy Act, the NRC prepares environmental impact statements and 

environmental assessments, as warranted, for each approved license, CoC, and amendment.

A general license is granted by regulation (i.e., 10 CFR Part 72) to 10 CFR Part 50 and 

10 CFR Part 52 power reactor licensees. To make use of an approved SNF dry storage 

system, the general licensee completes an evaluation in accordance with the requirements 

in 10 CFR 72.212 to determine that the selected system is appropriate for its site. These 

evaluations are not reviewed by the NRC licensing staff; instead, they are reviewed as part 

of an inspection of the general license ISFSI (e.g., pre-operational or dry-run inspection). 

Conservative requirements in dry storage system safety analyses (e.g., design-basis events 

and margins of safety), the evaluation of siting factors, the application of quality assurance to 

cask design and fabrication, and physical protection at the site provide the NRC with reasonable 

assurance that this regulatory oversight approach will ensure SNF is stored in an adequately 

safe manner under the general license.

The above review guidance does not apply to one facility: the GE Morris wet-storage, or 

pooltype, ISFSI in Illinois. This facility is the only wet-storage ISFSI licensed in 10 CFR Part 72. 

In 2004, the NRC renewed the license for this facility using appropriate guidance from its other 

regulatory programs (NRC, 2004).

Inspection

The NRC inspects 10 CFR Part 72 licensees, CoC holders, and system component fabricators 

to ensure safe and secure storage of SNF. Inspections focus on risksignificant activities, 

such as fabrication of important-to-safety components by fabricators. Baseline inspection 

frequencies differ for the types of 10 CFR Part 72 entities listed above. For ISFSIs, there is 

no set frequency; instead, inspections are eventdriven (e.g., before loading begins at a new 

ISFSI). For the other entities, frequencies are based on factors such as available resources, the 

population of the entity type, the level of production, and the frequency of design modifications. 

Other factors may result in increased inspection frequency or additional reactive inspections.

Security

Physical security plans for all licensees are designed to protect against sabotage and to 

promote the common defense and security. After September 11, 2001, security assessments 

were performed on four ISFSI/dry storage system designs (or casks) considered representative 

of the current fleets of dry storage systems. The purpose of the studies was to identify if 

licensees needed to take immediate actions to mitigate potential security concerns above and 

beyond the requirements in the ISFSI security orders issued in 2002. None was identified. Risk 

was considered as one factor in the selection of security scenarios to be analyzed as part of the 

assessments. As part of any licensing review for a specific-license ISFSI, an assessment of 

the ISFSI’s security plans is performed to ensure that the licensee can provide high assurance 

of adequate protection of public health and safety and the common defense and security. For 

general-license ISFSIs, the licensee updates its security plans to reflect the presence of the 

ISFSI and verifies that the ISFSI does not decrease the effectiveness of the reactor security 

plans. These revised plans are subject to onsite inspection by the NRC staff. While the 

assessments did not indicate the need for immediate actions, they did indicate there was a 

need to consider changes to the ISFSI security regulations. The staff is currently engaging 

with stakeholders and gathering further information to support the regulatory basis for an ISFSI 

security rulemaking.
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Previous and Current Risk-Informed, Performance-Based Efforts

The NRC has updated and is currently updating its guidance in this area to make it more riskinformed and performance-based. These efforts include prioritization of the review procedures 

in the SRPs, NUREG-1536, and NUREG-1567. The prioritization method is based on a 

qualitative assessment, based on staff judgment, of the risk and impact on defense-in-depth, 

where applicable, of each review procedure. For improvements to guidance related to burnup 

credit, the NRC produced a report investigating the probabilities of improper loading of contents 

into a dry storage system, known as misloads, to inform guidance addressing these events 

with respect to nuclear criticality safety (NRC, 2011a). Staff judgment regarding risk is being 

used to evaluate the appropriate margins for criticality safety for misloads. The NRC is also 

assessing the level of review effort needed for dry storage system shielding and radiation 

protection evaluations and the types of CoC and license conditions necessary to adequately 

capture the system design and operations requirements. While this assessment was initiated 

based on concerns derived from a particular licensee’s and dry storage CoC holder’s activities, 

interactions with other stakeholders indicated the need to evaluate staff guidance in shielding 

and radiation protection. Guidance related to SNF condition (e.g., intact, damaged) recently 

was modified from a prescriptive set of definitions to a framework for defining the SNF condition 

in terms of fuel-specific or system-related functions that must be performed, if any, to ensure 

compliance with applicable regulatory requirements (NRC, 2007).

Changes have been made to 10 CFR Part 72 based on licensing experience and scientific 

and technological advancements. A notable example is the 2003 modification of the 

requirements addressing the geological and seismological site characteristics for specificlicense facilities storing SNF in dry casks. For facilities licensed before the regulation change 

in October 2003, probabilistic techniques were allowed for use as a site selection criterion but 

not for determining the design earthquake (DE). The DE for sites in areas with known potential 

seismic activity was to be equal to the safe shutdown earthquake for a nuclear power plant. 

Considering the approach to the dry storage system (i.e., dry cask) designs—particularly their 

robustness against events that challenge cask integrity more than a seismic event as well 

as the parameters and conditions that lead to lower radiological consequences of a release 

versus a release from a power plant—the NRC believes the seismically induced radiological 

risk associated with facilities using dry casks is significantly less than the risk associated with 

a nuclear power plant. Therefore, the regulation was changed to allow use of a DE level more 

commensurate with the risk associated with an ISFSI or MRS. Requirements for determining 

the appropriate DE level include addressing the uncertainties of the DE estimate using a 

probabilistic seismic hazard analysis or suitable sensitivity analyses.

The NRC published NUREG-1864, “A Pilot Probabilistic Risk Assessment of a Dry Cask Storage 

System at a Nuclear Power Plant” in 2007 (NRC, 2007a). The study’s focus was limited in the 

way it was conducted and in its scope of applicability; it was geared toward developing and 

demonstrating an assessment method. Several aspects of a risk assessment, such as human 

factors and sensitivity and uncertainty analyses, were not performed. The study evaluated 

a particular cask design at a specific site and relied on several assumptions, including the 

storage system or cask being fabricated as designed and operated correctly (i.e., no operator 

error). Since the quantitative results were quite low (given in terms of latent cancer fatality 

due to radioactive release), further studies were not pursued. However, the results have 

been considered by inspection staff to reinforce staff’s understanding as to those aspects of 

operations that are more risksignificant, and the results are used to inform review evaluations of 

fuel rod and storage canister integrity under accident conditions. 
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4.7-4 | Risk Management Task Force Section 4.7: Interim Spent Fuel Storage

A separate probabilistic risk assessment (PRA) entitled “Probabilistic Risk Assessment of Bolted 

Storage Casks: Quantification and Analysis Report,” was performed by the Electric Power 

Research Institute (EPRI) and published in December 2003 (EPRI, 2003). The method of 

evaluation, the storage system design, and the site evaluation parameters differ from the NRC 

study, but the quantitative results for it are also quite low.

In addition to these two PRAs, a human reliability analysis (HRA) study was recently completed 

and published as NUREG/CR-7016, “Human Reliability Analysis—Informed Insights on Cask 

Drops” (NRC, 2012), and NUREG/CR-7017, “Preliminary, Qualitative Human Reliability Analysis 

for Spent Fuel Handling” (NRC, 2012a). The goals of this study were: to identify what should 

be included in a qualitative HRA for SNF and cask handling operations, to demonstrate that 

“A Technique for Human Event Analysis” (ATHEANA) HRA methods can be used for these 

types of operations, and to serve as a starting point for building a technical basis for potential 

improvements to procedures and practices in SNF handling operations. Though the studies 

were not done in the context of a larger or plant-specific PRA, they use the NRC’s pilot PRA 

study and EPRI’s PRA to provide information about SNF and cask handling operations. The 

HRA study attempted to be generic in scope; it was not tied to a specific nuclear power plant. 

The study did focus on two SNF cask designs, but this was done to encompass the scope 

of complexity in handling operations. The study included consideration of cask misloads 

in addition to cask drops and how human performance can plausibly lead to radiological 

consequences that affect plant personnel and, to a lesser extent, the public and 

the environment.

4.7.2 RMTF Findings and Recommendations for Spent Fuel Storage

To a certain extent, 10 CFR Part 72 requirements are informed by risk insights. For example, 

10 CFR 72.48, “Changes, Tests, and Experiments,” allows some changes to be made without 

NRC review and approval. The requirement for radioactive releases and direct radiation 

exposure to be maintained “as low as is reasonably achievable” implies the use of risk 

to determine if effluent and direct radiation levels are sufficiently low. The Backfit Rule in 

10 CFR Part 72.62 requires that the increase in protection of personnel and the public health 

and safety be sufficient to justify the cost of implementing the backfit.

SNF dry storage systems are designed to be robust, passive systems. The systems are 

designed to withstand the effects of “worst case” events or designbasis events and phenomena 

while maintaining the capabilities to provide adequate shielding and confinement of radioactive 

contents and to prevent nuclear criticality. The systems are designed to perform these 

functions while requiring minimal maintenance or repair. This design approach is the basis for 

the current regulatory approach, with requirements aimed at ensuring the systems are robust 

and that consequences from any radiological release are low. For specific-license ISFSIs, 

where the characteristics of the site and its surroundings are known, consideration of these 

characteristics can allow for certain events and phenomena not to be considered. Dry storage 

systems are designed, however, to be useable at as many potential sites as possible. Exclusion 

of consideration for certain events and phenomena is not permitted unless the design CoC 

includes a condition limiting the design’s use to those sites where the excluded events and 

phenomena are not credible. Further, specific-license ISFSIs may use a dry storage design that 

significantly differs from the approved (by CoC) dry storage system designs. An example of this 

scenario is the ISFSI at Ft. St. Vrain, which is a vault-type design.
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The Commission has directed the NRC staff to revisit the paradigm for SNF storage and 

transportation to include evaluating the dry storage of SNF for periods significantly in excess 

of those envisioned previously. Therefore, the licensing and inspection programs are being 

reevaluated to determine the enhancements needed to provide adequate regulatory oversight 

of extended storage. Any risk assessment and risk-informing efforts should adequately account 

for these activities. 

Consideration also should be given to the nature of the dry storage industry. This industry 

basically consists of four major vendors in competition to provide dry storage systems, which 

sets up a dynamic in which design margins are reduced to increase system capacity and 

reduce costs. In response, the NRC has used a conservative approach in its SFS regulations, 

guidance, and licensing practices to minimize the likelihood of adverse consequences to public 

health and safety.

The current situation and trends should influence the direction taken to implement the proposed 

approach to risk-inform the interim SNF storage program. The following summarizes the 

situation and trends relevant to risk-informing considerations:

•	 No additional wet storage applications are expected; thus, licensing activity for wet 

storage is expected to be very limited and no specific guidance has been developed for 

licensing wet storage facilities.

•	 While no applications have been submitted for a MRS, the future of MRS is tied to the 

U.S. policy under deliberation.

•	 The trend in dry storage is toward the general license option; therefore, few, if any, 

specific-license applications are expected beyond amendment applications and license 

renewal of the few existing facilities.

•	 The trend toward general licenses could change based on national policy changes or 

any transition of general license to specific license by decommissioning 10 CFR Part 50 

licensees or other factors.

•	 Amendment applications for specific licenses are few and limited in scope; therefore, 

most of the review is for dry storage system designs.

•	 Guidance for specific-license reviews refers substantially to guidance for dry 

storage systems.

•	 Dry storage system design CoC and CoC amendment applications (predominantly 

amendment applications) continue to be submitted frequently.

•	 Licensees and Coc holders have started to submit applications for license and CoC 

renewals as the initial license and CoC terms have begun to expire.

As such, the implementation of the approach should be influenced by the potential for benefit in 

the oversight of the different interim storage entities.

Beyond the current situation and trends in SNF storage, there are other factors that require 

consideration. Some factors may be driven by changes in technology in other aspects of the 
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4.7-6 | Risk Management Task Force Section 4.7: Interim Spent Fuel Storage

fuel cycle. For example, changes in fuel designs—whether for current or future reactor design 

generations—may influence what may be necessary to store such fuel after its use. Reactions 

to current events, such as national or regulatory policy, also should be considered. At present, 

there is some discussion on the need to move SNF more quickly from SNF pools into dry 

storage. This discussion was prompted by the recent events associated with the Fukushima 

Dai-ichi nuclear station. Also, at some point, the SNF will have to be transported. Additional 

discussion on transportation of nuclear material is provided in Section 4.8.

In making its findings and recommendations, the Risk Management Task Force also considered 

comments on spent fuel storage submitted in response to its November 22, 2011, Federal 

Register notice. These comments included the following:

•	 The risk presented by dry cask storage is much lower than for power reactors, but 

the regulatory framework for SFS calls for a relatively high degree of information and 

involves an unusually protracted approval process.

•	 PRAs performed to date—both by the NRC and EPRI—reached the same conclusion: 

the risks of dry storage ranged from very low to extremely low. However, no subsequent 

action was taken to review the regulatory framework based on these findings. 

•	 Regulations for SNF storage and transport need to be cohesively linked since many of 

the casks are dual purpose. Ideally, this would be done under one set of regulations.

•	 The regulations in 10 CFR Part 71 and Part 72 are not risk-informed. As a result, they 

are not conducive to efficient and timely movement of SNF from pools to storage. The 

ability to move fuel from spent fuel pools to storage, or ideally a final repository, is a 

critical lesson learned from the event at Fukushima Dai-ichi in Japan.

•	 The regulations in 10 CFR Part 71 and 72 are performance-based. As such, RGs are 

highly dependent on identifying methods acceptable to the NRC. Presently, however, 

many requirements are implemented through ISG, which lacks the rigor, reliability, and 

predictability of RGs. 

•	 The NRC should apply lessons learned from PRAs performed to date in its 

comprehensive review of the regulatory framework for spent fuel storage 

and transportation. 

As already noted, the SNF dry storage industry is a fairly small and competitive industry with 

basically four major vendors. Their focus is to provide dry storage systems that meet the needs 

of current and potential customers with the most costeffective designs. This means there are 

great incentives to find ways to meet regulatory requirements that allow the greatest flexibility. 

Additionally, any flexibility in the regulatory requirements that might be achieved through riskinforming of the regulations would be of interest to the industry. Activities such as the EPRI PRA 

are indicative of industry interest in and objectives for pursuing risk-informed regulations and 

regulatory practices in the interim SNF dry storage area. 

The regulations in 10 CFR Part 72 include performance-based requirements. The annual 

dose limits and accident dose limits in 10 CFR 72.104 and 72.106, “Controlled Area of an 

ISFSI or MRS,” are examples of these performance-based requirements. However, other NRC 

programs have developed specific riskacceptance goals. For example, the reactors program 
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uses core damage frequency as a goal in RG1.174, “An Approach for Using Probabilistic Risk 

Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing-basis.” 

In a letter to the Chairman in May 2000, the Chairman of the Advisory Committee on Reactor 

Safeguards (ACRS) and the Advisory Committee on Nuclear Waste stated that implementation 

of regulations within a risk-informed framework requires the establishment of risk-acceptance 

goals (NRC, 2000a). No such goals have been developed for regulation of dry storage 

activities, although probabilities have been used in specific licensing actions to inform decisions 

on the need to consider certain events and phenomena. Development of a consensus on an 

appropriate goal and the supporting basis will require a significant effort and will have to be set 

by a Commission Policy Statement.

The NRC revises its guidance periodically or as other circumstances (e.g., development of new 

information) warrant. With these revisions, engineering judgment and insights from experience 

related to risk inform decisions on guidance changes. Also, changes to 10 CFR Part 72, such 

as a recent rulemaking to extend the license and CoC term limits, involve consideration of 

risk. Often, however, this process is not guided or informed by risk information gained from risk 

assessments. This was pointed out by the ACRS in reference to the prioritization of the review 

procedures in the latest revision of NUREG-1536 when it was presented to the Committee. 

As previously stated, some risk studies have recently been performed, but they are few in 

number and limited in scope, applicability, and purpose. However, these studies could aid in 

the identification of risk-information needs. Some information useful to the development of risk 

information is available, but it has not been collected in a systematic way geared toward riskinforming. This information includes the staff’s technical knowledge and operational, licensing, 

and inspection experience.

Finding S-F-1: The regulatory approach for SNF storage is largely based on 

meeting applicable industry consensus standards and conservative guidance to 

ensure adequate safety margins in the facility and cask designs and operations.

More recently, insights from a limited number of risk studies have been gradually 

factored into this regulatory approach. Furthermore, though qualitative, a systematic 

approach that parallels answering the risk triplet was used in the latest revision of 

the Standard Review Plan.

Recommendation S-R-1: While elements of the proposed risk management 

approach have been used in the SNF storage regulatory approach to evaluate the 

acceptable level of risk and the sufficiency of defense-in-depth (physical barriers, 

controls or margins) more consistently, the NRC should develop the necessary risk 

information, the corresponding decision metrics, and numerical guidelines. This is 

important in guiding further changes to the existing SNF storage regulatory approach 

and the evaluation of strategies for extended SNF storage activities.

As noted in earlier portions of this report, defense-in-depth is an important part of the NRC’s 

regulatory program. The concept is most notably incorporated in 10 FR 72.124(a), the double 

contingency principle to prevent nuclear criticalities. In addition to the current licensing 
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approach, defense-in-depth may also be inherent in the designs and operations of the various 

dry storage systems. However, these aspects are not explicitly identified or recognized as 

defense-in-depth considerations. Therefore, while there are implicit applications of defensein-depth consideration in the SNF storage regulatory program, more explicit consideration and 

application of that philosophy is warranted. 

Finding S-F-2: The concept of defense-in-depth is not explicitly or consistently 

applied in the SNF storage regulatory program.

Recommendation S-R-2: As part of the implementation of the proposed risk 

management regulatory framework, the NRC should more consistently consider the 

concept of defense-in-depth explicitly and evaluate its proper use in the SNF storage 

regulatory program. The NRC should also improve appropriate parts of staff training 

to make this concept a central part of such training.

4.7.3 Implementation Options for Risk Management in Interim SNF Storage

The introduction of Chapter 4 identifies three options for implementation of the proposed risk 

management regulatory framework. The following are what those options might look like for the 

interim SNF storage program.

Option A: Continue the current approach 

The interim storage program would continue its current approach to risk-informed, performancebased programs under this option. Insights from operational and enforcement data would be 

considered in programmatic development, including rulemaking, guidance development, and 

inspection and enforcement. In instances where it is deemed necessary, the data may be 

evaluated within the bounds of a very narrow scope to inform specific, targeted actions, such 

as in the case of the study of misload probabilities for informing the revision of guidance related 

to burnup credit implementation. No significant new initiatives, including development of a 

Commission Policy Statement, would be undertaken based on risk considerations. Periodic 

self-assessments of specific program areas or lessons learned in response to events or other 

factors would continue to be performed as necessary but not under the direction provided by 

an overall risk management framework. An example of such assessments is the current work 

on extended storage and transportation of SNF, which has been initiated at the direction of the 

Commission because of the evolving national policy regarding final disposal of SNF.
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Option B: Implement the proposed risk management regulatory framework through selected 

guidance and rule changes

This option would emphasize specific rule and guidance changes to the existing SNF storage 

regulatory approach to implement the Risk Management Regulatory Framework contained 

in the proposed Commission Policy Statement developed to adopt this framework. It would 

also apply the proposed risk management regulatory framework to the Commission-directed 

review of the paradigm of spent fuel storage. This would include substantive outreach to interim 

SNF storage stakeholders. It might involve forming a working group to determine where risk 

management might offer the greatest opportunities. These opportunities might include staff 

training on risk concepts, DID and practical applications of risk methods (including development 

of qualified risk assessment staff), development of guidance and tools to support risk-informing 

regulatory activities, and guidance and rule changes that may result from activities related to 

extended storage and transportation. 

The actions taken as part of this option would necessitate implementation of the previously 

noted recommendations. Development of risk assessment tools and methods would require a 

significant amount of external technical support because of the lack of qualified risk assessment 

staff in the storage program area. This development should use the previously noted efforts 

already expended toward development of risk assessment tools appropriate for interim SNF 

storage activities. Feasibility and benefit also would have to be considered in determining which 

opportunities to pursue.

Option C: Implement the proposed risk management regulatory framework through broad-scale 

regulatory framework changes

This option would involve a fundamental revision of the basic framework and regulations 

used for licensing and oversight for interim SNF storage facilities and dry storage systems. 

Appropriate consideration of risk management concepts and defense-in-depth—a potentially 

different approach to licensing and oversight that maintains reasonable assurance of adequate 

public health and safety and protection of the environment—could be identified. This effort 

would require quantitative risk assessments to support such an approach. Only two such 

studies have been performed to date, and they are are limited in scope and applicability. The 

considerations described for implementing Option B would also apply to this option. Significant 

outreach to stakeholders also would be an important aspect of this option.

The RMTF recommends Option B. It provides a way to transition the agency to the proposed 

risk management regulatory framework and process in a reasonable amount of time and 

reflects ongoing operational priorities and resource constraints. It will also ensure continuity 

and stability in the regulated community through the continued use of existing regulations and 

guidance while implementing the risk management framework and process.

4.7.4 References

(EPRI, 2003) Electric Power Research Institute, “Probabilistic Risk Assessment 

(PRA) of Bolted Storage Casks: Quantification and Analysis 

Report,” EPRI, Palo Alto, CA: 2003. 1002877.
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4.7-10 | Risk Management Task Force Section 4.7: Interim Spent Fuel Storage

(NRC, 2000) U.S. Nuclear Regulatory Commission, “Standard Review Plan for 

Spent Fuel Dry Storage Facilities,” NUREG-1567, March 2000, 

Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML003686776.

(NRC, 2000a) U.S. Nuclear Regulatory Commission, Letter from B.J. Garrick and 

D.A. Powers to R.A. Meserve, “Use of Defense-in-depth in Risk 

Informing NMSS Activities,” ACRS Letter, May 25, 2000, ADAMS 

Accession No. ML003718610.

(NRC, 2004) U.S. Nuclear Regulatory Commission, “Encl 2 (Safety Evaluation 

Report) to 12/21/04 Ltr J E Ellis, General Electric Co, Issuance 

of Renewed Materials License No. SNM-2500, Morris Operation 

Independent Spent Fuel Storage Installation (72-1) (L23091),” 

December 21, 2004, ADAMS Accession No. ML043630514.

(NRC, 2007) U.S. Nuclear Regulatory Commission, “SFST ISG-1, Rev. 2, 

Division of Spent Fuel Storage and Transportation Interim Staff 

Guidance No. 1, Revision 2, ‘Classifying the Condition of Spent 

Nuclear Fuel for Interim Storage and Transportation Based on 

Function,’” May 11, 2007, ADAMS Accession No. ML071420268.

(NRC, 2007a) U.S. Nuclear Regulatory Commission, NUREG-1864, “A Pilot 

Probabilistic Risk Assessment of a Dry Cask Storage System 

at a Nuclear Power Plant,” March 2007, ADAMS Accession 

No. ML071340012.

(NRC, 2010) U.S. Nuclear Regulatory Commission, “Standard Review Plan for 

Spent Fuel Dry Storage Systems at a General License Facility,” 

NUREG-1536, Revision 1, July 2010.

(NRC, 2011) U.S. Nuclear Regulatory Commission, “Standard Review Plan 

for Renewal of Spent Fuel Dry Cask Storage System Licenses 

and Certificates of Compliance,” NUREG-1927, March 2011, 

ADAMS Accession No. ML111020115.

(NRC, 2011a) U.S. Nuclear Regulatory Commission, “Estimating the Probability 

of Misload in a Spent Fuel Cask,” November 30, 2011. ADAMS 

Accession No. ML113191144.

(NRC, 2012) U.S. Nuclear Regulatory Commission, “Human Reliability 

Analysis—Informed Insights on Cask Drops,” NUREG/CR-7016, 

February, 2012, ADAMS Accession No. ML110610673.

(NRC, 2012a) U.S. Nuclear Regulatory Commission, “Preliminary, Qualitative 

Human Reliability Analysis for Spent Fuel Handling,” 

NUREG/CR-7017, February 2012, ADAMS Accession 

No. ML110590883.
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4.8 Transportation 

4.8.1 Regulatory Background and Framework

The transportation of radioactive materials within the United States is regulated jointly by the 

U.S. Department of Transportation (DOT), the NRC, the U.S. Department of Energy (DOE), 

and State and local governments. In general, DOT has primary responsibility for carrier safety, 

hazardous communications, the definition of radioactive materials used for transportationrelated 

regulations, and highway routing. State and local governments have a significant role in 

overseeing shippers’ and carriers’ compliance with DOT regulations for carrier safety. They also 

have lead responsibility for responding to transportation emergencies within their jurisdictions. 

Local and State governments also can impose additional requirements on the shipment of 

radioactive materials within their jurisdictions, provided such requirements are compatible with 

applicable DOT requirements. 

The approval or certification of shipping package designs for radioactive materials is shared 

jointly between the NRC, DOT, and DOE. NRC and DOT responsibilities for certification of 

shipping package designs are delineated in a 1979 Memorandum of Understanding (MOU) 

between DOT and the NRC (NRC, 1979). The MOU assigns the NRC lead responsibility for 

certification of shipping package designs used to transport Type B quantities (i.e., quantities 

requiring accidentresistant packaging) and fissile material. DOT oversees safety standards 

for packages not required to be certified as accident resistant (e.g., Type A and industrial 

packages). These packages generally are self-certified by shippers or package vendors to 

withstand normal conditions experienced during shipments. DOE has the authority, under DOT 

regulations, to approve transportation package designs for DOEowned material, except where 

congressional legislation has specified otherwise (e.g., packages used to ship transuranic 

wastes to the Waste Isolation Pilot Plant and to a national repository that require NRC 

certification under the Nuclear Waste Policy Act). 

Since the NRC, DOT, and DOE all have authority to approve Type B shipping packages, it is 

important that the safety standards used by these agencies be nearly identical. The United 

States, a participating International Atomic Energy Agency (IAEA) member state, has also 

endorsed the concept that its domestic transportation regulations in Title 49, “Transportation,” 

of the Code of Federal Regulations (49 CFR) Part 173, “Shippers General Requirements for 

Shipments and Packages,” need to be compatible with the IAEA’s transportation regulations 

to the greatest extent practicable. The practical result is that the NRC and DOT periodically 

undertake a joint rulemaking effort to revise their respective regulations to be compatible with 

the latest revision to IAEA transportation regulations. 

Licensing

The NRC’s primary licensing role in transportation safety is the review and certification of 

Type B and fissile material shipping package designs under 10 CFR Part 71, “Packaging 

and Transportation of Radioactive Material.” Under 10 CFR Part 71, NRC reviews individual 

shipping package designs submitted by an applicant to certify that the design can withstand 

a series of specified hypothetical accident conditions and subsequently meet specific safety 

criteria dealing with criticality safety, containment, and package dose limits. In making this 

certification, the package design must consider both the packaging hardware and the contents. 

The tests include a 30-foot drop onto an unyielding surface, a puncture test consisting of 

a 40 inch drop onto a metal spike, a half-hour fire test at 1,475 degrees Fahrenheit and an 
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4.8-2 | Risk Management Task Force Section 4.8: Transportation

immersion test for fissile material. After the test, the package (packaging and contents) must 

remain subcritical, have a dose rate of less than 1 rem per hour at 1 meter from the package 

surface, and a release rate not exceeding an A2 quantity per week. An A2 quantity is the quantity 

below which an accident-resistant package is not required. If the package design complies with 

the safety criteria, it is issued an NRC certificate of compliance (COC). 

In general, individual shipments by NRC licensees do not need specific approval. The 

NRC grants its possession licensees a general license pursuant to 10 CFR 71.17, “General 

License: NRC Approved Package,” to deliver material to a private or contract carrier for 

shipment. The NRC licensee is responsible under the general license for ensuring that the 

material is properly prepared for shipment, including the selection of an appropriate DOT- or 

NRCapproved package. NRC licensees also may obtain a specific license for transportation 

under 10 CFR Part 71. However, this option has rarely been used (not since the mid-1980s). 

The NRC COC enables the package to be used under the general license in 10 CFR 71.17 as 

well as under DOT regulations for non-NRC licensees.

Inspection

The NRC inspects 10 CFR Part 71 certificate holders and package component fabricators to 

ensure that transportation casks are fabricated and tested in accordance with the package 

specifications in the NRC certificate. These inspections are conducted under Inspection Manual 

Chapter 2690 using inspection procedures applicable to those entities. Inspections focus on 

risk-significant activities, such as fabrication of important-to-safety components. The frequency 

of inspections is based on factors that include the population of the entity type, level of 

production, frequency of design modifications, and available resources. Other factors, including 

performance history, allegations, significant licensing issues, or previous significant inspection 

findings may result in increased inspection frequency or additional, reactive inspections.

Previous and Current Risk-Informed, Performance-Based Efforts

The current framework for certifying Type B and fissile material shipping packages has 

evolved from the efforts undertaken in the 1960s at the IAEA to achieve a set of internationally 

acceptable safety standards that could be applicable for both domestic and transboundary 

shipments. In theory, acceptance of common safety standards would ensure that bordering 

Nations were using a minimum set of safety standards for radioactive material shipments. It 

would also ensure that package approvals by one Nation could be accepted by other Nations 

through which a package may travel. 

In 1961, the IAEA issued safety standards for transportation requiring packages that carry over 

certain quantities of radioactive material (Type B) to withstand “worst-case credible accidents” 

(IAEA, 1961). Not surprisingly, there was considerable disagreement among IAEA member 

states on what constituted a worst-case credible accident. By 1964, IAEA member states 

defined a worst-case credible accident in terms of a series of physical tests that would bound, 

but not mimic, the potential forces seen in real-life accidents (IAEA, 1964). The physical tests 

were defined to be easily reproducible in all member states. These were the origin of the 30foot 

drop, punctures, and fire tests used to certify shipping packages today. Because there was not 

much information available on the effects that occur during transportation accidents, the tests 

were based on the collective engineering judgment of transportation experts representing 

IAEA member states. Implementation of the physical tests was accompanied by a 
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A Proposed Risk Management Regulatory Framework

recommendation for member states to acquire more accident data and revise the IAEA Safety 

Standards as needed. 

The basic physical tests implemented in the 1964 IAEA regulations are still the primary tests 

used today for approving Type B and fissile material shipping packages. The continued use of 

these physical tests has been supported by numerous risk studies done in the United States, 

as well as in other IAEA member states (AEC, 1972; NRC, 1977; NRC, 1988; NRC, 2000). 

All of the studies have shown that the risk of shipping spent fuel is very low. In addition, the 

United Kingdom (1984), Germany (1999), and the United States (1970s) have conducted 

fullscale demonstration tests by staging real-life accidents for spent fuel casks. In addition, the 

IAEA continually reviews and updates its transportation safety regulations and has repeatedly 

concluded that major changes in the core physical tests were unnecessary.

Consequently, most revisions to IAEA regulations over the years have dealt with areas other 

than the core test requirements for Type B and fissile material shipping packages. A notable 

example was the inclusion of new requirements for Type C shipping packages for shipments 

of large quantities by air in 1995. The development of Type C standards was based on a risk 

analysis of data on aircraft accidents. The impact speed that a Type C package would have to 

withstand was based on a “knee of the curve”-type analysis, a point where a large increase in 

impact speed would capture an increasingly small percentage of air crashes. At the time, the 

United States argued for a much larger impact speed based on its implementation of a Federal 

law in the 1980s that required air shipments of plutonium to withstand the crash of a highflying 

aircraft. The United States has not adopted Type C standards.

In 1995, the United States presented a working paper to IAEA’s Transportation Unit proposing 

a risk-informed approach to future revisions of its Transportation Safety Regulations. The 

proposal advocated accepting the current version of the regulations (1996) as a consensual 

baseline and using risk information and safety goals to screen future revisions. The intent of 

the paper was to slow down the adoption of amendments to regulations that were not riskinformed. The paper was considered at an IAEA Technical Committee Meeting, but it was not 

pursued further.

In 2003, the NRC adopted a provision for special package approvals, similar to the provision for 

special arrangements in IAEA regulations. The provision allows the NRC to approve packages 

for domestic shipments that do not meet some of the prescribed Type B tests if an applicant can 

prove equivalent safety.

4.8.2 RMTF Findings and Recommendations

While the term “defense-in-depth” is not explicitly used, the current regulatory approach 

for approving and inspecting radioactive shipping packages follows the risk-informed and 

performance-based defense-in-depth approach in a general sense. For example, the safety 

requirements for different types of shipping packages become more stringent with the quantity 

(radioactivity), or hazard, contained. The threshold for an accidentresistant package is based 

on an A1 (special form or encapsulated material) or A2 (normal form) quantity. In turn, the 

A1 and A2 quantities are based on accident models that keep the anticipated dose to first 

responders below the occupational exposure limit of 5 rem. If a package contains greater than 

an A1 or A2 quantity (i.e., has a potential to cause an exposure greater than 5 rem), it is required 

to meet Type B accident conditions. The current system also allows shipments of quantities that 
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4.8-4 | Risk Management Task Force Section 4.8: Transportation

would normally require Type B packages to be made in less robust packages that take credit for 

the low, specific activity of the material being shipped.

The current regulatory framework for the certification of radioactive material shipping packages 

is based on a 50-year consensus of IAEA member states. That consensus incorporates the 

acceptance of a very-low risk based on the assumption that “every transportation accident 

occurs offsite” and that transportation packages often are used by third parties not under the 

nuclear regulator’s control. In addition, the need to maintain compatibility with IAEA member 

states and among U.S. regulatory agencies makes changes to regulations resource intensive 

and requires the consent of multiple parties. The current regulatory framework has resulted in 

an outstanding safety record.

In making its findings and recommendations, the RMTF also considered comments on 

transportation submitted in response to its November 22, 2011 Federal Register notice. These 

comments included the following:

•	 In the case of transportation, the NRC’s regulatory framework appears to be graded in 

an inverse manner from what the risk insights would suggest. 

•	 The risk presented by transportation is much lower than that of power reactors; however, 

the regulatory framework calls for a relatively high degree of information and involves an 

unusually protracted approval process.

•	 Regulations for storage and transportation of spent fuel need to be cohesively linked 

since many casks are dual purpose. Ideally, this would be done under one set 

of regulations.

•	 The regulations in 10 CFR Parts 71 and 72 are not risk-informed. As a result, they are 

not conducive to efficient and timely movement of spent fuel from pools to storage and 

ultimate disposal. The need to be able to move fuel from spent fuel pools to storage, 

or ideally a final repository, in an efficient manner is a critical lesson learned from the 

accident at Fukushima Daiichi in Japan.

•	 The regulations in 10 CFR Parts 71 and 72 are performance-based. As such, there is a 

large dependence on regulatory guides to identify methods acceptable to the NRC. 

•	 The NRC should apply the lessons learned from probabilistic risk assessments 

performed to date in its comprehensive review of the regulatory framework for spent fuel 

storage and transportation. 

Much of the conservatism in approving shipping packages under 10 CFR Part 71 stems from 

assumptions used in implementing the regulations rather than in the regulations themselves. 

Risk-informing guidance for implementing current regulations may be desirable in some cases. 

While some aspects of the staff guidance are prescriptive, other aspects have been made more 

performance-based (e.g., definition of fuel condition). Some areas of the guidance and the 

overall review process can be revised to incorporate risk insights based on staff experience and 

judgment. For example, risk-informing to allow packages designed to current standards to ship 

other types of materials (e.g., high-burnup spent fuel) may be desirable. 
Section 4.8: Transportation Risk Management Task Force | 4.8-5

A Proposed Risk Management Regulatory Framework

In addition to operational history, since 10 CFR Part 71 applies to all Type B shipping packages 

in order to more systematically implement the proposed risk management framework for 

transportation, risk analyses of shipping radioactive materials other than spent fuel should be 

developed to complement the existing spent fuel transportation risk studies.

Finding T-F-1: While the U.S. transportation regulatory approach is governed by the 

IAEA transportation regulations, the current NRC transportation regulatory approach 

uses several elements of the proposed risk management framework.

Finding T-F-2: Risk assessments have been conducted on the safety of 

transportation of spent fuel. However, there is a lack of risk information on the 

transportation of other radioactive materials.

Recommendation T-R-1: Considering the strong international regulatory basis for 

transportation and the need to conform U.S. standards to those of the IAEA and 

other member states, application of the proposed risk management regulatory 

framework should focus on implementation guidance.

Recommendation T-R-2: The risk management process should be used to influence 

the future outcome of IAEA deliberations on proposed changes in international 

transportation regulations.

Recommendation T-R-3: The NRC should explore the value of using risk insights 

to justify regulations different from the IAEA’s for domestic use only, such as 

regulations dealing with domestic storage and transportation of high burnup 

fuel. Risk information could be used to develop a more flexible approach toward 

implementing and making gradual changes to current transportation regulations.

4.8.3 Implementation Options for Risk Management in Transportation

The introduction of Chapter 4 identifies three options for implementation of the proposed risk 

management regulatory framework. The following are what those options might look like for the 

approval of transportation shipping containers.

Option A: Continue the current approach 

The transportation program would continue its current approach of using consentbased 

standards. These standards are based on compatibility with DOT and IAEA regulations. Risk 

studies would be used to verify that current standards and guidance continue to protect against 

severe real-life accidents. Risk studies also could be used to influence the future outcome 

of IAEA deliberations on proposed changes in international regulations. Risk concepts could 

be used to justify regulations different from IAEA’s for domestic use only, such as regulations 

dealing with domestic storage and transportation of high burnup fuel.
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4.8-6 | Risk Management Task Force Section 4.8: Transportation

Option B: Implement the proposed risk management regulatory framework through selected 

guidance and rule changes

This option would emphasize specific guidance changes to implement the proposed risk 

management regulatory framework contained in the proposed Commission Policy Statement. 

This would include substantive outreach to industry, State, and local governments and the 

public. It might involve the formation of a working group to determine where risk management 

might offer the greatest opportunities. It would focus on using risk information to develop a 

more flexible approach toward implementing and making gradual changes to international 

standards and, eventually, to domestic regulations. It could be based on accepting the current 

version of the regulations as a consensual baseline and using risk information and safety goals 

to screen future revisions.

Option C: Implement the proposed risk management regulatory framework through broad-scale 

regulatory framework changes

This option would involve a fundamental revision of the basic framework and regulations 

used for licensing and oversight of transportation package approvals. This effort would 

require quantitative risk assessments to support such an approach. Significant outreach to 

stakeholders would also be an important aspect of this option. The greatest challenge would 

be to overcome resistance to abandoning a worldwide consent-based system perceived to work 

fairly well and that has an outstanding safety record.

The RMTF recommends Option B. The transportation regulations are unique in their 

international conformance. As such, changes to the regulations need to be done in a manner 

consistent with IAEA standards and those of member states. Option B would allow the NRC 

to focus its risk management efforts, in the near term, on improving the guidance used in 

implementation of the current transportation regulations. In the long term, it would allow the 

NRC to apply risk insights to international standards development. Option B provides a way to 

transition the agency to the proposed risk management regulatory framework and process in a 

reasonable amount of time and reflects ongoing operational priorities and resource constraints. 

It will also ensure continuity and stability in the regulated community through the continued use 

of existing regulations while implementing the proposed risk management regulatory framework 

and process to improve the transportation guidance.

4.8.4 References

(AEC, 1972) U.S. Atomic Energy Commission, “Environmental Survey of 

Transportation of Radioactive Materials to and from Nuclear 

Power Plants,” WASH-1238, December 1972.

(IAEA, 1961) International Atomic Energy Agency, “Regulations for the Safe 

Transport of Radioactive Materials,” Safety Series No. 6, 1961.

(IAEA, 1964) International Atomic Energy Agency, “Regulations for the Safe 

Transport of Radioactive Materials,” Safety Series No. 6, Revised 

Edition, 1964.
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(NRC, 1977) U.S. Nuclear Regulatory Commission, “Final Environmental 

Statement on the Transportation of Radioactive Material by Air and 

Other Modes,” NUREG-0710, December 1977.

(NRC, 1979) U.S. Nuclear Regulatory Commission, “Transportation of 

Radioactive Materials, Memorandum of Understanding,” published 

in the Federal Register on July 2, 1979, (44 FR 38690).

(NRC, 1988) U.S. Nuclear Regulatory Commission, “Shipping Container 

Response to Severe Highway and Railway Accident Conditions, 

NUREG/CR-4829,” Lawrence Livermore National Laboratory, 

October 1988, Agencywide Documents Access and Management 

System (ADAMS) Accession No. ML070810403.

(NRC, 2000) U.S. Nuclear Regulatory Commission, “Reexamination of 

Spent Fuel Shipments Risk Estimates,” NUREG/CR-6672, 

Sandia National Laboratory, February 2000, ADAMS Accession 

No. ML003698324.

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A Proposed Risk Management Regulatory Framework

5. CONCLUSIONS

The NRC has made progress in its efforts to implement risk-informed and performance-based 

approaches into its regulation of the various uses of byproduct, source, and special nuclear 

materials. Nevertheless, it is necessary to re-assess that progress and the underlying strategic 

vision from time to time.

The Risk Management Task Force (RMTF) has found that the NRC’s programs do not require 

radical or revolutionary changes, but they could benefit from continuing the evolution that has 

been embraced throughout the agency’s history. To that end, a Risk Management Regulatory 

Framework, as depicted in Figure 5-1, is being recommended as the next logical step for the 

NRC. This proposed framework uses a disciplined risk management process to identify and 

evaluate issues and make decisions about appropriate defense-in-depth protections for various 

radiological hazards. The risk-informed and performance-based defense-in-depth protections 

provide sufficient barriers, controls, and personnel to prevent, contain, and mitigate the 

exposure of workers or the public to radioactive materials. The appropriate barriers, controls, 

and personnel are based on the hazards present, the relevant scenarios leading to possible 

exposures, and the associated uncertainties to ensure that the risks resulting from the failure of 

some or all of the established barriers are maintained acceptably low.

Figure 5-1 A Proposed Risk Management Regulatory Framework 
A Proposed Risk Management Regulatory Framework

5-2 | Risk Management Task Force Chapter 5: Conclusions

In addition to the findings and recommendations in Chapter 4, the RMTF proposed options 

for implementation. Generally, the RMTF recommends an approach of systematically 

implementing the risk management framework through guidance and rule changes. This 

option is an evolutionary one that focuses on specific guidance and rule changes, as well as 

related activities (such as training) to implement a proposed Risk Management Regulatory 

Framework that would be rooted in a Commission Policy Statement. The RMTF recognizes the 

considerable work the NRC has done in risk-informing its programs as well as the need to 

do more. 

The first step in implementation of the recommendations in this report, if adopted, would 

be to initiate a thorough stakeholder engagement process. The process would begin 

with the issuance of a draft Commission Policy Statement on risk management for review 

and comments. Interactions on the draft policy statement could be supported by a series 

of workshops similar to those held as part of the Commission’s Safety Culture policy 

statement development. As reflected in the summary of comments provided in Appendix I, 

several stakeholders stated a strong preference for workshops as a way to better 

understand and help shape Commission direction in implementing the framework. The 

workshops would afford licensees, nongovernmental organizations, industry groups, States, 

and the public at large an opportunity for early involvement in shaping the agency’s risk 

management approach. 

The results of this stakeholder engagement would be considered by the NRC staff in 

preparing a final Commission Policy Statement on risk management and developing a longterm action plan to shape future direction of implementation. The staff would also need to 

consider existing commitments or initiatives, budget and resource constraints, and impacts to 

licensees and other stakeholders. 

During the period needed to formulate the proposed policy statement, the NRC’s regulatory 

practices will be changing, as will the designs and operations of U.S. power reactors, as a 

result of the lessons learned from the Fukushima Dai-ichi accident in Japan. It is important 

that these shorter-term actions are compatible with the longer-term changes envisioned by the 

Commission, whether they are related to the RMTF recommendations or a different course 

of action. Consideration of the proposed design-enhancement category for power reactors is 

particularly important in this regard.

None of this will be easy, as the agency’s experience with past risk-informed initiatives has 

demonstrated. However, subsequent NRC and industry experience with those initiatives, 

such as the Maintenance Rule and Regulatory Guide 1.174, and the Phase II Byproduct 

Material Review, has shown their net positive value.

Challenges that would have to be addressed to implement the vision proposed in this report 

are broadly summarized below:

•	 A change would be required within the agency and externally to increase 

understanding of the value and use of risk concepts and risk management 

language.

The proposed framework involves change for the agency, its licensees, and other 

stakeholders. While it is considered to be evolutionary, it is nonetheless a change that 

would involve agency and licensee programs and resources. Accordingly, the NRC 
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A Proposed Risk Management Regulatory Framework

would need to assure buy-in across the agency, develop a communications strategy, 

and prepare key messages. The proposed stakeholder workshops would be an 

important part of that communication strategy.

•	 The proposed risk-informed and performance-based concept of defense-in-depth may 

require the development of additional decision metrics and numerical guidelines.

The NRC has developed a working set of risk-informed decision metrics for power 

reactors. Examples include measures on an overall acceptable level of risk (safety 

goals) and guidance to the NRC staff on addressing incremental risk increases. Similar 

metrics do not exist in other program areas, but they may be needed to fully implement 

the proposed defense-in-depth framework.

•	 The approach would likely require developing new or revised risk-assessment 

consensus codes and standards.

The increased use of consensus standards is a Governmentwide goal intended to 

improve the efficiency of agency decisionmaking. The practical difficulties in achieving 

this goal have been well demonstrated as NRC has worked to implement risk 

assessment standards for power reactor applications.

•	 Consideration of cost in the proposed design-enhancement category in the power 

reactor regulatory program would necessitate a reconsideration of the agency’s tools 

for performing cost-benefit analysis.

Under some implementation alternatives considered by the RMTF, cost-benefit analyses 

would be performed (perhaps by licensees) on a potentially larger set of accident 

scenarios and on a more frequent (periodic) basis. The current set of tools used by 

agency staff for cost-benefit analyses was not designed with such uses in mind and 

would require improvements to make them easier to use, if such alternatives were 

chosen. In addition, the NRC staff is currently assessing possible updates to the actual 

criteria and factors incorporated into the NRC’s cost-benefit analyses. 

•	 A long-term commitment from the Commission and senior agency management 

would be required for implementation.

Development and issuance of a Commission Policy Statement on risk management 

would signal the commitment of the agency to the proposed framework. However, 

transitioning to a risk-management framework would take time. To successfully 

transition to this framework, the NRC would need to continue finding ways to bring risk 

concepts into the daily activities of the staff and investigate how risk analysis works 

in concert with the traditional deterministic analysis, such as relying on engineering 

margins, to achieve the risk management goal. This would begin with a top-down 

commitment to the concept, would be reinforced with training, and would be made a part 

of routine practices. Development of guidance, agencywide training, and a consistent 

emphasis from the highest levels of the agency to the risk management framework are 

key activities that the agency would need to undertake. 
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5-4 | Risk Management Task Force Chapter 5: Conclusions

Despite these challenges, there are substantial benefits to be realized by transitioning to the risk 

management framework proposed in this report. These include:

•	 Updated knowledge from contemporary studies, such as risk analysis, would be 

incorporated into the regulations and guidance, thereby improving their realism and 

technical basis.

While stability in a regulatory authority is generally considered to be a positive 

attribute, that authority also has a responsibility to examine new information and, when 

appropriate, make changes. Important aspects of the NRC’s regulatory practices were 

established more than 30 years ago, before an extensive base of operating experience 

and methods existed and risk assessments were developed and successfully applied. 

The proposed Risk Management Regulatory Framework and the program-area-specific 

recommendations are intended to build upon the successes of the past 30 years while 

improving the efficiency and effectiveness of the NRC’s regulatory program.

•	 Implementation of a systematic approach would foster a consistent regulatory 

decisionmaking process throughout the agency and improve resource allocation.

The proposed Risk Management Regulatory Framework and process are intended to 

be used universally for NRC actions involving the regulation and oversight of licensed 

activities. They take into account that the risk environment varies substantially across 

the wide variety of uses of radioactive materials that the NRC regulates. This common 

process would be used to identify and evaluate issues, make decisions, and allocate the 

appropriate amount of resources to implement actions to manage these diverse risks. 

While the details of licensing decisions, inspections, and other regulatory actions would 

continue to be different, a consistent decisionmaking process would be used.

•	 Consistency in language and communication would be improved across the agency 

and externally.

Several stakeholders have noted there is not a common understanding and use of terms 

such as risk-informed, performance-based, and defense-in-depth within the NRC and by 

external stakeholders. This makes effective communication a challenge. By focusing on 

the common use and applicability of these terms across the agency and the difference 

in approaches that various activities entail, the proposed framework would help improve 

consistency and understanding of the agency’s internal and external communications.

•	 Support of issue resolution would be achieved in a systematic, consistent, and 

efficient manner.

The proposed Risk Management Regulatory Framework provides a vehicle to 

systematically integrate relevant recent information into the decisionmaking process. 

Under this framework, a common decisionmaking process and the risk management 

goal for risk-informed and performance-based defense-in-depth would support 

developing program-specific criteria for the identification and disposition of issues. This 

common approach, with specific guidance for different types of licensees, could help 

ensure a more systematic, consistent, and efficient resolution of issues across the 

NRC’s regulatory programs.
Chapter 5: Conclusions Risk Management Task Force | 5-5

A Proposed Risk Management Regulatory Framework

•	 The design-enhancement category proposed for the power reactor regulatory 

program would clarify the attributes of all requirements established as substantial 

safety (beyond-design-basis) improvements. This approach may contribute to the 

resolution of the “patchwork” issue identified by the Fukushima Near-Term 

Task Force.

It should not be surprising that requirements, such as the rules for station blackout 

and aircraft impact, which emerged at different times and were considered 

“additional” protection, were not established with a consistent approach. The RMTF 

recommendation for a design enhancement category is intended to remedy this 

“patchwork” approach. Although there would be some costs incurred to establish this 

rule, these costs would be more than balanced by the long-term benefit of having a 

clear, consistent description of treatment for beyond-design-basis events and accidents.

To a certain degree, some resistance to change is natural and perhaps even desirable for the 

NRC to maintain a clear and stable regulatory environment. In addition, the ongoing activities 

within the agency and limited resources can impede the development and implementation 

of acknowledged process improvements. However, a patchwork of regulatory requirements 

has been created as a result of addressing problems on a case-by-case basis over many 

years. The RMTF has concluded that cost-effective changes are possible and recommends 

they be undertaken in a holistic manner across the NRC’s programs. With this in mind, the 

implementation of the proposed Risk Management Regulatory Framework can be pursued in a 

planned and deliberate manner so that it does not disrupt the NRC’s mission, but ensures that 

the NRC continues to improve on how it protects the public health and safety, promotes the 

common defense and security, and protects the environment.

Appendix A Risk Management Task Force | A-1

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APPENDIX A

RISK MANAGEMENT SYSTEMS

A.1 Introduction

As discussed in Chapters 2 and 3 of the report, “A Proposed Risk Management Regulatory 

Framework,” the objective of the U.S. Nuclear Regulatory Commission (NRC) is to manage 

the risks from the use of byproduct, source, and special nuclear materials by establishing and 

maintaining appropriate performance-based regulatory controls and oversight. In its simplest 

form, this objective is met by ensuring that sufficient barriers (supported by associated controls 

and personnel) are placed between radioactive materials and members of the workforce and 

public to minimize the risks of adverse effects on health and safety. These barriers, controls, 

and personnel must provide protection during routine operations, as well as for challenges such 

as equipment failures and external hazards (e.g., fires and floods). The NRC has longstanding 

positions and guidance related to the consideration of risk-insights and performance-based 

concepts in its decisionmaking processes. As discussed in a March 1999 (NRC, 1999) 

guidance document issued by the Commission, the NRC’s risk definition takes the view that 

when one asks, “What is the risk?” one is really asking three questions: 

•	 What can go wrong? 

•	 How likely is it?

•	 What are the consequences? 

These three questions can be referred to as the “risk triplet” (Kaplan and Garrick, 1981). In 

simple terms, when one deals with risk, one must consider both consequences and likelihoods. 

The traditional definition of risk—that is, probability times consequences—is fully embraced by 

the “triplet” characterization of risk, although it is more limited. The consequences of concern 

to the NRC are related to the use of byproduct, source, and special nuclear materials and to the 

possible exposure of workers or the public to radiation from the normal use of those materials 

or as a result of accidents. Although the NRC’s regulatory programs range from the control 

of simple sealed sources to complex nuclear power plants, the RMTF described in the main 

body of this report the way in which most of the agency’s regulatory decisions involve asking 

the above questions and determining an appropriate course of action based on the answers 

derived from analyses or judgment. The RMTF defined a risk-informed and performance-based 

defense-in-depth concept to serve as the risk management goal for NRC-regulated activities. In 

addition, a move to a common, structured decisionmaking model would ensure consistency and 

improved communications throughout the NRC’s regulatory programs.

Before focusing on risk management as the logical evolution of the NRC’s current riskinformed, performance-based approach to regulation, the task force considered a number of 

approaches to accomplishing its charter. In doing this, the RMTF considered the agency’s 

historic approach to consideration of risk, researched and reviewed literature across the field 

of risk, and held discussions with internal and external stakeholders to develop options for a 

more comprehensive and holistic approach for its regulatory programs. These included: 1) no 

action, 2) a purely deterministic approach, 3) a single risk-based numerical criterion, and 4) an 

approach based on the concept of risk management. 
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The no-action option would continue the agency’s current risk-informed, performance-based 

approach. This approach has been used successfully since the late 1990s to greater or lesser 

degrees in the various regulatory programs of the NRC. Its attributes are by and large known 

to the regulated community and are therefore predictable. The approach has provided the 

NRC the ability to make its regulatory decisionmaking more systematic, more objective, more 

consistent, and more transparent. However, it does not position the agency to address the 

challenges of the future, specifically the prospect of flat or declining budgets for most, if not 

all, Federal agencies. Nor does it specifically address the direction of Executive Order 13563, 

“Improving Regulation and Regulatory Review” (Exec. Order, 2011).

The second option considered by the RMTF was a purely deterministic one. Under this 

approach, the agency would discontinue its risk-informed, performance-based approach and 

return, over time, to deterministicbased regulations and guidance for its regulatory programs. 

Risk insights and performance considerations would no longer be considered in developing 

regulatory standards and guidance. This approach would arguably be clearer and easier 

to comply with than standards that allow licensees to choose one of a number of ways to 

meet a performance-based regulation. The approach would also be less flexible than the 

current approach, not allowing for site or facility-specific considerations, and would represent 

a departure from the trend of international standardssetting approaches, which have an 

increasingly robust risk basis. Finally, such an approach would more likely than not drive up the 

costs of regulatory compliance with no compelling increase in safety or security. 

The third option considered by the task force was the concept of a risk-based numerical criterion 

to be used across all agency regulatory programs. This approach is used by some other 

countries and is attractive from the prospect of providing a single unifying basis for protection 

of public health and safety and the environment. It could also be useful in communicating to 

the public the primary basis for the NRC’s various regulatory programs. However, it would not 

take into account the wide variety of hazards posed by various uses of nuclear materials and 

could incur unnecessary and perhaps unacceptable costs to the regulated community. It would 

also represent a significant change in the NRC’s approach to regulation that would necessitate 

an overhaul of the regulations and guidance for reactors, materials, fuel cycle, waste, and 

transportation programs at a substantial cost to the agency.

The fourth option, risk management, is being widely used in various sectors, including 

Government agencies, financial institutions, and technology companies, to address the kinds 

of challenges the NRC faces and that the RMTF was tasked to address. Risk management 

allows for various approaches to address risks as part of the decisionmaking process, 

including the use of both quantitative and qualitative tools, which is essential in the broad 

range of NRC regulatory programs. It represents a logical evolution from the risk-informed, 

performance-based philosophy that has governed NRC activities for many years. It may also 

provide program managers with a more systematic approach to resource allocation, whether 

in budget formulation, response to events, or licensing decisions. The RMTF believes that risk 

management offers the potential for an improved regulatory framework, which is described 

generally and for specific regulatory programs in other parts of this report. 

Given the NRC’s mission and the fact that most regulatory decisions involve an evaluation 

of and control of the risks posed by the use of radioactive materials, the RMTF recommends 

(Recommendation 2.2) that “risk management” be used as the framework for a comprehensive 

and holistic risk-informed, performance-based regulatory approach for reactors, materials, 

waste, fuel cycle, and transportation. Risk management concepts, process descriptions, and 
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decisionmaking models are widely used in various sectors, including Government agencies, 

financial institutions, and technology companies. The U.S. Department of Homeland Security 

Risk Lexicon (DHS, 2010) defines risk management as follows:

Risk management is the process for identifying, 

analyzing, and communicating risk and accepting, avoiding, 

transferring, or controlling it to an acceptable level considering 

associated costs and benefits of any actions taken.

A.2 Risk Management Frameworks

There are many general descriptions of risk management processes or methodologies 

that include similar key points. A review of many of the frameworks and processes for risk 

management reveals that most share common steps or phases. Not surprisingly, these risk 

management frameworks are a variation of traditional decisionmaking models. The steps to a 

basic decisionmaking model are as follows:

•	 define the problem and the desired outcome 

•	 research and identify options 

•	 analyze alternatives

•	 make a decision, (i.e., choose an alternative) 

•	 implement the decision

•	 monitor the results 

Descriptions of several risk management frameworks and how they support the decisionmaking 

process are described in the following sections.

A.2.1 Framework for Environmental Risk Management

 The Presidential/Congressional Commission on Risk Assessment and 

Risk Management; Final Report 1997

In the 1990 Clean Air Act Amendments, Congress mandated that a Commission on Risk 

Assessment and Risk Management be formed to:

...make a full investigation of the policy implications and 

appropriate uses of risk assessment and risk management in 

regulatory programs under various Federal laws to prevent 

cancer and other chronic human health effects which may 

result from exposure to hazardous substances.
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A-4 | Risk Management Task Force Appendix A

The Commission, consisting of specialists in various fields of science and public policy, was 

assembled in May 1994. Following its deliberations and interactions with various stakeholders, 

the Commission issued its final report in 1997, which included a framework for incorporating 

risk management concepts into the decisionmaking processes for the U.S. Environmental 

Protection Agency (EPA) (Commission, 1997). The Commission’s Framework defines a clear, 

sixstage process for risk management that can be scaled to the importance of a public health or 

environmental problem and that:

•	 Enables risk managers to address multiple relevant contaminants, sources, and 

pathways of exposure so that threats to public health and the environment can be 

evaluated more comprehensively than is possible when only single chemicals in single 

environmental media are addressed.

•	 Engages stakeholders as active partners so that different technical perspectives, public 

values, perceptions, and ethics are considered.

•	 Allows for incorporation of important new information that may emerge at any stage of 

the risk management process.

The basic structure of the framework developed by the Presidential/Congressional Commission 

is shown in Figure A-1.

Figure A-1 Framework for Environmental Health Risk Management 

The Presidential/Congressional Commission on Risk 

Assessment and Risk Management, Final Report 1997 

Figure A-1 Framework for Environmental Health Risk Management

The Presidential/Congressional Commission on Risk

Assessment and Risk Management. Final Report 1997
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A.2.2 Understanding Risk: Informing Decisions in a Democratic Society

National Research Council, 1996

The National Resource Council has performed several studies related to risk assessments and 

risk management as they apply to various government functions and public health concerns. In 

1996, the Council documented (National Research Council, 1996) a study it had undertaken to 

address the following task statement:

“Risk characterization” is a complex and often controversial 

activity that is both a product of analysis and dependent 

on the processes of defining and conducting analysis. The 

study committee will assess opportunities to improve the 

characterization of risk so as to better inform Decisionmaking 

and resolution of controversies over risk. The study will 

address: technical issues such as the representation of 

uncertainty; issues relating to translating the outputs of 

conventional risk analysis into non-technical language; and 

social, behavioral, economic, and ethical aspects of risk that 

are relevant to the content or process of risk characterization.”

The report offers seven principles to increase the likelihood of achieving sound and acceptable 

decisions. These principles are as follows:

1. Risk characterization should be a decisiondriven activity, directed toward informing 

choices and solving problems.

2. Coping with a risk situation requires a broad understanding of the relevant losses, 

harms, or consequences to the interested and affected parties.

3. Risk characterization is the outcome of an analyticdeliberative process. Its success 

depends critically on systematic analysis that is appropriate to the problem, responds to 

the needs of the interested and affected parties, and treats uncertainties of importance to 

the decision problem in a comprehensible way. Success also depends on deliberations 

that formulate the decision problem, guide analysis to improve decision participants’ 

understanding, seek the meaning of analytic findings and uncertainties, and improve 

the ability of interested and affected parties to participate effectively in the risk decision 

process. The process must have an appropriately diverse participation or representation 

of the spectrum of interested and affected parties, of decisionmakers, and of specialists 

in risk analysis at each step.

4. The analytic-deliberative process leading to a risk characterization should include early 

and explicit attention to problem formulation; representation of the spectrum of interested 

and affected parties at this early stage is imperative. 

5. The analytic-deliberative process should be mutual and recursive. Analysis and 

deliberation are complementary and must be integrated throughout the process leading 
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A-6 | Risk Management Task Force Appendix A

to risk characterization: Deliberation frames analysis, analysis informs deliberation, and 

the process benefits from feedback between the two.

6. Those responsible for a risk characterization should begin by developing a provisional 

diagnosis of the decision situation so that they can better match the analyticdeliberative 

process leading to the characterization to the needs of the decision, particularly in terms 

of level and intensity of effort and representation of parties.

7. Each organization responsible for making risk decisions should work to build 

organizational capability to conform to the principles of sound risk characterization. At 

a minimum, an organization should pay attention to organizational changes and staff 

training efforts that might be required, to ways of improving practice by learning from 

experience, and to both costs and benefits in terms of the organization’s mission 

and budget.

The Council’s report provided the following schematic representation of a risk decision process:

A.2.3 Risk Governance Framework 

International Risk Governance Council

 Risk and Regulatory Policy – Improving the Governance of Risk

OECD Reviews of Regulatory Reform, 2010

The International Risk Governance Council (IRGC) is a private, independent, notforprofit 

foundation based in Geneva, Switzerland. It was founded in 2003 with a mission to support 

governments, industry, nongovernmental organizations, and other organizations in their 

efforts to deal with major and global risks facing society and to foster public confidence in risk 

governance. The IRGC was established to address widespread concern within the public 

sector, the corporate world, academia, the media, and society at large that the complexity 

and interdependence of an increasingly large number of risk issues was making it ever more 

difficult for risk managers to develop and implement adequate risk governance strategies. 

Figure A-2 Understanding Risk: Informing Decisions in a Democratic 

Society, National Resource Council, 1996 

Figure A-2 Understanding Risk: Informing Decisions in a Democratic

Society, National Resource Council, 1996
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They produced a white paper, entitled “Risk Governance – Towards an Integrative Approach” 

(IRGC, 2005) to describe a framework for an integrated, holistic, and structured approach for 

improving the ways risk is identified, assessed, managed, monitored, and communicated. The 

IRGC framework is depicted in Figure A-3.

The IRGC risk governance framework and its sequence of pre-assessment, risk appraisal, 

risk characterization, risk evaluation and risk management is also discussed in a report by 

the Organization for Economic Cooperation and Development (OECD). The report, entitled 

“Risk and Regulatory Policy – Improving the Governance of Risk” (OECD, 2010) is one of a 

series of OECD reports on regulatory reforms. A premise of the report is that there is a gap 

between the level of risk that is aspired to by policymakers and the level that is achievable 

through regulation. In addition, the report acknowledges that since not all risks can be reduced 

to zero, tradeoffs in risk reduction measures are inevitable. The OECD studied areas for the 

improvement of risk governance through an analysis of the legal, procedural, and practical 

challenges for risk regulation.

A.2.4 Strengthening the Use of Risk Management Principles in Homeland

Security; Government Accountability Office, June 2008

Risk Management: Further Refinements Needed to Assess Risks and

Prioritize Protective Measures at Ports and Other Critical Infrastructure;

Government Accountability Office, December 2005

The U.S. Government Accountability Office (GAO) convened a forum of 25 national and 

international experts on October 25, 2007, to advance a national dialogue on applying risk 

management to homeland security. Participants included Federal, State, and local officials and 

risk management experts from the private sector and academia. Forum participants identified 

(1) what they considered to be effective risk management practices used by organizations from 

Figure A-3 International Risk Governance Council (IRGC) 

Risk Governance Framework 

 Figure A-3 International Risk Governance Council (IRGC)

Risk Governance Framework
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A-8 | Risk Management Task Force Appendix A

the private and public sectors, and (2) key challenges to applying risk management to homeland 

security, and actions that could be taken to address them. 

The report (GAO, 2008) refers to the risk management framework developed during a previous 

GAO activity on assessing risks related to ports and other critical infrastructure (GAO, 2005). 

The framework was prepared based on industry best practices and other criteria. This 

framework, shown in Figure A-4, divides risk management into five major phases: (1) setting 

strategic goals and objectives and determining constraints, (2) assessing risks, (3) evaluating 

alternatives for addressing these risks, (4) selecting the appropriate alternatives, and 

(5) implementing the alternatives and monitoring the progress made and results achieved.

A.2.5 National Aeronautics and Space Administration

Risk-Informed Decisionmaking Handbook, 2010

The preface to the U.S. National Aeronautics and Space Administration (NASA) handbook 

observes the following:

The NASA risk-informed decisionmaking (RIDM) handbook (NASA, 2010) was prepared to 

provide a framework and guidance for risk-informed decisionmaking at that agency. RIDM at 

NASA is intended to ensure decisions between alternatives are made with an awareness of the 

Risk management (RM) is an integral aspect of virtually every challenging 

human endeavor, but well-defined RM processes have only recently 

begun to be developed and implemented as an integral part of systems 

engineering at NASA, given the complex concepts that RM encapsulates 

and the many forms it can take. However, few will disagree that effective 

risk management is critical to program and project success. 

Figure A-4 GAO Risk Management Framework
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risks associated with each, thereby helping to prevent late design changes, which can be key 

drivers of risk, cost overruns, schedule delays, and cancellation. The framework is 

shown below:

A.2.6 Commandant Instruction 3500.3

Operational Risk Management, U.S. Coast Guard

In response to various mishaps and related investigations, the U.S. Coast Guard prepared 

training material to emphasize risk management principles that outline a systematic process 

to continuously assess and manage risks: the operational risk management (ORM) process 

(USCG, 1999). The ORM process is a continuous, systematic process of identifying and 

controlling risks in all activities according to a set of preconceived parameters by applying 

appropriate management policies and procedures. This process includes detecting hazards, 

assessing risks, and implementing and monitoring risk controls to support effective, riskbased decisionmaking. The ORM process can be represented by the seven steps shown in 

Figure A-6.

Figure A-5 NASA Framework
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A.2.7 International Organization for Standardization, Standard 31000, 

“Risk Management – Principles and Guidelines”

The International Organization for Standardization (ISO) is a worldwide federation of national 

standards bodies. ISO prepared its 31000 standard, “Risk Management – Principles and 

Guidelines” (ISO, 2009) and supporting standards in recognition that all types of organizations 

need to address factors and uncertainties that challenge their objectives (i.e., risks). The 

standard defines various principles regarding the integration of risk management into 

organizations’ processes and the expectations for risk management evaluations. The consistent 

use of risk management requires a management framework to embed, support, and improve the 

processes throughout the organization. The implementation of a risk management approach 

is, in turn, supported by a structured process that includes establishing the context (e.g., the 

specific objectives and decisions, internal and external factors to consider, and decisionmaking 

criteria), assessing the risks, selecting options, monitoring results and revising actions and 

processes, and communicating with various stakeholders. The relationship between the 

principles, framework, and risk management process are shown in the following figure. 

Figure A-6 U.S. Coast Guard Framework 

Figure A-6 U.S. Coast Guard Framework
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A.2.8 Department of Homeland Security, Risk Management Fundamentals

In response to The Policy of Integrated Risk Management, established by the Secretary of 

Homeland Security, the Office of Risk Management and Analysis at DHS published a guide, 

“Risk Management Fundamentals” (DHS, 2011), to promote a common understanding of and 

approach to risk management. Figure A-8 shows the DHS risk management process.

Figure A-7 ISO Framework 

Figure A-8 

Figure A-7 ISO Framework

Figure A-8 DHS Risk Management Process
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A-12 | Risk Management Task Force Appendix A

The DHS process is comprised of the following:

•	 Defining and framing the context of decisions and related goals and objectives. 

•	 Identifying the risks associated with the goals and objectives. 

•	 Analyzing and assessing the identified risks.

•	 Developing alternative actions for managing the risks and creating opportunities, 

and analyzing the costs and benefits of those alternatives.

•	 Making a decision among alternatives and implementing that decision.

•	 Monitoring the implemented decision and comparing observed and expected 

effects to help influence subsequent risk management alternatives and decisions. 

A discussion of DHS’s implementation of its Integrated Risk Management (IRM) Framework is 

provided below.

A.2.8.1 Implementation of Risk-Informed Security at the Department of Homeland Security

The Department of Homeland Security has begun the implementation of a risk management 

process similar to that recommended by the RMTF and described in the body of this report. The 

DHS process is guided by the concept of IRM, which has the following goals:

•	 Unify efforts among all homeland security partners to ensure that strategies and actions 

are informed by a common understanding of homeland security risk.

•	 Ensure that information and analysis about homeland security risks are incorporated into 

strategic and operational decisionmaking processes.

•	 Build a common understanding of risk management through development of a risk 

lexicon, risk-informed planning process, training, and standards of practice.

•	 Provide mechanisms to share risk data, risk assessments, and risk management 

decision support and analysis tools across the homeland security enterprise.

In alignment with the IRM goals, DHS has performed risk assessments at the strategic and 

operational level to make better use of risk information in departmental decisionmaking. 

The DHS approach to each of the components of the risk triplet (Kaplan and Garrick, 

1981) is summarized below based on the more detailed descriptions in “Risk Management 

Fundamentals” (DHS, 2011). Each summary is followed by specific examples of the concepts 

applied in current DHS risk assessments.
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What can go wrong?

In determining which risks are to be considered for a particular risk assessment, the DHS

Risk Management Fundamentals suggest focusing only on the risks that are relevant to the 

underlying decision. These can include strategic, operational, and institutional risks. They 

should generally be divided into scenarios that can be analyzed individually, covering the full 

scope of the decision context while avoiding overlaps.

National Risk Profile Development

As defined in the National Infrastructure Protection Plan (NIPP), the goal of infrastructure 

protection at DHS is to “build a safer, more secure, and more resilient America by preventing, 

deterring, neutralizing, or mitigating the effects of deliberate efforts by terrorists to destroy, 

incapacitate, or exploit elements of our Nation’s Critical Infrastructure and to strengthen national 

preparedness, timely response, and rapid recovery of critical infrastructure in the event of an 

attack, natural disaster, or other emergency (DHS, 2009).”

As part of the efforts to meet this goal, the DHS National Protection and Programs Directorate’s 

Office of Infrastructure Protection works with the critical infrastructure community to determine 

the risk to the Nation’s infrastructure from natural hazards and terrorism. The results of this 

discussion are used to create the National Risk Profile (“the Profile”), which “every year 

identifies the highest relative risks to critical infrastructure and those critical infrastructure 

sectors (18 total) at a higher risk from the greatest number of hazards. The Profile also identifies 

other risk management concerns, such as high-likelihood risks and low-likelihood/highconsequence infrastructure protection priorities (DHS, 2009).”

How likely is it?

The “Risk Management Fundamentals” (DHS, 2011) notes that common sources for frequencies 

and probabilities include historical records, models, simulations, and elicitations of subjectmatter experts. These frequencies and probabilities may be qualitative or quantitative with the 

choice to quantify risk information based on the needs of the decisionmaker.

Integrated Chemical, Biological, Radiological, and Nuclear Terrorism Risk Assessment 

Threat Estimation

The Science and Technology Directorate’s Integrated Chemical, Biological, Radiological, and 

Nuclear (CBRN) Terrorism Risk Assessment (ITRA) is conducted by DHS as directed under 

Homeland Security Presidential Directive 18 (HSPD-18) to provide a risk-based decision 

support tool to agencies across the Federal Government responsible for reducing and mitigating 

CBRN terrorism risk. To determine the frequencies associated with rare, highly uncertain 

terrorist attacks, the ITRA relies on the elicitation of subject-matter experts from the intelligence 

community (DHS, 2012).

Risk Assessment Process for Informed Decisionmaking (RAPID) Vulnerability Estimation

RAPID is a multihazard probabilistic risk assessment created to inform the DHS budget cycle. 

The assessment includes the calculation of scenario-dependent vulnerabilities to terrorist 

attacks given the current U.S. security alignment. RAPID uses structured surveys to determine 

how the different elements of the homeland security enterprise work together to prevent terrorist 
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attacks. The data are collected for each DHS budget program and input into fault trees, 

allowing for the calculation of the risk reduced by each program. This information can identify 

potential gaps in the security structure and compare programmatic risk reduction to help guide 

the Department’s strategic direction and influence the allocation of resources.

What are the consequences?

As with frequency and probability, consequences may also be qualitative or quantitative. 

Depending on the analysis, they include loss of life, injuries, economic impacts, psychological 

consequences, environmental degradation, and inability to execute essential missions. To 

determine the relevant consequences for an individual assessment, “Risk Management 

Fundamentals” (DHS, 2011) recommends using structured techniques, such as value 

focused thinking.

Strategic National Risk Assessment Consequences

The Strategic National Risk Assessment (SNRA), which was developed and implemented by 

DHS in support of Presidential Policy Directive 8 for National Preparedness, calls for national 

preparedness to be based on core capabilities that support “strengthening the security and 

resilience of the United States through systematic preparation for the threats that pose the 

greatest risk to the security of the Nation, including acts of terrorism, cyber attacks, pandemics, 

and catastrophic natural disasters.” 

The assessment examined the consequences associated with six categories of harm: loss 

of lives, injuries and illnesses, direct economic costs, and, for the first time at DHS, social 

displacement, psychological distress, and environmental impact. The SNRA drew upon input 

from the Federal interagency, including data and information from Government models and 

assessments, historical records, structured analysis, and judgments of experts from different 

disciplines (DHS, 2011a).

A.2.8.2 Risk in Regulation at DHS

Executive Order 12866, “Regulatory Planning and Review,” published in the Federal Register

on October 4, 1993 (58 FR 51735), stipulates that Federal agencies must make a reasoned 

determination that the benefits of an intended regulation justify its costs. In many DHS 

analyses, the primary benefit of a regulation is the reduction in the probability or consequence 

of terrorist attacks. In these regulations, the calculation of benefits can be determined by 

qualitative or quantitative risk reduction assessments.

U.S. Customs and Border Protection Break-even Analyses

U.S. Customs and Border Protection (CBP) has proposed several regulations involving the 

transit of people and goods across U.S. borders. The costs of the regulations are determined 

by examining the measures that travelers and businesses will take as a direct result of the 

proposed rules, estimating the costs of those activities, and adding the costs for the government 

to implement the regulation. In a typical cost-and-benefit analysis, the economic benefits would 

then be compared with the costs. For several CBP regulations, the regulatory benefits include 

the risk reduction associated with preventing certain types of terrorist attacks. With the difficulty 

associated with determining the probability of major terrorist attacks occurring, CBP had to forgo 

the calculation of benefits from the expected probability reduction and instead calculated the 
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reduction in probability of an attack necessary to balance the costs. This technique is called 

break-even analysis and has been used in several of CBP’s regulatory decisions, (IEI, 2008 

and 2008a). 

While the context of the DHS risk management process has fundamental differences with that 

of the NRC’s, the continuing experience of integrating risk management at DHS can provide 

important insights into the implementation of an agencywide risk management effort and 

specifically into the potential structure for risk-informing security threats. 

A.2.9 Center for Strategic and International Studies

The above sections provide short descriptions of risk management systems developed by 

organizations and other Federal agencies. There are countless other examples of such 

programs, and the use of risk management in both Government and private sector activities. A 

study entitled, “Risk Management in Non-DoD U.S. Government Agencies and the International 

Community: Best Practices and Lessons Learned” (CSIS, 2011), provides insights and case 

studies of risk management practices at various agencies (including the NRC) and within 

several countries. 

A.3 Recommended NRC Risk Management Framework

Drawing upon the insights from the various risk management frameworks described above, the 

RMTF adopted a very similar process for NRC decisionmaking. The steps of the process are 

similar to those defined by each of the above frameworks and are shown in Figure A-9.

The steps of this proposed risk management framework for the NRC include the following:

(1) Identification and framing (characterization) of an issue, proposal, or problem:

o Define the nature of the hazard.

o Identify the potential worker or public exposure.

o Determine the source of the issue or proposal (e.g., event, application).

Figure A-9 The Regulatory Decisionmaking Process
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(2) Identification of options and alternatives:

o Identify possible alternatives, including no action.

o Consider performance-based approaches.

o Prepare defense-in-depth proposals to be assessed against tolerability or 

acceptability criteria and other factors.

(3) Analysis:

o Select risk evaluation technique(s) given hazard, issue, and decisionmaking 

criteria (can include deterministic, risk assessments, judgment, etc.).

o Evaluate options related to appropriate barriers and controls given: 

•	 What can go wrong?” 

•	 “How likely is it?” and 

•	 “What are the consequences?”

(4) Deliberation (integrated decisionmaking):

o Evaluate options, considering uncertainties and acceptability criteria.

o Consider external factors.

o Determine appropriate risk-informed and performance-based defense-in-depth 

protection.

(5) Actions and implementation of selected option:

o Follow NRC processes.

•	 rulemaking, licensing, environmental reviews

•	 oversight

o Apply regulatory activities (Life Cycle)

•	 design, operations, decommissioning

(6) Monitoring and feedback:

o Gather information from oversight activities.

o Gather information from events, stakeholders, and other sources.

o Assess problems and identify corrective actions.
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(7) Communicate with stakeholders throughout the process:

o Gather information.

o Deliberate.

o Implement and solicit feedback.

Additional discussions related to the deliberative process are provided in Chapter 2 and 

Appendix B, “Risk Management – Analysis and Deliberation.”

A.4 Applicability

The risk management goal and related risk management process are intended to be used 

universally for NRC decisions regarding the regulation and oversight of licensed activities, 

including reactors, materials, waste, fuel cycle, and transportation. In many cases, the current 

NRC decisionmaking processes (e.g., integrated decisionmaking for nuclear reactors and riskinformed decisionmaking for materials licensees) are similar to the risk management approach 

described in this report. Major actions, such as licensing of reactor or fuel cycle facilities, 

agency level initiatives, and rulemakings, include project plans and processes that are often 

explained in terms of the six steps of the risk management process (or in equivalent terms and 

steps). The handling of some routine activities may at first glance appear simpler than the risk 

management process, but this is likely because some steps, such as selecting risk evaluation 

techniques and decisionmaking criteria, are already incorporated into procedures or guidance 

documents. Provided the existing guidance and processes ensure that appropriate barriers are 

established and risks are maintained at an acceptably low level, the move to a risk management 

framework is unlikely to require significant changes to existing practices. It is not expected that 

the risk management process will complicate matters such that staff or managers should avoid 

its adoption. Instead, it is expected that a comprehensive and holistic approach across NRC 

programs will provide long-term efficiencies, consistency, and other benefits.

Although beyond the immediate scope of the RMTF, the general concepts of risk management 

and methodical decisionmaking are also applicable to other agency activities. For example, 

a consistent risk management approach across NRC programs could support changes in 

management practices such that program managers would have increased flexibility to allocate 

resources to address risk or safety concerns. Insights from inspections, operating experience, 

scientific studies, or risk assessments could inform timely changes to programs and focus areas 

as part of the implementation of budgeted resources. Another example in which the NRC uses 

risk management processes is in the area of providing security to the agency’s 

information systems. 
A Proposed Risk Management Regulatory Framework

A-18 | Risk Management Task Force Appendix A

A.5 References

(Commission, 1997) The Presidential/Congressional Commission on Risk Assessment 

and Risk Management, “Framework for Environmental Health Risk 

Management,” Vols. 1 and 2, 1997.

(CSIS, 2011) Center for Strategic and International Studies, “Risk Management 

in Non-DoD U.S. Government Agencies and the International 

Community: Best Practices and Lessons Learned,” A Report of 

the CSIS Defense and National Security Group, March 2011.

(DHS, 2009) U.S. Department of Homeland Security, “National Infrastructure 

Protection Plan,” 2009.

(DHS, 2010) U.S. Department of Homeland Security, “DHS Risk Lexicon,” 

September 2010.

(DHS, 2011) U.S. Department of Homeland Security, “Risk Management 

Fundamentals,” November 2011.

(DHS, 2011a) U.S. Department of Homeland Security, “The Strategic National 

Risk Assessment in Support of PP 8: A Comprehensive RiskBased Approach toward a Secure and Resilient Nation,” 2011.

(DHS, 2012) U.S. Department of Homeland Security, “Companion to the 

Integrated CBRN Terrorism Risk Assessment (ITRA),” 2012.

(Exec. Order, 2011) Executive Order 13563, “Improving Regulation and Regulatory 

Review,” January 18, 2011, published in the Federal Register on 

January 21, 2011 (76 FR 821).

(GAO, 2005) U.S. Government Accountability Office, “Risk Management: 

Further Refinements Needed to Assess Risks and Prioritize 

Protective Measures at Ports and Other Critical Infrastructure,” 

December 2005.

(GAO, 2008) U.S. Government Accountability Office, “Strengthening the Use of 

Risk Management Principles in Homeland Security,” June 2008.

(IEI, 2008) Industrial Economics, Inc. “The Western Hemisphere Travel 

Initiative Implemented in the Land Environment,” 2008.

(IEI, 2008a) Industrial Economics, Inc., “Importer Security Filing and Additional 

Carrier Requirements,” 2008.

(IRGC, 2005) International Risk Governance Council, “White Paper on 

Risk Governance – Towards an Integrative Approach,” 

September 2005.
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A Proposed Risk Management Regulatory Framework

(ISO, 2009) International Organization for Standardization, “Risk 

Management – Principles and Guidelines,” ISO-31000, 2009.

(Kaplan and Garrick, 1981) S. Kaplan and B.J. Garrick, “On the Quantitative Definition of 

Risk,” Risk Analysis, 1:11–27.

(National Research National Resource Council, “Understanding Risk: Informing 

Council, 1996) Decisions in a Democratic Society,” National Academy 

Press, 1996.

(NASA, 2010) U.S. National Aeronautics and Space Administration, “RiskInformed Decisionmaking Handbook,” NASA/SP2010-576, 

Version 1.0, April 2010.

(NRC, 1999) U.S. Nuclear Regulatory Commission, Staff Requirements 

Memorandum Regarding SECY-98-144, “White 

Paper on Risk-informed and Performance-Based 

Regulation,” March 1, 1999, Agencywide Documents 

Access and Management System (ADAMS) Accession 

No. ML003753601.

(OECD, 2010) Organization for Economic Cooperation and Development, “Risk 

and Regulatory Policy – Improving the Governance of Risk,” 2010.

(USCG, 1999) U.S. Coast Guard, “Operational Risk Management,” Commandant 

Instruction 3500.3, November 1999.

Appendix B Risk Management Task Force | B-1

A Proposed Risk Management Regulatory Framework

APPENDIX B

RISK MANAGEMENT – ANALYSES AND DELIBERATION

B.1 Introduction

As discussed in other sections of this report, the risk management framework shown in 

Figure B-1 is primarily an approach to provide structure and logic to the decisionmaking 

process. Major elements of the process are the technical analyses and the deliberations to 

determine what risk-informed, performance-based, defense-in-depth protections are appropriate 

for a given radiological hazard, relevant scenarios, and the associated uncertainties. 

B.2 Analysis

A key input into the deliberations within a risk management program is the technical analysis 

or actual evaluation of risks associated with the subject system or activity. A challenge for 

introducing risk management at the NRC will be the terminology associated with the “analyze” 

step in the process and how risk assessments and the “traditional engineering analyses” have 

been used in the licensing of most materials, devices, and facilities. The efforts to improve 

the use of risk insights in NRC processes were sometimes interpreted as being separate and 

distinct from previous engineering and licensing activities. An important part of developing a 

risk management framework at the NRC will be communicating that the traditional engineering 

analyses and related acceptance criteria are actually part of risk evaluation and management of 

risk through establishing certain conventions and standards. NRC regulatory programs for all 

of the various materials, devices, and facilities have been based, either explicitly or implicitly, on 

identifying radiological hazards and asking:

•	 What can go wrong? 

•	 How likely is it?

•	 What are the consequences? 

These questions are commonly referred to as the “risk triplet” (Kaplan and Garrick, 1981). In 

some traditional approaches, the “what can go wrong” question was addressed by identifying 

certain events or conditions (e.g., maximum credible accidents, design-basis events) and 

Figure B-1 The Regulatory Decisionmaking Process
A Proposed Risk Management Regulatory Framework

B-2 | Risk Management Task Force Appendix B

evaluating structures or systems using acceptance criteria contained in NRC regulations, 

guidance, and industry codes and standards. Establishing the scenarios and acceptance 

criteria considered at least a subjective estimation of the “how likely is it” question. The “what 

are the consequences” question also considered the likelihood of an event or condition in 

establishing the acceptance criteria, whether it was an actual dose limit or a limit on a system, 

structure, or a component that helped to mitigate an accident or contain radioactive material 

(see Appendix F for reactor-related examples). So, while the Risk Management Task Force 

(RMTF) does not see a move to a risk management system as necessarily at odds with existing 

NRC regulatory programs, it foresees communication issues with both internal and external 

stakeholders regarding some aspects of the proposed regulatory framework and the 

related terminology.

As presented in the main body of this report, the RMTF has found that a risk management 

framework is a logical way to fulfill the NRC’s mission. Figure 2-2 (repeated below) shows the 

transition or flow from the mission to ensure adequate protection of public health and safety to a 

framework to manage risks to workers and the public from the use of radioactive materials. 

Figure B-2 A Proposed Risk Management Regulatory Framework
Appendix B Risk Management Task Force | B-3

A Proposed Risk Management Regulatory Framework

Within this framework, a risk-informed and performance-based approach is not an alternative 

to defense-in-depth concepts, but instead is simply a means to determine, in concert with 

other techniques, an appropriate level of protection to prevent, contain, and mitigate possible 

releases of radioactive material from NRC-licensed activities. The selection of a technique or 

combination of techniques to evaluate risks (i.e., analyze a problem or proposal) considers the 

radiological hazard and the relative strengths and limitations of specific techniques to address 

the material, device, or facility during relevant offnormal scenarios. It is also likely that different 

techniques may better support specific portions of the life cycle or specific concerns in terms 

of internal and external events, design and subsequent configuration management, and the 

importance of the man and machine interface.

Given the importance of terminology in the development and implementation of a risk 

management approach, the RMTF offers the following definitions:1

Technical Analysis 

An evaluation to support NRC decisionmaking that includes the 

identification of potential radiological hazards and addresses: 

(1) what can go wrong, (2) how likely is it, and (3) what are 

the consequences. The evaluation process uses one or 

more techniques to address these questions. The possible 

techniques include expert judgment, reliance on industry or 

international standards, traditional engineering analyses, 

hazards or scenario analyses, and risk assessments.

Risk Assessment

An analysis technique that uses estimates of frequencies and 

consequences to systematically calculate risk and present 

the results (usually in numerical form). In this context, risk 

assessments include probabilistic risk assessments (PRAs), 

probabilistic safety assessments (PSAs), and risk analyses in 

the reactor programs and with risk-informed decisions in the 

materials programs. 

A key concept in the development of a risk management process at the NRC is that the 

evaluation of risks can be done any number of ways—including traditional engineering 

analyses, risk assessments, and other techniques selected to support specific decisions 

related to particular issues and hazards. This point is reinforced by discussions in International 

Standards Organization (ISO) 31010, “Risk Management – Risk Assessment [technical analysis] 

Techniques” (ISO, 2009), which describe a variety of techniques to analyze a problem and 

inform the decisionmaking process. 

1 The RMTF terminology differs somewhat from generic risk management terminology, which generally refers 

to risk assessment as any technique used to support deliberations (i.e., the RMTF equivalent of technical 

analysis). The above terminology was chosen because of the longstanding practice at the NRC and within 

the broader nuclear community to differentiate between probabilistic risk assessments and traditional 

deterministic type analyses.
A Proposed Risk Management Regulatory Framework

B-4 | Risk Management Task Force Appendix B

These techniques include the following:

•	 discussions or brainstorming

•	 expert elicitation

•	 hazard analyses

•	 scenario analyses

•	 failure mode and effects analyses 

•	 fault tree*

•	 event tree*

•	 decision trees

•	 Monte Carlo simulations

•	 cost-benefit analyses

•	 frequency-consequence curves*

* Usually associated with “risk assessment” type technique

Within the proposed Risk Management Regulatory Framework, technical analyses will be used 

to support decisions on appropriate regulatory controls and oversight, which are described 

in terms of risk-informed, performance-based defensein depth and related measures to 

prevent, contain, and mitigate conditions and accidents that could result in radiation exposures 

to workers or the public. The processes currently in place for operating nuclear power 

reactors include a balancing of traditional engineering approaches (sometimes referred to 

as deterministic or mechanistic analyses) and risk assessments. This balancing of technical 

analysis techniques to support decisions regarding appropriate defense-in-depth measures is 

shown in Figure B-3.

Figure B-3 Balancing Risk Assessments and Deterministic Techniques
Appendix B Risk Management Task Force | B-5

A Proposed Risk Management Regulatory Framework

From the previous discussions, the purpose of the technical analysis is to support subsequent 

deliberations on what are appropriate barriers and controls to prevent, contain, and mitigate 

possible releases of radioactive material and ensure that the risks from events that degrade 

or challenge the barriers are maintained acceptably low. The appropriate barriers range 

from simple containers for some radioactive sources to complex structures for nuclear power 

plants. Likewise, the systems and actions taken to maintain barriers can range from labels 

and administrative controls to complex mitigation systems. Controls established for materials 

licensees can also address circumstances where the use of a device requires the temporary 

bypass of physical barriers (e.g., radiography and irradiation devices). The possible need for 

emergency preparedness requirements is based on the risk that barriers might be compromised 

such that the public could be exposed to radioactive materials, and protective actions, such as 

sheltering or evacuation, might be warranted. 

The selection of the technical analysis technique considers the nature of the hazard, the 

possible challenges to barriers, and the complexity of barriers and supporting systems. In 

general, decisions regarding simpler devices and frequent events (e.g., activities with many 

materials licensees) can be supported by technical analyses based on traditional engineering 

approaches, operating experience, and qualitative risk assessments. Decisions related to more 

complex facilities and infrequent events (e.g., nuclear power plants) can benefit from analyses 

that include more robust risk assessments. This relationship between various technical analysis 

techniques and the nature of the possible radiological hazard is shown in Figure B-4:

The discussions of traditional approaches and risk assessments shown in Figures B-3 and B-4 

are mostly defined for nuclear power reactors but have some relevance to other regulatory 

program areas. As used here, the traditional approach evaluates risk by defining certain events 

that could challenge barriers and performing engineering analyses on systems and structures 

to ensure that defined acceptance criteria are satisfied. An example is to define a designbasis weather event (e.g., wind or precipitation) and evaluate the loads on structures using 

Figure B-4 Technical Analysis Techniques and Deliberation
A Proposed Risk Management Regulatory Framework

B-6 | Risk Management Task Force Appendix B

established industry codes and standards. Another example is to assume a mechanical failure, 

such as a specific pipe break, and perform analyses on the capabilities of safety systems to 

remove heat from the reactor core and maintain temperatures below regulatory requirements. 

These types of engineering analyses tend to define design-basis events (sometimes 

referred to as maximum credible or maximum hypothetical events) based on historical data 

and engineering judgment; include upper bound assumptions; use stylized, conservative 

analytical models and assumptions (e.g., failure of mitigation systems); and incorporate safety 

margins. These kinds of analyses are well suited for design activities where engineers are 

making decisions on structures, materials, cooling capabilities, and other factors affecting 

the selection and construction of plant systems, structures, and components. Limitations of 

traditional approaches can include the (1) “over-design” and resultant increase in cost of some 

plant features resulting from conservative assumptions and (2) “underdesign” and resultant 

compromise of barriers if the design-basis event is exceeded or a vulnerability is not identified. 

The attributes of more detailed risk assessments (e.g., PRAs for power reactors) include 

the analysis of numerous event sequences, the use of best estimate analytical models, the 

identification of system importance and establishment of reliability goals, the identification 

of vulnerabilities, and the interactions and dependencies between systems and operators. 

The risk assessments have been better suited to support decisions related to configuration 

management and accident management for events exceeding the traditional design-basis 

events. Limitations of the risk assessment techniques include (1) completeness of scenario 

lists, and (2) the difficulties in defining typical design specifications, including safety margins, 

using existing PRA methodologies. It should be noted that risk assessments have been 

introduced into the NRC’s regulatory system in areas such as the requirements for an integrated 

safety assessment for fuel cycle facilities, the performance of PRAs for new reactors, and the 

optional use of PRAs for nuclear reactors to justify changes to special treatment requirements 

for systems, structures, and components (Title 10 Code of Federal Regulations (10 CFR) 50.69, 

“Risk-informed Categorization and Treatment of Structures, Systems, and Components for 

Nuclear Power Reactors”). 

The key to the selection of a technical analysis technique or combination of techniques is to 

ensure that the analysis will support the associated deliberation or decisionmaking process in 

an effective and efficient manner. This, in turn, means that there needs to be some acceptance 

criteria or other rationale for decisionmaking that are established and understood by the 

decisionmaker and agency stakeholders. 

B.3 Deliberate

The technical analyses are a key input to the process but can support a successful outcome 

only if there is an equally disciplined approach to the actual decisionmaking. The NRC’s 

decisions related to its core mission can be described in terms of ensuring that its licensees 

provide and maintain the appropriate risk-informed and performance-based defense-in-depth 

protections to achieve the following:

•	 Ensure appropriate barriers, controls, and qualified personnel to prevent, contain, and 

mitigate possible inadvertent exposure to radioactive material according to the hazard 

present, the relevant scenarios, and the associated uncertainties.

•	 Ensure the risks resulting from the failure of some or all of the established barriers and 

controls are maintained acceptably low.
Appendix B Risk Management Task Force | B-7

A Proposed Risk Management Regulatory Framework

The technical analysis is an important input into the deliberation process, but it is not the only 

factor that influences the final decision. The decisionmaker needs to consider the uncertainty 

and sensitivity analyses associated with the technical analysis, as well as the analytical 

results and how they compare to decision criteria established for mechanistic approaches, risk 

assessment approaches, or a combination thereof. The uncertainties to be considered include 

the degree of understanding of hazards, scenarios, phenomena, safety margins, and other 

tangible and intangible factors. There have been numerous cases in which emergent technical 

issues have been resolved, at least in part, by safety margins or other actions previously taken 

to address uncertainties, including conservatisms, to cover the so-called “unknown unknowns.” 

The deliberative process also includes consideration of resources and schedules for the agency 

and its licensees and the input received from various internal and external stakeholders. Other 

factors include legal requirements and the desire to maintain consistency with various guidance 

documents, treaties, or standards.

A representation of the deliberative process, including consideration of the various factors, is 

shown in Figure B-5.

In a risk management regime, the decisionmaking process involves discussions of what risks 

are tolerable or acceptable to the NRC, key stakeholders, and the public at large. Various 

approaches have been developed to represent tolerable or acceptable risks from nuclear plants 

and other modern industrial facilities. It should be noted that the NRC has established and, 

in some cases, codified acceptance criteria (deterministic and risk-informed) for regulatory 

decisions related to license applications, amendments to licenses, and other actions for specific 

types of licensees. 

Figure B-5 Deliberations
A Proposed Risk Management Regulatory Framework

B-8 | Risk Management Task Force Appendix B

Examples include the following:

•	 Reactors

o NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports 

for Nuclear Power Plants: LWR Edition” (NRC, 2007)

o Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment 

in Risk-informed Decisions on Plant-specific Changes to the Licensing-basis,” 

Revision 2 (NRC, 2011)

o Regulatory Guide 1.187, “Guidance for Implementation of 10 CFR 50.59, 

“Changes, Tests, and Experiments” (NRC, 2000)

o Regulatory Guide 1.201, “Guidelines for Categorizing Structures, Systems, 

and Components in Nuclear Power Plants According to Their Safety 

Significance” (NRC, 2006)

o Regulatory Guide 1.205, “Risk-informed, Performance-based Fire Protection for 

Existing Light-Water Nuclear Power Plants,” Revision 1 (NRC, 2009)

•	 Materials

o NUREG-1556 (Volumes 1 – 21), “Consolidated Guidance About Materials 

Licensees” (NRC, 1998)

o “Risk-Informed Decisionmaking for Nuclear Material and Waste Applications,” 

Revision 1 (NRC, 2008)

o NUREG/CR-6642, “Risk Analysis and Evaluation of Regulatory Options for 

Nuclear Byproduct Material Systems” (NRC, 2000a)

•	 Fuel Cycle Facilities

o NUREG-1520, “Standard Review Plan for the Review of a License Application for 

a Fuel Cycle Facility,” Revision 1 (NRC, 2010)

o NUREG-1513, “Integrated Safety Analysis Guidance Document” (NRC, 2001)

•	 Storage and Transportation Systems

o NUREG-1536, “Standard Review Plan for Dry Cask Storage Systems,” Initial 

Report, (NRC, 1997)

o NUREG-1927, “Standard Review Plan for Renewal of Spent Fuel Dry Cask 

Storage System Licenses and Certificates of Compliance” (NRC, 2011a)

o NUREG-1609, ” Standard Review Plan for Transportation Packages for 

Radioactive Material,” Initial Report (NRC, 1999)
Appendix B Risk Management Task Force | B-9

A Proposed Risk Management Regulatory Framework

o NUREG-1617, ”Standard Review Plan for Transportation Packages for Spent 

Nuclear Fuel,” Initial Report (NRC, 2000b)

These guidance documents often outline a decisionmaking process that is similar to the one 

described in this report. For example, the report entitled, “Risk-informed Decisionmaking for 

Nuclear Material and Waste Applications,” Revision 1, (NRC, 2008), includes steps to define the 

regulatory issue, determine the appropriate techniques to perform evaluations, and apply the 

risk-informed decisionmaking process. 

In some cases, the technical analysis described in the NRC guidance emphasizes the traditional 

engineering approaches used in the initial licensing of most NRC facilities. While traditional 

engineering approaches, sometimes referred to as deterministic analyses, often omit specific 

discussion of event frequencies, probabilities, and consequences, they are nevertheless an 

evaluation technique that supports risk management decisions. Defining acceptance criteria 

in terms of limits on specific barriers (with added safety margins) is a common approach. The 

traditional regulations and guidance may also define specific assumptions regarding initiating 

events, such as earthquakes, floods, and other challenges to one or more of the defense-indepth barriers. A common approach supporting the NRC deliberative processes includes the 

development and use of consensus codes and standards. This practice is consistent with 

the National Technology Transfer and Advancement Act of 1995 (Public Law 104113), which 

requires government agencies to use consensus standards where possible. In the reactor 

area, the NRC has incorporated specific sections of the ASME Code and some Institute of 

Electrical and Electronics Engineers (IEEE) Standards by reference into its regulations. Other 

consensus codes and standards are referenced in material or facility licenses. Increasingly, the 

incorporation of risk assessments and performance-based approaches has been introduced to 

various codes and standards. The transition from the current guidance to the decisionmaking 

process proposed by the RMTF is described in Chapter 4 for the various NRC 

regulatory programs. 

As previously mentioned, a concern with the traditional or deterministic approach is that the 

establishment of “maximum credible accidents” as the sole basis for defining defense-in-depth 

protections may not account for unidentified vulnerabilities. Examples of such insights from 

risk assessments related to nuclear reactors include the identification of concerns regarding 

intersystem loss-of-coolant accidents, anticipated transients without scram, and station 

blackout events. Attempts to introduce risk assessments as a way to supplement the traditional 

approaches gained support following the publication of WASH-1400, “The Reactor Safety 

Study,” and the accident at Three Mile Island (TMI) in 1979. Specific milestones include the 

issuance of the following NRC policy statements:

•	 “Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing 

Plants,” (NRC, 1985)

•	 “Safety Goals for the Operations of Nuclear Power Plants,” (NRC, 1986)

•	 “Policy Statement on the Regulation of Advanced Reactors,” (NRC, 2008a)

•	 “Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final 

Policy Statement” (NRC 1995)
A Proposed Risk Management Regulatory Framework

B-10 | Risk Management Task Force Appendix B

Each of these policy statements addressed operating reactors and future reactors and 

maintained that specific activities for the operating reactors (e.g., TMI action items) were a 

viable alternative to incorporating risk assessments more formally into the decisionmaking 

process. NRC actions were informed by various activities and risk assessments, including 

the “Individual Plant Examination Program: Perspectives on Reactor Safety and Plant 

Performance” (NRC, 1997a), individual plant examinations for external events (NRC, 2001a), 

and NUREG-1150, “Severe Accident Risks: An Assessment for Five U.S. Nuclear Power 

Plants—Final Summary Report” (NRC,1990). The policy statements supported improved use 

of risk assessment techniques in the design and licensing of future reactors. The PRA policy 

statement of 1995 encouraged, but did not require, the increased use of risk assessment 

techniques in all NRC regulatory programs. The absence of decisionmaking criteria 

complicated the implementation of these policy statements. A step forward in the integration 

of risk assessments into the decisionmaking process was provided by Regulatory Guide 1.174 

in 1998. This Regulatory Guide clarified how insights from risk assessments should be used 

within an integrated decisionmaking process to support applications for and NRC review of 

proposed changes to the licensing-basis for operating reactors. Similar guidance was provided 

for materials, transportation, and waste programs in the risk-informed decisionmaking process 

developed by FSME and NMSS. The integrated decisionmaking process from Regulatory 

Guide 1.174 and associated guidance to support decisionmaking are shown below: 

Figure B-6 RG 1.174 Integrated Decisionmaking Process
Appendix B Risk Management Task Force | B-11

A Proposed Risk Management Regulatory Framework

A risk-informed approach has also been developed for the NRC’s oversight of commercial 

nuclear power plants (NRC, 2006a). The oversight process uses a significance determination 

process (SDP) to help assess licensees’ performance and determine appropriate agency 

responses. The guidance related to the SDP is shown in Figure B-8.

The guidance documents include a variety of acceptance criteria for different regulatory 

programs, specific facilities, and for different types of events or operating conditions. There 

have been several notable attempts to develop acceptance criteria that could support consistent 

decisionmaking for different types of nuclear facilities—and in some cases for any type of 

industrial facility. Such criteria are often presented in terms of some measure of consequence 

(e.g., dose, fatalities, health effects, economic costs) versus probability or estimated frequency 

of an event or condition. The NRC introduced the possible use of a frequency consequence 

(F-C) curve in NUREG-1860, “Feasibility Study for a Risk-Informed and Performance-Based 

Regulatory Structure for Future Plant Licensing” (NRC, 2007a). The F-C curves proposed in 

NUREG-1860 and in proposals related to the next generation nuclear plant (NGNP, 2010) are 

provided in Figure B-9.

Figure B-7 RG 1.174 Process and Decision Criteria 

Figure B-7 RG 1.174 Process and Decision Criteria

Red (high safety or security significance) is quantitatively greater than 

10-4ΔCDF or 10-5 ΔLERF. 

Qualitatively, a Red significance indicates a decline in licensee performance that is 

associated with an unacceptable loss of safety margin. Sufficient safety margin still 

exists to prevent undue risk to public health and safety. 

Yellow (substantial safety or security significance) is quantitatively greater than 

10-5 and less than or equal to 10-4 ΔCDF or greater than 10-6 and less than or equal 

to 10-5 ΔLERF. Qualitatively, a Yellow significance indicates a decline in licensee 

performance that is still acceptable with cornerstone objectives met, but with 

significant reduction in safety margin. 

Green (very low safety or security significance) is quantitatively less than or equal to 

10-6 ΔCDF or 10-7 ΔLERF. 

Qualitatively, a Green significance indicates that licensee performance is acceptable 

and cornerstone objectives are fully met with nominal risk and deviation. 

10-6

10-5

10-4

ΔLERF 

10-7

10-6

10-5

ΔCDF Reactor Oversight Process – Significance Determination Process 

White (low to moderate safety or security significance) is quantitatively greater than 

10-6 and less than or equal to 10-5ΔCDF or greater than 10-7 and less than or equal to 

10-6 ΔLERF. Qualitatively, a White significance indicates an acceptable level of 

performance by the licensee, but outside the nominal risk range. Cornerstone 

objectives are met with minimal reduction in safety margin. 

Figure B-8 ROP Significance Determination Guidance 

Figure B-8 ROP Significance Determination Guidance
A Proposed Risk Management Regulatory Framework

B-12 | Risk Management Task Force Appendix B

Some F-C diagrams include multiple regions to differentiate between risks generally viewed as 

tolerable, risks viewed as unacceptable, and risks that should be reduced as practical or costeffective to do so (e.g., the “as low as is reasonably achievable” (ALARA) region in Figure B-10). 

A useful discussion of various approaches to decisionmaking using F-C concepts is provided in 

the report, “Societal Risks,” prepared for the Health and Safety Executive (HSE) in the United 

Kingdom (Ball, 1998). 

Figure B-10 Generic F-C Curve, with ALARA region 

Figure B-9 F-C Curves from NUREG-1860 and NGNP 

Figure B-9 F-C Curves from NUREG-1860 and NGNP

Figure B-10 Generic F-C Curve, with ALARA region
Appendix B Risk Management Task Force | B-13

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Although not proposing to incorporate a specific F-C curve into the risk management regulatory 

framework, the RMTF encourages the inclusion of broader topics, such as the HSE reports, in 

the NRC’s training programs. The development and use of the F-C curve similar to Figure B-10 

is perceived by some as a dramatic departure from past practices, but many NRC programs 

already incorporate aspects of the approach by differentiating between high frequency–low 

consequence activities and the potential for low frequency–high consequence accidents. In 

the analyses to support licensing of nuclear power reactors, events have traditionally been 

defined within the categories of (1) normal operation, (2) anticipated operational occurrences, 

(3) design-basis accidents, and (4) beyond-design-basis accidents. The primary criteria for 

placing scenarios within the above categories are related to event frequencies. The allowable 

consequences (defined in terms of degree of fuel damage) are defined for the categories, and 

generally more damage is acceptable for scenarios with lower frequencies (see Appendix F). 

The general concept is also evident in the graded treatment of different licensees, with those 

presenting lesser risks requiring fewer barriers and controls while facilities such as nuclear 

power plants are required to have multiple barriers, mitigating systems and procedures, and 

emergency plans. The NRC’s safety goals for nuclear power reactors are actually defined 

in terms of a measure of consequence (prompt fatalities and cancer fatalities) and frequency 

(as a percentage of other societal risks). The introduction section to the policy statement 

documented the NRC’s view that radiation protection requirements, such as 10 CFR Part 20, 

“Standards for Protection Against Radiation,” ensured that the consequences from higher 

frequency events (e.g., normal operation) did not warrant additional measures to reduce the 

frequency of routine releases. 

The use of different regions or categories to delineate importance or regulatory treatment has 

also been used in various NRC programs. The ALARA concept has long been a major part of 

the radiation protection programs that the NRC requires of its licensees. The backfit provisions 

of the NRC’s reactor regulations and the guidance on regulatory analyses for rulemakings have 

likewise recognized that there are those barriers and controls that the NRC requires to ensure 

adequate protection of public health and safety, those that provide a substantial increase in 

public protection such that the costs of implementation are justified, and those in which the 

safety benefit does not justify the associated costs (NRC, 2004). Similar categories (regions) 

using changes in core damage frequency (CDF) and large early release frequency (LERF) 

were defined in the above figures from Regulatory Guide 1.174 for defining proposed licensing 

actions that would be considered independent of the total CDF/LERF, considered in relation to 

total CDF/LERF, or not considered. A similar construct shown in Figure B-11 was proposed in 

the risk-informed decisionmaking approach for licensing of byproduct material applications. 

Figure B-11 Risk-Informed Decisionmaking Figure B-11 Risk-Informed Decisionmaking
A Proposed Risk Management Regulatory Framework

B-14 | Risk Management Task Force Appendix B

The NRC goal to move to more performance-based approaches to regulation would be 

supported by expanding the ALARA principle from radiation protection to the broader area 

of risk management (as shown in Figure B-10). As discussed in Appendix D, “PerformanceBased Regulation,” deterministic and prescriptive approaches can limit the flexibility of both the 

regulated industries and the NRC to respond to lessons learned from operating experience and 

support the adoption of improved designs or processes. An ALARA approach within the risk 

management regulatory framework could define performance criteria in terms of risk-informed, 

performance-based defense-in-depth, while providing licensees with the flexibility to determine 

the most cost-effective means to provide and maintain appropriate protections to prevent, 

mitigate, and contain the possible release of radioactive materials. Previous activities—such 

as the maintenance rule (10 CFR 50.65, “Requirements for Monitoring the Effectiveness of 

Maintenance at Nuclear Power Plants”) for nuclear power reactors—have been successful in 

establishing objective and measurable criteria for risk-informed, performance-based regulation 

and oversight of NRC licensees. The adoption of the ALARA principle would reinforce the 

stated objective in the Advanced Reactor Policy Statement (NRC, 2008a) for new designs to 

provide enhanced margins of safety and use simplified, inherent, passive, or other innovative 

means to accomplish their safety functions. SECY-10-0121, “Modifying the Risk-informed 

Regulatory Guidance for New Reactors” (NRC, 2010a) discusses several issues related to a 

possible discrepancy between the Advanced Reactor Policy Statement goals for enhanced 

safety for new plants and the actual regulation and oversight of new reactors based on the 

same rules and guidance applied to operating reactors. 

In addition to comparing the results from the technical analyses to the appropriate decision or 

acceptance criteria, the deliberative process needs to consider other factors, such as those 

shown in Figure B-5. A key factor included in the definition of risk-informed, performance-based 

defense-in-depth is the consideration of uncertainties. Uncertainties include gaps in knowledge 

regarding the radiological materials and their behavior during scenarios that could result in the 

exposure of workers or the public. There may also be uncertainties associated with the abilities 

of barriers and controls to prevent or mitigate accident scenarios or to contain the radioactive 

materials following such scenarios. Computational models are used to simulate devices and 

complex systems for infrequent scenarios to help determine appropriate design features and 

controls, but these models introduce another group of uncertainties to consider during the 

decisionmaking process. Such uncertainties should be addressed, as practical, within the 

technical analyses and should be identified and accounted for as part of deliberations on a 

particular issue or problem. Sensitivity studies can be a useful tool to evaluate uncertainties 

and otherwise inform the decisionmaking process. 

B.4 Implementation Factors

Decisions resulting from deliberations on specific issues or proposals address whether the 

defense-in-depth protections associated with the subject radiological hazard are adequate, 

can be relaxed, or need to be strengthened. It is important to recognize that the NRC has a 

limited and relatively fixed set of regulatory processes by which it can define requirements 

for its licensees. The implementation processes include preparing regulations and guidance; 

reviewing proposed licensing actions; performing environmental reviews; and executing 

oversight programs. 

It is likewise necessary to understand that the actual defense-in-depth protections are 

implemented and maintained by licensees who process and control the radioactive materials 

or related facilities. However, the use of radioactive materials introduces particular risks and 
Appendix B Risk Management Task Force | B-15

A Proposed Risk Management Regulatory Framework

public concerns, and so the NRC’s regulatory regime, adopted pursuant to the Atomic Energy 

Act of 1954, as amended, and other statutes, provides the legal basis for ensuring that NRC 

licensees provide reasonable assurance of adequate protection of the health and safety of their 

workers and the public. Licensees within the NRC’s various regulatory programs provide the 

appropriate defense-in-depth protections for workers and the public through: 1) design features 

incorporated into devices or facilities, 2) operating practices and procedures, 3) maintenance 

of barriers and supporting systems, 4) inspections and surveillance programs, 5) radiation 

monitoring, 6) reporting events and other information to their organizations, the NRC, and other 

agencies, and, finally 7) the decommissioning of facilities. 

As discussed throughout this report, appropriate evaluations and actions will vary depending on 

the type and amount of radioactive material, as well as on the complexity of the device or facility. 

For some materials and devices, concerns regarding an inadvertent release is small and few 

controls beyond routine radiation protection programs (time, distance, shielding) are necessary. 

On the other hand, facilities such as commercial nuclear power plants contain a large amount 

of radioactive materials that warrants multiple levels of protection to prevent, contain, and 

mitigate possible releases. These same recognitions are reflected in how the NRC implements 

actions (e.g., regulations vs. guidance) and how licensees provide appropriate protection (e.g., 

engineered design features or administrative controls).

A representation of a risk management framework combining NRC processes and associated 

licensee controls is shown below:

The means by which the NRC might define requirements or expectations and the means 

available for licensees to provide appropriate defense-in-depth protections are actually part of 

the deliberative process, as well as being the vehicles for implementing decisions. An example 

Figure B-12 Implementation of Decisions
A Proposed Risk Management Regulatory Framework

B-16 | Risk Management Task Force Appendix B

is the NRC goal to define performance-based approaches as an alternative to traditional or 

prescriptive requirements in which a rule or license condition would define specified engineered 

design features or specific maintenance and surveillance practices. Consideration of a 

performance-based option could influence the analysis and deliberation steps of the risk 

management framework. 

B.5 References

(Ball, 1998) D. J. Ball and P. J. Floyd, “Societal Risks,” Final Report for the 

Health and Safety Executive, United Kingdom, 1998

(ISO, 2009) International Organization for Standardization, “Risk Management 

– Risk Assessment Techniques,” ISO-31010, 2009.

(Kaplan and Garrick, 1981) S. Kaplan and B.J. Garrick, “On the Quantitative Definition of Risk, 

Risk Analysis,” 1:11−27.

(NGNP, 2010) Idaho National Laboratory, “Next Generation Nuclear Plant 

Licensing-basis Event Selection White Paper,” INL/EXT-1019521, 

September 2010, Agencywide Documents Access Management 

System (ADAMS) Accession No. ML102630246.

(NRC, 1985) U.S. Nuclear Regulatory Commission, “Policy Statement 

on Severe Reactor Accidents Regarding Future Designs 

and Existing Plants,” August 1985, ADAMS Accession 

No. ML003711521.

(NRC, 1986) U.S. Nuclear Regulatory Commission, “Safety Goals for the 

Operations of Nuclear Power Plants; Policy Statement,” 

August 1986, ADAMS Accession No. ML011210381.

(NRC, 1990) U.S. Nuclear Regulatory Commission, “Severe Accident 

 Risks: An Assessment for Five U.S. Nuclear Power 

Plants — Final Summary Report,” NUREG-1150, Volume 1,

December 1990, ADAMS Accession No. ML040140729.

(NRC, 1995) U.S. Nuclear Regulatory Commission, “Use of Probabilistic

Risk Assessment Methods in Nuclear Regulatory 

Activities; Final Policy Statement,” August 1995, ADAMS

Accession No. ML021980535.

(NRC, 1997) U.S. Nuclear Regulatory Commission, “Standard Review 

Plan for Dry Cask Storage Systems,” Initial Report, NUREG-1536,

January 1997, ADAMS Accession No. ML010040237

(NRC, 1997a) U.S. Nuclear Regulatory Commission, “Individual Plant 

Examination Program: Perspectives on Reactor Safety 

and Plant Performance,” NUREG-1560, December 1997,

ADAMS Accession No. ML063550244.
Appendix B Risk Management Task Force | B-17

A Proposed Risk Management Regulatory Framework

(NRC, 1998) U.S. Nuclear Regulatory Commission, “Consolidated 

Guidance About Materials Licenses,” NUREG-1556, 

1998 and later, multiple volumes.

(NRC, 1999) U.S. Nuclear Regulatory Commission, “Standard Review Plan for 

Transportation Packages for Radioactive Material,” Initial Report, 

NUREG-1609, March 1999. 

(NRC, 2000) U.S. Nuclear Regulatory Commission, “Guidance for

Implementation of 10 CFR 50.59, Changes, Tests, and 

Experiments,” Regulatory Guide 1.187, November 2000,

ADAMS Accession No. ML003759710.

(NRC, 2000a) U.S. Nuclear Regulatory Commission, “Risk Analysis and 

Evaluation of Regulatory Options for Nuclear Byproduct

Material Systems,” NUREG/CR-6642, February 2000, 

ADAMS Accession No. ML003684878 (nonpublic).

(NRC, 2000b) U.S. Nuclear Regulatory Commission, “Standard Review 

Plan for Transportation Packages for Spent Nuclear Fuel,” 

Initial Report, NUREG-1617, March 2000, ADAMS Accession 

No. ML003696262.

(NRC, 2001) U.S. Nuclear Regulatory Commission, “Integrated Safety 

Analysis Guidance Document,” NUREG-1513, May 2001, 

ADAMS Accession No. ML011440260.

(NRC, 2001a) U.S. Nuclear Regulatory Commission, “Perspectives 

Gained From the Individual Plant Examination of

External Events (IPEEE) Program,” NUREG-1742,

September 2001, ADAMS Accession No. Ml021270132.

(NRC, 2004) U.S. Nuclear Regulatory Commission, “Regulatory Analysis

Guidelines of the U.S. Nuclear Regulatory Commission,”

NUREG/BR-0058, Revision 4, September 2004, ADAMS

Accession No. ML042820192.

(NRC, 2006) U.S. Nuclear Regulatory Commission, “Guidelines for 

Categorizing Structures, Systems, and Components in 

Nuclear Power Plants According to Their Safety 

Significance,” Regulatory Guide 1.201, May 2006,

ADAMS Accession No. ML061090627.

(NRC, 2006a) U.S. Nuclear Regulatory Commission, “Reactor Oversight

Process,” NUREG-1649, Revision 4, December 2006,

ADAMS Accession No. ML070890365.

(NRC, 2007) U.S. Nuclear Regulatory Commission, “Standard Review 

Plan for the Review of Safety Analysis Reports for Nuclear 

Power Plants: LWR Edition, NUREG-0800, March 2007.
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B-18 | Risk Management Task Force Appendix B

(NRC, 2007a) U.S. Nuclear Regulatory Commission, “Feasibility Study for 

a Risk-informed and Performance-Based Regulatory 

Structure for Future Plant Licensing,” NUREG-1860, Volume 1, 

December 2007, ADAMS Accession No. ML080440170.

(NRC, 2008) U.S. Nuclear Regulatory Commission, “Risk-informed 

Decisionmaking for Nuclear Material and Waste 

Applications,” Revision 1, February 2008, ADAMS

Accession No. ML080720238.

(NRC, 2008a) U.S. Nuclear Regulatory Commission, “Policy Statement 

on the Regulation of Advanced Reactors,” October 2008, 

ADAMS Accession No. ML082750370.

(NRC, 2009) U.S. Nuclear Regulatory Commission, “Risk-informed, 

Performance-Based Fire Protection for Existing Light-Water

Nuclear Power Plants,” Regulatory Guide 1.205,

December 2009, ADAMS Accession No. ML092730314.

(NRC, 2010) U.S. Nuclear Regulatory Commission, “Standard Review 

Plan for the Review of a License Application for a Fuel 

Cycle Facility,” NUREG-1520, Revision 1, May 2010,

ADAMS Accession No. ML101390110.

(NRC, 2010a) U.S. Nuclear Regulatory Commission, “Modifying the 

Risk-Informed Regulatory Guidance for New Reactors,”

SECY-10-0121, September 14, 2010, ADAMS Accession

No. ML102230076.

(NRC, 2011) U.S. Nuclear Regulatory Commission, “An Approach for 

 Using Probabilistic Risk Assessment in Risk-Informed 

Decisions on Plant-Specific Changes to the Licensing 

Basis,” Regulatory Guide 1.174, Revision 2, May 2011, 

ADAMS Accession No. ML100910006.

(NRC, 2011a) U.S. Nuclear Regulatory Commission, “Standard Review 

Plan for Renewal of Spent Fuel Dry Cask Storage System

Licenses and Certificates of Compliance,” NUREG-1927, 

March 2011, ADAMS Accession No. ML111020115.
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APPENDIX C

DEFENSE-IN-DEPTH

C.1 Introduction

The term “defense-in-depth” has been used since the 1960s in the context of ensuring nuclear 

reactor safety. The concept was developed and applied to compensate for the recognized 

lack of knowledge of nuclear reactor operations and the consequences of potential accidents. 

While both experience with reactor operations and the knowledge base of potential (and actual) 

consequences have grown considerably since that time, the concept continues to be highly 

relevant today.

The Risk Management Task Force (RMTF) has reviewed a number of documents that 

historically have helped to shape the characterization of defense-in-depth. Since the 

characterizations provided in these documents are not completely consistent and are focused 

on operating power reactors, the RMTF concluded that clarifying what the U.S. Nuclear 

Regulatory Commission (NRC) means by defense-in-depth is a necessary part of the 

development of a holistic strategic vision. This appendix describes the proposed RMTF 

characterization, summarizes other descriptions of defense-in-depth (which have focused, as 

noted above, on power reactors), and compares them with the RMTF characterization. In one 

case (power reactors), the RMTF has extended this characterization to provide 

additional perspective.

C.2 Task Force Characterization of Defense-in-depth

The RMTF characterizes defense-in-depth as follows:

Risk-informed and performance-based defense-in-depth 

protections:

•	 Ensure appropriate barriers, controls, and personnel to 

prevent, contain, and mitigate exposure to radioactive 

material according to the hazard present, the relevant 

scenarios, and the associated uncertainties; and 

•	 Ensure that the risks resulting from the failure of some 

or all of the established barriers and controls, including 

human errors, are maintained acceptably low.

Key terms in this characterization include:

•	 Barriers: Barriers can take a number of forms, such as a container, a wall, or a restricted 

area of land. In some cases, engineered systems such as cooling systems may exist to 

help ensure barrier functionality. The number and types of barriers needed should be 
A Proposed Risk Management Regulatory Framework

C-2 | Risk Management Task Force Appendix C

commensurate with and proportional to the extent of the hazard and reflect the current 

understanding of important associated uncertainties.

•	 Controls: Administrative actions are established to ensure the performance of a barrier.

•	 Personnel: Programs are established to ensure that people are properly trained and 

qualified for the jobs they have and that periodic training is provided to keep those 

people focused on safety. When performance problems arise, specific lessons-learned 

training is provided to ensure, in the best way possible, that problems are not repeated.

•	 Hazard: A hazard is an assemblage of radioactive material licensed by the NRC (e.g., a 

nuclear reactor core, a well-logging device containing radioactive material, a uranium 

mine). The extent of the hazard is measured in terms of the amount of radioactive 

material present and the expected length of time it would exist.

•	 Associated uncertainties: Uncertainties exist in a number of aspects of facility design 

and operation, from the properties of materials used to the likelihood of events that could 

challenge that facility.

•	 Risks: Risk assessment methods provide an effective means for measuring the 

adequacy of defense-in-depth. That is, risk assessment measures the likelihood 

of individual barriers being compromised, the probabilities that additional barriers 

would be compromised if the previous barrier were to be compromised, and how 

poorly understood phenomena or new information would affect these likelihoods and 

probabilities. This information is useful in ensuring an appropriate balance among barrier 

capabilities, in ensuring the low likelihood that multiple barriers would be compromised, 

and ensuring that the overall result is an adequately low risk to the potentially harmed 

individual or group.

C.3 Historical Characterizations

NRC white paper, as modified in 10 CFR 50.69, Statements of Consideration

The NRC’s “White Paper on Risk-Informed and Performance-Based Regulation” was 

completed in 1999 and contained descriptions of many terms, including defense-in-depth 

(NRC,1999). This description was modified somewhat and was issued in 2004 in the 

Statements of Consideration for the NRC’s final rule, Title 10 of the Code of Federal Regulations

(10 CFR) 50.69, “Risk-Informed Categorization and Treatment of Structures, Systems, and 

Components for Nuclear Power Reactors” (NRC, 2004), as follows:

Defense-in-depth is an element of the NRC’s safety philosophy that employs 

successive measures to prevent accidents or mitigate damage if a malfunction, 

accident, or naturally caused event occurs at a nuclear facility. Defense-indepth is a philosophy used by the NRC to provide redundancy as well as the 

philosophy of a multiple-barrier approach against fission product releases. The 

defense-in-depth philosophy ensures that safety will not be wholly dependent 

on any single element of the design, construction, maintenance, or operation of 

a nuclear facility. The net effect of incorporating defense-in-depth into design, 

construction, maintenance, and operation is that the facility or system in question 

tends to be more tolerant of failures and external challenges.
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A Proposed Risk Management Regulatory Framework

The RMTF characterization of defense-in-depth is intended to achieve the same intent of the 

2004 characterization and extends the 2004 characterization in two ways. First, while it seems 

to be implicit in the 2004 characterization, the RMTF has made more explicit that the number 

and types of barriers are dependent on the radioactive hazard present. Second, the RMTF 

characterization explicitly uses risk assessment to measure the outcome of defense-in-depth, 

which the 2004 characterization describes more generally as “more tolerant.”

International Atomic Energy Agency

In INSAG-10 (IAEA, 1996), the International Atomic Energy Agency (IAEA) provided a 

description of defense-in-depth for operating reactors as follows:

Defence in depth is generally structured in five levels. Should one level fail, the 

subsequent level comes into play. The objective of the first level of protection 

is the prevention of abnormal operation and system failures. If the first level 

fails, abnormal operation is controlled or failures are detected by the second 

level of protection. Should the second level fail, the third level ensures that 

safety functions are further performed by activating specific safety systems 

and other safety features. Should the third level fail, the fourth level limits 

accident progression through accident management, so as to prevent or mitigate 

severe accident conditions with external releases of radioactive materials. 

The last objective (fifth level of protection) is the mitigation of the radiological 

consequences of significant external releases through the off-site 

emergency response.

The RMTF characterization is intended to apply to a broader set of regulated activities, some 

of which would not need to have the five levels discussed in INSAG-10. However, the RMTF 

characterization should achieve the same outcomes as the INSAG-10 characterization when 

applied to operating reactors.

Next Generation Nuclear Plant (NGNP)

In 2009, Idaho National Laboratory, under contract to the U.S. Department of Energy, published 

a paper, “Next Generation Nuclear Plant Defense-in-Depth Approach” (NGNP, 2009), that 

reviewed the existing literature on defense-in-depth and outlined a definition of defense in 

depth that would be used in the design of new power reactors. The definition was summarized 

as follows:

• Plant Capability Defense-in-Depth reflects the decisions made by the designer 

in the selection of functions, structures, systems, and components for the 

design that ensure defense-in-depth in the physical plant. 

• Programmatic Defense-in-Depth reflects the decisions made regarding the 

processes of manufacturing, constructing, operating, maintaining, testing, and 

inspecting the plant and the processes undertaken that ensure plant safety 

throughout the lifetime of the plant. 
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C-4 | Risk Management Task Force Appendix C

• Risk-Informed Evaluation of defense-in-depth reflects the development and 

evaluation of strategies that manage the risks of accidents, including the 

strategies of accident prevention and mitigation. This aspect of defense-indepth also provides the framework for performing deterministic and probabilistic 

safety evaluations, which help determine how well various Plant Capability 

Defense-in-Depth and Programmatic Defense-in-Depth strategies have 

been implemented. 

As noted previously, the RMTF characterization is intended to apply to a broader set of 

regulated activities. The RMTF characterization, extended for operating power reactors 

as discussed in Section C.4 of this appendix, has important similarities to the NGNP 

characterization. More specifically, both the NGNP and RMTF characterizations include the 

concept of using risk assessment methods as a measure of effectiveness.

Davis Besse petition— Director’s decision

The report of NRC’s Near-Term Task Force (NTTF) (NRC, 2011) on the Fukushima accident 

cites and discusses a 2003 NRC description of defense-in-depth. The NTTF discussion is 

as follows:

An instructive discussion of the defense-in-depth philosophy also appears in 

director’s decisions relating to a petition on Davis-Besse (FirstEnergy Nuclear 

Operating Company (Davis-Besse Nuclear Power Station, Unit 1), DD033, 

58 NRC 151, 163 (2003)).

The decision described defense-in-depth as encompassing the following 

requirements:

(1) require the application of conservative codes and standards to establish 

substantial safety margins in the design of nuclear plants; 

(2) require high quality in the design, construction, and operation of nuclear 

plants to reduce the likelihood of malfunctions, and promote the use of 

automatic safety system actuation features;

(3) recognize that equipment can fail and operators can make mistakes and, 

therefore, require redundancy in safety systems and components to reduce 

the chance that malfunctions or mistakes will lead to accidents that release 

fission products from the fuel;

(4) recognize that, in spite of these precautions, serious fuel-damage accidents 

may not be completely prevented and, therefore, require containment 

structures and safety features to prevent the release of fission products; and

(5) further require that comprehensive emergency plans be prepared and 

periodically exercised to ensure that actions can and will be taken to notify 

and protect citizens in the vicinity of a nuclear facility.
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A Proposed Risk Management Regulatory Framework

The [Near-Term] Task Force has found that the defense-in-depth philosophy is 

a useful and broadly applied concept. It is not, however, susceptible to a rigid 

definition because it is a philosophy. For the purposes of its review, the 

[Near-Term] Task Force focused on the following application of the defense-indepth concept:

• protection from external events that could lead to fuel damage

• mitigation of the consequences of such accidents should they occur, with a 

focus on preventing core and spent fuel damage and uncontrolled releases of 

radioactive material to the environment 

• emergency preparedness (EP) to mitigate the effects of radiological releases to 

the public and the environment, should they occur

These levels of defense-in-depth are appropriate for significant external 

challenges to a facility. In applying these defense-in-depth features, the 

[Near-Term] Task Force sought to ensure that the Commission’s regulatory 

requirements, processes, and programs effectively address each layer of 

protection while maintaining appropriate balance among them. The [Near-Term] 

Task Force notes that this approach is also consistent with Levels of Defense 3, 

4, and 5 in IAEA Draft Safety Standard DS 414, “Safety of Nuclear Power Plants: 

Design,” dated January 2010.

The 2003 characterization was focused on operating reactors, and the NTTF discussion of it 

has a similar focus. The RMTF characterization, extended for operating power reactors as 

discussed later in this appendix, encompasses the elements of the 2003 characterization. As 

part of this extension, it uses reliability assessment to address more quantitatively concepts 

such as redundancy and the use of automatic safety system actuations, and risk assessment, 

as means to measure the effectiveness of the set of barriers.

NUREG-1860

With the advent of new reactor licensing activities and the potential development of reactor 

designs quite different from current operating reactors, the NRC decided to assess the feasibility 

of implementing a new regulatory structure as an alternative to 10 CFR Part 50. This new 

structure would take advantage of the extensive experience with 10 CFR Part 50 and would 

align better with alternative reactor designs. NUREG-1860, “Feasibility Study for a RiskInformed and Performance-Based Regulatory Structure for Future Plant Licensing” (NRC, 

2007), documents this feasibility assessment and provides a framework “that provides an 

approach, scope and criteria that could be used to develop a set of requirements that would 

serve as an alternative to 10 CFR [Part] 50 for licensing future” nuclear power plants.

One key part of the NUREG-1860 framework is defense-in-depth, characterized as having the 

following principles:

•	 Measures against intentional acts as well as inadvertent events are provided. 

•	 The design provides accident prevention and mitigation capability. 
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C-6 | Risk Management Task Force Appendix C

•	 Accomplishment of key safety functions is not dependent upon a single element of design, 

construction, maintenance, or operation.

•	 Uncertainties in SSCs and human performance are accounted for in the safety analyses.

•	 The design has the capability to prevent an unacceptable release of radioactive material. 

•	 Plants are sited at locations that facilitate the protection of public health and safety.

Like the other characterizations discussed above, NUREG-1860 focused on power reactors. As 

noted previously, the RMTF characterization is intended to apply to a broader set of regulated 

activities. The RMTF characterization, as extended for power reactors as discussed in 

Section C.4, encompasses the NUREG-1860 characterization.

C.4 Extension of RMTF Characterization for Power Reactors

The RMTF concluded that its characterization of defense of depth could be extended for power 

reactors to be more specific and to reflect the availability of quantitative methods (probabilistic 

risk assessments). This extension (with additions in italics) is:

Provide risk-informed and performance-based defense-in-depth protections to:

•	 Ensure appropriate barriers, controls, and personnel to prevent, contain, and 

mitigate exposure to radioactive material according to the hazard present, the 

relevant scenarios, and the associated uncertainties.

o Each barrier is designed with sufficient safety margins to maintain its 

functionality for relevant scenarios and account for uncertainties.

o Systems that are needed to ensure a barrier’s functionality are designed 

to ensure appropriate reliability for relevant scenarios.

o Barriers and systems are subject to performance monitoring.

and

•	 Ensure that the risks resulting from the failure of some or all of the established 

barriers and controls, including human errors, are maintained acceptably low.

In addition to those discussed in Section C.2, key terms in this characterization for power 

reactors include:

•	 Safety margins: The traditional engineering approach of including margins in equipment 

design to account for uncertainties.

•	 Appropriate reliability: The likelihood of successful operation needed by a system such that, 

considering the frequency of systems challenges, the risk is acceptably low.
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•	 Performance monitoring: Periodically assessing the performance of a barrier or system 

to ensure its continued functionality. Barrier monitoring can be more qualitative in nature 

while system monitoring can be accomplished by programs such as the mitigating system 

performance indicators used in the NRC’s Reactor Oversight Process.

C.5 References

(IAEA, 1996) International Atomic Energy Agency, “Defence in Depth in Nuclear 

Safety,” INSAG-10, A Report by the International Nuclear Safety 

Advisory Group, 1996.

(NGNP, 2009) Idaho National Laboratory, “Next Generation Nuclear Plant 

Defense-in-Depth Approach,” INL/EXT-0917139, December 2009. 

(NRC,1999) U.S. Nuclear Regulatory Commission, Staff Requirements 

Memorandum Regarding SECY-98-44, “White Paper on RiskInformed and Performance-Based Regulation,” March 1, 1999, 

Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML003753601.

(NRC, 2004) U.S. Nuclear Regulatory Commission, “Risk-Informed 

Categorization and Treatment of Structures, Systems, and 

Components for Nuclear Power Reactors,” 10 CFR 50.69, 

Published in the Federal Register on November 22, 2004 

(69 FR 68008).

(NRC, 2007) U.S. Nuclear Regulatory Commission, “Feasibility Study 

for a Risk-Informed and Performance-Based Regulatory 

Structure for Future Plant Licensing,” NUREG-1860, Volume 1, 

December 2007, ADAMS Accession No. ML080440170.

(NRC, 2011) U.S. Nuclear Regulatory Commission, “Recommendations for 

Enhancing Reactor Safety in the 21st Century; The Near-Term 

Task Force Review of Insights from the Fukushima Dai-Ichi 

Accident,” July 2011, ADAMS Accession No. ML112510271.

Appendix D Risk Management Task Force | D-1

A Proposed Risk Management Regulatory Framework

APPENDIX D

PERFORMANCE-BASED REGULATION

D.1 Background

The U.S. Nuclear Regulatory Commission (NRC) has a longstanding goal to move toward 

more risk-informed and performance-based approaches in its regulatory programs. The Risk 

Management Task Force (RMTF) has found that such efforts have been useful, and it has 

incorporated “risk-informed and performance-based defense-in-depth” as a major part of the 

proposed Risk Management Regulatory Framework. The Commission has previously directed 

the NRC staff to solicit input from industry and other stakeholders on performance-based 

initiatives, including areas that are not amenable to risk-informed approaches, to supplement 

the agency’s traditional deterministic system of licensing and oversight. It should be noted 

that deterministic1 and prescriptive2 regulatory requirements were based mostly on experience, 

testing programs, and expert judgment, considering factors such as engineering margins and 

the principle of defense-in-depth. These requirements are viewed as being successful in 

establishing and maintaining adequate safety margins for NRC-licensed activities. The NRC 

has recognized, however, that deterministic and prescriptive approaches can limit the flexibility 

of both the regulated industries and the NRC to respond to lessons learned from operating 

experience and to adopt improved designs or processes. 

The agency has as one of its primary safety goal strategies the use of sound science and 

state-of-the-art methods to establish, where appropriate, risk-informed and performance-based 

regulations. The NRC issued a white paper on risk-informed and performance-based regulation 

to define the terminology and expectations for evaluating and implementing the initiatives 

related to risk-informed, performance-based approaches (NRC, 1999). The white paper defines 

a performance-based approach as follows:

A performance-based regulatory approach is one that establishes performance and results as 

the primary basis for regulatory decisionmaking, and incorporates the following attributes: 

1. measurable (or calculable) parameters (i.e., direct measurement of the physical 

parameter of interest or of related parameters that can be used to calculate the 

parameter of interest) exist to monitor system, including facility and 

licensee, performance, 

1 A deterministic approach to regulation establishes requirements for engineering margin and for quality 

assurance in design, manufacture, and construction. In addition, it assumes that adverse conditions can 

exist and establishes a specific set of design-basis events and related acceptance criteria for specific 

systems, structures, and components based on historical information, engineering judgment, and desired 

safety margins. An example is a defined load on a structure (e.g., from wind, seismic events, or pipe 

rupture) and an engineering analysis to show that the structure maintains its integrity. 

2 A prescriptive requirement specifies particular features, actions, or programmatic elements to be included 

in the design or process, as the means for achieving a desired objective. An example is a requirement 

for specific equipment (e.g., pumps, valves, heat exchangers) needed to accomplish a particular function 

(e.g., remove a defined heat load).
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D-2 | Risk Management Task Force Appendix D

2. objective criteria to assess performance are established based on risk insights, 

deterministic analyses, and/or performance history, 

3. licensees have flexibility to determine how to meet the established performance 

criteria in ways that will encourage and reward improved outcomes, and 

4. a framework exists in which the failure to meet a performance criterion, while 

undesirable, will not in and of itself constitute or result in an immediate 

safety concern.3

Performance-based approaches can be pursued either independently or in combination 

with risk-informed approaches. The white paper mentioned above was a followup to the 

NRC’s assessment of Direction Setting Issue (DSI) 12, “Risk-Informed Performance-Based 

Regulation,” in the mid-1990s (NRC, 1996). The NRC staff and Commission continued to 

make progress on developing policies and guidance related to performance-based approaches 

and subsequently issued documents such as SECY-00-0191, “High Level Guidelines for 

Performance-based Activities” (NRC, 2000), and NUREG/BR-0303, “Guidance for Performancebased Regulation” (NRC, 2002). 

D.2 Nuclear Power Reactors

Within the nuclear power reactor arena, performance-based approaches have been proposed 

and implemented for regulatory programs related to designrelated activities and operational 

activities. As with the risk-informed initiatives, there are meaningful distinctions between 

performance-based activities that primarily deal with requirements for plant design and those 

that deal with plant operation.

D.2.1 Operation

The NRC has traditionally considered the performance of equipment and personnel in 

its development of regulatory and oversight programs. Examples of NRC activities that 

incorporated some aspects of performance-based approaches before defining a more structured 

methodology include the incremental changes to surveillance test intervals (e.g., Generic Letter 

93-05, “Line Item Technical Specifications Improvements to Reduce Surveillance Requirements 

for Testing During Power Operation” (NRC, 1993)), implementation of the leakbeforebreak 

methodology for nuclear power reactors (NRC, 2007), and the “as low as is reasonably 

achievable” (ALARA) provisions in Title 10 of the Code of Federal Regulations (10 CFR) 

Part 20, “Standards for Protection Against Radiation.” 

Although initiated prior to the development of the formal performance-based approach, 

the maintenance rule (10 CFR 50.65, “Requirements for Monitoring the Effectiveness of 

Maintenance at Nuclear Power Plants”) is often cited as a good example of a risk-informed, 

performance-based approach. The maintenance rule requires monitoring of the overall 

continuing effectiveness of a licensee’s maintenance programs to ensure that (1) safety-related 

and certain non-safety-related structures, systems, and components (SSCs) are capable of 

performing their intended functions, and (2) for non-safety-related equipment, failures will not 

3 Using the previous example (footnote 2), a performance-based approach might provide additional flexibility 

to a licensee on plant equipment and configurations used to accomplish a safety function (e.g., removing a 

heat load), but the performance criteria could not be the actual loss of a safety function that would result in 

the release of radioactive materials. 
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A Proposed Risk Management Regulatory Framework

occur that prevent the fulfillment of safety-related functions, and failures resulting in scrams and 

unnecessary actuations of safety-related systems are minimized. The rule requires licensees 

to monitor the performance of the defined SSCs against licensee-established goals, which are 

established based on safety significance and operating experience, and take corrective action 

when the goals are not met. In addition, the rule requires that licensees assess and manage 

the increase in risk that might result from proposed maintenance activities (e.g., taking safety 

significant systems out of service). The scope of the maintenance rule therefore brought more 

SSCs into the NRC’s regulatory programs but did so while providing licensees with flexibility to 

define methods of monitoring, performance criteria, and reliability goals for specific SSCs. 

In 1995, the NRC amended 10 CFR Part 50, “Domestic Licensing of Production and Utilization 

Facilities,” Appendix J, “Primary Reactor Containment Leakage Testing for WaterCooled 

Power Reactors” (NRC, 1995), to provide a performance-based Option B for the containment 

leakage testing requirements. Option B requires that test intervals for various leakage tests 

be determined by using a performance-based approach. Under this approach, performancebased test intervals are based on consideration of the operating history of the component (e.g., 

an isolation valve or the containment structure) and resulting risk from its failure. Option B, in 

concert with technical reports and guidance documents, has been used to allow licensees with 

a satisfactory integrated leak rate testing (ILRT) performance history (i.e., two consecutive, 

successful Type A tests) to reduce the test frequency for the Type A containment ILRT from 

three tests in 10 years to one test in 15 years. This relaxation was based on assessments 

performed by the nuclear industry and the NRC that showed that the risk increase associated 

with extending the ILRT surveillance interval was very small.

A recent initiative that adopts a risk-informed and performance-based approach is the 

incorporation of the National Fire Protection Association (NFPA) standard NFPA 805, 

“Performance-based Standard for Fire Protection for Light-Water Reactor Electric Generating 

Plants,” into NRC’s regulations (NRC, 2009). NFPA 805 provides deterministic requirements 

that are very similar to those in the NRC’s traditional fire protection regulations, but also 

includes performance-based methods for evaluating plant configurations that would not 

meet the conservative deterministic requirements. The performance-based methods allow 

engineering analyses to demonstrate that the changes in overall plant risk that result from 

these plant configurations is acceptably small and that fire protection defense-in-depth is 

maintained.4 Defense-in-depth, as applied to fire protection, means that an appropriate balance 

is maintained between (1) preventing fires from starting, (2) timely detection and extinguishing 

of fires that might occur, and (3) protection of SSCs important to safety from a fire that is not 

promptly extinguished. The adoption of NFPA 805 provides a licensee with flexibility on how to 

implement its fire protection program while maintaining an acceptable level of fire safety.

4 Building upon the guidance in Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk 

Assessment in Risk-informed Decisions on Plant-specific Changes to the Licensing-basis” (NRC, 2011), 

Regulatory Guide 1.205, “Risk-informed, Performance-based Fire Protection for Existing Light-Water 

Nuclear Power Plants” (NRC, 2009), states:

Prior NRC review and approval is not required for individual changes that result in a risk 

increase less than 1×10-7/year (yr) for CDF [core damage frequency] and less than 1×10-8/ yr 

for LERF [large early release frequency]. The proposed change must also be consistent 

with the defense-in-depth philosophy and must maintain sufficient safety margins. The 

change may be implemented following completion of the plant change evaluation.
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D-4 | Risk Management Task Force Appendix D

D.2.2 Design

The concept of performance-based approaches in the design of nuclear plant SSCs is harder 

to grasp than is its incorporation into operational requirements. Much of the focus in this area 

relates to improvements in the consideration of seismic hazards. The NRC issued Regulatory 

Guide 1.208, “A Performance-based Approach To Define the Site-specific Earthquake Ground 

Motion” (NRC, 2007a) to provide an alternative to the traditional approach for addressing the 

requirements in 10 CFR 100.23, “Geologic and Seismic Siting Criteria.” The regulatory guide 

provides guidance on the development of site-specific ground motion response spectrum 

(GMRS). The goal of the site-specific, performance-based GMRS is to achieve approximately 

consistent performance for SSCs, across a range of seismic environments, annual probabilities, 

and structural failure frequencies. The approach is termed performance-based because 

it defines a process built on establishing design criteria for SSCs that are derived from 

performance goals related to withstanding site-specific seismic hazards (e.g., less than about 

1 percent probability of unacceptable performance for the site-specific response spectrum 

ground motion). While this effort establishes alternate design criteria for use with traditional 

or risk-informed engineering analyses, other design efforts could consider performance-based 

approaches by coupling the design efforts with operational information, such as inspections, 

performancegoals, and riskinsights. The use of inspections, tests, analyses, and acceptance 

criteria within 10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear Power 

Plants,” as the means of confirming nuclear plant construction conforms to licensing-basis 

documents can be considered such a performance-based approach. 

D.3 Nonreactor Activities

In the late 1990s to early 2000s, the staff undertook a number of programmatic activities to 

better include performance considerations into the nuclear materials program. In addition 

to DSI-12 mentioned above, these activities were documented in SECY-99-062, “Nuclear 

Byproduct Material Risk Review” (NRC, 1999a), SECY-99-100, “Framework for Risk-informed 

Regulation in the Office of Nuclear Material Safety and Safeguards” (NRC, 1999b), NUREG/

CR-6642, “Risk Analysis and Evaluation of Regulatory Options for Nuclear Byproduct Material 

Systems” (NRC, 2000a), and the “Phase II – Byproduct Materials Review” (NRC, 2001). As a 

result of this work, significant changes were made to the licensing and inspection program to 

better incorporate performance considerations. 

The NUREG-1556 series, Volumes 1-21, “Consolidated Guidance About Materials Licensees,” 

was developed in the late 1990s to pull together into one place the various guidance documents 

written over the years for the wide variety of materials licensees (NRC, 1998). These 

documents allow license applicants to find the applicable regulations, guidance, and acceptance 

criteria used in granting a materials license. Operational experience (performance) and risk 

insights guided the development of these documents—higher risk activities with significant 

performance challenges have more prescriptive regulations and guidance than do lower risk 

activities. Over time, the guidance in NUREG-1556 has been revised to further incorporate 

performance considerations, and a new revision to the series is under development to address 

security issues.

The materials inspection program was fundamentally revised in 2001—both in terms of 

approach and frequency—based on performance. The inspection approach was modified to 

emphasize licensee knowledge and performance of NRC-licensed activities over document 

review. Inspectors now review a licensee’s program against focus areas that reflect those 
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A Proposed Risk Management Regulatory Framework

attributes that are considered (based on insights from NUREG/CR-6642) to be most risksignificant. If a licensee’s performance against a given focus element during the inspection is 

considered to be acceptable, the inspector moves on to the next focus element. Performance 

concerns or questions lead an inspector to go deeper into that area. Inspection frequencies 

were also revised to take into account the risk of a given operation and the performance 

of licensees over time. At the present time, the Office of Federal and State Materials and 

Environmental Management Programs has tasked a working group with conducting a 

reevaluation of the inspection program to determine whether further adjustments to the 

program are warranted to make it more efficient and effective. A key part of that reevaluation is 

consideration of operational performance, based on event reporting, enforcement history, and 

stakeholder interviews. 

The materials program routinely—at least annually—assesses licensee performance-based 

on events reported by NRC and Agreement State licensees to the Nuclear Materials Events 

Database (NMED). NMED provides a tool for capturing, assessing, and trending licensee 

performance in terms of reportable events. This trending function is an important part of the 

feedback loop for changes to the materials regulatory program: rulemaking, licensing, 

and inspection. 

D.4 NRC Activities

The NRC has implemented some performance-based elements into areas other than 

establishing regulatory requirements, such as its inspection program and reactor oversight 

process. Inspection procedures have been changed to increase the focus on performance, 

reliability, and availability, and comparison of performance to requirements and goals. The 

reactor oversight process complements the NRC’s inspection program by using performance 

indicators related to the performance of SSCs and licensee personnel. Although not involving 

revisions to NRC regulations, these activities have not only improved the effectiveness of the 

NRC, but also focused licensee activities on improving performance and overall plant safety.

D.5 Challenges and Recommendations

Efforts to increase the consideration of performance-based approaches for NRC-licensed 

facilities and activities face many of the same obstacles as those identified for risk-informed 

approaches (see Appendix E) and, as a general matter, any proposal to change wellestablished 

ways of doing business. This is true within the NRC and also within the regulated communities, 

which have sometimes expressed their preference for well defined, prescriptive requirements 

versus performance-based proposals. A specific challenge for performance-based approaches 

relates to terminology and perception. The phrase “risk-informed, performance-based” may 

have already become an overused buzzword as all parties seek to meet expectations that 

their work activities or proposals support the NRC’s policy statement. In addition, the phrase 

has generally become synonymous with the risk-informed aspects of the agency’s goal since 

riskrelated initiatives have received much more attention and resources. It should be noted, 

however, that performance-based approaches often complement and support risk-informed 

approaches, in terms of maintaining validity of assumptions (e.g., availability and reliability), 

and there have been significant successes in using them together. A question going forward 

is how much attention should be given to advancing performance-based approaches separate 

from possible risk-informed initiatives with related performance elements. The proposed 

Risk Management Regulatory Framework being recommended by the RMTF may provide an 

opportunity to help resolve the issues related to terminology and implementation.
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D-6 | Risk Management Task Force Appendix D

Another potential issue facing performance-based approaches relates to the perception that 

problems or negative trends are to be avoided and that reaching a threshold for corrective 

actions is indicative of failure. This “zero-failure” expectation is sometimes applied independent 

of realistic safety or health considerations. A common example is the public perception 

regarding the release of or exposure to very low amounts of radioactive material. This may 

be an unsolvable communication problem, which also applies to risk-informed initiatives, but 

it should be recognized as a challenge when developing proposals for performance-based 

approaches. This issue was raised in DSI-12, which included the question of “What should be 

NRC’s strategy and philosophy with respect to changing NRC’s responsibilities and authority 

in areas of little public risk?” Specific examples of such difficulties include the NRC’s efforts to 

address the release of materials contaminated at very low levels through the Below Regulatory 

Concern Policy Statement (NRC, 1993a) or the later Clearance rulemaking effort (NRC, 2005). 

The risk management approach being proposed by the RMTF may facilitate a graded approach 

for the regulation of NRC-licensed materials, devices, and facilities (e.g., introduction of the 

“design-enhancement” category for nuclear power reactors). This could help in matters where 

the NRC and its stakeholders perceived choices as either “full regulation” or “no regulation,” and 

therefore had difficulty in crafting compromises amenable to all parties. 

The development and use of performance-based approaches have introduced some other 

policy and legal issues. These were evident in the implementation of the maintenance rule, 

as well as in the rollout of the NRC’s reactor oversight process. Each of these efforts required 

numerous meetings, preparation of guidance, revisions to guidance documents, use of the 

NRC Web site for frequently asked questions and other communications, and extensive 

training of the NRC staff and industry personnel. NRC evaluations included in activities such 

as the Risk-Informed Environment initiative (Clark, 2008) and Office of Inspector General 

audits (NRC, 2006) have confirmed the need for additional guidance and training in this area. 

Specific implementation challenges would exist for the various types of facilities and activities 

regulated by the NRC. The challenges facing the proposed Risk Management Regulatory 

Framework would likely be similar to those identified and overcome during the development and 

implementation of these previous initiatives.

D.6 References

(Clark, 2008) Theresa V. Clark, Mark Caruso, Garreth Parry, and Lynn Mrowca, 

“Fostering a Risk-informed Environment in Nuclear Reactor 

Regulation,” American Nuclear Society, International Topical 

Meeting on Probabilistic Safety Assessment and Analysis (PSA 

200), Knoxville, TN, September 2008.

(NRC, 1993) U.S. Nuclear Regulatory Commission, “Line Item Technical 

Specifications Improvements to Reduce Surveillance 

Requirements for Testing During Power Operation,” Generic 

Letter 93-05, September 27, 1993, .Agencywide Documents 

Access and Management System (ADAMS) Accession 

No. ML031070342.

(NRC, 1993a) U.S. Nuclear Regulatory Commission, “NRC Withdraws Below 

Regulatory Concern Policy Statements,” Press Release, 

August 18, 1993.
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(NRC, 1995) U.S. Nuclear Regulatory Commission, “Primary Reactor 

Containment Leakage Testing for WaterCooled Power Reactors,” 

10 CFR Part 50, Appendix J, Final Rule, published in the Federal 

Register on September 26, 1995 (60 FR 49495).

(NRC, 1996) U.S. Nuclear Regulatory Commission, “Staff 

Requirements – COMSECY-96-061– Risk-informed, 

Performance-based Regulation (DSI 12),” Staff 

Requirements Memorandum, April 15, 1997, ADAMS 

Accession No. ML003671740.

(NRC, 1998) U.S. Nuclear Regulatory Commission, “Consolidated 

Guidance About Materials Licensees,” NUREG-1556, 

Multiple Volumes, 1998 and later.

(NRC, 1999) U.S. Nuclear Regulatory Commission, Staff Requirements 

Memorandum Regarding SECY-98-144, “White Paper 

on Risk-informed and Performance-based Regulation,” 

March 1, 1999, (ADAMS) Accession No. ML003753601.

(NRC, 1999a) U.S. Nuclear Regulatory Commission, “Nuclear Byproduct 

Material Risk Review,” SECY-99-062, Commission Paper, 

March 1, 1999, ADAMS Accession No. ML003671237.

(NRC, 1999b) U.S. Nuclear Regulatory Commission, “Framework for 

Risk-informed Regulation in the Office of Nuclear Material 

Safety and Safeguards,” SECY-99-100, Commission 

Paper, March 31, 1999, ADAMS Accession 

No. ML003671288.

(NRC, 2000) U.S. Nuclear Regulatory Commission, “High Level Guidelines for 

Performance-based Activities,” SECY-00-191. Commission Paper, 

September 1, 2000, ADAMS Accession No. ML003742883.

(NRC, 2000a) U.S. Nuclear Regulatory Commission, “Risk Analysis and 

Evaluation of Regulatory Options for Nuclear Byproduct Materials 

Systems,” NUREG/CR-6642, February 2000, ADAMS Accession 

No. ML003693052. (Not publically available)

(NRC, 2001) U.S. Nuclear Regulatory Commission, “Phase II – Byproduct 

Materials Review,” NRC Report, August 2001, ADAMS Accession 

No. ML012270095.

(NRC, 2002) U.S. Nuclear Regulatory Commission, “Guidance for PerformanceBased Regulation,” NUREG/BR-0303, December 2002, ADAMS 

Accession No. ML023470659.
A Proposed Risk Management Regulatory Framework

D-8 | Risk Management Task Force Appendix D

(NRC, 2005) U.S. Nuclear Regulatory Commission, “Staff Requirements 

– SECY-05-0054, “Proposed Rule: Radiological Criteria for 

Controlling the Disposition of Solid Materials (RIN 3150-AH18)”,” 

Staff Requirements Memorandum, June 1, 2005, ADAMS 

Accession No. ML051520185.

(NRC, 2006) U.S. Nuclear Regulatory Commission, “Evaluation of NRC’s 

Use of Probabilistic Risk Assessment (PRA) in Regulating the 

Commercial Nuclear Power Industry,” Office of the Inspector 

General, Audit Report OIG-06-A-24, September 29, 2006, . 

ADAMS Accession No. ML062720275.

(NRC, 2007) U.S. Nuclear Regulatory Commission, “LeakBeforeBreak 

Evaluation Procedures, NUREG-0800, Standard Review Plan for 

the Review of Safety Analysis Reports for Nuclear Power Plants: 

LWR Edition, Section 3.6.3, March 2007, ADAMS Accession 

No. ML063600396

(NRC, 2007a) U.S. Nuclear Regulatory Commission, “A Performance-based 

Approach To Define the Site-specific Earthquake Ground Motion,” 

Regulatory Guide 1.208, March 2007, ADAMS Accession 

No. ML070310619. 

(NRC, 2009) U.S. Nuclear Regulatory Commission, “Risk-informed, 

Performance-based Fire Protection for Existing Light-Water 

Nuclear Power Plants,” Regulatory Guide 1.205, December 2009, 

ADAMS Accession No. ML092730314.

(NRC, 2011) U.S. Nuclear Regulatory Commission, “An Approach for Using 

Probabilistic Risk Assessment in Risk-informed Decisions on 

Plant-specific Changes to the Licensing-basis,” Regulatory 

Guide 1.174, Revision 2, May 2011, ADAMS Accession 

No. ML100910006.
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APPENDIX E

EXPERIENCE WITH RISK-INFORMED REGULATION

E.1 Background

The Energy Research, Inc., report ERI 02-001, “Implementation of Risk-informed Regulation 

at U.S. Nuclear Power Plants” (Dube, 2002), included a survey of a representative sample of 

probabilistic risk assessment (PRA) managers (corresponding to 25 percent of the nuclear units) 

regarding their views on risk-informed initiatives. The report noted that substantial progress 

had been made regarding the implementation of risk-informed regulation. However, the report 

also noted that this progress was rather uneven across programs and across licensees. A few 

licensees, such as the South Texas Project (STP), had made significant strides, while a number 

of plants, particularly single-unit sites, continued to linger. A universal theme expressed by 

PRA managers in the 2002 timeframe was that maintaining the PRA models to the necessarily 

high standards was resource-intensive. Additionally, a sizable fraction of in-house PRA staff 

supported day-to-day plant operations and maintenance, consistent with Title 10 of the Code of 

Federal Regulations (10 CFR) 50.65(a)(4), the “maintenance rule.” The result was an already 

too thin staff spread even thinner by the demands of the plant and existing regulations.

Figure E-1 shows the results for several risk-informed applications from the 2002 survey. 

Figure E-1 Status of Risk-Informed Activities (2002) Figure E-1 Status of Risk-Informed Activities (2002)
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E-2 | Risk Management Task Force Appendix E

E.2 Update to 2011

In response to the Commission’s staff requirements memorandum on SECY-10-0121, 

“Modifying the Risk-informed Regulatory Guidance for New Reactors” (NRC, 2010; NRC 2011), 

the NRC staff, with the participation of stakeholders, conducted tabletop exercises regarding the 

implementation of a number of key risk-informed initiatives at operating power reactors. The 

adequacy of the current guidance, if and when applied to new reactor designs (with generally 

lower risk profiles), was the primary focus of the assessment. Although it was not the intent of 

these exercises to perform a formal survey of licensee attitudes toward risk-informed initiatives, 

the degree of interest in risk-informed applications for operating power reactors, as well as new 

reactors, was evident during the discussions.

Maintenance Rule (a)(4): 

The views from 2002 have not diminished over the past decade. Discussions with every PRA 

manager and lead engineer at the time reinforced the notion that this requirement has added 

measurably to nuclear safety. It has resulted in an unprecedented increase in the prestige (and 

workload) of PRA organizations, which heretofore may have been just obscure analysts in a 

remote part of the overall company. Nuclear stations have created a dedicated panel of plant 

personnel, called integrated decisionmaking panels (IDPs), to assess the safety significance 

of performing maintenance during all phases of plant operation. The IDP brings together PRA 

staff, reactor operators, maintenance staff, and planners whose goal is to exercise the PRA 

models (and judgment) to their fullest. This has fostered an awareness of the safety impact of 

working online on balance-of-plant equipment, such as circulating water pumps, which could 

measurably increase reactor trip frequency.

During the conduct of the recent tabletop exercises discussed above, staff observed that many, 

if not most, current licensees used presolved risk configuration tables or online risk monitors 

to assess risk before and during plant maintenance activities. The “blended” approach, 

whereby the PRA is combined with inputs on the degree of defense-in-depth and planttransient 

assessment was highlighted. Stakeholders demonstrated how factors other than PRA were 

often more limiting in terms of the risk management action level. While the staff noted during 

the exercise that there have been 116 violations of the (a)(4) provision of the maintenance 

rule since May 1, 2001, they all have been green findings, with 115 of those being noncited 

violations (NCV) and 1 cited as failure to correct a previous NCV. Implementation of 10 CFR

50.65(a)(4), by almost any measure, appears to be a successful risk-informed regulation. New 

reactors will be required by regulation to have and maintain an essentially fullscope PRA (all 

operating modes and external hazards for which staff has endorsed consensus PRA standards), 

so the online risk management tools are expected to be even more robust. 

Risk-Informed Technical Specifications: 

The scope of risk-informed technical specification initiatives (RITS) span the entire spectrum—

from wholesale changes that have been proposed to highly strategic isolated changes in 

allowed outage times (AOT). Most licensees have participated in owners group efforts to 

increase surveillance test intervals (STI), such as has been done on instrumentation and reactor 

trip channels by the now superseded Westinghouse Owners Group (WOG). Others, such as 

the previous Combustion Engineering Owners Group, coordinated proposed AOT changes to 

lowpressure safety injection pumps, safety injection tanks, containment spray, and emergency 

diesel generators (EDG). Most plants that have applied for extensions in AOT for EDGs have 
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been approved for as much as 14 days. Since 2002, the importance of this application has 

not diminished, but it is fair to say that over the past decade much of the “low-hanging fruit” in 

terms of greatest gain in safety and operational flexibility for the lowest implementation cost 

has been gathered. Nonetheless, nearly half of the reactor units have received NRC approval 

to implement RITS 5b (surveillance frequency control program). The more comprehensive 

RITS 4b (completion times) has to-date received limited interest with only STP currently 

applying the technology and several other plants expressing near-term interest. For new 

reactor designs, the USAPWR design center has developed a 4b and 5b option in the technical 

specifications, with Comanche Peak Units 3 and 4 pursuing the technology. Other combined 

license (COL) applicants have taken a “wait-and-see” approach, not willing to risk holding up 

their COL approval until the groundwork has been established by pilot applications. Many 

applicants have stated informally that once they obtain their COLs, they intend to pursue a wide 

range of risk-informed initiatives, including 4b and 5b.

Appendix J, Integrated Leak Rate Testing (ILRT): 

Another application, where nearly all licensees who have tried have had some degree of 

success, is risk-informed containment leak rate testing in its various forms. Requirements 

are found in Appendix J of 10 CFR Part 50, “Domestic Licensing of Production and Utilization 

Facilities. Type A testing relates to the overall measurement of containment air mass and 

leakage rate by pressurizing the containment during an outage (i.e., ILRT). Type B testing is 

a series of pneumatic tests to detect and measure local leak rates across specified pressure 

retaining boundaries, such as the personnel air lock door seal. Type C testing is a set of 

pneumatic tests to measure containment leakage rates through containment isolation valves. 

Current requirements call for Type A ILRT once in 10 years. Based in part on successful past 

performance on these tests, as well as risk assessments, many licensees have requested 

extensions of the test interval to once every 15 years. Both the WOG and Nuclear Energy 

Institute have had major efforts in this area. Again, depending on the particulars of a refueling 

outage, the ILRT may often be a critical path item because of the necessity to isolate and 

pressurize the containment. While Appendix J was not one of the topical areas that was 

exercised during the aforementioned tabletop workshops in 2011, discussions with industry 

representatives and staff indicate a large fraction of the nuclear power plant units have used 

this initiative. 

In-Service Inspection (ISI) of piping: 

Operating nuclear power plants must inspect piping in accordance with the ASME Boiler and 

Pressure Vessel Code, Section XI. The original inspection programs are based on deterministic 

analyses using design stress reports. These analyses are extremely conservative. Service 

experience has shown that historical pipe failures occur caused by corrosion and fatigue and in 

locations not addressed by the existing inspection programs. American Society of Mechanical 

Engineers (ASME) Code Cases have been providing much of the technical basis for proceeding 

with these revisions. About half the nuclear plant PRA managers surveyed in 2002 had a 

major submittal for relief, either awaiting approval or already approved, and that percentage 

has increased over the past decade to about 90 percent of the U.S. nuclear units. Those who 

have had success with risk-informed ISI (RI-ISI) identify this activity as “low hanging fruit,” 

in that the risk benefits are realized while reducing burden and worker radiation exposures. 

During the 2011 tabletop exercises, the NRC staff concluded that implementation of RI-ISI was 

“riskneutral” for the current reactor fleet, and likely to be so for the new reactor designs. The 

Electric Power Research Institute (EPRI) has a research program underway to implement the 
A Proposed Risk Management Regulatory Framework

E-4 | Risk Management Task Force Appendix E

EPRI approach to RI-ISI at many of the new reactor design centers. EPRI has also proposed 

the implementation of risk-informed preservice inspection, but the staff has not approved this 

proposed technology at this time.

Alternative Source Term: 

In 1995, more than 30 years after the Technical Information Document TID-14844 (AEC, 1962), 

the revised or “alternative” source term was published as NUREG-1465 (NRC, 1995). This 

alternative source term assumes the timedependent rather than instantaneous releases of 

fission products, and it envelops all light-water reactor plants with the representative accident 

source terms. A licensee may choose to implement all aspects of the alternative source 

term, including composition and magnitude, timing of release, and chemical and physical 

form of the radiological source term; or the licensee may choose to implement only selected 

aspects of the alternate source term. Guidance for licensees of operating nuclear power 

reactors in applying the alternative source term was provided in Regulatory Guide (RG) 1.183, 

“Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power 

Reactors,” (NRC, 2000) and Standard Review Plan, Section 15.0.1, Revision 0, “Radiological 

Consequences Analysis Using Alternate Source Terms,” (NRC, 2000a).

Alternative source terms have the potential benefit primarily in the relaxing of design 

requirements for a number of systems, such as control room heating, ventilation, and air 

conditioning and secondary containment. To meet control room dose requirements, stringent 

design criteria regarding rate of room pressurization, leakage rates, and other factors often 

must be made. Design deficiencies or system performance issues can place an operating unit 

in a condition outside the design basis of the plant, in some cases jeopardizing continued plant 

operation. Realistic source terms can relax the timing of the actuation of certain systems and 

pose less of a problem in terms of issues, such as emergency diesel generator loading. At the 

time of the original survey in 2002, about two-thirds of the PRA managers were planning to 

have some sort of initiative to credit alternative source terms, but it is not evident that this has 

come to full fruition. The 2011 tabletop exercises did not assess alternative source terms, but 

the standard design applications for certification, including the AP1000, Economic Simplified 

BoilingWater Reactor, USAPWR, and U.S. EPR have availed themselves of the alternative as 

described in RG 1.183 (the ABWR has not). 

In-Service Testing (IST): 

Much like ISI, IST is an ASME Code requirement to provide periodic performance testing of 

check valves, motoroperated valves, airoperated valves, and pumps. RI-IST could potentially 

improve plant safety by assigning higher priority to key equipment, reducing the duration of 

plant outage, and allowing for more efficient use of plant resources. NRC guidance is set 

forth in RG 1.175, “An Approach for Plant-specific Risk-informed Decisionmaking: In-Service 

Testing,” (NRC, 1998). However, the response of the industry in this arena has been less than 

enthusiastic. In the 2002 survey, PRA managers representing about two-thirds of the nuclear 

units had no intention of issuing a licensing submittal in the near future, and that attitude has not 

changed significantly in the past decade. The major argument for not making a submittal is that 

the licensees do not see a major payback. The high cost of the analysis outweighs what little 

savings they project to achieve from the revised IST program. Because there was no expressed 

interest in RI-IST on the part of the new reactor applicants, this topical area was not evaluated 

during the 2011 tabletop exercises.
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Categorization of SSCs (10 CFR 50.69), (Also Referred to as Graded QA): 

The implementation of risk-informed categorization and treatment of structures, systems, and 

components (SSCs) under 10 CFR 50.69, “Risk-informed Categorization and Treatment of 

Structures, Systems and Components for Nuclear Power Reactors,” remains an underachieving 

initiative for currently operating reactors. PRA managers surveyed in 2002 did not see a clear 

and immediate benefit, either in reduction of costs nor in terms of increased safety, and that 

view has not changed over the past decade. Only STP has implemented an earlier form of 

10 CFR 50.69 under a waiver of regulations. Since then, only one other licensee has expressed 

any interest in the program. While the potential payback would not seem to justify the effort 

and cost that must be put into such a program for many if not most operating reactors, new 

reactor applicants see large potential savings if the program is implemented before equipment 

procurement for new build. However, during the 2011 tabletop exercises, none of the early COL 

applicants expressed interest in applying 10 CFR 50.69 before receiving their COLs. By then, 

the COL holder would have already issued most of the design specifications for equipment to be 

procured. Hence, some of the greatest benefits of using 10 CFR 50.69 will not be realized for 

the first round of COL holders.

Fire Protection (NFPA 805):

One other risk-informed activity, not included in the 2002 survey, deserves mentioning. 

Changes to fire protection programs (NRC, 2009) through implementation of 10 CFR 50.48(c) 

(National Fire Protection Association (NFPA) 805) is, at this point, one of the major riskinformed initiatives in the industry, with over 40 units pursuing this alternative approach to 

fire protection. In the views of the industry, the effort is consuming most of the industry’s 

“spare” PRA resources. The activity can be considered “high cost, high payback.” High cost 

means that a high quality PRA that substantially conforms to RG 1.200 (NRC, 2009a) for both 

internally initiated events and internal fire is needed. The cost of implementing fire protection 

enhancements can run into the tens of millions of dollars. High payback—in that the 

alternative, continuing with the current fire protection measures, or plant shutdown—is even 

more prohibitive. 

Summary

The NRC staff summarized the results of the tabletop exercises in a public meeting held on 

February 28, 2012 (NRC, 2012). The NRC staff did not identify potential significant decreases 

in the enhanced safety margins for new reactors arising from the various initiatives and 

practices discussed above. The tabletop exercises likewise concluded that the current risk 

thresholds used in the reactor oversight process are appropriate for new reactors. 

E.3 References

(AEC, 1962) U.S. Atomic Energy Commission, “Calculation of Distance Factors 

for Power and Test Reactor Sites,” TID-14844, March 23, 1962.

(Dube, 2002) D.A. Dube, “Implementation of Risk-informed Regulation at 

U.S. Nuclear Power Plants,” ERI 02-001, Energy Research, Inc., 

May 2002.
A Proposed Risk Management Regulatory Framework

E-6 | Risk Management Task Force Appendix E

(NRC, 1995) U.S. Nuclear Regulatory Commission, “Accident Source Terms for 

Light-Water Nuclear Power Plants,” NUREG-1465, February 1995, 

Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML041040063.

(NRC, 1998) U.S. Nuclear Regulatory Commission, “An Approach for Plantspecific, Risk-informed Decisionmaking: In-Service Testing,” 

Regulatory Guide 1.175, August 1998, ADAMS Accession 

No ML003740149.

(NRC, 2000) U.S. Nuclear Regulatory Commission, “Alternative Radiological 

Source Terms for Evaluating Design Basis Accidents at Nuclear 

Power Reactors,” Regulatory Guide 1.183, July 2000, ADAMS 

Accession No ML003716792.

(NRC, 2000a) U.S. Nuclear Regulatory Commission, “Radiological Consequence 

Analysis Using Alternate Source Terms,” NUREG-0800, Standard 

Review Plan, Section 15.0.1, Revision 0, July 2000, ADAMS 

Accession No ML003734190.

(NRC, 2009) U.S. Nuclear Regulatory Commission, “Risk-informed, 

Performance-based Fire Protection for Existing Light-Water 

Nuclear Power Plants,” Regulatory Guide 1.205, December 2009, 

ADAMS Accession No ML092730314.

(NRC, 2009a) U.S. Nuclear Regulatory Commission, “An Approach for 

Determining the Technical Adequacy of Probabilistic Risk 

Assessment Results for Risk-informed Activities,” Regulatory 

Guide 1.200, Revision 2, March 2009, ADAMS Accession 

No ML090410014

(NRC, 2010) U.S. Nuclear Regulatory Commission, “Modifying the Riskinformed Regulatory Guidance for New Reactors,” SECY-10-0121, 

September 14, 2010, ADAMS Accession No ML102230076.

(NRC, 2011) U.S. Nuclear Regulatory Commission, “Modifying the RiskInformed Regulatory Guidance for New Reactors”,” 

SECY-10-0121, Staff Requirements Memorandum, March 2, 2011, 

ADAMS Accession No ML110610166.

(NRC, 2012) U.S. Nuclear Regulatory Commission, “Risk-informed Regulatory 

Framework for New Reactors,” Public Meeting Presentation, 

February 28, 2012, ADAMS Accession No ML12062A198
Appendix F Risk Management Task Force | F-1

A Proposed Risk Management Regulatory Framework

APPENDIX F

NUCLEAR POWER REACTORS – LICENSING-BASIS EVENTS

F.1 Introduction

From the beginning, the licensing and oversight of commercial nuclear reactors has focused 

on potential hazards and the establishment of requirements to prevent, contain, and mitigate 

possible releases of radioactive material. The safety analyses for plants have considered 

both the failure and malfunction of plant equipment (e.g., pumps, piping, and control rods) and 

external hazards (e.g., earthquakes and floods) that might initiate transients and compromise 

barriers. The failures and malfunctions of plant equipment have traditionally been described 

in Chapter 15 of each plant’s final safety analysis report (FSAR). The larger-scale events that 

might affect multiple plant areas or structures are usually discussed in Chapters 2 and 3 of the 

FSAR, as well as in specific sections on systems, structures, and components (SSCs) and 

related operating programs. 

F.2 Internal Events

In the analyses of the failure or malfunction of specific plant SSCs, many plants use a version 

of an American Nuclear Society (ANS) standard that includes classification of plant conditions 

(ANS, 1983). This approach divides plant conditions into four categories, in accordance with 

anticipated frequency of occurrence and potential radiological consequences to the public. The 

four categories are as follows:

Condition I - Normal operation and operational transients

Condition II - Faults of moderate frequency

Condition III - Infrequent faults

Condition IV - Limiting faults

The basic principle applied in relating design requirements to each of the conditions is that 

the most probable occurrences should yield the least radiological risk to the public, and 

those extreme situations having the potential for the greatest risk to the public shall be those 

least likely to occur. The events are analyzed using detailed computer codes that have 

been benchmarked to tests and experiments. Reactor trip and engineered safety systems 

are modeled in the analyses according to established conventions (e.g., the single failure 

criterion and the crediting of only safety-related SSCs). The safety analysis typically includes 

acceptance criteria similar to the following for the various events, in keeping with the established 

relationships between expected event frequency and radiological consequences:

•	 Fuel damage (defined as penetration of the fission product barrier, i.e., the fuel rod clad) 

is not expected during normal operation and operational transients (Condition I) or any 

transient conditions arising from faults of moderate frequency (Condition II). It is not 

possible, however, to preclude a very small number of rod failures. These will be within 

the capability of the plant cleanup system and are consistent with the plant 

design bases.
A Proposed Risk Management Regulatory Framework

F-2 | Risk Management Task Force Appendix F

•	 The reactor can be brought to a safe state following a Condition III event with only 

a small fraction of fuel rods damaged (see above definition), although sufficient fuel 

damage might occur to preclude resumption of operation without considerable 

outage time.

•	 The reactor can be brought to a safe state and the core can be kept subcritical with 

acceptable heat transfer geometry following transients arising from Condition IV events.

The analysis of SSC failures and malfunctions can be described in the above terms of 

postulated initiating events (PIEs), analytical models and assumptions, and acceptance criteria. 

As described in Appendix B, this traditional approach can generally be summarized as follows:

Traditional Approach

•	 limited set of design-basis accidents

o upper bound assumptions

o stylized, conservative models

o single failure criterion

•	 safety margins

•	 mechanistic and physical models

•	 analytical conservatisms 

•	 well suited to support design activities

Routine plant operations are generally controlled by established radiation protection practices 

and requirements. For power reactors, radiation controls for the protection of the public are 

provided in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, “Standards for 

Protection Against Radiation”; 10 CFR Part 50, “Domestic Licenses of Production and Utilization 

Facilities,” Appendix I, “Numerical Guides for Design Objectives and Limiting Conditions for 

Operation to Meet the Criterion ‘As Low is As Reasonably Achievable’ for Radioactive Material 

in Light-WaterCooled Nuclear Power Reactor Effluents”; and effluent controls in the plant 

license and procedures (e.g., 10 CFR 50.36a and offsite dose calculation manual). 

Condition II events are also referred to as anticipated operational occurrences (AOOs) and 

are evaluated to ensure limited fuel failures and, thereby, minimal releases of radioactive 

material. The protection against AOOs is provided primarily by the reactor protection system, 

and analyses are performed to confirm that safety limits are not exceeded. Safety limits are 

defined for adequate heat removal from the reactor’s fuel rods (e.g., specified acceptable fuel 

design limits (SAFDLs)) and pressure relief to protect the reactor coolant pressure boundary (for 

transients resulting in increased reactor coolant system (RCS) pressure). Condition II events 

are selected through evaluation of various events, such as changes (increases or decreases) 

in heat removal by secondary systems, decreases in reactor coolant system flow, reactivity 

and power distribution anomalies, and changes (increases or decreases) in reactor coolant 

inventory. Assuming plant systems perform as expected, safety functions are not challenged by 

AOOs, and a plant would be able to perform repairs, if needed, and return to operation within a 

short time.
Appendix F Risk Management Task Force | F-3

A Proposed Risk Management Regulatory Framework

Infrequent and limiting faults are also referred to as design-basis accidents (DBAs). These 

events are evaluated using acceptance criteria that may include limited fuel failures but maintain 

barriers such that core cooling functions are maintained. This, in turn, would demonstrate that 

the calculated offsite dose would not exceed established criteria related to short-term health 

effects. In general, the actual occurrence of a DBA-related initiating event would be expected 

to challenge plant personnel and require actuation of engineered safety features, such as the 

emergency core cooling system. Although plant conditions may warrant entry into emergency 

operating procedures and would trigger emergency preparedness activities when emergency 

action levels were exceeded, the plant response would be expected to stabilize, and it is unlikely 

to warrant evacuations because of offsite doses challenging the protection action guidelines.

F.3 External Hazards

In addition to the failure or malfunction of plant equipment, the challenges to and protections of 

nuclear power plants from various external hazards are described in plant licensing documents 

(e.g., FSAR Chapter 2). External hazards are those events that could initiate a variety of plant 

transients or accidents while also challenging one or more of the barriers provided to mitigate 

or contain potential releases of radioactive materials. The failure of a specific SSC may be 

evaluated using failure modes and effects analysis and mechanistic models for the thermalhydraulic response of the plant for comparison to established acceptance criteria. However, 

the large number of SSCs that could be affected by a major external event led to a different 

approach for the evaluations and protections from such events. Examples of external 

hazards are:

•	 nearby industrial, transportation, and military facilities

•	 meteorological events (e.g., precipitation, wind, storms)

•	 flooding events (external or internal)

•	 seismic events

•	 fires

•	 malevolent acts

The general approach for external hazards is to incorporate protections or qualification 

requirements related to the hazard to all safety-related components, or at least to a large 

set of equipment subject to the event. Seismic qualification requirements are, for example, 

generally applied to safety-related SSCs. A protection feature added to a subset of safetyrelated components might be flood protection for those rooms possibly affected by the flooding 

of a nearby river. As with the evaluation of AOOs and DBAs, an important part of selecting 

and evaluating the external hazards is the evaluation of expected probability or frequency of 

challenges from external events. Information and estimates related to various external hazards 

are provided in NRC regulatory guides and by agencies such as the U.S. Geological Survey and 

U.S. Army Corps of Engineers. The goal in these evaluations is to provide protections against 

events that can reasonably be predicted to occur at a site. For some PIEs, the NRC guidance 

has established acceptance criteria related to ensuring a low probability of exceeding the 

assumptions (e.g., wind speeds) for external hazards. Site specific information is used to select 

meteorological events, such as the probable maximum precipitation and the potential for wind 
A Proposed Risk Management Regulatory Framework

F-4 | Risk Management Task Force Appendix F

speeds from tornadoes and hurricanes exceeding design values. Regional information is also 

used to determine the maximum probable flood. Plant structures are evaluated to ensure they 

can withstand and provide the necessary protections against such events. Geological records 

are evaluated to estimate the potential effects of seismic events on the plant site and its specific 

SSCs. Several internal events, such as fire and internal flooding scenarios, are addressed in 

a similar manner. Existing guidance related to selected plant level events is summarized in 

Table F-1.

Table F-1 Approaches to External Hazards

Plant-level 

Event Guidance Frequency Discussion

Precipitation

(rain, snow)

Regulatory Guide (RG) 1.206, 

“Combined License Applications for 

Nuclear Power Plants” (NRC, 2007) 

and “Interim Staff Guidance on 

Assessment of Normal and Extreme 

Winter Precipitation Loads on 

the Roofs of Seismic Category I 

Structures” (NRC, 2009)

n/a (historical information)

Hurricanes

RG 1.221, “Design Basis Hurricane 

and Hurricane Missiles for Nuclear 

Power Plants” (NRC, 2011)

design-basis hurricane wind speeds should 

correspond to an exceedance frequency of 

10-7 per year (calculated as a best estimate)

Tornadoes

RG 1.76, “design-basis Tornado and 

Tornado Missiles for Nuclear Power 

Plants” (NRC, 2007a)

design-basis tornado wind speeds should 

correspond to an exceedance frequency of 

10-7 per year (calculated as a best estimate)

Nearby 

Industrial

RG 1.206 (NRC, 2007) and 

RG 1.91, “Evaluation of 

Explosions Postulated To Occur on 

Transportation Routes Near Nuclear 

Power Plants” (NRC, 1978)

Exceedance frequency of going over 

the 10 CFR Part 100, “Reactor Site 

Criteria,” dose limits on the order of 

magnitude of 10-6 per year 

External 

Flooding

RG 1.59, “Design Basis Floods for 

Nuclear Power Plants” (NRC, 1977) n/a (historical information)

Seismic 

RG 1.208, “A Performance-based 

Approach to Define the Site-specific 

Earthquake Ground Motion” (NRC, 

2007b)

The desired performance being expressed as 

the target value of 10-5 for the mean annual 

probability of exceedance (frequency) of the 

onset of significant inelastic deformation 

Aircraft Impact 

(intentional)

RG 1.217, “Guidance for the 

Assessment of beyond-design-basis 

Aircraft Impacts“ (NRC, 2011a)

n/a (defined aircraft impact characteristics)

Security 

Assaults (various) n/a (defined design-basis threat)

Internal Fires

RG 1.205, “Risk-informed, 

Performance-based Fire Protection 

for Existing Light-Water Nuclear 

Power Plants” (NRC, 2009a)

Prior NRC review and approval is not 

required for individual changes that result in a 

risk increase less than 10-7/year (yr) for CDF 

and less than 10-8/yr for LERF

Loss of large 

areas caused 

by fires or 

explosions

Interim Staff Guidance DC/

COLISG-016 – Compliance 

with 10 CFR 50.54(hh)(2) and 

10 CFR 52.80(d) (NRC, 2010)

n/a (defined mitigation strategies)
Appendix F Risk Management Task Force | F-5

A Proposed Risk Management Regulatory Framework

F.4 Deterministic Methods

The terminology related to nuclear plant licensing and relationships between design basis, 

design-basis events, beyond-design-basis accidents, and licensing-basis is confusing. 

The complexity of the terminology has increased over the last several decades as new 

methodologies, such as PRA, were introduced and as the NRC and industry responded to 

specific issues or concerns (e.g., station blackout (SBO)). As explained in “A Short History of 

Nuclear Regulation, 1946–2009” (NRC, 2010a), the initial design and licensing of nuclear power 

plants were approached as follows:

Regulators using a deterministic approach simply tried to imagine “credible” 

mishaps and their consequences at a nuclear facility and then required the 

defense-in-depth approach—layers of redundant safety features—to guard 

against them. 

These “credible mishaps” were, in turn, used to define design-basis events, which were, then, 

used to determine the safety classification of SSCs, the contents of licensing-basis documents, 

such as FSARs and technical specifications, and supporting documents, such as plant 

procedures. The licensing efforts for early plants focused therefore on “design-basis events.” 

Regulator and licensee attention were centered on the mitigation of AOOs and DBAs and on 

ensuring that plant structures and layouts addressed design-basis external hazards such that 

safety-related equipment was protected and plants could proceed from operations to a safe 

shutdown condition following a design-basis event.

Safety Margins

The traditional or deterministic approach addressed uncertainties and provided conservatisms 

through the establishment of design limits, acceptance criteria, and safety margins. There 

are several descriptions of the term “safety margin,” with each recognizing that plants and 

specific barriers are designed and constructed such that actual failures are not expected until 

key parameters well exceed the values assumed in the supporting engineering evaluations. 

However, like many other terms commonly used in licensing processes, “safety margin” is not 

officially defined. 10 CFR 50.59, “Changes, Tests, and Experiments,” once included a criterion 

related to a reduction in the “margin of safety,” but the criterion was replaced specifically 

because of the differing interpretations of the phrase (NRC, 1999). The supplementary 

information associated with the rulemaking included the following characterization:

...Margins within the plant design and in the established licensing-basis 

exist on many levels. There are margins from the assumptions of initial 

conditions, conservatisms such as computer modeling and codes to account 

for uncertainties, allowances for instrument drift and system response time, 

redundancy, and independence of components. Margins are built into the facility 

to account for routine plant fluctuations and transients and response to accident 

conditions. Margins also exist in the established regulatory acceptance criteria to 

be met for response to various accidents and transients. As a result, substantial 

margins are provided by the regulatory envelope within which a plant has 

demonstrated its ability to respond to a spectrum of design basis accidents...
A Proposed Risk Management Regulatory Framework

F-6 | Risk Management Task Force Appendix F

A representation of various margins and how they contribute to global plant safety is shown in 

Figure F-1 (recreated from NEA/CSNI/R(2007)9, “Task Group on Safety Margin Action Plan 

(SMAP), Safety Margins Action Plan – Final Report” (NEA, 2007). 

NUREG-1860 (NRC, 2007c) proposed a definition for safety margin that basically equates to 

the combination of licensing and barrier margins from the above figure. There are continuing 

efforts, such as the Idaho National Laboratory’s Risk-informed Safety Margin Characterization 

project, which are attempting to better define and integrate safety margins into risk assessment 

methodologies to support nuclear plant life extensions or other initiatives (INL, 2009).

F.5 Risk Assessments

Additional insights regarding the risks related to nuclear power plants (what can go wrong, 

how likely is it, what are the consequences) were developed in the last several decades, 

and questions arose as to how to address them in the design and licensing processes. The 

increasing use of risk assessment methodologies resulted in additional events being considered 

during licensing (e.g., SBO, anticipated transient without scram (ATWS)), which informed 

areas such as emergency planning, and led to various initiatives to incorporate risk-informed, 

Figure F-1 Nuclear Event Analysis Margins (NEA, 2007) Figure F-1 Nuclear Event Analysis Margins (NEA, 2007)
Appendix F Risk Management Task Force | F-7

A Proposed Risk Management Regulatory Framework

performance-based activities into the NRC regulation and oversight of nuclear power plants. As 

discussed in Appendix B, the risk assessment approach has the following characteristics:

Risk Assessment Characteristics

•	 numerous event sequences

o systems approach best estimate models

o uncertainty analysis

•	 vulnerability determinations

•	 human and system interactions

•	 realism

•	 well suited to support operational decisions

Increasingly, the NRC’s regulations began to address and be informed by risk assessments 

and consideration of beyond-design-basis events, which involve the failure of safety-related 

functions or PIEs not included in the design-basis events. Although not using the terms “designenhancement events or conditions,” the NRC did establish special regulatory requirements for 

selected events, such as SBO, ATWS, and aircraft impact. In addition, regulations such as the 

maintenance rule and programs such as the Reactor Oversight Process used risk insights to 

require or encourage licensees to manage the overall risks associated with plant operation. 

Other beyond-design-basis scenarios were evaluated to identify potential vulnerabilities (i.e., 

events dominating the risk profile) and to support aggregate estimates of plant risk. Similarly, 

the NRC approach to environmental reviews evolved to consider risk insights and possible 

severe accident mitigation alternatives that should be included in applicants’ environmental 

reports. The addition of the beyond-design-basis events is shown in the lower right region 

of Figures F-2 and F-3. As discussed in Chapter 4 and Appendix H, the distinction between 

“adequate protection” and “safety enhancement” is also an issue when discussing the 

categorization of events and related events and programs.

As a general matter, the frequency of events shown in Figure F-2 decreases as they go from 

normal operations to AOOs to DBAs to the design-enhancement and severe-accident events 

within the beyond-design-basis region. The potential consequences in terms of challenges to 

barriers (primarily fuel cladding) and ultimate release of radioactive material increases across 

event categories, from little to no damage for AOOs to significant core damage for some severe 

accidents. The result is a relationship that is similar to the F-C curve discussed in Appendix B. 

This similarity is important in the identification and evaluation of possible options to revise 

the licensing and oversight of nuclear power plants to better incorporate risk insights and 

performance-based approaches. Figure F-2 addresses the relationship of SSC classification 

within the licensing-basis event categories, but similar associations could be shown for 

procedures or other regulatory programs. Figure F-3 provides a simplified version of the figure 

and includes an estimation of the International Atomic Energy Agency (IAEA) defense-indepth scale (IAEA, 2010) to show that there are similarities between the U.S. and international 

approaches. Figure F-4 provides a simplified depiction of various regulatory documents and 

programs and their relationship to the various event categories.
A Proposed Risk Management Regulatory Framework

F-8 | Risk Management Task Force Appendix F

Figure F-2 Event Categories
Appendix F Risk Management Task Force | F-9

A Proposed Risk Management Regulatory Framework

Event Regulatory Document/Program 

Category 

L

i

c

e

n

s

e 

M

a

i

n

t

e

n

a

n

c

e 

R

u

l

e 

F

S

A

R

Q

A

R

O

P 

~10-2 

~10-4

Postulated 

Initiating 

Event 

Freq 

Normal 

Ops 

AOOs 

DBAs 

DECs 

Severe 

Accidents 

1 

Rad

Release 

2 

E

P 

(1) Selected events 

 (SBO, ATWS, AIA) 

(2) New Reactors 

 (Severe Accident Features & 

 PRA information (Chap 19)) 

P

r

o

c

e

d

u

r

e

s 

E

O

P

s 

S

A

M

G

s 

& 

E

D

M

G

s 

E

P 

Figure F-4 Licensing Basis Events, Documents and Programs 

Figure F-3 Event Categories and IAEA Defense-in-depth

Figure F-4 Licensing-basis Events, Documents and Programs
A Proposed Risk Management Regulatory Framework

F-10 | Risk Management Task Force Appendix F

As discussed throughout this report, the existing construct has generally provided sufficient 

levels of protection. To better integrate the traditional engineering approaches and risk 

assessment methodologies, the RMTF has developed the risk-informed and performance-based 

defense-in-depth definition for all NRC regulatory programs and the additional clarification of the 

terminology for nuclear power reactors that was presented in Chapter 4.

In attempting to define options and possible implementation transitions, the concept of riskinformed and performance-based defense-in-depth provides general guidance, but to be more 

useful, it needs to be translated into the actual parts of the licensing process shown in Figure 

F-2 and the related general construct of the event analyses summarized in Table F-2:

Table F-2 Licensing Approach to Events (current approach)

PIEs Key Assumptions Acceptance Criteria

Normal Ops n/a n/a

Maintain initial conditions, standby equipment

Part 20, ALARA

AOOs

Functional list

(e.g., ANS 51.1/52.1 

Cond II)

(PIE frequency 

< ~10-2/yr)

Credit SR SSCs

single failure

Maintain safety limits, 

such as SAFDLs, RCS 

pressure

DBAs

Functional list

(e.g., ANS 51.1/52.1 

Cond III/IV)

(PIE frequency

 > ~10-4/yr)

Credit SR SSCs

single failure

Maintain coolable core, 

containment

Design-basis

External

Events

Maximum credible scenario (e.g., frequency of 

exceedance > ~10-7 for tornado conditions)

Engineering analysis, 

safety margins

Codes and standards 

related to barrier 

design and equipment 

qualification

Design-enhancement

(in concept but term not 

currently used)

Specific events:

ATWS

SBO

AIA

As defined in each rule – generally best estimate

As defined in each rule 

(alternate shutdown, 

coping time, AIA 

functions)

Residual Risks

(in concept but term not 

currently used)

n/a n/a n/a

Beyond-design-basis 

External Events n/a n/a n/a
Appendix F Risk Management Task Force | F-11

A Proposed Risk Management Regulatory Framework

A hurdle to developing a more coherent structure for licensing-basis events and the related 

analyses (deterministic and risk assessments) is the lack of a requirement for licensees to have 

and maintain plant-specific PRAs. 

Operating Reactors

Licensees for operating nuclear power reactors are not required to have a PRA, but they 

have generally completed assessments of internal and external events in response to 

Generic Letter 88-20, “Individual Plant Examinations for Severe Accident Vulnerabilities” 

(NRC, 1988), including Supplement 4, “Individual Plant Examination of External Events 

(IPEEE) for Severe Accident Vulnerabilities” (NRC, 1991). In addition, most licensees have 

performed and maintain some form of a plant-specific PRA to support assessments required 

by the maintenance rule (10 CFR 50.65), environmental reviews for license renewal, and the 

significance determination process within the reactor oversight program. These assessments 

might have led to plant modifications or other actions to address identified risk concerns, but 

the assessments themselves were not incorporated into NRC’s licensing-related activities. 

Many operating reactor licensees have performed PRAs that comply with NRC-endorsed codes 

and standards (NRC, 2009b) to support risk-informed licensing actions or voluntary adoption 

of risk-informed regulations related to fire protection (10 CFR 50.48) or treatment of SSCs 

(10 CFR 50.69). These activities may include NRC reviews or audits of the licensee’s PRAs 

and establishment of regulatory programs for maintaining PRAs, but they do not usually result 

in the incorporation of the PRA or its results into the licensing-basis documents. Most of the risk 

assessments performed for operating reactors have been Level 1 PRAs, with some licensees 

performing Level 2 PRAs.1 

The NRC has developed standardized plant analysis risk (SPAR) models for all operating 

plants in the United States, with most of the models limited to internal events, although some 

include external hazards, a few including shutdown conditions, and three progress to Level 2 

PRAs. The NRC has also performed two studies that have included Level 3 PRAs for selected 

operating plants (NRC, 1975; NRC, 1990).

New Reactors

New reactors are licensed under the provisions of 10 CFR Part 52 and are subject to the 

requirements in 10 CFR 50.71, “Maintenance of Records, Making of Reports,” which states:

(h)(1) No later than the scheduled date for initial loading of fuel, each holder of a 

combined license under subpart C of 10 CFR part 52 shall develop a level 1 and a 

level 2 probabilistic risk assessment (PRA). The PRA must cover those initiating 

events and modes for which NRC-endorsed consensus standards on PRA exist 

one year prior to the scheduled date for initial loading of fuel.

1 For the type of nuclear plant currently operating in the United States, a PRA can estimate three levels of risk:

• A Level 1 PRA, which estimates the frequency of accidents that cause damage to the nuclear reactor 

core. This is commonly called core damage frequency.

• A Level 2 PRA, which starts with the Level 1 core damage accidents and estimates the frequency of 

accidents that release radioactivity from the containment.

• A Level 3 PRA, which starts with the Level 2 radioactivity release accidents and estimates the 

consequences in terms of injury to the public and damage to the environment. 
A Proposed Risk Management Regulatory Framework

F-12 | Risk Management Task Force Appendix F

(2) Each holder of a combined license shall maintain and upgrade the PRA required 

by paragraph (h)(1) of this section. The upgraded PRA must cover initiating events 

and modes of operation contained in NRC-endorsed consensus standards on PRA 

in effect one year prior to each required upgrade. The PRA must be upgraded every 

four years until the permanent cessation of operations under § 52.110(a) of 

this chapter.

(3) Each holder of a combined license shall, no later than the date on which the 

licensee submits an application for a renewed license, upgrade the PRA required by 

paragraph (h)(1) of this section to cover all modes and all initiating events.

Summaries of PRAs are submitted by reactor designers seeking certification of their design, as 

well as applicants and holders of combined licenses. The NRC performs reviews and audits, 

and the PRA summaries are included in Chapter 19 of final safety analysis reports (FSARs). 

The PRAs support decisions related to the regulatory treatment of non-safety systems (RTNSS), 

availability programs, severe accident mitigation alternatives, and various risk-informed 

proposals included in applications (e.g., risk-informed technical specifications). The NRC has 

developed SPAR models for new reactor designs certified by the NRC.

Generation IV Reactor Designs

As discussed in Chapter 4 and Appendix H, Generation IV reactors introduce technologies such 

as liquidmetal and helium as coolants, different fuel forms, and other significant differences from 

light-water reactors. In addition, international activities and goals for Generation IV reactors 

include the proposed use of inherent safety features and a reduced reliance on offsite emergency 

preparedness. The technology differences and deployment goals have led the designers and 

proponents of Generation IV reactors to include Level 3 PRAs as part of their design and 

licensing plans. An example is the set of activities related to the next generation nuclear plant 

(NGNP, 2011).

F.5 Possible Changes to or Additions of Event Categories

The proposed Risk Management Regulatory Framework is intended to improve the NRC’s 

regulation and oversight of nuclear power reactors by providing a logical way to determine 

the appropriate barriers and controls needed to provide risk-informed and performance-based 

defense-in-depth. To accomplish this, the concept of risk-informed and performance-based 

defense-in-depth needs to be integrated into the overall assessment of events and hazards 

that is shown in Figures F-2 and F-3. As discussed in Chapter 4 and Appendix H, the most 

notable change in the RMTF proposals is the formalization of the design-enhancement category 

within the traditional beyond-design-basis regime. Although it might be beneficial to simply add 

the design-enhancement category, the best approach would use this opportunity to ensure an 

overall construct of events, analyses, and resultant plant safety features and programs within 

categories and logical transitions between categories.

A longstanding issue related to the regulation of nuclear power plants relates to the distinction 

between actions needed to ensure “adequate protection” and those pursued as “safety 

enhancements.” The RMTF has included recommendations in Chapter 4 and Appendix H 

to clarify this issue by relating the two tiers (adequate protection and safety enhancements) to 

design-basis events and the design-enhancement category. This is represented in Figure F-5:
Appendix F Risk Management Task Force | F-13

A Proposed Risk Management Regulatory Framework

Unfortunately, it is very difficult to consistently align defense-in-depth, core safety functions, 

licensing-basis events, safety classification, and other topics with the tiers of adequate 

protection and safety enhancements. Other organizations (WENRA, 2008; IAEA, 2010) 

have developed a more coherent set of standards that, although similar in many respects to 

the NRC’s processes, would be difficult to adopt without a significant change in the overall 

regulatory framework (i.e., Option C in Chapter 4). Several alternatives for constructing the 

design-enhancement category with limited changes to the existing regulatory framework 

(i.e., Option B in Chapter 4) are provided in Appendix H. 

It is possible that the creation of the design-enhancement category could help efforts related 

to improving the identification, analysis, and protections against design-basis events. 

Although there has been a general trend of improving computer models and addressing 

some overconservatisms in analyzing the design-basis events (e.g., statistical treatment of 

uncertainties), the assumptions and analyses in Chapter 15 of the FSAR for a specific plant or 

reactor design have been basically unchanged for decades. Revised analyses to support power 

uprates or major plant modifications have taken advantage of limited changes in computer 

codes but are otherwise similar to analyses performed in the 1970s. Previous attempts to revise 

the handling of design-basis events have encountered significant resistance, partly because it 

was difficult to reach agreement on the regulatory treatment of barriers and controls related to 

the revised scenarios. The scenarios in this case are comprised of specified initiating events, 

behavior of mitigating systems, inclusion of coincident failures and single failure criterion, and 

the limits established on fission product barriers, such as the fuel cladding, reactor coolant 

pressure boundary, and containment structure. Several of the previous initiatives are briefly 

described below:

Figure F-5 Regulatory Framework for Nuclear Power Reactors
A Proposed Risk Management Regulatory Framework

F-14 | Risk Management Task Force Appendix F

•	 “Best-Estimate” or Realistic Loss-of-coolant Accident Analysis: The NRC revised its 

regulations (10 CFR 50.46) in 1988 to allow methods other than those prescribed 

in Appendix K to 10 CFR Part 50 for the evaluation of emergency core cooling 

systems (NRC, 1989). The changes were made to credit research that showed the 

Appendix K methods resulted in estimates of system performance that were significantly 

more conservative than originally thought. The revised rule allowed, but did not require, 

use of more realistic models, with sufficient supporting justifications and treatment of 

uncertainties. Reactor vendors have developed related topical reports, and a significant 

number of licensees have taken advantage of the revised analyses to support power 

uprates or other plant changes.

•	 Changes to Single Failure Criterion: The NRC adopted the single failure criterion as a 

means to promote redundancy and reliability of safety systems needed to respond to 

plant transients and accidents. In this regard, the single failure criterion is an important 

element of the NRC’s traditional approach to defense-in-depth (NRC, 1977a). The need 

to identify the most limiting single failure also resulted in improved analyses of failure 

modes and effects, which in turn influenced plant designs. Some have argued, however, 

that more modern analytical methods can better support plant designs by identifying risksignificant failure combinations and that alternatives to the single failure criterion may do 

better at ensuring system reliability than does the established approach (NRC, 2005). 

•	 Development of Risk-informed Safety Analysis Approach and Pilot Application: 

A response to the RMTF solicitation of public comments on a possible Risk Management 

Regulatory Framework identified a previous industry proposal related to taking a more 

risk-informed approach to transient and accident analyses. The Westinghouse Owners 

Group (WOG) submitted WCAP-16084, “Development of Risk-informed Safety Analysis 

Approach and Pilot Application” (WOG, 2003), to address the classification of transients 

and accidents into realistic categories by considering the frequency of occurrence of the 

event combination (i.e., the initiating event in combination with coincident occurrences 

and a single failure). The WOG subsequently withdrew the topical report based on 

feedback from the NRC staff that the scope of the report was too broad for effective 

review and approval. 

F.6 References

(ANS, 1983) American National Standards Institute/American Nuclear Society, 

ANSI /ANS 51.1-1983, “Nuclear Safety Criteria for the Design of 

Stationary Pressurized Water Reactor Plants,” ANS, LaGrange 

Park, IL.

(IAEA, 2010) International Atomic Energy Agency, “Draft – Safety of Nuclear 

Power Plants: Design,” Draft Safety Guide DS-414, 

September 2010.

(INL, 2009) Idaho National Laboratory (Hess, et. al.), “Risk-Informed Safety 

Margin Characterization,” INL/CON-09-15549, July 2009.

(NEA, 2007) Nuclear Energy Agency, “Task Group on Safety Margin Action 

Plan (SMAP), Safety Margins Action Plan – Final Report,” 

NEA/CSNI/R (2007)9, 2007.
Appendix F Risk Management Task Force | F-15

A Proposed Risk Management Regulatory Framework

(NGNP, 2011) Idaho National Laboratory, “Next Generation Nuclear PlantProbabilistic Risk Assessment White Paper,” 

INL/EXT-11-21270, September 2011.

(NRC, 1975) U.S. Nuclear Regulatory Commission, “Reactor Safety Study: An 

Assessment of Accident Risks in U.S. Commercial Nuclear Power 

Plants,” NUREG-75-014 (WASH-1400), October 1975.

(NRC, 1977) U.S. Nuclear Regulatory Commission, “Design Basis Floods for 

Nuclear Power Plants,” Regulatory Guide 1.59, August 1977, 

Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML003740388.

(NRC, 1977a) U.S. Nuclear Regulatory Commission, “Single Failure 

Criterion,” SECY-77-439, August 17, 1977, ADAMS Accession 

No ML060260236.

(NRC, 1978) U.S. Nuclear Regulatory Commission, “Evaluations of Explosions 

Postulated To Occur on Transportation Routes Near Nuclear 

Power Plants,” Regulatory Guide 1.91, February 1978, ADAMS 

Accession No ML003740286.

(NRC, 1988) U.S. Nuclear Regulatory Commission, “Individual Plant 

Examination for Severe Accident Vulnerabilities,” Generic 

Letter 88-20, November 23, 1988, ADAMS Accession 

No ML031470299.

(NRC, 1989) U.S. Nuclear Regulatory Commission, “Best Estimate Calculations 

of Emergency Core Cooling System Performance,” Regulatory 

Guide 1.157, May 1989, ADAMS Accession No ML003739584.

(NRC, 1990) U.S. Nuclear Regulatory Commission, Severe Accident Risks: An 

Assessment for Five U.S. Nuclear Power Plants - Final Summary 

Report,” NUREG-1150, Volume 1, December 1990, ADAMS 

Accession No. ML040140729.

(NRC, 1991) U.S. Nuclear Regulatory Commission, “Individual Plant 

Examination of External Events (IPEEE) for Severe Accident 

Vulnerabilities,” Generic Letter 88-20, Supplement 4, 

June 28, 1991, ADAMS Accession No ML031150485.

(NRC, 1999) U.S. Nuclear Regulatory Commission, “Changes, Tests, and 

Experiments (10 CFR. 50.59),” Final Rule, published in the 

Federal Register on October 4, 1999 (64 FR 53582)

(NRC, 2005) U.S. Nuclear Regulatory Commission, “Risk-Informed 

and Performance-Based Alternatives to the Single Failure 

Criterion,” SECY-05-0138, August 2, 2005, ADAMS Accession 

No. ML051950619.
A Proposed Risk Management Regulatory Framework

F-16 | Risk Management Task Force Appendix F

(NRC, 2007) U.S. Nuclear Regulatory Commission, “Combined License 

Applications for Nuclear Power Plants,” Regulatory Guide 1.206, 

June 2007, ADAMS Accession No. ML070720184.

(NRC, 2007a) U.S. Nuclear Regulatory Commission, “Design-basis Tornado 

and Tornado Missiles for Nuclear Power Plants,” Regulatory 

Guide 1.76, March 2007, ADAMS Accession No. ML070360253.

(NRC, 2007b) U.S. Nuclear Regulatory Commission, “A Performance-based 

Approach To Define the Site-specific Earthquake Ground Motion,” 

Regulatory Guide 1.208, March 2007, ADAMS Accession 

No. ML070310619.

(NRC, 2007c) U.S. Nuclear Regulatory Commission, “Feasibility Study 

for a Risk-informed and Performance-based Regulatory 

Structure for Future Plant Licensing,” NUREG-1860, Volume 1, 

December 2007, ADAMS Accession No. ML080440170.

(NRC, 2009) U.S. Nuclear Regulatory Commission, “Interim Staff Guidance on 

Assessment of Normal and Extreme Winter Precipitation Loads 

on the Roofs of Seismic Category I Structures,” DC/COL-ISG-7, 

July 1, 2009, ADAMS Accession No. ML091490556.

(NRC, 2009a) U.S. Nuclear Regulatory Commission, “Risk-informed, 

Performance-based Fire Protection for Existing Light-Water 

Nuclear Power Plants,” Regulatory Guide 1.205, December 2009, 

ADAMS Accession No. ML092730314.

(NRC, 2009b) U.S. Nuclear Regulatory Commission, “An Approach for 

Determining the Technical Adequacy of Probabilistic Risk 

Assessment Results for Risk-Informed Activities,” Regulatory 

Guide 1.200, Revision 2, March 2009, ADAMS Accession 

No. ML090410014.

(NRC, 2010) U.S. Nuclear Regulatory Commission, “Interim Staff Guidance 

DC/COL-ISG-016 – Compliance with 10 CFR 50.54(hh)(2) and 

10 CFR 52.80(d),” June 2010 (Not publically available)

(NRC, 2010a) U.S. Nuclear Regulatory Commission, “A Short History of 

Nuclear Regulation, 1946 – 2009,” NUREG/BR-0175, Revision 2, 

October 2010, ADAMS Accession No. ML102980443.
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(NRC, 2011) U.S. Nuclear Regulatory Commission, “Design-basis 

Hurricane and Hurricane Missiles for Nuclear Power Plants,” 

Regulatory Guide 1.221, October 2011, ADAMS Accession 

No. ML110940300.

(NRC, 2011a) U.S. Nuclear Regulatory Commission, “Guidance for the 

Assessment of Beyond-design-basis Aircraft Impacts,” Regulatory 

Guide 1.217, August 2011, ADAMS Accession No. ML092900004.

(WENRA, 2008) Western European Nuclear Regulators’ Association, “WENRA

Reactor Safety Reference Levels,” Reactor Harmonization 

Working Group, January 2008.

(WOG, 2003) Westinghouse Owners Group, “Development of Risk-informed 

Safety Analysis Approach and Pilot Applications,” 

WCAP-16084NP, Revision 0, September 2003.

Appendix G Risk Management Task Force | G-1

A Proposed Risk Management Regulatory Framework

APPENDIX G

SAFETY CLASSIFICATION OF SYSTEMS, STRUCTURES,

AND COMPONENTS

G.1 Background

Nuclear power plants are large industrial facilities that include systems, structures, and 

components (SSCs) to generate electricity and to ensure that the nuclear reactor does not pose 

an undue risk to public health and safety. It is in the economic interests of the plant operator 

for the power generation equipment to be manufactured and maintained such that it supports 

reliable plant operation. The need to define requirements on SSCs that help ensure plant safety 

has also been recognized from the early development of nuclear power. A logic or system to 

classify equipment according to its role in ensuring plant safety is a natural part of the process 

to define requirements for design, operation, and other features of SSCs with safety functions. 

The term “safety-related” has been used to define requirements for the protection of SSCs from 

safe shutdown earthquakes (Title 10 of the Code of Federal Regulations (10 CFR) Part 100, 

“Reactor Cite Criteria”) and is more widely used to distinguish those SSCs warranting special 

treatment in terms of quality assurance, environmental qualification, and applicability of various 

industry codes and standards. The terminology of “safety-related,” “important to safety,” and 

related phrases have also been intertwined with the broader debates about the role of the NRC 

and how it regulates commercial nuclear power plants. 

Safety-related SSCs are defined in 10 CFR 50.2, “Definitions,” as follows:

Safety-related structures, systems, and components 

means those structures, systems, and components that are 

relied upon to remain functional during and following designbasis events to assure:

(1) the integrity of the reactor coolant pressure boundary

(2) the capability to shut down the reactor and maintain 

it in a safe shutdown condition; or

(3) the capability to prevent or mitigate the 

consequences of accidents that could result 

in potential offsite exposures comparable to 

the applicable guideline exposures set forth 

in 10 CFR 50.34(a)(1) or 10 CFR 100.11 

[“Determination of Exclusion Area, Low Population 

Zone, and Population Center Distance,”] of this 

chapter, as applicable.

Design-basis events are defined in 10 CFR 50.49, “Environmental Qualification of Electric 

Equipment Important to Safety for Nuclear Power Plants,” as follows:
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G-2 | Risk Management Task Force Appendix G

Design-basis events are defined as conditions of normal operation, 

including anticipated operational occurrences, design-basis 

accidents, external events, and natural phenomena for which the 

plant must be designed to ensure functions (b)(1)(i) (A) through (C) 

[see above items 1, 2 and 3] of this section.

The terminology associated with SSC classification and related special treatment has been a 

longstanding source of confusion and discussions between the staff and industry. This is in 

part because of the frequent use of the terms “safety-related” and “important to safety” in similar 

discussions and sometimes without distinction. Although attempts to clarify the terminology 

have been ongoing for several decades, there continue to be some communication issues 

regarding the classification of SSCs. As will be discussed later, the introduction of RiskInformed approaches and creation of additional categories and classifications have created 

additional complexity to the discussions. For the purpose of this report, the distinction between 

“safety-related” and “important to safety” is taken from Generic Letter 8401, “NRC Use of the 

Terms ‘Important to Safety’ and ‘Safety Related’” (NRC,1984), and a related memorandum 

entitled “Standard Definitions for Commonly-Used Safety Classification Terms” (NRC,1981). 

The clarification of these terms was also related to policy issues on whether the NRC should 

limit its regulations and oversight to safety-related equipment or the path taken, which was to 

broaden its scope to other matters that contributed to the risks posed by nuclear power plants to 

the public health and safety.

The term “important to safety” was defined as follows:

Definition (from 10 CFR Part 50, “Domestic Licensing of Production 

and Utilization Facilities, Appendix A, “General Design Criteria for 

Nuclear Power Plants”) Important to Safety

“Those structures, systems, and components that provide 

reasonable assurance that the facility can be operated without undue 

risk to the health and safety of the public.”

Additional insights related to “important to safety” include:

•	 Encompasses the broad class of plant features, covered (not necessarily explicitly) 

in the General Design Criteria, that contribute in important ways to safe operation 

and protection of the public in all phases and aspects of facility operation (i.e., normal 

operation and transient control as well as accident mitigation)

•	 Includes Safety-Grade (or Safety-Related) as a subset

•	 Note that “important to safety” in 10 CFR 50.49 includes safety-related, non-safetyrelated whose failure could prevent accomplishment of safety functions defined in 

definitions of safety-related, and certain post-accident monitoring equipment.
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The discussions of safety-related reinforced the definition from NRC regulations (subsequently 

incorporated into 10 CFR 50.2, “Definitions,” and made consistent across various regulations) 

and clarified that “safety-grade” was synonymous with “safety-related.” The class of “safetyrelated” SSCs was described as a subset of “important to safety” SSCs, and reference was 

made to Regulatory Guide (RG) 1.29, “Seismic Design Classification” (NRC, 2007), for a typical 

listing of “safety-related” SSCs for light-water reactors.

Various regulatory guides, industry codes and standards, and other guidance documents include 

discussions of safety-related SSCs and have created subcategories within the broader class 

of safety-related SSCs. For example, the Standards Committee within the American Nuclear 

Society (ANS) has identified safety classes (ANS, 1983) that are summarized as follows:

Table G-1 ANS/ASME Safety Classes

Safety Class 1 Major reactor coolant pressure boundary functions

(reactor pressure vessel, connected piping, reactor coolant loops 

(including primary side of steam generators), pressurizers, and 

recirculation loops)

Safety Class 2 More important pressure boundary functions

(primary containment, containment penetrations, sprays and auxiliary 

systems, emergency core cooling systems, residual heat removal 

systems, secondary side of steam generators, standby liquid control, 

and reactor core isolation cooling)

Safety Class 3 All other nuclear safety-related functions

(nuclear safety-related cooling water, residual heat removal secondary 

side, nuclear safety-related ventilation, chemical and volume control 

system, fuel pool, nuclear safety-related electrical equipment, 

secondary containments)

Non-safety

Related 

Class(es) 

Specified nonnuclear safety-related functions

(radwaste, fire protection, fuel pool cooling and cleanup, effluent 

monitoring, fuel storage and handling, reactor water cleanup, 

turbinegenerator, condenser) 

These or similar classifications are used to define appropriate requirements related to design, 

manufacture, and quality assurance for different types of SSCs or by different standarddeveloping organizations. For example, the American Society of Mechanical Engineers (ASME) 

Boiler and Pressure Vessel Code defines requirements for various quality group classifications 

using Quality Groups A through D that generally align to the above ANS safety classes (see 

RG 1.26, “Quality Group Classifications and Standards for Water, Steam, and RadioactiveWaste-Containing Components of Nuclear Power Plants” (NRC, 2007a)). The safety-related 

designation is likewise used directly or by reference in standards issued by the Institute of 

Electrical and Electronics Engineers, American Concrete Institute, and the American Society for 

Testing and Materials. Much of the specific guidance provided in the various regulatory guides 

and industry standards is related to the special requirements for an SSC given its safety-related 

designation and possible service conditions associated with nuclear power plants.

As implied by the wide application of the concept and terminology of safety-related SSCs, the 

classification of SSCs follows from the definition in the NRC regulations and has evolved, for the 

currently operating reactors, to a point of general consensus and stability. The related concept 
A Proposed Risk Management Regulatory Framework

G-4 | Risk Management Task Force Appendix G

of design-basis events has likewise become fairly well established, and discussions usually 

involve details about assumptions or analyses. However, the Fukushima accident in Japan has 

resulted in some renewed discussions regarding the identification and maintenance of designbasis events. The subjects of design-basis and beyond-design-basis events are discussed in 

more detail in Appendix F.

G.2 Additional Regulatory Requirements (“Design-Enhancement Events”)

The evolution of nuclear plant designs, operating practices, and regulatory requirements has 

involved various approaches and compromises that have complicated the relationships between 

design-basis events, safetyclass designations, and special treatment requirements, such as the 

applicability of 10 CFR 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants 

and Fuel Reprocessing Plants.” Such changes have resulted from risk insights produced 

by probabilistic risk assessments (PRAs) and operating experience. They also include 

the development of 10 CFR 50.62, “Requirements for Reduction of Risk from Anticipated 

Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants,” 

and 10 CFR 50.150, “Aircraft Impact Assessment.” The use of PRAs has also challenged the 

traditional designations and treatment of safety-related SSCs and greatly increased discussions 

regarding beyond-design basis events. 

Although not used in the context of NRC regulations, the terms “design extension events” or 

“design extension conditions” have been introduced by some international organizations and 

other national regulatory bodies (WENRA, 2008; IAEA, 2011). In the RMTF proposed risk 

management framework, this category of events has been called the “design enhancement 

category.” This category is used to provide a transition between design-basis events and 

beyond-design-basis events, provide a more logical approach for the regulatory treatment 

of some events, and alleviate concerns about uncertainties where risk insights are used as 

part of the decisionmaking process for regulatory requirements. A reason for introducing the 

concept here is that the terminology may be helpful in discussing events such as ATWS, station 

blackout (SBO), and aircraft impact, which have introduced regulatory requirements for nonsafety-related SSCs. In some sense, these are simply examples of the NRC determining that 

SSCs needed for certain events or conditions fall within the broader category of important to 

safety without needing to be classified as safety-related. The advantage of adopting a term 

like “design-enhancement events” is that it supports a logical construct of events and related 

non-safety-related SSCs that are important to safety instead of thinking of ATWS, SBO, aircraft 

impact, and possible future actions as individual outliers from the traditional approach of designbasis events and safety-related SSCs. 

A short discussion of the several events or requirements falling within the potential category of 

design-enhancement events is provided below.

Anticipated Transients Without Scram (10 CFR 50.62)

The requirements for addressing ATWS were published on June 26, 1984, in the Federal 

Register (49 FR 26036) and were issued as a result of concerns identified in PRAs and 

operating events. The Commission intended the ATWS rule requirements to provide further 

assurance that failure of the reactor to scram following an anticipated operational transient 

would not adversely affect public health and safety. In recognizing that the ATWS scenarios 

were expected to be infrequent occurrences, the regulation and related guidance acknowledged 

that the related SSCs need not be categorized as safety-related (see Generic Letter 85-06, 
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A Proposed Risk Management Regulatory Framework

“Quality Assurance Guidance for ATWS Equipment That Is Not Safety Related” (NRC, 1985)). 

Instead, the Commission required that the equipment be designed to perform its function in a 

reliable manner and encouraged reliability assurance programs for reactor protection systems to 

minimize the chances of an ATWS event. 

The requirements related to ATWS highlight the difficulty in explaining clearly NRC regulations. 

For example, the NRC’s acknowledgment that ATWS-related SSCs need not be safety-related 

would seem consistent with the scenario not being a “design-basis event,” and many (but 

not all) references to ATWS refer to it as a “beyond-design-basis accident.” However, certain 

SSCs (e.g., ATWS mitigation system actuation circuitry (AMSAC) or redundant reactivity control 

system (RRCS)) are required by the regulations to address ATWS events. Therefore, they 

do perform a design-basis function as that term is defined in Appendix B to Nuclear Energy 

Institute (NEI) 97-04, “Design Bases Program Guidelines” (NEI, 2000), which is endorsed in 

RG 1.186, “Guidance and Examples for Identifying 10 CFR 50.2 Design Basis”1 (NRC, 2000). 

This apparent contradiction (beyond-design-basis event, yet serving a design-basis function) 

is applicable to the SBO and aircraft impact examples and might be alleviated by adopting a 

concept similar to “design-enhancement category.”

Station Blackout (10 CFR 50.63)

In 1988, the Commission concluded that additional regulatory requirements were justified 

to address concerns about SBO events (NRC, 1988). The term “station blackout” refers to 

the complete loss of alternating-current (ac) electric power to the essential and nonessential 

switchgear buses in a nuclear power plant. SBO, therefore, involves the loss of offsite power 

concurrent with turbine trip and failure of the onsite emergency ac power system, but not the 

loss of available ac power to buses fed by station batteries through inverters or the loss of 

power from “alternate ac sources.” As with the ATWS rule, SBO requirements were established 

as a result of insights from risk assessments and some operational events. 

The rule requires plants to be able to withstand for a specified duration and recover from an 

SBO. The SBO duration for each plant is determined based on:

•	 the redundancy of the onsite emergency ac power sources

•	 the reliability of the onsite emergency ac power sources

•	 the expected frequency of loss of offsite power

•	 the probable time needed to restore offsite power

1 The definition of design basis in 10 CFR 50.2 is:

Design bases means that information which identifies the specific functions to be performed by a structure, 

system, or component of a facility, and the specific values or ranges of values chosen for controlling 

parameters as reference bounds for design. These values may be (1) restraints derived from generally 

accepted “state of the art” practices for achieving functional goals, or (2) requirements derived from analysis 

(based on calculation and/or experiments) of the effects of a postulated accident for which a structure, 

system, or component must meet its functional goals.specific functions to be performed by a structure, 

system, or component of a facility, and the specific values or ranges of values chosen for controlling 

parameters as reference bounds for design. These values may be (1) restraints derived from generally 

accepted ‘state of the art’ practices for achieving functional goals, or (2) requirements derived from analysis 

(based on calculation and/or experiments) of the effects of a postulated accident for which a structure, 

system, or component must meet its functional goals.”
A Proposed Risk Management Regulatory Framework

G-6 | Risk Management Task Force Appendix G

RG 1.155, “Station Blackout” (NRC, 1988a) provides quality assurance guidance for non-safetyrelated systems and equipment used to meet the requirements of 10 CFR 50.63, “Loss of 

Alternating Current Power.”

Aircraft Impact

As a result of the terrorist events of September 11, 2001, the NRC developed and issued a 

new regulation, 10 CFR 50.150, “Aircraft Impact Assessment” (NRC, 2009), which requires 

applicants for new nuclear power plants to perform a designspecific assessment of the effects 

on the facility of the impact of a large, commercial aircraft. The applicant is required to use 

realistic analyses to identify and incorporate design features and functional capabilities to 

show, with reduced use of operator actions, that either the reactor core remains cooled or the 

containment remains intact and that either spent fuel cooling or spent fuel pool integrity 

is maintained. 

As discussed in the supplementary information for the aircraft impact assessment rulemaking, 

the Commission determined that the intentional crash of a large, commercial aircraft into a 

nuclear power plant is a beyond-design-basis event, and the NRC’s requirements that apply to 

the design, construction, testing, operation, and maintenance of design features and functional 

capabilities for design-basis events will not apply to design features or functional capabilities 

selected by the applicant solely to meet the requirements of the aircraft impact rule. 

Consideration of a rule to require applicants for new nuclear power reactors to perform an 

aircraft impact assessment and describe design features and functional capabilities addressing 

such impacts, which are beyond-design-basis scenarios, is similar to the Commission’s 

consideration in the mid-1980s of new rules addressing accidents more severe than designbasis accidents. The Policy Statement on Severe Reactor Accidents (NRC, 1985a) explained 

the Commission’s conclusion that, although it was proposing criteria to show new reactor 

designs to be acceptable for severe accident concerns, thenexisting plants posed no undue risk 

to public health and safety, and thus, there was no need for action on operating reactors based 

on severe accident risks.

While clearly stating that the aircraft impacts are considered beyond-design-basis events, 

the Commission also defined requirements to evaluate and report changes to the design 

features and functional capabilities credited with meeting the acceptance criteria in the rule. 

Such controls were defined for the aircraft impact rulemaking based on similar requirements 

for severe accident features that are described in the design description documents for new 

reactors. The controls added for aircraft impact and severe accident features (e.g., change 

controls and reporting requirements) are an improvement over those defined for the ATWS and 

SBO rules and provide an example of how the NRC might address updating of information for 

items considered “important to safety” but not safety-related. 

G.3 Regulatory Treatment of Non-safety Systems

In the early 1990s, the NRC developed an approach to address the proposed increased 

use of passive safety features in advanced reactor designs. Unlike operating reactors, the 

passive advanced light-water reactor designs (ALWR), such as the AP600 design, proposed 

extensive use of safety systems that rely on the driving forces of buoyancy, gravity, and stored 

energy sources. In addition to the active systems used during normal plant operations, the 

passive ALWR designs proposed non-safety-grade active systems to provide defense-in-depth 
Appendix G Risk Management Task Force | G-7

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capabilities for reactor coolant makeup and decay heat removal. These systems would be the 

first line of defense to reduce challenges to the passive systems in the event of transients or 

plant upsets. The licensing-related analyses proposed by the industry for the passive designs 

rely solely on the passive safety systems to demonstrate compliance with the acceptance 

criteria of various design-basis transients and accidents. To incorporate the defense-in-depth 

measures into the licensing process, while recognizing the passive safety features’ role in 

responding to design-basis events, the staff and industry developed a process for the regulatory 

treatment of non-safety systems (RTNSS) (NRC, 1995). 

To further illustrate the concept, the criteria used to identify SSCs within the RTNSS regime for 

the economic simplified boilingwater reactor (ESBWR) design are as follows:

•	 SSC functions relied upon to meet deterministic NRC performance requirements, such 

as 10 CFR 50.62 (ATWS) and 10 CFR 50.63 (SBO),

•	 SSC functions relied upon to ensure long-term safety (beyond 72 hours) and to address 

seismic events,

•	 SSC functions relied upon under power operating and shutdown conditions to meet the 

Commission’s safety goal guidelines of a CDF of less than 1×10-4 per reactor year and a 

large release frequency of less than 1×10-6 per reactor year;

•	 SSC functions needed to meet the containment performance goal, including containment 

bypass, during severe accidents, and

•	 SSC functions relied upon to prevent significant adverse systems interactions.

This can be viewed as a more formal system to identify and address SSCs determined to be 

important to safety but not safety-related, including the design-enhancement events such as 

ATWS and SBO. The remaining criteria address the characteristics of passive plants (e.g., use 

of non-safety-related systems to replenish passive systems or perform cooling functions after 

72 hours), and Risk-Informed insights (for both CDF and containment performance goals). A 

key feature of the RTNSS regime is the determination of appropriate regulatory oversight for 

the RTNSS SSCs based on evaluations of risk significance. The controls are defined in the 

licensing documents such as the reliability assurance program. 

G.4 Risk-Informed Categorization (10 CFR 50.69)

The NRC staff prepared SECY-98-300, “Options for Risk-Informed Revisions to 

10 CFR Part 50 – “Domestic Licensing of Production and Utilization Facilities” (NRC, 1998) as 

part of the continuing efforts to increase the use of Risk-Informed approaches in the regulation 

of nuclear power plants. The staff proposed to change the regulatory scope of SSCs needing 

special treatment in such areas as quality assurance, environmental qualification, technical 

specifications, and the ASME Code. The paper described how this could be accomplished, 

in part, by developing Risk-Informed definitions (i.e., classifications) for safety-related and 

safety-important SSCs and “grading” special treatment requirements for SSCs based upon 

their risk importance. This effort included many interactions with the Commission and various 

stakeholders and ultimately led to the issuance of a new regulation in 10 CFR 50.69, “RiskInformed Categorization and Treatment of Structures, Systems and Components for Nuclear 

Power Reactors.” The rule defines four classes of SSCs based on the traditional categories 
A Proposed Risk Management Regulatory Framework

G-8 | Risk Management Task Force Appendix G

of safety-related and non-safety-related and consideration of risk insights, including severe or 

beyond-design-basis accidents, to define SSCs as being either safety-significant or low-safetysignificant. The representation provided below for the four Risk-Informed safety class (RISC) 

categories in 10 CFR 50.69 is taken from RG 1.201, “Guidelines for Categorizing Structures, 

Systems, and Components in Nuclear Power Plants According to Their Safety Significance” 

(NRC, 2006).

Regulatory Guide 1.201 and the related industry guidance in NEI 00-04, “10 CFR 50.69 SSC 

Categorization Guideline” (NEI, 2005), provides detailed guidance on the considerations and 

process for determining the classification of SSCs into the RISC 1-4 categories. The process 

includes risk characterizations, defense-in-depth characterizations, risk sensitivity studies, and 

reviews by an integrated decisionmaking panel. Much of the focus of the guidance documents 

related to 10 CFR 50.69 is on the categorization of SSCs. Less discussion is provided for 

the actual treatment programs being implemented under 10 CFR 50.69, which has raised 

implementation questions, especially for equipment in RISC categories 2 and 3. At this time, 

no licensee has submitted an application requesting to implement 10 CFR 50.69, although at 

least one licensee is preparing a pilot application. Following the initial pilot application, lessons 

learned from the application review will be used to revise the associated industry guidance and 

RG 1.201.

G.5 SSC Classification for Advanced Reactors (non-LWRs)

The need to revise various regulatory requirements for non-LWR reactor technologies has 

resulted in some more dramatic proposals for changing the selection of design-basis events and 

the classification of SSCs. Two proposals are described below:

NUREG-1860

The staff issued NUREG-1860, “Feasibility Study for a Risk-Informed and PerformanceBased Regulatory Structure for Future Plant Licensing” (NRC, 2007b), to document a possible 

Figure G-1 10 CFR 50.69 Risk Informed Safety 

 Class (RISC) Categories 

 Figure G-1 10 CFR 50.69 Risk-Informed Safety 

Class (RISC) Categories
Appendix G Risk Management Task Force | G-9

A Proposed Risk Management Regulatory Framework

framework that (1) could be used to develop an alternative set of technical requirements to 

10 CFR Part 50 applicable for future non-LWR nuclear power plants (the framework includes a 

proposed draft set of technical requirements), and (2) could be used to establish Risk-Informed 

licensing-basis events and the safety classification of structures, systems, and components. 

The approaches discussed in NUREG-1860 are sometimes referred to as the technology 

neutral framework.

NUREG-1860 proposes a safety classification scheme in which all the plant SSCs fall into 

two categories—safety-significant or nonsafety-significant—distinguished by whether the 

SSCs need special treatment or not. Consistent with the framework’s discussion of selecting 

licensing-basis events (LBEs), the term “safety-significant” is assigned to those SSCs whose 

functionality plays a role in meeting the acceptance criteria imposed on the LBEs. The term 

“special treatment” is used to designate requirements imposed on SSCs that go beyond 

industry-established requirements for equipment classified as “commercial grade.” These 

requirements provide additional confidence that the equipment is capable of meeting its 

functional requirements under PRAanalyzed conditions. 

The type of special treatment varies depending on the function and importance of the SSC. The 

treatment helps to ensure that the SSCs will perform reliably (as postulated in the PRA) under 

the conditions (temperature, pressure, radiation,) assumed to prevail in the event sequences 

for which the SSC’s successful function is credited in the risk analysis. A basic special 

treatment requirement for all safety-significant SSCs will be the establishment and monitoring 

of reliability and availability goals. All safety-significant SSCs will have reliability and availability 

consistent with the values assumed in the PRA. During operation, a process similar to the 

monitoring of the performance and condition of structures, systems, or components, against 

licensee-established goals of 10 CFR 50.65, “Requirements for Monitoring the Effectiveness 

of Maintenance at Nuclear Power Plants” (the maintenance rule), is expected to be an integral 

part of the monitoring program for this special treatment requirement. Monitoring will consist 

of periodically gathering, trending, and evaluating information pertinent to the performance 

and availability of PRA-related SSCs and comparing the result with the established goals and 

performance criteria to verify that the goals are being met. 

Next Generation Nuclear Plant Program

The NRC staff is assessing various white papers submitted as part of the preapplication 

interactions with the U.S. Department of Energy (DOE) related to the Next Generation Nuclear 

Power (NGNP) program. The NGNP program is adopting a licensing approach similar to earlier 

proposals related to high-temperature gas reactors, such as the modular high-temperature gascooled reactor and the pebble bed modular reactor. The same process for selecting licensingbasis events and determining equipment classifications has also been incorporated into industry 

and international standards (e.g., ANS 53.1, “Nuclear Safety Design Standard for Modular 

Helium-Cooled Reactor Plants” (ANS, 2011)). 

The NGNP safety classification approach (NGNP, 2010) results in the classification of SSCs as 

safety-related, non-safety-related with special treatment, or non-safety-related. Based upon 

the selected licensing-basis events and considering defense-in-depth attributes, the plant’s 

SSCs are evaluated for the safety significant role they may play in preventing or mitigating 

the radiological consequence of such events to classify which SSCs are safety related. With 

its event selections derived, in part, from frequency-and-consequence curves in a process 
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G-10 | Risk Management Task Force Appendix G

similar to NUREG-1860, the NGNP approach is a mix of the classification approaches from the 

technology neutral framework and 10 CFR 50.69. The classes are described as follows:

•	 Safety-Related:

o SSCs relied on to perform required safety functions to prevent or mitigate 

the consequences of design-basis events (DBE) to comply with the top level 

regulatory criteria (TLRC)

o SSCs relied on to perform required safety functions to prevent the frequency 

of beyond-design-basis events (BDBE) with consequences greater than the 

10 CFR 50.34, “Contents of Applications; Technical Information,” dose limits from 

increasing into the DBE region of the frequency-consequence curve

•	 Non-Safety-Related with Special Treatment:

o SSCs relied on to perform functions to mitigate the consequences of anticipated 

operational occurrences (AOO) to comply with the TLRC

o SSCs relied on to perform functions to prevent the frequency of DBEs with 

consequences greater than the 10 CFR 20, “Standards for Protection Against 

Radiation,” offsite dose limits from increasing into the AOO region of the 

frequency-and-consequence curve

•	 Non-Safety-Related:

o All other SSCs (with no special treatment required)

The NRC staff is currently interacting with DOE on the NGNP white paper on SSC classification, 

and so the process has not yet been finalized or endorsed by the NRC.

G.6 International Approaches

International Atomic Energy Agency

Several International Atomic Energy Agency (IAEA) documents discuss the safety classification 

of SSCs, including Draft Safety Guide DS-367, “Safety Classification of Structures, Systems, 

and Components in Nuclear Power Plants” (IAEA, 2011). This draft safety guide updates 

previous discussions on this topic in documents such as the “IAEA Safety Glossary” 

(IAEA, 2007), which describes plant equipment in terms of “items important to safety,” “safety 

systems,” and “safety-related items.”2 Draft Safety Guide DS-367 describes a more structured 

evaluation and classification of SSCs, which starts from the higher level requirement defined in 

Draft Safety Requirements No. SSR 2/1, “Safety of Nuclear Power Plants: Design” (IAEA, 2010).

2 In the IAEA Safety Glossary, a “safety system” is a system important to safety, provided to ensure the safe 

shutdown of the reactor or the residual heat removal from the core, or provided to limit the consequences of 

anticipated operational occurrences and design-basis accidents. A safety-related item is important to safety 

but is not part of a safety system. In terms of NRC terminology, the IAEA safety system would be safetyrelated and the categories of RTNSS or RISC-2 would correspond to safety-related items as discussed in 

the IAEA Safety Glossary.
Appendix G Risk Management Task Force | G-11

A Proposed Risk Management Regulatory Framework

The method for classifying the safety significance of items important to safety shall be based 

primarily on deterministic methodologies complemented where appropriate by probabilistic 

methods, with account taken of factors such as the following:

•	 the safety function(s) to be performed by the item

•	 the consequences of failure to perform the safety function

•	 the frequency at which the item will be called upon to perform a safety function

•	 the time following a postulated initiating event at which, or the period for which, it will be 

called upon to operate.

The draft safety classification guide categorizes safety functions, which can be either preventive 

or mitigative, based on the severity of the consequences of failure of that specific function. The 

severity is divided into three levels: high, medium, and low, as follows:

o The severity should be considered “high” if a failure of the safety function could:

−	 lead directly to a release of radioactive material that exceeds the limits for 

design-basis accidents accepted by the regulatory body, or

−	 cause the values of key physical parameters to challenge or exceed 

design limits for design-basis accidents.

o The severity should be considered “medium” if a failure of the safety function 

could, at worst:

−	 lead to a release of radioactive material below the limits for design basis 

accidents accepted by the regulatory body, or 

−	 cause the values of key physical parameters to exceed the design limits 

for anticipated operational occurrences, but remain within the specified 

design limits for design-basis accidents

o The severity should be considered “low” if a failure of the safety function could, 

at worst:

−	 lead to a release of radioactive material below the limits for the plant 

conditions for anticipated operational occurrences, or 

−	 cause the values of key physical parameters to exceed the specified 

design limits for normal operation, but remain within the specified design 

limits for anticipated operational occurrences

The IAEA guide calls for the safety significance of all plantspecific safety functions to be 

established and categorized according to the risk. The approach recommended in the safety 

guide has three categories (which align with the approaches adopted in many Member States 

for existing nuclear power plants), supplemented by a fourth category (reflecting the approach 

followed for some newer reactor designs) to cater to additional features in the design to control 
A Proposed Risk Management Regulatory Framework

G-12 | Risk Management Task Force Appendix G

consequences in design extension conditions.3 It should be noted that the safety classification 

guide discusses design extension events or conditions as follows:

Safety functions for the mitigation of “design extension conditions” are intended 

to limit accident progression (e.g. in-vessel mitigation before significant core 

degradation occurs) and to mitigate the consequences of a severe accident 

(e.g., ex-vessel mitigation to control the remains of a significantly degraded core).

Design extension conditions are thereby equated to the fourth level of defense-in-depth, as 

described in INSAG-10, “Defense-in-depth in Nuclear Safety” (IAEA, 1996), which is the control 

of severe conditions, including the prevention of accident progression and mitigation of the 

consequences of a severe accident.

The four safety categories, in relation to the safety functions, are summarized in Table G2.

•	 Level A mitigative safety functions for design-basis accidents should establish a 

controlled state following a design-basis accident. A controlled state should be reached 

as soon as possible.

•	 Level B mitigative safety functions for design-basis accidents should:

a) after a controlled state is reached, achieve and maintain a safe shutdown state 

following a design-basis accident, and

b) minimize the challenge to the remaining barriers from the design-basis accident.

Table G-2 IAEA Safety Categories

Type of safety function

Severity of the consequences of the failure of plantspecific 

safety functions

High Medium Low

Preventive safety functions Safety Category 1 Safety Category 2 Safety Category 3

Safety functions for mitigation 

of anticipated operational 

occurrences

Safety Category 1 Safety Category 2 Safety Category 3

Safety functions for control 

or mitigation of design-basis 

accidents (level A)

Safety Category 1 Safety Category 2 Safety Category 3

Safety functions for control 

or mitigation of design-basis 

accidents (level B)

Safety Category 2 Safety Category 3 Safety Category 3

Safety functions for mitigation 

of consequences in design 

extension conditions

Safety Category 4 Safety Category 4 No Safety 

Category

3 Design extension events or conditions are generally the same as the design enhancement category 

proposed by the RMTF.
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A Proposed Risk Management Regulatory Framework

An evaluation is performed for all plant SSCs, and the SSCs are placed in safety functional 

groups based on the functions they provide (e.g., control of reactivity, removal of heat from 

the core and spent fuel, and confinement of radioactive material). The assignment of SSCs 

to safety categories also considers factors such as the availability of other SSCs to fulfill the 

specific-safety function.

Annex II to the draft IAEA safety classification guide provides examples of engineering design 

rules (i.e., special treatment requirements) for the various safety categories 

Western European Nuclear Regulators’ Association

Western European Nuclear Regulators’ Association (WENRA) has developed “reference levels,” 

with the objective to attain a common approach to nuclear safety within Europe. Although 

not as detailed as the IAEA draft safety classification guidance, the WENRA Reactor 

Harmonization Working Group (WENRA, 2008) has defined the following reference level related 

to safety classification:

1. Objective 

1.1 All SSCs important to safety shall be identified and classified on 

the basis of their importance for safety. 

2. Classification process 

1.2 The classification of SSCs shall be primarily based on 

deterministic methods, complemented where appropriate by 

probabilistic methods and engineering judgment.

1.3 The classification shall identify for each safety class: 

− The appropriate codes and standards in design, 

manufacturing, construction and inspection; 

− Need for emergency power supply, qualification to 

environmental conditions; 

− The availability or unavailability status of systems serving 

the safety functions to be considered in deterministic 

safety analysis; 

− The applicable quality requirements. 

3. Ensuring reliability 

3.1 SSCs important to safety shall be designed, constructed and 

maintained such that their quality and reliability is commensurate 

with their classification. 
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G-14 | Risk Management Task Force Appendix G

3.2 The failure of a SSC in one safety class shall not cause the 

failure of other SSCs in a higher safety class. Auxiliary systems 

supporting equipment important to safety shall be 

classified accordingly. 

4. Selection of materials and qualification of equipment 

4.1 The design of SSCs important to safety and the materials used 

shall consider the effects of operational conditions over the 

plant lifetime and the effects of design basis accidents on their 

characteristics and performance. 

4.2 A qualification procedure shall be adopted to confirm that SSCs 

important to safety meet throughout their design operational lives 

the demands for performing their function, taking into account 

environmental conditions over the lifetime of the plant and when 

required in anticipated operational occurrences and 

accident conditions. 

Similar to the discussions of design extension events in the IAEA guidance, WENRA has 

summarized design extension events as follows:

•	 Design Extension Conditions (DECs): The plant can be brought into a controlled 

state and the integrity of the containment is maintained (the containment shall cope with 

a core melt situation). Significant releases are practically eliminated. DECs may be 

analyzed using the best-estimate approach:

o For those conditions that are not practically eliminated, design provisions shall 

be made such that only protective measures that are of limited scope in terms of 

area and time are necessary for the protection of the public, and sufficient time is 

available to implement these measures 

o The possibility of certain conditions occurring is considered to have been 

practically eliminated if it is physically impossible for the conditions to occur 

or if the conditions can be considered with a high degree of confidence to be 

extremely unlikely to arise.

A more detailed discussion in the report of the WENRA Reactor Harmonization Working Group 

is provided in the following reference level for “Design Extension of Existing Reactors:”

1. Objective 

1.1 The design extension analysis shall examine the performance 

of the plant in specified accidents beyond the design basis, 

including selected severe accidents, in order to minimize as far as 

reasonably practicable radioactive releases harmful to the public 

and the environment in cases of events with very low probability 

of occurrence. 
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A Proposed Risk Management Regulatory Framework

2. Selection and analysis of Beyond Design Basis Events 

2.1 Beyond design basis events shall be selected and considered 

in the safety analysis to determine those sequences for which 

reasonable practicable preventive or mitigative measures can be 

identified and implemented. 

2.2 Realistic assumptions and modified acceptance criteria may be 

used for the analysis of the beyond design basis events. 

3. Instrumentation for the management of beyond design basis 

accident conditions 

3.1 Adequate instrumentation shall exist which can be used in severe 

accident environmental conditions in order to manage such 

accidents according to guidelines/procedures for 

severe accidents. 

3.2 Necessary information from instruments shall be relayed to the 

control room as well as to a separately located supplementary 

control room/post and be presented in such a way to enable a 

timely assessment of the plant status and critical safety functions 

in severe accident conditions. 

Where:

● Design extension is understood as measures taken to cope with 

additional events or combination of events, not foreseen in the 

design of the plant. Such measures need not involve application of 

conservative engineering practices but could be based on realistic, 

probabilistic or best estimate assumptions, methods and 

analytical criteria. 

● These reference levels aim at providing protection at the level 

4 of the defense-in-depth. Such protection could be provided 

by existing equipment that has been assessed, and if needed 

modified, to perform the relevant function in a severe accident 

condition or additional equipment on a best estimate basis

● Special attention needs to be given for certain reactor types to the 

analysis of severe accident conditions with an open containment 

during certain shutdown states. Should such an accident occur, 

it should be possible to achieve timely containment isolation or 

implement equally effective compensatory measures. Therefore 

consideration has to be given to the time needed for the 

restoration of containment isolation and effective leaktightness, 

taking into account factors such as the progression of the 

accident sequences. 
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G-16 | Risk Management Task Force Appendix G

4. Protection of the containment against selected beyond design 

basis accidents

4.1 Isolation of the containment shall be possible in a beyond design 

basis accident. However, if an event leads to bypass of the 

containment, consequences shall be mitigated. 

4.2 The leaktightness of the containment shall not degrade 

significantly for a reasonable time after a severe accident. 

4.3 Pressure and temperature in the containment shall be managed in 

a severe accident. 

4.4 Combustible gases shall be managed in a severe accident. 

4.5 The containment shall be protected from overpressure in a 

severe accident. 

4.6 High pressure core melt scenarios shall be prevented. 

4.7 Containment degradation by molten fuel shall be prevented or 

mitigated as far as reasonably practicable. 

The WENRA description states that the types of events to be analyzed for design extension (if 

not already analyzed as a design-basis accident) would include: 

•	 ATWS 

•	 SBO

•	 total loss of feedwater 

•	 loss-of-coolant accident, together with the complete loss of one emergency core cooling 

system; uncontrolled level drop during midloop operation (PWR) or during refueling 

•	 total loss of the component cooling water system 

•	 loss of core cooling in the residual heat removal mode 

•	 loss of fuel pool cooling 

•	 loss of ultimate heat sink function 

•	 uncontrolled boron dilution (PWR) 

•	 multiple steam generator tube ruptures (PWR) 

•	 loss of required safety systems in the long term after a postulated initiating event 
Appendix G Risk Management Task Force | G-17

A Proposed Risk Management Regulatory Framework

G.7 Discussion

Additional information related to the development of the current NRC structure was provided 

to the Commission in SECY-11-0156, “Feasibility of Including Risk Information in Categorizing 

Structures, Systems, and Components as Safety-related or Non-Safety-related” (NRC, 2011). 

A simplified representation of the various categorization schemes is shown in Figure G-2.

 

The classification of SSCs is also closely related to the introduction of a design-enhancement 

category and the overall approach to risk-informed and performance-based defense-in-depth, 

which is discussed in Chapter 4 and Appendix H. The previous discussions in this appendix of 

various ways to classify SSCs according to their safety role and significance identify several key 

points that should be considered in developing such schemes for currently operating plants or 

future plants. These key points include the following:

•	 Clearly define the reason for the classification system. In general, the safety 

classification systems are used to differentiate special treatment requirements for SSCs 

in areas such as quality assurance, environmental qualification, and applicability of 

various industry codes and standards. Special treatment implies requirements imposed 

on SSCs that go beyond industry-established requirements for equipment classified 

as “commercial grade.” More recent proposed evaluations of safety classification 

systems (e.g., NUREG-1860 and IAEA DS-367) include specific discussions on safety 

functions; safety categories; and the expectations for equipment in terms of capability, 

dependability, robustness, and reliability. It is important to think of these matters early 

Figure G-2 Representations of Safety

Classification Concepts
A Proposed Risk Management Regulatory Framework

G-18 | Risk Management Task Force Appendix G

since the safety classification of SSCs is merely a tool to achieve the desired outcomes. 

The task of changing the current system should not be underestimated since the 

existing infrastructure (e.g., NRC regulations, licensee procedures, industry codes and 

standards) has been developed to define special treatment requirements on the basis 

of the definition of safety-related SSCs and the premise that the designation “safetyrelated” helps ensure the desired outcomes.

•	 The definition of safety-related SSCs includes that equipment “relied upon to remain 

functional during and following design-basis events” to address the critical safety 

functions of (1) maintaining the reactor coolant pressure boundary, (2) shutting down 

the reactor, and (3) preventing or mitigating accidents that could result in significant 

offsite radiation exposures. These are very similar to the three fundamental safety 

functions defined by IAEA, which are to (1) control reactivity, (2) remove heat from the 

reactor core, and (3) confine radioactive material. Much of the current discussions 

revolve around how to address distinctions between “design-basis events” and “beyonddesign-basis events.” This sometimes translates into discussions of how to consider 

PRAs since “design-basis events” are often thought of as those addressed by traditional 

deterministic analyses and “beyond-design-basis events” are associated with PRAs. As 

discussed elsewhere in this report, the concept of “design-enhancement conditions” has 

been introduced to address some concerns regarding the transition between design and 

beyond-design-basis events. Various activities (e.g., NUREG-1860, NGNP, IAEA 

DS-367) have highlighted the relationships between licensing-basis event selection and 

SSC safety classification and the need to consider and resolve these topics as part of a 

more holistic regulatory approach.

•	 The ability to develop and implement alternatives is complicated by having plants in 

various stages of their life cycle (e.g., operation, construction, design). Operating plants 

have available an option to adopt the classification scheme defined in 10 CFR 50.69. 

However, the degree to which the traditional system has been ingrained into codes and 

standards, procurement processes, and other areas of plant design and operation is an 

obstacle to the implementation of 10 CFR 50.69. New plant designs undergoing review 

have either adopted the existing system or introduced the RTNSS concept to address 

passive safety features. The resolution of some implementation issues with RTNSS 

during the reviews of the AP1000 and ESBWR designs should make it easier for small 

modular light-water reactors to adopt the concept for this class of plants (see SECY11-0024, “Use of Risk Insights to Enhance the Safety Focus of Small Modular Reactor 

Reviews” (NRC, 2011a)). Generation IV reactor designs generally involve significant 

differences from the currently operating fleet in terms of reactor coolant, fuel design, 

and safety features. Many of these designs have also been developed considering 

international safety standards and could therefore more easily adopt approaches 

discussed in NUREG-1860, the NGNP white papers, or IAEA DS-367. 

•	 It may be an opportune time to evaluate possible changes to the safety classification 

approach in concert with the development of a design and licensing strategy for small 

modular light-water reactors and Generation IV reactor designs. Although NRC staff 

activities are underway for assessing a longer-term Risk-Informed approach (as 

described in SECY-11-0024), the various industry players are acting independently and 

considering slightly different licensing approaches within the general framework that 

has been used for other new reactor designs. The small modular light-water reactors 

appear to prefer the stability, predictability, and existing infrastructure of the current 
Appendix G Risk Management Task Force | G-19

A Proposed Risk Management Regulatory Framework

processes developed for new passive plant designs (i.e., safety-related and RTNSS 

classifications). It may be possible to use the flexibility within the existing definitions 

(e.g., design-basis events and design-basis accidents) to improve the consideration 

of risk insights and construct a more logical set of licensing-basis events and related 

equipment special treatment requirements within the current classification construct 

(or a variation supported by a limited number of exemptions). The preferred approach 

of the gas-cooled reactor community is fairly well established and is reflected in the 

NGNP white papers and ANS 53.1. The sodium-cooled fast reactor community is 

currently developing a standard (ANS 54.1) that will propose a licensing and safety 

classification scheme for its reactor designs. International efforts are likewise underway, 

and while they have some commonalities with U.S. proposals, the IAEA and WENRA 

approaches are likely to use designations and criteria that are different from the various 

U.S. proposals. 

G.8 References

(ANS, 1983) American National Standards Institute/American Nuclear Society, ANSI/

ANS 51.1 – 1983, “Nuclear Safety Criteria for the Design of Stationary 

Pressurized Water Reactors,” ANS, LaGrange Park, IL.

(ANS, 2011) American National Standards Institute/American Nuclear Society, ANSI/

ANS53.1 – 2011, “Nuclear Safety Design Process for Modular HeliumCooled Reactor Plants,” ANS, LaGrange Park, IL.

(IAEA, 1996) International Atomic Energy Agency, “Defence in Depth in Nuclear 

Safety,” INSAG-10, A Report by the International Nuclear Safety Advisory 

Group, 1996.

(IAEA, 2007) International Atomic Energy Agency, “IAEA Safety Glossary, Terminology 

Used In Nuclear Safety and Radiation Protection,” 2007 Edition.

(IAEA, 2010) International Atomic Energy Agency, “Draft – Safety of Nuclear Power 

Plants: Design,” Draft Safety Guide DS-414, September 2010.

(IAEA, 2011) International Atomic Energy Agency, “Draft – Safety Classification of 

Structures, Systems, and Components in Nuclear Power Plants,” Draft 

Safety Guide DS-367, April 2011.

(NEI, 2005) Nuclear Energy Institute, “10 CFR 50.69 SSC Categorization Guideline,” 

NEI 0004, July 2005, Agencywide Documents Access and Management 

System (ADAMS) Accession No. ML052910035.

(NEI, 2000) Nuclear Energy Institute, “Guidance and Examples for Identifying 

10 CFR 50.2 Design Bases,” NEI 97-04, Revised Appendix B, 

November 2000, ADAMS Accession No. ML003771698.

(NGNP, 2010) Idaho National Laboratory, “Next Generation Nuclear Plant Structures, 

Systems, and Components Safety Classification White Paper,” 

INL/EXT-10-19509, September 2010.
A Proposed Risk Management Regulatory Framework

G-20 | Risk Management Task Force Appendix G

(NRC, 1981) U.S. Nuclear Regulatory Commission, “Standard Definitions for 

Commonly-used Safety Classification Terms,” memorandum dated 

November 20, 1981, ADAMS Accession No. ML111230450.

(NRC, 1984) U.S. Nuclear Regulatory Commission, “NRC Use of the Terms ‘Important 

to Safety’ and ‘Safety Related,’” Generic Letter 84-01, January 5, 1984, 

ADAMS Accession No. ML031150515.

(NRC, 1985) U.S. Nuclear Regulatory Commission, “Quality Assurance Guidance for 

ATWS Equipment That Is Not Safety Related,” Generic Letter 85-06, 

April 16, 1985, ADAMS Accession No. ML031140390.

(NRC, 1985a) U.S. Nuclear Regulatory Commission, “Policy Statement on Severe 

Reactor Accidents Regarding Future Designs and Existing Plants,” 

August 1985, ADAMS Accession No. ML003711521.

(NRC, 1988) U.S. Nuclear Regulatory Commission, “Loss of All Alternating Current 

Power,” Final Rulemaking, published in the Federal Register on June 21, 

1988 (53 FR 23215). 

(NRC, 1988a) U.S. Nuclear Regulatory Commission, “Station Blackout,” Regulatory 

Guide 1.155, August 1988, ADAMS Accession No. ML003740034.

(NRC, 1995) U.S. Nuclear Regulatory Commission, “Policy and Technical Issues 

Associated with the Regulatory Treatment of Non-Safety Systems 

(RTNSS) in Passive Plant Designs (SECY 94-084),” SECY-95-132, May 

22, 1995, ADAMS Accession No. ML003708005.

(NRC, 1998) U.S. Nuclear Regulatory Commission, “Options for Risk-Informed 

Revisions to 10 CFR Part 50 – “Domestic Licensing of Production 

and Utilization Facilities,” SECY-98-300, December 23, 1998, ADAMS 

Accession No. ML992870048. 

(NRC, 2000) U.S. Nuclear Regulatory Commission, “Guidance and Examples for 

Identifying 10 CFR 50.2 Design Bases,” Regulatory Guide 1.186, 

December 2000, ADAMS Accession No. ML003754825.

(NRC, 2006) U.S. Nuclear Regulatory Commission, “Guidelines for Categorizing 

Structures, Systems, and Components in Nuclear Power Plants According 

to Their Safety Significance,” Regulatory Guide 1.201, May 2006, ADAMS 

Accession No. ML061090627.

(NRC, 2007) U.S. Nuclear Regulatory Commission, “Seismic Design Classification,” 

Regulatory Guide 1.29, Revision 4, March 2007, ADAMS Accession 

No. ML070310052.

(NRC, 2007a) U.S. Nuclear Regulatory Commission, “Quality Group Classifications 

and Standards for Water-, Steam- and Radioactive-Waste-Containing 

Components of Nuclear Power Plants,” Regulatory Guide 1.26, 

Revision 4, March 2007, ADAMS Accession No. ML070290283.
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(NRC, 2007b) U.S. Nuclear Regulatory Commission, “Feasibility Study for a RiskInformed and Performance-Based Regulatory Structure for Future 

Plant Licensing,” NUREG-1860, December 2007, ADAMS Accession 

No. ML080440170.

(NRC, 2009) U.S. Nuclear Regulatory Commission, “Consideration of Aircraft Impacts 

for New Nuclear Power Reactors,” Final Rule, published in the Federal 

Register on June 12, 2009 (74 FR 28111).

(NRC, 2011) U.S. Nuclear Regulatory Commission, “Feasibility of Including Risk 

Information in Categorizing Structures, Systems, and Components 

as Safety-Related or Non-Safety-Related,” SECY-11-0156, 

November 2, 2011, ADAMS Accession No. ML112690353.

(NRC, 2011a) U.S. Nuclear Regulatory Commission, “Use of Risk Insights to Enhance 

the Safety Focus of Small Modular Reactor Reviews,” SECY-11-0024, 

February 18, 2011, ADAMS Accession No. ML110110691.

(WENRA, 2008) Western European Nuclear Regulators’ Association, “WENRA Reactor 

Safety Reference Levels,” Reactor Harmonization Working Group, 

January 2008.

Appendix H Risk Management Task Force | H-1

A Proposed Risk Management Regulatory Framework

APPENDIX H

NUCLEAR POWER REACTORS – CHAPTER 4 ALTERNATIVES

H.1 Introduction

The main report identifies three options for implementing a risk management framework 

and adopting a risk-informed and performance-based defense-in-depth approach for the 

U.S. Nuclear Regulatory Commission’s (NRC’s) various regulatory programs. Option B involves 

adopting the proposed Risk Management Regulatory Framework through Commission policy 

statements, rule changes, and revisions to guidance documents. There are many alternatives 

that could be pursued under Option B, ranging from selected adjustments of guidance 

documents to the development of new requirements and regulatory programs (i.e., graduated 

changes between guidance changes and developing a significantly different regulatory 

framework). Several possible alternatives for operating and new power reactors are provided in 

this appendix as a means to support discussions and show the range of choices available under 

Option B. 

A question that inevitably arises when considering a regulatory framework and categorizing 

events or equipment is whether requirements are needed for adequate protection of the health 

and safety of the public. The NRC establishes regulatory requirements for protecting public 

health and safety and common defense and security using a two-tier structure. This two-tier 

structure is consistent with the statutory requirements in the Atomic Energy Act of 1954, as 

amended. (See Union of Concerned Scientists v. NRC, 824 F.2d 108, 120 (D.C. Cir. 1987)). 

The top tier consists of those requirements needed to ensure adequate protection of public 

health and safety and to be in accord with the common defense and security. The adequate 

protection standard is also described in terms of ensuring that licensed activities pose no undue 

risk to public health and safety or to common defense and security. In the above cited case, the 

Court of Appeals for the D.C. Circuit recognized that adequate protection (no undue risk) does 

not equate to “zero risk.” The Court stated:

Similarly, under the adequate-protection standard of section 182(a), the NRC 

need ensure only an acceptable or adequate level of protection to public health 

and safety; the NRC need not demand that nuclear power plants present no risk 

of harm.

Measures taken to prevent, contain, and mitigate events or concerns in this tier historically 

correspond to the traditional design-basis accidents described in Appendix F and several 

programmatic requirements for dealing with beyond-design-basis events (e.g., emergency 

planning and loss of large areas because of fires or explosions). NRC requirements to address 

concerns of adequate protection are developed and imposed without consideration of cost. 

The second tier of the NRC’s safety structure was described by the Court as follows:

If it so desires, however, the Commission may impose safety measures on 

licensees or applicants over and above those required by section 182(a)’s 

adequate-protection standard. As we have noted, section 161 of the Act 

empowers the Commission to issue rules, regulations, or orders to “protect 

health or to minimize danger to life or property.” 42 U.S.C. Sec. 2201(b), (i). This 

section cannot be read simply to permit the Commission to provide adequate 
A Proposed Risk Management Regulatory Framework

H-2 | Risk Management Task Force Appendix H

protection; another section of the Act requires the Commission to do that much. 

We therefore must view section 161 as a grant of authority to the Commission to 

provide a measure of safety above and beyond what is “adequate.” The exercise 

of this authority is entirely discretionary. If the Commission wishes to do so, it 

may order power plants already satisfying the standard of adequate protection 

to take additional safety precautions. When the Commission determines whether 

and to what extent to exercise this power, it may consider economic costs or 

any other factor. The Commission, after all, need not exercise the authority 

granted by section 161 at all; given this fact, the Commission certainly may use 

cost-benefit analysis to decide whether exercising the authority conferred by 

section 161 makes economic or policy sense.

In the development of the alternatives presented within this appendix, the Risk Management 

Task Force (RMTF) considered how some requirements would be associated with the first tier 

mission of the NRC to ensure an adequate protection of public health and safety and common 

defense and security, and how other requirements would address the second tier by going 

beyond adequate protection in an attempt to minimize danger to life or property. As described 

by the Court of Appeals and within longstanding NRC regulations and guidance documents, the 

requirements established to minimize risk (beyond measures needed for adequate protection) 

will not attempt to eliminate all risk, but will instead pursue reasonable reductions. An evaluation 

of the costs and benefits of proposals falling within the second tier will be used as part of the 

determination of what is a reasonable requirement to minimize risks to public health and safety 

and common defense and security. 

The RMTF has incorporated the two-tier safety structure discussed above, along with the risk 

management goal described in Chapter 2, into the alternatives discussed in this appendix. The 

risk management goal is to provide risk-informed and performance-based defense-in-depth 

protections to:

1. Ensure appropriate barriers, controls, and personnel to prevent, contain, and mitigate 

exposure to radioactive materials, according to the hazard present, the relevant 

scenarios, and the associated uncertainties.

2. Ensure that the risks resulting from the failure of some or all of the established barriers 

and controls, including human errors, are maintained acceptably low. 

Taken together, the above criteria address both tiers of the safety structure. However, it is not 

the case that the first criterion aligns with the first tier (adequate protection) and the second 

criterion aligns with the second tier (additional protections). Instead, a specific design feature 

or operating control can be seen as addressing one or more of the factors within the risk 

management goal, and the importance of the protection provided will determine whether it 

serves to ensure adequate protection or provides additional protections to further minimize 

dangers to life or property. 

As discussed in Appendix F, “Nuclear Power Reactors – Licensing-basis Events,” the selection 

and analysis of relevant scenarios play an important role in the design and operation of nuclear 

power reactors. For the purpose of this appendix, the interrelated processes used to define 

the proposed Risk Management Regulatory Framework are discussed in terms of the following 

phases of developing and maintaining design features and operating controls:
Appendix H Risk Management Task Force | H-3

A Proposed Risk Management Regulatory Framework

•	 selection of events and conditions to be considered in establishing plant 

design features and operating controls

•	 selection and implementation of plant design features and operating 

controls to meet the defined acceptance criteria for selected events

•	 preservation of the risk-informed, performance-based defense-in-depth outcomes 

provided by the selected plant design features and operating controls

Each of the above activities involves the regulatory decisionmaking process discussed in 

Chapter 3. Event selection is an engineering exercise and includes defining the types of 

internal transients and external challenges to be analyzed in support of establishing design 

features and operating controls. An important part of event selection is defining categories, 

such as design-basis and design-enhancement; related standards and conventions, such 

as applying the single-failure criterion and determining equipment safety classifications; and 

defining expectations for the appropriate combination of mechanistic and risk assessment 

analyses techniques. The output of the event selection phase includes relevant scenarios and 

related conditions, forces, and other parameters used in determining the appropriate design 

features and operating controls. The selected design features, operating controls, or other 

actions provide appropriate risk-informed and performance-based defense-in-depth protections 

against the scenarios and associated uncertainties. The implemented actions can, in turn, 

be considered the starting point for future decisions on maintaining or changing designs or 

operating practices. This third part of the interrelated activities is intended to ensure that the 

outcome of the previous steps—providing risk-informed defense-in-depth protections—is 

maintained even though changes to facilities and procedures will occur over the lifetime of each 

nuclear power reactor. 

The event selection, action selection, and outcome preservation terms are used to describe 

roles and responsibilities of the NRC and licensees and other attributes of the alternatives 

described in this appendix. The alternatives will also be presented consistent with the event 

categories described in Appendix F and elsewhere in this report, namely:

•	 Design-basis Events

o normal operation

o anticipated operational occurrences

o design-basis accidents

o design-basis external hazards

•	 Beyond-design-basis Events

o design-enhancement events 

−	 internal events

−	 external hazards
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H-4 | Risk Management Task Force Appendix H

o residual risk scenarios

−	 internal events

−	 external hazards

Design-basis events have traditionally been associated with mechanistic analyses, reliance 

on and protection of safety-related equipment, and use in establishing technical specifications 

and other licensing-related operational controls. Beyond-design-basis events have more often 

included the use of best-estimate-type analyses, probabilistic risk assessments (PRAs), and 

use in establishing additional plant protections to further minimize risks to the public health and 

safety and common defense and security (e.g., additional protections for station blackout (SBO) 

conditions and aircraft impacts).

A specific example is the selection of a weather event, such as a hurricane, with related wind 

speeds and surge levels. Licensees address the selected event by actions, such as designing 

and building structures with appropriate strengths and installing watertight doors to protect 

safety-related equipment installed at elevations below the defined surge level. Preserving the 

outcome is achieved through programs to evaluate plant changes to ensure that the desired 

protections against a hurricane are maintained over the life of the facility. Event categories can 

be used to distinguish between “design-basis” forces and surge levels and storm conditions 

considered less likely or otherwise different from the design-basis scenario. The evaluations of 

design features and operating controls for the design-basis event currently credit only safetyrelated systems, structures, and components. The proposed design-enhancement category 

would capture other storm conditions and could allow licensees to credit other available 

equipment and procedures.

Two key concepts related to event selection, categorization, and subsequent design features 

and operating limits are: 1) the threshold to define when an event needs to be considered within 

a category, and 2) the acceptance criteria to define when a design feature or operating limit 

provides the desired protection from the defined scenario(s). Differences in how these concepts 

are incorporated into the event selection, action selection, and outcome preservation phases 

define the three alternatives described in this appendix for improving the NRC’s development of 

the proposed category of design-enhancement events. 

As discussed in Chapter 4 and elsewhere in this report, the RMTF recommends that the 

NRC recognize “design-enhancement events” as a specific category of beyond-design-basis 

events. The purpose of the design-enhancement category is to address gaps that exist 

between the regulatory controls that are appropriate to address the risk management goal 

(e.g., risk-informed, performance-based defense-in-depth) and current controls involving a 

combination of design-basis events and ad hoc requirements added in reaction to specific 

events or other concerns. The goal would be to define a consistent approach for such events 

in terms of analysis techniques, safety classification, change control, reporting, and other 

regulatory requirements that have been defined previously on a case-specific basis. Any of 

these alternatives could also be used to clarify the role of programs, such as severe accident 

management guidelines (SAMGs) within the revised or added regulations for beyond-designbasis events. The Task Force envisions that the combination of design-basis events, designenhancement events, and various programs such as emergency preparedness collectively 

define the risk-informed and performance-based defense-in-depth protections that are the 

centerpiece of the proposed Risk Management Regulatory Framework. 
Appendix H Risk Management Task Force | H-5

A Proposed Risk Management Regulatory Framework

The RMTF’s proposed creation of a design-enhancement category is generally consistent with 

recommendations by the Fukushima Near-Term Task Force (NRC, 2011) and the Western 

European Nuclear Regulators’ Association (WENRA, 2008). The WENRA design extension 

category was discussed in Appendix G, “Nuclear Power Reactors – Safety Classification,” and is 

described as follows:

The design extension analysis shall examine the 

performance of the plant in specified accidents beyond 

the design basis, including selected severe accidents, 

to minimize, as far as reasonably practicable, releases 

harmful to the public and the environment in cases of 

events with very low probability of occurrence. 

Beyond-design-basis events shall be selected (based 

on a combination of deterministic and probabilistic 

assessments, as well as engineering judgment) and 

considered in the safety analysis to determine those 

sequences for which reasonable, practicable, preventive, 

or mitigative measures can be identified and implemented. 

Realistic assumptions and modified acceptance criteria 

may be used for the analysis of the beyond-design-basis 

events. (WENRA)

The concept of design-extension events or conditions has also been incorporated into draft 

standards prepared by the International Atomic Energy Agency (IAEA) related to the design 

of nuclear power plants (IAEA, 2010) and the safety classification of structures, systems, and 

components (IAEA, 2011). For example, the draft standard related to the safe design of nuclear 

power plants states:

Requirement 21: Design extension conditions (IAEA)

A set of design extension conditions shall be derived from 

engineering, deterministic and probabilistic considerations 

for the purpose of establishing further boundary conditions 

for the plant to withstand without acceptable limits for 

radiological consequences being exceeded. These 

conditions shall be used to identify the additional accident 

scenarios to be addressed in the design and to plan 

practicable provisions for the prevention and mitigation of 

such accidents.

Within the IAEA discussions of design extension conditions are related NRC requirements and 

activities to address beyond-design-basis events (e.g., severe accident design features for new 

reactors and severe accident management guidelines for operating reactors). 
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H-6 | Risk Management Task Force Appendix H

H.2 Design-Enhancement Events – Possible Alternatives

The following figure represents a general framework for the proposed risk-informed, 

performance-based defense-in-depth approach:

Integrating the above discussions, the RMTF developed several alternatives for moving forward 

on the proposed Risk Management Regulatory Framework for nuclear power reactors. The 

alternatives considered variations in the following three questions:

•	 Who identifies and decides on the appropriate actions to address design-enhancement 

events?

•	 What thresholds are defined for inclusion in the design-enhancement category?

•	 What acceptance criteria are defined for actions taken to address events and conditions 

in the design-enhancement category?

Figure H-1 Regulatory Framework for Nuclear Power Reactors
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A Proposed Risk Management Regulatory Framework

Possible variations in handling the above choices result in the three alternatives depicted below:

A discussion of the three selected approaches to design-enhancement is provided below. An 

additional discussion (Section H.3) is provided following the description of the three alternatives 

regarding possible changes to the design-basis events that may be facilitated by the introduction 

of the design-enhancement category for operating and new power reactors.

H.2.1 Alternative 1: NRC Identifies Design-enhancement Events

In the first alternative, the identification of a new category of design-enhancement events takes 

advantage of existing studies and lessons learned from Fukushima and other historical events. 

The NRC would identify specific events for inclusion in the design-enhancement category. The 

NRC would likewise define the acceptance criteria and other requirements associated with 

each of the specified events within this new category. This basic process (without defining 

common attributes such as reporting and change control) was used when the NRC imposed 

requirements for events or conditions, such as SBO and aircraft impact assessments, for new 

reactors. A possible approach for the event selection phase of the overall process to address 

design-enhancement events is shown in Figure H-2 as the following steps:

1. The NRC identifies initiating event(s)

•	 Considers internal transients, external hazards, and 

specific events and conditions

2. The NRC performs technical analyses 

•	 Considers safety benefits, operating experience, and 

regulatory analysis

Initial scenario 

categorization 

(who and what) 

Licensee 

following 

NRC-defined 

thresholds and 

acceptance 

criteria 

NRC 

Scenario 

disposition 

Licensee 

implements 

actions: 

design features 

and 

operating controls 

ALARA 

Risk Limits 

Alternative 1 

Alternative 2 

Alternative 3 

Figure H-2 Possible Alternatives for Nuclear Power Reactors Figure H-2 Possible Alternatives for Nuclear Power Reactors
A Proposed Risk Management Regulatory Framework

H-8 | Risk Management Task Force Appendix H

3. The NRC defines design-enhancement events

•	 Defines specific events or conditions (e.g., SBO)

4. The NRC defines acceptance criteria

•	 Considers examples such as fuel integrity, coping time, 

and codes and standards

The above steps would provide a methodical approach and consistent logic to the identification 

of design-enhancement events. The identification of and defining acceptance criteria for events 

on an ad hoc basis is considered to be the current practice and is described in Chapter 4 as 

Option A. While informed by risk assessments and the NRC’s historical practice of treating 

design-enhancement events within the second tier of its safety structure (i.e., added protection), 

the determination of whether protection from a particular event is important for adequate 

protection or as a safety enhancement would continue to be made on a case-by-case basis. 

Note that the acceptance criteria and evaluation techniques could be prescriptive such as they 

are in the SBO rule or performance-based such as in the aircraft impact rule for new reactors. 

In accordance with the proposed Risk Management Regulatory Framework promoted in the 

main body of this report, the NRC’s objective is to allow performance-based approaches 

wherever practical. In addition, the design-enhancement events should be part of the overall 

combination of events and related actions that ensure that risk-informed and performancebased defense-in-depth protections are provided. Given the significant role of the NRC in 

defining events within this example, it would not be necessary to require designers or licensees 

Figure H-3 Alternative 1a (NRC identified, NRC approval)
Appendix H Risk Management Task Force | H-9

A Proposed Risk Management Regulatory Framework

to perform additional studies or identify design or site-specific hazards for possible inclusion 

as design-enhancement events. The possible need for plant-specific PRAs or other analytical 

tools would depend on the requirements defined for the specific design-enhancement events 

(e.g., the analyses to support licensee implementation (step 5) and change control (step 13) in 

Figure H-3). The lack of site-specific risk assessments and evaluations of appropriate actions is 

a significant shortcoming of this alternative. 

A possible approach to the action selection and implementation phase consists of the 

licensee performing a facility-specific implementation analysis for the NRC defined event and 

acceptance criteria, and reporting its results and planned actions to the NRC for review and 

approval (steps 6 and 7 in Figure H-3). This approach has been used for past and ongoing 

activities, such as implementing the SBO rule and the adoption of National Fire Protection 

Association 805, “Performance-based Standard for Fire Protection for Light-Water Reactor 

Electric Generating Plants” (NRC, 2009). This approach may be required in those cases 

involving amendments to licenses. However, the NRC review and approval of licensees’ 

analyses and implementation plans can add significantly to the resources and schedules needed 

to resolve the addition of or changes to events incorporated into the licensing-basis for nuclear 

power reactors. An approach taken for aircraft impact assessments for new reactors is shown in 

Figure H-4. This approach emphasizes NRC reviews and inspections during the implementation 

phase and is usually associated with a more performance-based approach to issue resolution. 

An historical example of event and action selections relates to the availability of electrical 

power for plant equipment. The design-basis events include several events, such as the loss 

of offsite power scenario as an anticipated operational occurrence, and assumptions regarding 

offsite power for the loss of coolant design-basis accident. These events, in turn, led to actions 

such as the design and installation of safety-related diesel generators, sequencing of loads 

Figure H-4 Alternative 1b (NRC identified, performance-based)
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H-10 | Risk Management Task Force Appendix H

to be added to the electrical busses, and technical specifications to maintain diesel generator 

availabilities and capabilities. To address insights from PRAs and operational experience, 

the NRC imposed additional requirements for plants to withstand for a specified duration 

and recover from a station blackout event, which assumes the loss of both offsite and onsite 

emergency power sources. In the aftermath of the attacks of September 11, 2001, the NRC and 

stakeholders developed additional measures to provide electrical power to selected equipment 

in the event of a loss of a large area of a plant because of explosions or fires. The Fukushima 

accident is resulting in additional studies and may possibly identify additional risks, events, or 

conditions related to responding to a prolonged loss of electrical power. All of these scenarios 

and related actions form part of the construction of risk-informed and performance-based 

defense-in-depth for each nuclear power reactor licensed by the NRC.

The third phase shown in the above figures involves licensee and NRC actions to preserve 

the risk-informed defense-in-depth protections for the relevant scenarios and associated 

uncertainties. The needed elements to maintain defense-in-depth protections include reporting 

requirements, inspection activities, configuration control, change control, and corrective action 

programs. This phase has sometimes been referred to as maintaining the design basis for 

facilities. The movement to a risk management strategy provides an opportunity to revise the 

emphasis from “maintaining design basis,” which can be perceived as a general resistance to 

change, to emphasizing performance-based outcomes in terms of preserving risk-informed and 

performance-based defense-in-depth through ensuring appropriate barriers and controls.

Several possible changes to regulations to implement Alternative 1 are described below:

50.2 Definitions Add definition for design-enhancement events, risk 

management, and risk-informed and performance-based 

defense-in-depth. Add or clarify other terms to clearly 

distinguish event categories and related regulatory 

treatments.

50.x Design-Enhancement Add specific regulation(s) or revise existing regulations to

Events introduce the design-enhancement event category. In 

this example, the rule would likely define common 

attributes, such as change control and reporting (e.g., 

placement within final safety analysis report (FSAR)) and 

would designate specific events for the category. The 

appropriate treatment of equipment and operating controls 

would also be defined. Existing requirements, such as 

SBO, anticipated transient without scram (ATWS), and 

aircraft impact assessment (AIA) could be incorporated or 

referenced in the new regulation. 

This alternative might be the most straightforward and perhaps quickest to implement of the 

three described. It would not, however, easily address site-specific hazards and the “generic 

requirements” could require actions at some plants to address very unlikely scenarios for their 

specific locations. The NRC could periodically assess new information and pursue regulatory 

changes through existing programs, such as the operating experience program, generic safety 

issue process, generic communication program, and rulemaking. 
Appendix H Risk Management Task Force | H-11

A Proposed Risk Management Regulatory Framework

The RMTF has presented Alternative 1 as a limited improvement over the NRC’s existing 

approach. Historically, this approach has been able to address emerging regulatory issues, but 

it has had problems in areas of timeliness and consistency (e.g., the patchwork mentioned in 

the Fukushima Near-Term Task Force (NTTF) report) (NRC, 2011)). Although not the favored 

approach of the RMTF, the system described as Alternative 1 could be used as a transition to 

the more risk-informed, performance-based alternatives. As an example, the NRC could identify 

specific events or conditions identified following the Fukushima accident, and these could be 

pursued to ensure a timely response to concerns such as possible flooding scenarios. In the 

longer term, plant-specific risk assessments would be performed and additional actions would 

be taken as called for in approaches such as those described as Alternative 2 or 3. 

H.2.2 Alternative 2: NRC Identifies Thresholds for Event Sequences,

Acceptance Criteria Are Based on ALARA Principles

As discussed in other sections of this report, a key to the implementation of risk management 

systems is the definition of thresholds and acceptance criteria to support the deliberative 

process. Under this alternative, the NRC would define the threshold for events falling within 

the design-enhancement category using the risk-informed and performance-based defense-indepth principle to address possible shortfalls (or excesses) in design-basis events and ensure 

that the risks resulting from the failure of established barriers and controls are maintained 

acceptably low. The threshold would, as much as possible, build upon existing practices, such 

as the requirements and guidance for regulatory analyses, backfits, severe accident mitigation 

alternatives (SAMAs), and risk-informed licensing (e.g., Regulatory Guide 1.174, “An Approach 

for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-specific Changes 

to the Licensing-basis,” (NRC, 2011a)). 

There are many considerations in the identification of appropriate thresholds, including the 

analysis techniques (e.g., scope and level of PRAs), the handling of uncertainties, and the 

definition of relevant scenarios. One element of the thresholds could, for example, be defined in 

terms of relevant scenarios (defined by a single or group of PRA sequences) with an estimated 

core damage frequency (CDF) greater than 10-5 per year or a large early release frequency 

(LERF) greater than 10-6 per year, which is generally consistent with the NRC’s guidelines for 

performing regulatory analyses (NRC, 2004)1. At the point where Level 3 PRAs are available, 

the NRC’s Quantitative Health Objectives (QHO) or other societal measures could be directly 

considered as part of the event categorization criteria. Other criteria or relevant scenarios 

could be included, as needed, to complement the frequency-related parameters (e.g., security 

events, aircraft impact for new reactors). It would also be necessary to define approaches for 

addressing initiating events (e.g., seismic) for which either the conditional failure probability or 

the uncertainties increase dramatically as frequencies are reduced to levels approaching the 

thresholds for inclusion or the acceptance critieria defined for the design-enhancement category 

(e.g., 10-5 per year) (Johnson and Apostolakis, 2012).

Alternative 2 would require licensees to perform periodic reviews and analyses to identify 

relevant scenarios and determine appropriate actions to address design-enhancement events. 

A requirement similar to 10 CFR 50.71(h) to have and upgrade PRAs would be necessary 

1 The values used in this report (e.g., criteria such as CDF greater than 10-5 per year) are used to describe 

the alternatives and are not specific RMTF recommendations. Although the values used are generally 

consistent with current NRC guidance documents, the development of a more detailed proposal would need 

to carefully consider defining the thresholds and acceptance criteria and ensure the criteria support the 

overall risk management goal.
A Proposed Risk Management Regulatory Framework

H-12 | Risk Management Task Force Appendix H

for operating reactors, and a new or revised regulation would require the periodic analysis 

to identify relevant scenarios, considering plant changes, operating experience, and new 

information. As such, this alternative changes the focus from simply clarifying and expanding 

the design-enhancement category and would require licensees to adopt a risk management 

approach for plant design and operation (i.e., the identification and resolution of events falling 

within the design-enhancement category). 

The acceptance criteria for the periodic analysis performed by licensees for designenhancement events for this alternative would be defined primarily in terms of the “as low as 

is reasonably achievable” or ALARA principle, similar to that used in the radiation protection 

arena. The process would employ the decisionmaking process described in Chapter 2. It could 

use much of the existing guidance (updated, as needed) mentioned previously for regulatory 

analyses, backfits, SAMAs, and risk-informed licensing actions. For example, existing guidance 

includes several criteria, one of which is a factor of dollars per person-rem (roentgen equivalent 

man) avoided through the installation of additional design features to address severe accident 

scenarios. The development of this alternative would include defining, an appropriate, 

periodicity for the analyses to identify and address the new design-enhancement events. 

Adoption of this alternative would require an update of regulatory analysis guidelines to ensure 

an effective cost-benefit analysis is performed by licensees when considering ways to address 

design-enhancement events. As with the discussion on thresholds, it may be appropriate 

to include considerations for different types of design-enhancement events. For example, 

the AIAs required for new reactors would be an example of design enhancement, but they 

have acceptance criteria defined in terms of maintaining barrier functions and not specifically 

frequency of core damage or radiological release. Considerations within the ALARA criteria 

also depend on assessment techniques with current Level 1 PRA-based criteria using surrogate 

measures, such as CDF, and future Level 3 PRAs expected to support criteria based on 

estimated health effects or other societal risk measures. Other factors might be considered to 

supplement those identified for various scenarios (e.g., CDF or health objectives) by including 

measures based on the importance of design features or operating controls in preventing 

or mitigating multiple scenarios (e.g., PRA importance measures) or in addressing core 

safety functions (i.e., reactivity control, heat removal, confinement of radioactive materials). 

These types of consideration are integral to current programs addressing special treatment 

requirements for different risk-informed equipment categories defined in 10 CFR 50.69, “Riskinformed Categorization and Treatment of Structures, Systems and Components for Nuclear 

Power Reactors,” and the regulatory treatment of non-safety systems (RTNSS) for new reactors 

(NRC, 1995). 

The current requirements in 10 CFR 50.71(h) require new reactor licensees to upgrade their 

PRA every 4 years, with the upgraded PRA covering initiating events and modes of operation 

contained in NRC-endorsed consensus standards on PRA in effect 1 year before each 

required upgrade. The RMTF considers the 4-year interval to be a reasonable starting point for 

discussions about the proposed risk-management rule. In addition, the current requirements 

include provisions to improve risk assessments using the NRC endorsement of consensus 

standards as the vehicle to upgrade the PRA to cover additional modes and initiating events. 

A similar approach could be taken for the PRA requirements within this alternative and could 

support the ultimate movement to Level 3 PRAs, which could better support the site-specific 

ALARA evaluations for design-enhancement conditions.
Appendix H Risk Management Task Force | H-13

A Proposed Risk Management Regulatory Framework

This alternative is illustrated in Figure H-5. It is readily apparent that this alternative changes 

the emphasis from the NRC identifying specific scenarios to the NRC defining criteria for design 

and site-specific risk management to be performed by licensees or applicants. 

Several possible changes to regulations to implement Alternative 2 are described below:

50.2 Definitions Add definition for design-enhancement events, risk 

management, and risk-informed and performance-based 

defense-in-depth. Add or clarify other terms to clearly 

distinguish event categories and related 

regulatory treatments.

50.x Design-enhancement Add specific regulation or revise existing regulations to 

Events introduce the design-enhancement event category 

and define the thresholds and acceptance criteria for 

the scenarios. The rule would likely define common 

attributes, such as change control, documentation, and 

reporting. It would also define the appropriate treatment of 

equipment and operating controls. Existing requirements, 

such as SBO, ATWS, and AIA, could be superseded by, 

incorporated (with either existing or revised requirements), 

or referenced in the new regulation.

Figure H-5 Alternative 2, Risk Management (ALARA)
A Proposed Risk Management Regulatory Framework

H-14 | Risk Management Task Force Appendix H

50.y Risk Management A requirement would be included for licensees to 

periodically assess and address potential events 

meeting thresholds derived from the risk-informed and 

performance-based defensein depth definition. The 

technical analyses would involve risk assessments (e.g., 

PRAs) and other techniques, as necessary, to address 

relevant scenarios. This approach could address matters 

such as GSI-199, “Implications of Updated Probabilistic 

Seismic Hazard Estimates in Central and Eastern United 

States on Existing Plants,” and the periodic assessments 

recommended in the Fukushima NTTF report.

50.71(h) Probabilistic risk This requirement exists for new reactors and would be 

assessment (PRA) added for operating reactors. This rule could be revised to 

address the risk management requirements 

mentioned above.

In constructing the requirements under this alternative, it would be made clear that specific, 

NRC regulations (existing or future) define the necessary requirements for adequate protection 

of the public health and safety. The design-enhancement and risk management regulations 

discussed here would be confined to the second tier of requirements that provide protection 

beyond that level. It would be possible under this approach to maintain the NRC’s longstanding 

preference to not have a specific definition of adequate protection in terms of event frequencies 

and consequences. The thresholds (e.g., 10 CFR Part 20, “Standards for Protection Against 

Radiation,” requirements for high frequency–low consequence events and CDF greater than 

10-5 per year for low frequency–high consequence events) are a logical extension of current 

practice. Licensees would be required to assess scenarios within the design-enhancement 

category to reduce risk to ALARA levels. The possible design features or other measures 

would be evaluated against their efficacy in reducing the identified risk in a cost-effective 

manner. The whole approach would be based on the decisionmaking process described in 

Chapter 2. The determination of “reasonable” or cost-effective would use existing or future 

guidance used by the NRC in its own determinations (e.g., analyses for SAMAs, rulemakings, 

or backfits).2 This should help provide consistency between evaluations performed by licensees 

and the NRC staff. 

The proposed additional risk management requirement would involve licensees periodically 

(e.g., every 4 years or upon new information becoming available) assessing postulated initiating 

events, evaluating riskreduction measures, and, where cost-effective, implementing those 

measures to address relevant scenarios exceeding the defined threshold. A summary of the 

assessment and related actions would be provided to the NRC, possibly in an existing reporting 

requirement, such as updating of the FSARs. It should be noted that the periodic assessments 

and consideration of cost-effective measures to reduce risk would not replace routine (i.e., dayto-day) risk management functions, which would continue to be supported by technical 

specifications, maintenance rule, and other programs related to configuration management and 

plant operations. 

2 A question to address during the development of the design-enhancement category and ALARA approach 

would be the threshold for including relevant scenarios for consideration. For example, 10 CFR 50.109, 

“Backfitting,” requires that the NRC impose a backfit only when it determines that the new requirements 

would result in a “substantial increase in the overall protection of public health and safety.” NRC guidance 

documents (NRC, 2004) equate a substantial increase to a reduction in CDF of greater than 10-5 per year.
Appendix H Risk Management Task Force | H-15

A Proposed Risk Management Regulatory Framework

Although arguably consistent with existing requirements and practices related to thresholds 

and acceptance criteria, Alternative 2 would likely present a communications challenge given 

that it requires specific discussion of the distinction between adequate protection and added 

protection. Whereas some would see this alternative as an imposition on licensees to do 

additional assessments and take additional actions, others would likely see it as allowing 

licensees to forgo risk reductions based on financial considerations. A recent case that 

highlights the agency’s longstanding difficulty in explaining the subtleties of these concepts is 

the discussions surrounding some recommendations from the Fukushima NTTF. 

Regarding new plant designs and licensing, the adoption of the ALARA principle would reinforce 

the stated objective in the Advanced Reactor Policy Statement (NRC, 2008) for new designs to 

provide enhanced margins of safety and use simplified, inherent, passive, or other innovative 

means to accomplish their safety functions. SECY-10-0121, “Modifying the Risk-informed 

Regulatory Guidance for New Reactors” (NRC, 2010), discusses several issues related to a 

possible discrepancy between the Advanced Reactor Policy Statement goals for enhanced 

safety for new plants and the actual regulation and oversight of new reactors based on the 

same rules and guidance applied to operating reactors. The proposed ALARA approach for 

the design-enhancement conditions could be a practical means to require new plants to define 

specific measures to reduce risks since such measures could be developed in a more costeffective way before completing plant designs or construction. This approach would be similar 

to that taken for AIAs and recognizes that integrating protections into plant designs can be 

cost effective even if major plant changes to an operating reactor would be impractical. This 

alternative has the advantage that the desired outcome is achieved without having different 

regulatory requirements for new and existing reactors, and yet the risk reductions achieved 

by newer plant designs would be maintained by the proposed risk management rule. Some 

new reactor designs have even included specific measures to address the RTNSS as a way to 

reduce risks identified in PRAs while controlling costs by crediting existing systems 

and programs.

Additional improvements might be pursued along with the specific changes needed to implement 

this alternative to ensure consistency between various regulations for event analyses, change 

control, and reporting. For example, various external events (e.g., weather, industrial facilities, 

seismic, and flood) that include event frequencies as a key parameter use different thresholds 

and acceptance criteria for those frequencies. Other regulations governing activities such as 

changes, tests, and experiments; maintenance; and fire protection also use risk-insights to 

govern plant operations or interactions with the NRC. While not necessarily essential for this 

alternative, the development of the design-enhancement and risk management requirements 

would be an opportunity to review and revise various requirements to ensure a consistent 

treatment of the relevant scenarios. The development of the risk management requirements 

and associated requirements for the treatment of equipment and operating controls could 

also improve the alignment between the licensing and regulations for power reactors and the 

reactor oversight process (ROP), which already relies heavily on risk assessment techniques. 

For example, plants currently install safety enhancements to improve plants’ risk profiles and 

thereby avoid adverse performance indicators or reduce the significance of possible violations 

when evaluated using the ROP significance determination process. The NRC encourages 

such enhancements, but it is not always clear what, if any, regulatory controls are applicable to 

them. This alternative would likely capture these enhancements within the proposed designenhancement and risk-management regulations. A similar discrepancy has been identified 

between the NRC’s environmental regulations, which require an assessment of severe accident 

mitigation alternatives, and the “safety regulations,” which do not require the implementation of 
A Proposed Risk Management Regulatory Framework

H-16 | Risk Management Task Force Appendix H

cost-justified enhancements. In addition to the opportunities to improve consistency, there would 

likely be some necessary changes to peripheral regulations to maintain an effective regulatory 

system. For example, the requirements within the proposed risk management rule would need 

to be developed to work with the change control provisions of other NRC regulations (e.g., 

10 CFR 50.59, “Changes, Tests, and Experiments”) to avoid conflicting findings regarding a 

proposed plant modification.

H.2.3 Alternative 3: NRC Identifies Thresholds for Initiating Event Frequencies, 

Acceptance Criteria Are Fixed 

Alternative 3 is similar to Alternative 2 except it replaces the ALARA principle for the designenhancement analyses with more definitive acceptance criteria in terms of event frequencies 

and consequences. The process shown in Figure H-3 would therefore represent this alternative, 

except the thresholds and acceptance criteria would be revised. Although this alternative, when 

compared to Recommendation 2, does not as clearly address Recommendation PR-R-2 by 

considering costs in establishing design-enhancement events, a proposal similar to Alternative 3 

has been made in various forums and is presented here for the sake of discussion. Note that 

in this alternative, the design-basis events continue to be addressed by existing regulations 

and traditional approaches and the risk management requirements focus on the designenhancement category.

As with Alternative 2, there are numerous ways to define the thresholds and acceptance criteria, 

depending on the chosen risk assessment techniques and other considerations. In the example 

included here to describe this alternative, the threshold for design-enhancement events is based 

on postulated initiating event (PIE) frequency, and the acceptance criteria are defined in terms 

of parameters, such as the QHO, or, as in Alternative 2, surrogate measures, such as CDF (and 

possibly adding LERF as another measure). Additional considerations in defining the thresholds 

could bring in measures based on the importance of design features or operating controls in 

preventing or mitigating multiple scenarios (e.g., PRA importance measures) or in addressing 

core safety functions (i.e., reactivity control, heat removal, confinement of radioactive materials). 

For the purpose of explaining this alternative, the example uses a PIE frequency of 10-5 per year 

as the threshold and a CDF of 10-6 per year as the acceptance criteria. In other words, 

licensees in this example would be required to:

1. Perform the periodic reassessment discussed in Alternative 2.

2. Identify scenarios (e.g., single or multiple PRA sequences) in which the PIE frequency 

exceeds 10-5 per year and the related CDF exceeds 10-6 per year (or equivalent if using 

LERF or QHO).

3. Identify and implement measures to reduce the frequency of the initiating event or 

resultant core damage for the relevant scenarios from Step 2 such that threshold or 

acceptance criteria are not exceeded.

To differentiate between those requirements needed for adequate protection (Tier 1) and those 

pursued for additional safety (Tier 2), this alternative would emphasize differences in analysis 

techniques and special treatment requirements for credited equipment and procedures. As 

with Alternative 2, adequate protection would be addressed primarily through existing specific 

rules, traditional design-basis accidents, safety-related equipment, and well established 
Appendix H Risk Management Task Force | H-17

A Proposed Risk Management Regulatory Framework

procedures (e.g., emergency operating procedures).3 Design-enhancement events would be 

addressed similarly to Alternative 2 and would involve best-estimate risk assessments, reliable 

but not necessarily safety-related equipment, and response guidelines for severe accidents 

and extreme damage conditions. While not specifically including a cost-benefit assessment, 

licensees would be able to develop the most cost-effective approaches to address the 

relevant scenarios.

As with Alternative 2, pursuit of this option could also include assessments and revisions to 

other regulations to improve the consistency between various thresholds and decision criteria. 

This example would also present communication challenges given that it involves discussions 

of the differences in treatment of design-basis and design-enhancement events. In addition, the 

use of risk-informed acceptance criteria that include a specific CDF (e.g., 10-6 per year) or other 

risk measure might be interpreted as a new and clear definition of adequate protection since 

this alternative does not include a cost-benefit analysis for the features and controls added to 

address design-enhancement events. Such cost-benefit analyses are normally performed for 

the added protections associated with the second tier of the NRC’s safety structure, as defined 

by the Court of Appeals. As previously mentioned, the NRC has historically avoided a specific 

definition of adequate protection to ensure it maintains a needed degree of latitude in the 

regulation of nuclear power reactors. 

H.3 More on Design-Basis and Design-Enhancement Events

The focus of the first three alternatives was on establishing a category for certain beyonddesign-basis events. Requirements and processes would be established to identify the relevant 

scenarios, to define and implement measures to meet the acceptance criteria set for those 

scenarios, and to preserve the outcome of the effort by maintaining appropriate barriers and 

controls to satisfy the risk-informed defense-in-depth goal. Existing regulatory programs that 

relate to the design-enhancement concept include the RTNSS category of equipment in the new 

reactor arena and the ROP for operating reactors, which includes the significance determination 

process and performance indicators that consider beyond-design-basis events. There is also 

an overlap between some aspects of the alternatives discussed in this appendix and regulatory 

initiatives such as 10 CFR 50.69, and proposed changes to 10 CFR 50.46, “Acceptance Criteria 

for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors.” 

The previous alternatives could, in theory, be pursued without addressing the current 

handling of design-basis events. The alternatives could, however, support or be developed 

in conjunction with changes contemplated for regulations, such as 10 CFR 50.46. One of the 

difficulties in developing changes to that regulation is that the initiative attempts to separate 

the spectrum of loss of coolant accidents, all of which are currently considered design-basis 

events, into those remaining as design-basis events and those to become beyond-design-basis 

events. The development of the design-enhancement category could facilitate this change 

by providing a regulatory structure for the breaks beyond the break-size kept in the traditional 

design-basis event category. In the absence of a common category with defined acceptance 

criteria and rules for change control and reporting, the draft regulations for the 10 CFR 50.46 

initiative included specific new regulations in these areas for the breaks being removed from 

the design-basis event category. This, in turn, is seen by some as defeating the underlying 

3 Note that the RMTF proposed approach for Alternative 3 would maintain design enhancements as measures 

taken beyond that needed for adequate protection. This approach differs from the NTTF recommendation 

for the design-extension category to be imposed on the basis of adequate protection.
A Proposed Risk Management Regulatory Framework

H-18 | Risk Management Task Force Appendix H

goal of removing unnecessary regulatory burden associated with extremely unlikely events and 

has stymied the initiative. Performance-based approaches like Alternatives 2 and 3 could be 

developed to provide the necessary combination of regulatory structure and licensee flexibility to 

enable progress on 10 CFR 50.46 and similar initiatives. 

The development of and subsequent experience with 10 CFR 50.69 highlight the difficulty of 

trying to introduce risk-informed methodologies to address previously unregulated scenarios, 

leaving traditional design-basis events intact, and acknowledging that relief may be warranted 

from requirements for some equipment credited in responding to design-basis events. As 

discussed in Appendix G, “Nuclear Power Reactors – Safety Classification of Systems, 

Structures, and Components,” a major focus of 10 CFR 50.69 is the classification of equipment. 

The risk-informed safety class (RISC) categories developed for 10 CFR 50.69 are presented in 

Figure G-1, and repeated below:

There has been limited interest from the industry in adopting 10 CFR 50.69, largely because, 

although the available guidance details the categorization of equipment into the above classes, 

there does not appear to be agreement among parties as to the allowable relaxation from the 

special treatment requirements for the RISC-3 equipment. 

There is a correlation between the actions that would be taken to address design-enhancement 

events under the proposed risk management requirements and the actions taken to address 

the RISC-2 equipment under 10 CFR 50.69 and the RTNSS equipment for some new reactors. 

In each case, the resultant requirements improve the assessment of and the actions taken 

to respond to relevant scenarios that are not addressed well by the design-basis events but 

that nevertheless contribute to the risk profile of a nuclear power reactor. The previously 

discussed three alternatives do not address the possible changes to specific areas of nuclear 

power plant design and operation that result from the existing approach to design-basis events. 

Possible changes in this area have been recognized in various agency initiatives, including the 

Figure H-6 10 CFR 50.69 Risk-Informed Safety Class Categories Figure H-6 10 CFR 50.69 Risk-Informed Safety Class Categories
Appendix H Risk Management Task Force | H-19

A Proposed Risk Management Regulatory Framework

creation of the RISC-3 category in the above figure. The pursuit of Alternative 2 or 3 without 

some changes to the set of design-basis events and related regulatory requirements could be 

perceived as a one-sided application of risk insights, which is the perception that has limited the 

acceptance of 10 CFR 50.69.

A simplified summary of the current approach to design-basis events is provided within the 

following figure and table (Figure H-7 and Table H-1, repeated from Appendix F):

Figure H-7 Event Categories and IAEA Defense-in-depth
A Proposed Risk Management Regulatory Framework

H-20 | Risk Management Task Force Appendix H

Table H-1 Licensing Approach to Events (current approach)

PIEs Key Assumptions Acceptance 

Criteria

Normal Ops n/a

n/a 

Maintain initial conditions, standby 

equipment

10 CFR Part 20, 

ALARA

AOOs

Functional list

(e.g., ANS 

51.1/52.1 Cond II)

(PIE frequency < 

~10-2/yr)

Credit SR SSCs

single failure

Maintain safety 

limits such as 

DNB, RCS 

pressure

DBAs

Functional list

(e.g., ANS 

51.1/52.1 Cond 

III/IV)

(PIE frequency < 

~10-4/yr)

Credit SR SSCs single failure Maintain coolable 

core, containment

Design-basis

External

Events

maximum credible scenario (e.g., probability of 

exceedance < ~10-7 for tornado conditions)

Engineering 

analysis, safety 

margins

Codes and 

standards related 

to barrier design 

and equipment 

qualification

Design-Enhancement

(in concept but term not 

currently used)

Specific events

ATWS

SBO

AIA

As defined in each rule – generally 

best estimate

As defined 

in each rule 

(alternate 

shutdown, 

coping time, AIA 

functions)

Residual Risks

(in concept but term not 

currently used)

n/a n/a

n/a

(related items 

include severe 

accident features 

(new rxs), 

SAMDA, ROP, 

and risk metrics)

Beyond-Design

Basis External Events n/a n/a n/a
Appendix H Risk Management Task Force | H-21

A Proposed Risk Management Regulatory Framework

The current construct includes (1) thresholds for including events (historically addressed 

by standards such as American Nuclear Society Standard ANS 51.1(ANS, 1983)), (2) key 

assumptions and analysis techniques, such as the single failure criterion, and (3) acceptance 

criteria, such as safety limits (e.g., departure from nucleate boiling (DNB), reactor coolant 

system (RCS) pressure, peak cladding temperature, and offsite dose). Although the specific 

events and acceptance criteria are often not specifically addressed within the NRC’s 

regulations, the alignment of events, analysis methodologies, and acceptance criteria with 

the related plant equipment, operational programs, and other measures has evolved over 

many years and is ingrained in guidance documents, codes and standards, and institutional 

memory within the NRC and other organizations. It should be noted that previous evaluations 

and initiatives have not identified a fundamental flaw with the makeup or handling of designbasis events, and the RMTF agrees that, overall, the regulations and guidance have ensured 

that nuclear power plants pose no undue risk to public safety. And so, although possible 

improvements have been developed for the consideration of events within the NRC’s licensing 

processes, a major obstacle is the difficulty in mapping out a strategy for changing from the 

current state, which is generally considered acceptable, to any revised construct of events and 

related protected measures. 

The addition of a design-enhancement category and related requirements for identifying 

events and developing measures to keep risks as low as is reasonably achievable or below 

an established level may change the landscape currently defined by design-basis events and 

several selected conditions (e.g., SBO, ATWS, and aircraft impact for new reactors). This 

change may enable changes to the current handling of design-basis events by providing a 

defined regulatory program for items that may not warrant the requirements associated with 

design-basis events but should have some measures to manage the risks introduced by nuclear 

power reactors. A review of the existing design-basis events and the overall combination of 

events within all categories could be performed to determine whether some elements of the 

current design-basis events could be better addressed within the design-enhancement category. 

Any relocations from the design-basis categories to the design-enhancement category would 

likely support a more performance based way of ensuring risk-informed and performance-based 

defense in depth. In some ways, this effort would resemble previous NRC initiatives, such as 

conversion to the improved standard technical specifications, which involved relocating various 

requirements based on the criteria defined in 10 CFR 50.36, “Technical Specifications,” from 

plants’ technical specifications to licensee-controlled documents. This limited approach would 

not include significant changes to the general structure of the design-basis events and would 

not require immediate or dramatic actions by licensees to reanalyze events, change licensingbasis documents, or revise plant equipment or procedures. It is possible that such an approach 

could enable licensees to proceed in a piecemeal fashion, revising the treatment of specific 

events as needed to address particular plant modifications, equipment procurement issues, or 

other specific problems.

A premise for this limited approach is that adequate protection for a nuclear power reactor 

is provided by a licensee’s compliance with the requirements derived from the current set of 

design-basis events and regulatory requirements for specific programs, such as emergency 

planning. These requirements define the minimum set of protective measures that the NRC 

deems necessary to allow the operation of a nuclear power reactor. The situation is complicated 

somewhat by the mix of regulatory requirements used to ensure adequate protection and 

the fact that some are prescriptive in nature and some are performance-based. An example 

of prescriptive requirements is plant technical specifications, with requirements for plant 

shutdowns within specified time periods for certain conditions involving inoperable equipment. 
A Proposed Risk Management Regulatory Framework

H-22 | Risk Management Task Force Appendix H

Performance-based approaches include 10 CFR 50.54(hh)(2), which defines adequate 

protection requirements to develop and implement strategies to address core safety functions 

under circumstances associated with the loss of large areas of a plant because of explosions 

or fire but provides licensees flexibility in selecting and controlling equipment used to support 

those strategies. 

The decisionmaking process will need to consider the relationship of the specific event with 

other design-basis and beyond-design-basis events and related regulatory requirements, an 

assessment of options, the various technical analyses, implementation issues, stakeholder 

views, and other factors as part of the deliberation and resultant decision on the handling of 

specific events or conditions. Although not proposing to make recommendations on specific 

design-basis and design-enhancement events, the RMTF offers the following observations 

regarding the event types, thresholds, assumptions, and acceptance criteria discussed in 

Appendix F.

H.3.1 Normal Operations

Current requirements regarding normal plant operations include standard radiation protection 

and effluent controls and ensuring that initial conditions (e.g., power level and shutdown 

margin) are consistent with assumptions in design-basis accident analyses and that safety 

equipment (e.g., diesel generators and emergency core cooling) are available and capable of 

performing safety functions. There are several ongoing initiatives, such as efforts to risk-inform 

technical specifications that are leading to changes in requirements for how licensees maintain 

plant readiness for possible events. An example is the extension of the allowable times that 

equipment can be out of service if a licensee implements a risk-informed safety function 

determination program. In addition to continuing such initiatives, the Task Force envisions some 

changes to requirements for normal operation as a result of adding design-enhancement events 

and the possible movement of some current design-basis events (discussed below) to the 

design-enhancement category. 

H.3.2 Anticipated Operational Occurrences

The current list of anticipated operational occurrences for nuclear power reactors has evolved 

over the course of licensing various reactors, the development of regulatory guides, and the 

availability of consensus codes and standards. These transients are usually grouped by 

initiating events, such as loss of coolant flow, reactivity addition, and loss of heat removal. 

Anticipated operational occurrences are generally considered to include initiating events with 

an approximate frequency greater than 10-2 per year. The initial conditions for the transients 

are established as limits on normal operation and include parameters such as power levels, 

temperatures and pressures, and power distributions within the reactor core. The performance 

of the reactor protection system is modeled for bounding types of events within the category to 

provide confidence that safety limits associated with certain barriers (e.g., fuel cladding, reactor 

coolant system) are not exceeded. Conservative assumptions on plant parameters, coincident 

failures, the inclusion of the single failure criterion, and the reliance on only safety-related 

equipment are incorporated into the analysis.

With the addition of the design-enhancement category, the NRC and stakeholders could revisit 

the handling of some or all of the current AOOs. Given that preventing minor fuel damage is 

not necessarily part of the requirements for safety-related equipment and is as much an asset 

protection feature as it is a public health concern, a review of current AOOs could result in 
Appendix H Risk Management Task Force | H-23

A Proposed Risk Management Regulatory Framework

some scenarios being moved to the design-enhancement category. AOOs are also evaluated 

using analysis assumptions that may go beyond that needed for adequate protection (e.g., 

coincident failures) and the design-enhancement category could support movement of the less 

likely scenarios out of the design-basis events. However, the current safety limits and analyses 

are well established and changes would need to be evaluated to determine if any significant 

changes to the handling of AOOs were warranted. In the absence of new evaluations on the 

handling of AOOs, insights might be available from PRA studies, evaluations of 10 CFR 50.69, 

or other existing technical analyses that could determine whether some AOOs result in 

unwarranted protective measures and would be candidates for deletion or movement to designenhancement events.

H.3.3 Design-Basis Accidents

Design-basis accidents generally address possible but unlikely initiating events, such as loss 

of coolant accidents. The analytical approach includes assumptions such as single failures, 

coincident failures, and reliance on only safety-related equipment. Acceptance criteria for 

design-basis accidents are established to limit fuel damage, maintain ability to cool the core, 

and maintain the integrity of containment structures. Ongoing activities related to design-basis 

accidents include proposed changes to 10 CFR 50.46 (differentiating between very unlikely 

large breaks and breaks below a transition break size), and GSI-191 (evaluating long-term 

cooling requirements and issues related to blockage of containment sumps). As previously 

mentioned, the creation of the design-enhancement category could facilitate these activities 

by providing an alternative to the regulatory controls associated with design-basis accidents. 

For example, IAEA Draft Safety Guide DS-367, “Safety Classification of Structures, Systems, 

and Components in Nuclear Power Plants” (IAEA, 2011), which is described in Appendix G, 

distinguishes between equipment used to reach a controlled state following a design-basis 

accident and equipment used to subsequently achieve and maintain a safe shutdown state. 

Such a distinction could support assessing specific phases within the current design-basis 

accidents. Risk assessments might inform decisions on other design-basis accidents, such as 

secondary-side high-energy line breaks and control rod ejection accidents, which could, in turn, 

support changes to limits and equipment associated with those events. 

H.3.4 Design-Basis External Events

External events included in the design basis were intended to address credible scenarios 

related to external hazards, such as from neighboring industrial facilities, transportation 

routes, weather, floods, earthquakes, and security events. The analytical approach generally 

consists of defining a conservative challenge to the plant based on site-specific considerations. 

Conservative values for weather-related forces (e.g., wind speeds), flood levels, and seismicinduced motions are used as inputs to determine the needed protections for safety-related 

equipment. For example, a room containing emergency core cooling equipment would be 

evaluated to ensure that it would withstand design-basis winds (including debris), design-basis 

floods, and could withstand a design-basis earthquake. Confidence is provided by showing that 

required safety-related equipment is protected against all of the identified external threats, up to 

the design-basis values. Plants are also evaluated to show that operational programs related 

to fire protection and security are able to provide the required levels of protection from fire or 

malevolent acts. 

As mentioned in Chapter 4, one issue with plant-level design-basis events relates to the 

differences in thresholds for defining those events. An example would be establishing 
A Proposed Risk Management Regulatory Framework

H-24 | Risk Management Task Force Appendix H

requirements for structures at a site to ensure they could withstand a design-basis tornado 

determined using a probability of exceeding the design value set at 10-7 while the same 

structures might be based on a historical review of precipitation and flooding levels that are 

not easily expressed in probabilistic terms. The selection of design-basis values for external 

hazards is admittedly made difficult by the limited availability of historical information and 

predictive models. The Task Force has recommended that the NRC should evaluate its 

handling of external hazards and, to the degree possible, improve the consistency of regulatory 

treatment between the various threats (including security). The addition of the designenhancement category would provide regulatory controls for any initiating events that, upon 

further review, would not warrant treatment as design-basis events. Should the evaluations 

associated with the Fukushima accident or other assessments determine that design-basis 

external hazards were not adequately defined; the NRC would pursue the needed regulatory 

reforms to require analyses and plant changes. Although independent from the efforts of this 

Task Force, redefining design-basis external hazards could consider the desire to improve 

consistency and the possible introduction of risk management requirements for licensees. 

Improved consideration of possible cliff-edge effects (i.e., catastrophic results from exceeding 

design values) is envisioned as a result of adopting the risk-informed defense-in-depth goal and 

will be discussed within the section on beyond-design-basis external events.

H.3.5 Design-Enhancement Events

Several approaches to identify, analyze, and protect against design-enhancement events were 

provided in this appendix as Alternatives 1–3. 

H.3.6 Residual Risks

There are few specific regulatory requirements related to residual risks, which for the purpose 

of this report will be considered as any scenario not meeting the threshold of a designenhancement or design-basis event (e.g., a scenario with estimated CDF below 10-6 per year 

in the previously discussed Alternatives 2 and 3). The limited consideration of these events 

currently include consideration of severe accident features in new plant designs and the 

contribution of these events to the overall risk profile for comparison to the NRC safety goals 

or guidance in Regulatory Guide 1.174. The alternatives considered by the Task Force would 

not significantly change the treatment or role of the events within the residual risk category. 

The risk-informed, performance-based defense-in-depth approach would include a careful 

assessment of possible cliff-edge effects and other insights from the risk assessment to ensure 

that uncertainties and other factors were adequately considered for those scenarios in the 

residual risk category. 

H.3.7 Beyond-Design-Basis External Events

Beyond-design-basis external events have not generally been addressed within the NRC’s 

regulatory programs. The evaluations performed under the Individual Plant Evaluation – 

External Events activities did identify some site-specific vulnerabilities related to external 

hazards, and licensees took some corrective actions. The degree to which such actions were 

incorporated into licensing-basis documents (e.g., FSARs) varies. 

The adoption of the design-enhancement category could be especially useful in addressing 

site-specific external hazards that are not currently addressed by (or that would be relocated 

from) the design-basis events category. The periodic reanalysis included in Alternatives 2 and 3 
Appendix H Risk Management Task Force | H-25

A Proposed Risk Management Regulatory Framework

would require licensees to address new information on hazards, such as revised seismic hazard 

information that currently requires handling through the NRC’s generic safety issues and generic 

communications programs. An important aspect of the envisioned risk assessments is the 

identification and resolution of possible cliff-edge effects in which an external hazard exceeding 

a certain level has a very high probability of causing core damage. The flooding and resultant 

core damage events at the Fukushima reactor facility are examples of such vulnerabilities. There 

would continue to be events, however, that would be considered too improbable to be addressed 

within the design-enhancement category and would therefore go unaddressed or be considered 

only so far as they contribute to the overall risk profile of the nuclear power reactor. In addition, 

whatever approach is selected will need to recognize and resolve the limitations of dealing 

with seismic risks or other external hazards at extremely low event frequencies (Johnson and 

Apostolakis, 2012).

Special attention should be given to events or conditions that, although considered beyond the 

design basis for nuclear power reactors, resulted in actions deemed necessary for adequate 

protection of the public health and safety. Emergency planning is an example of an operational 

program that might not be necessary to address design-basis events but was added as a 

defense-in-depth measure and has been deemed necessary for adequate protection for 

operating nuclear power reactors. The requirements for licensees to develop a strategy for the 

loss of large areas of a plant because of explosions or fires were likewise issued on the basis 

that they were required for adequate protection. However, given the performance-based nature 

of that requirement, the resultant extreme damage mitigation guidelines and related equipment 

are treated in a manner usually associated with design-enhancement events.

H.4 References

(ANS, 1983) American National Standards Institute/American Nuclear Society, 

ANSI/ANS 51.1 – 1983, “Nuclear Safety Criteria for the Design of 

Stationary Pressurized Water Reactors,” ANS, LaGrange, Park, IL.

(IAEA, 2010) International Atomic Energy Agency, “Draft – Safety of Nuclear 

Power Plants: Design,” Draft Safety Guide DS-414, 

September 2010.

(IAEA, 2011) International Atomic Energy Agency, “Draft – Safety Classification 

of Structures, Systems, and Components in Nuclear Power 

Plants,” Draft Safety Guide DS-367, April 2011.

(Johnson and B.C. Johnson and G.E. Apostolakis, “Seismic Risk Evaluation 

Apostolakis, 2012) within the Technology Neutral Framework,” Nuclear Engineering 

and Design, 242 (2012) 341– 352.

(NRC, 1995) U.S. Nuclear Regulatory Commission, “Policy and Technical 

Issues Associated with the Regulatory Treatment of Non-safety 

Systems (RTNSS) in Passive Plant Designs (SECY-94-084),” 

SECY-95-132, May 22, 1995, Agencywide Documents Access and 

Management System (ADAMS) Accession No. ML003708005.
A Proposed Risk Management Regulatory Framework

H-26 | Risk Management Task Force Appendix H

(NRC, 2004) U.S. Nuclear Regulatory Commission, “Regulatory Analysis 

Guidelines for the U.S. Nuclear Regulatory Commission,” 

NUREG/BR-0058, Revision 4, August 2004, ADAMS Accession 

No. ML042820192.

(NRC, 2008) U.S. Nuclear Regulatory Commission, “Policy Statement on the

Regulation of Advanced Reactors,” Final Policy Statement,

October 7, 2008, ADAMS Accession No. ML082750370.

(NRC, 2009) U.S. Nuclear Regulatory Commission, “Risk-Informed,

Performance-Based Fire Protection for Existing Light-Water 

Nuclear Power Plants,” Regulatory Guide 1.205, Revision 1, 

December 2009, ADAMS Accession No ML092730314.

(NRC, 2010) U.S. Nuclear Regulatory Commission, “Modifying the Riskinformed Regulatory Guidance for New Reactors,” SECY-10-0121 

September 14, 2010, ADAMS Accession No ML102230076.

(NRC, 2011) U.S. Nuclear Regulatory Commission, “Recommendations for 

Enhancing Reactor Safety in the 21st Century: The Near-Term 

Task Force Review of Insights from the Fukushima Dai-Ichi 

Accident,” July 2011, ADAMS Accession No. ML112510271.

(NRC, 2011a) U.S. Nuclear Regulatory Commission, “An Approach for Using 

Probabilistic Risk Assessment in Risk-informed Decisions 

on Plant-specific Changes to the Licensing-basis,” 

Regulatory Guide 1.174, Revision 2, May 2011, ADAMS 

Accession No. ML100910006.

(WENRA, 2008) Western European Nuclear Regulators’ Association, “WENRA

Reactor Safety Reference Levels,” Reactor Harmonization 

Working Group, January 2008.
Appendix I Risk Management Task Force | I-1

A Proposed Risk Management Regulatory Framework

APPENDIX I

STAKEHOLDER INTERACTIONS

I.1 Background

The value of soliciting insights from the U.S. Nuclear Regulatory Commission (NRC) staff 

and external stakeholders regarding the Risk Management Task Force (RMTF) activities 

was recognized from its formation. The RMTF charter included several questions that were 

considered in the development of surveys for internal and external stakeholders. The internal 

surveys were prepared and distributed to a cross-section of NRC staff and managers in 

several key program offices. A summary of the responses to the internal surveys is provided 

in Section I.2. The request for public input to the process took the form of a notice in the 

Federal Register (published on November 22, 2011 (76 FR 72220)). A summary of the 

responses to the Federal Register notice is provided in Section I.3. In developing its findings 

and recommendations, the RMTF also benefited greatly from numerous informal discussions 

with NRC staff and managers, and external stakeholders. RMTF members accompanied 

Commissioner George Apostolakis on visits to all of the NRC regional offices. These 

interactions with the regional staffs were especially useful because they provided the RMTF with 

an opportunity to hear firsthand the experiences and issues associated with the use of riskinformed and performance-based approaches in the NRC’s oversight process. 

I.2 Internal Stakeholders

An internal survey was developed and distributed to a cross-section of NRC staff and managers 

in several key program offices that deal with the regulation and oversight of byproduct, source, 

and special nuclear materials. A summary of the survey responses is provided below:

1. Do you believe there is a common understanding and usage of risk-informed, 

performance-based approaches, and defense in depth within the NRC, industry, 

and other stakeholders? Which concepts are especially unclear?

Reactor respondents noted that the five elements of a “risk-informed approach” are defined in 

Regulatory Guide (RG) 1.174, “An Approach for Using Probabilistic Risk Assessment in Riskinformed Decisions on Plant-specific Changes to the Licensing-basis,” and the four guidelines 

for establishing a “performance-based approach” are in SRM/SECY-98-144, “White Paper on 

Risk-informed and Performance-based Regulation (Revised).” 

While there was significant disagreement as to which terms are the most or the least 

understood, the majority of respondents─in both the reactor and non-reactor program areas─ 

believed that none of the terms are well understood. Several of those who believed there was 

a general conceptual understanding still stated that they did not believe this carried over to 

the implementation process. Two respondents suggested using the maintenance rule as an 

example of a successful implementation framework for how the concepts can work together. 

Many respondents noted the common misunderstanding between risk-informed and risk-based. 

One respondent claimed NRC staff and members of the public are generally comfortable using 

risk insights to tighten standards, while the industry is interested in using risk assessments 

to weaken standards, rather than applying risk in both directions. Another noted that the 

implementation of defense-in-depth and safety margins are within divisions with staff trained in 
A Proposed Risk Management Regulatory Framework

I-2 | Risk Management Task Force Appendix I

deterministic analyses and not probabilistic risk approaches (PRAs), so they do not have the 

training to appropriately consider risk.

Some of the conceptual difficulties with defense-in-depth included the definition of a barrier: 

How effective does it need to be to properly apply defense-in-depth? The relationship between 

defense-in-depth and safety margins is also not well-defined. Additionally, one respondent 

noted the danger of using defense-in-depth to apply any requirement that the NRC wants 

without a strong basis, undermining the risk-informed approach. Respondents from NRC 

materials areas noted that the concept of defense-in-depth, in particular, was not one that was 

understood or applied within that program. 

Finally, some respondents found a tendency to generalize the term performance-based to any 

regulations that are not prescriptive. In using performance-based measures, it is important 

that the measurement criteria are directly related to safety and security. There is concern that 

certain performance-based requirements are difficult to enforce.

2. Are the agency’s current practices, in terms of policies, procedures, and 

training, adequate for accomplishing the goal of a more holistic, risk-informed, 

performance-based regulatory structure?

The majority of reactor respondents believed that the current practices were insufficient. Most 

agreed that the current policies were effective for their areas of applicability, but there was 

concern that this did not go much beyond the use of quantitative risk estimates, as defined in 

RG 1.174, to address uncertainties, model quality, the interplay of risk and defense-in-depth, 

or a more general risk management framework. There was not as clear a consensus from 

respondents in the agency’s non-reactor programs. In particular, these respondents noted that 

that one of the challenges the agency faces is recognizing that one approach to risk does not fit 

the wide diversity of the NRC’s programs (e.g., reactors, materials, waste, and fuel cycle).

Several thought that the policies were sufficient but that procedures were not as well defined. 

One respondent noted that National Fire Protection Association 805 seemed to run into 

problems because it attempted to be more risk-based than risk-informed. 

Most agreed that additional training for the staff would be useful. While current training courses 

have been established, they have been sparsely attended. Additionally, the focus has been 

primarily on PRA rather than on risk management more broadly.

One respondent focused on the holistic aspect of the RMTF mission. The respondent 

discussed the need for flexibility to account for the differences in licensing decisions. The 

response continued by noting that some of the differences in the application of these concepts 

could also come from genuine differences in core beliefs about the industry’s viability and the 

NRC’s level of responsibility in improving safety rather than just maintaining it.

3. How effective have past and ongoing risk-informed initiatives been? What are the 

relevant lessons learned from these initiatives? For example:

a. Regulatory Guide 1.174 

b. Risk-informed technical specifications 
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c. Risk-informed In-Service inspection 

d. Risk-informed In-Service testing 

e. Title 10 of the Code of Federal Regulations (10 CFR) 50.69, “Risk-informed 

Categorization and Treatment of Structures, Systems and Components for 

Nuclear Power Reactors” 

f. NUREG-1556 licensing guidance

g. NUREG-1520, Fuel Cycle Standard Review Plan

h. Materials inspection frequencies

i. Enforcement policy

j. Rulemaking activities

k. Other

The general consensus seems to be that the first four were generally successful, but 

10 CFR 50.69 was not. Additional successful initiatives offered by the respondents included 

Generic Issues, station blackout rule, individual plant examinations and individual plant 

examinations of external events, combustible gas control requirements, 10 CFR 50.65(a)(4), 

reactor oversight process (ROP), and risk-informed performance-based (RIPB) pressurized 

thermal shock rule.

There were several attempts at describing why 10 CFR 50.69 failed, but there seemed to 

be general agreement that the resulting rule did not provide any incentive for the industry to 

adopt it. Furthermore, it may have been the large relaxation of requirements suggested by risk 

insights that led the staff to compensate by making the final approved structure too burdensome 

for voluntary implementation. One respondent also noted that the process was poorly defined, 

requiring a continual changing of requirements to satisfy concerns of the Commission, Advisory 

Committee on Reactor Safeguards, and the industry.

Several responders noted that the implementation of risk-informed approaches can be predicted 

based on whether or not they will lessen the burdens for the industry. If the industry does not 

see a cost benefit in a voluntary approach, it will not adopt it.

Even some of the successful approaches required a long time through a poorly defined process 

before they were implemented.

The NUREG-1556 series was cited as a particularly helpful series in improving the efficiency of 

the materials licensing process and greatly improved understanding of that process, but some 

noted that the documents were outdated and in need of revision. Risk-informing of materials 

inspection frequencies were considered to be a means to save resources and not really an effort 

at applying risk to the inspection program. More frequent inspections of higher risk activities 

such as industrial radiography and well-logging are needed. One respondent noted that riskinforming has been most effective and lasting when applied to rulemaking and 

guidance development. 
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I-4 | Risk Management Task Force Appendix I

Finally, two of the responders discussed the idea that just because an approach is used does 

not necessarily mean that it has been successful. One suggested that research needs to be 

conducted to look at safety outcomes of the risk-informed approaches that have 

been attempted.

4. How effective have past and ongoing performance-based initiatives been? What 

are the relevant lessons learned from these initiatives?

There were few detailed responses to this question, and the answers were split. Those who 

thought performance-based initiatives have been successful pointed to the requirements 

for looking at uncertainty, and the NRC’s flexibility in enforcement to allow better focus 

on maintaining safety. Also mentioned were the particular examples of maintenance and 

surveillance, which can focus resources on important concerns while also providing incentives 

to the industry through the potential for a reduced burden. Some noted that performancebased inspections in the materials area have been successful and focus on the right things, but 

expressed a concern over the frequency of inspections, especially in light of the 10 year license 

renewal cycle. 

The negative responses focused on the difficulty and time burden in defining appropriate 

criteria, which can make the respective initiatives inefficient and can have the potential that the 

industry can satisfy poorly defined criteria without addressing real safety concerns.

5. How effective have past and ongoing risk-informed and performance-based 

initiatives been? What are the relevant lessons learned from these initiatives?

There was general agreement among the respondents that the maintenance rule and the ROP 

have been successful, but several suggested that the ROP is currently overly risk-based and 

should move to a more risk-informed approach. One respondent credited the ROP’s success, 

in part, to the flexibility in its implementation. Many found it too early to judge the fire-protection 

initiatives. A couple found the early work promising, but others thought the early process 

has been too slow and overly contentious. Part 36 (pool irradiators) and the two-person 

rule were highlighted as rule changes that appropriately and effectively addressed safety. 

One respondent noted that revisions to Part 35 were risk-informed, but that the training and 

experience requirements for authorized users were difficult to understand and difficult 

to implement. 

One respondent conducted a survey in 2002 of the nuclear industry that found that the 

maintenance rule and risk-informed technical specifications had been the best received, and the 

respondent thought that these results would still apply today. (see Appendix E)

6. How have other major initiatives or program changes incorporated risk-informed, 

performance-based approaches? What are the relevant lessons learned from 

these activities in terms of how they could have incorporated RIPB approaches or 

the limitations of RIPB approaches

One respondent discussed the response to September 11, concluding that in the end, it was 

primarily based on the subjective opinion of NRC security experts on the assumption that 

PRA could not be applied to security. However, this person felt the B.5.b1 implementation 

1 B.5.b refers to an item within security-related Orders issued by the NRC following the events of 

September 11, 2001. This item was subsequently codified in Section 50.54(hh) of 10 CFR Part 50.
Appendix I Risk Management Task Force | I-5

A Proposed Risk Management Regulatory Framework

represented severe mitigation actions that would be useful for a range of beyond-design-basis 

events and was an impressive early example of risk-informed decisionmaking. A more detailed 

discussion of frequencies and consequences associated with security would be classified. Nonreactor respondents on the issue of security requirements were split, with some considering that 

risk was appropriately considered and others noting that the requirements have gone too far. 

The GSI-1912 was regarded by most respondents as a failing to properly use risk insights. They 

noted that there was no motivation for the industry to use it, and the uncertainties were so large 

that the regulations would need to be added to the full PRA structure for them to be defensible.

Most responses also thought the SAMAs were of limited use because only agerelated cost 

beneficial improvements could be requested.

Several general lessons learned were included by individual respondents. A selection of these 

is provided below:

•	 Ensure that defense-in-depth is not weakened when implementing a risk-informed 

performance-based approach.

•	 Because of the wide variety of regulated uses, there have been a variety of riskinformed, performance-based approaches used. What is needed is a more top-down 

effort that aligns RIPB through a broad agency wide policy statement. 

•	 General license program is one example of regulation which is not RIPB. The minimal 

requirements of the program combined with little or no inspection oversight creates a 

regulatory program that does not work well.

•	 Use care when risk assessments point to excessive margin because the margin may 

also be necessary to account for uncertainty elsewhere in the system.

•	 Build on risk-informed programs on previous work avoiding the revision or elimination of 

the current implementation.

•	 Think of what the NRC response would be if the regulatory chronology were reversed: 

If the current risk-informed suggested policies were in place for decades and 

somebody suggested increasing to the design basis, would it still make sense to add 

that conservatism? 

7. If you had to choose just one area to focus on for risk-informing your programs, 

what would it be and why? If none, why not?

The respondents had a variety of answers, depending on their program areas. There were 

multiple calls for looking at beyond-design-basis and severe accidents and for making greater 

use of more advanced PRA techniques. One response focused on the potential benefits of 

agerelated passive system degradation. Non-reactor respondents identified several areas, 

including the licensing process generally, the waste classification system in Parrt 61, and 

materials security requirements. 

2 GSI-191 is Generic Safety Issue 191, “Assessment of Debris Accumulation on PWR [pressurized water 

reactor]. Sumps.” The issue has been the subject of numerous generic communications, meetings, and 

reports. A report , NUREG-0933, “Resolution of Generic Safety Issues,” is periodically updated to provide a 

status and closure plans for generic safety issues.
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I-6 | Risk Management Task Force Appendix I

A more general response suggested that the focus should be on better understanding, as well 

as use, and communication of risk results. Two responses suggested that currently there was 

too much attention being paid to risk insights and more focus should be on defense-in-depth.

8. If you had to choose just one area to focus on using a performance-based 

approach, what would it be and why? If none, why not?

The few responses to this question focused on inspection procedures and qualification 

requirements for several applications. Positive responses considered the potential for fewer 

NRC resources in the future, and the ability of performance-based approaches to better focus 

those resources on public health and safety.

Two responses mentioned the concern that a performance-based approach that is not riskinformed must be carefully planned to avoid risk side effects. In using performance-based 

approaches for procedures with limited data, one response emphasized the importance of 

ensuring the data are sufficient to measure the relevant criteria; proxy data may be unsuitable 

for performance measurement. The general license program was again identified as an area 

for improvement given that many general licensees do not understand their responsibilities, 

the risks associated with the devices that they have and, as a result, material ends up being 

improperly disposed. 

9. If you had to choose just one area to focus on using a risk-informed and 

performance-based approach, what would it be and why? If none, why not?

Several responses looked at inspection, safety-grade components versus performance-based 

reliability, and the beyond-design-basis categories of events. One response also suggested 

the most important area to apply a risk-informed approach is security, especially any future 

rulemaking processes.

One response suggested rewriting the 1998 “white paper” on risk-informed efforts (versus riskbased) and performance-based approaches, with a focus on emphasizing the maintenance rule, 

addressing tactical aspects of implementing the policy statement on PRA, and establishing the 

framework as a new policy statement.

Two respondents noted that the fuel cycle oversight program needs to be more risk-informed 

and performance-based, especially given that the ISA requirements in Subpart H to Part 70 

already reflect an RIPB approach.. Another noted that the regulations overall need to be 

examined based on their contribution to safety, starting with those items that have the greatest 

overall contribution to risk. 

The respondent working with research and test reactors thought the application of risk-informed, 

performance-based approaches could be useful for licensing and license renewal of research 

and test reactors (RTRs) if it was found that the initial effort was justified.

The respondent working in emergency preparedness mentioned that an attempt to apply a riskinformed approach was already underway, and this person looks to further include risk insights 

as the program moves forward.
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A Proposed Risk Management Regulatory Framework

10. How would you go about transitioning from the current approach to the riskinformed, performance-based approach in the area you have chosen?

For the proposed projects, responses included ensuring high-level buy-in from the Commission 

through the managerial ranks, public meetings to discuss the new approach, and allowing 

fewer exceptions to the current rules (notices of enforcement discretion and emergency or 

exigent technical specification changes). There were conflicting answers about the need 

for industry collaboration. One response mentioned that RIPB approaches to safety-related 

equipment should only be done on a voluntary basis, while another said that, generally, 

voluntary approaches would not work. Also, some suggested low-level changes to start, while 

one suggested rewriting the entire framework, starting with the safety goal. Another response 

mentioned that the most realistic place to begin is with the implementation of the Fukushima 

Near-Term Task Force recommendations. One respondent recommended complementary 

“forward-looking” and “backward-looking” reviews of regulations. The former would look at 

unaccepatable event or outcomes and then identify those portions of the regulations that 

assurre those events or outcomes do not take place. The latter would involve developing a list 

of key regulatory requirements and the consequences that these requirements are designed 

to prevent. The information from these evaluations would provide a basis for determining the 

adequacy of regulatory requirements from a RIPB perspective. 

11. Taking all things into consideration (e.g., additional modeling development, data 

collection, skill sets, legal considerations, and organization factors), what needs 

to be in place for No. 10 to occur, and how long do you think it would take?

Several responses focused on the need for Commission-level buy-in, and the need for 

convincing the public and the industry of the need for a change. Other suggestions included the 

integration of additional risk staff into deterministic divisions, more training on the interpretation 

of PRA and risk in general, and allowing fewer industry exceptions to risk-informed rules. 

The time estimates were in the 5-year to 10-year range. Some suggested the approach is 

evolutionary and should continue for the foreseeable future.

12. How should the key concepts in the deterministic regulatory framework (e.g., 

design-basis accidents, dose limits, safety-related and defense in depth) be 

integrated with the key features of a risk-informed, performance-based approach?

There were several categories of responses. The first pointed out that RG 1.174 already 

discusses how to integrate probabilistic and deterministic approaches. One person noted that 

in developing alternative rule making, the guideline to comply with existing regulations may 

not make sense. The response suggested that the concept, however, should be preserved by 

looking at the motivation for relevant existing regulations and using risk insights to meet the 

intention of the regulation.

The second category discussed the use of defense-in-depth as a way to account for 

uncertainties in the model. One response suggested using risk-informed approaches to outline 

the scope, which would allow the use of defense-in-depth and other deterministic approaches to 

be used in a more rational and focused manner.

Materials respondents noted that deterministic requirements are attractive to many licensees 

and NRC staff because of their simplicity and clarity. One stated that dose limits, for example, 
A Proposed Risk Management Regulatory Framework

I-8 | Risk Management Task Force Appendix I

are by their very nature RIPB: they provide a limit which licensees have flexibility on the means 

to achieve that limit. 

One response suggested that risk-informed approaches should lead to the optimization of 

risk reduction given fixed resources rather than allowing only the loosening of specific rules. 

Any rule relaxation would lead to some increase in risk and would be difficult to justify, this 

respondent said.

13. Is it clear in what situations and circumstances the NRC should be receptive to, or 

on its own introduce, risk-informed approaches, performance-based approaches, 

or risk-informed and performance-based approaches?

The respondents pointed out that the NRC currently can do very little without industry buy-in, 

and it would be difficult to pursue a new approach without the backing of stakeholders. One 

suggested it might be possible for the NRC to introduce approaches on its own if they could 

be shown to lead to increased public safety and security, but this would be difficult to put into 

practice. Several respondents suggested the NRC should always be open to new approaches 

by the industry (one noted that there may be some exceptions), and when they prove to be cost 

beneficial and technically feasible, these approaches should be pursued.

14. Which external stakeholders do you consider critical for the Task Force to consult 

in its work? Suggestions on how best to engage them?

All but two of the reactor responses suggested talking to industry leaders early in the process. 

One suggested talking to specific industry experts, but not through the Nuclear Energy Institute 

(NEI); the other thought that for emergency preparedness, interagency partners and relevant 

contractors (offsite response organizations) should be contacted first. One commenter 

suggested that stakeholders need to know they have something to gain or lose in the project to 

become fully engaged. Non-reactor respondents considered NEI and NRC licensees as critical 

groups to be engaged. 

Most of the non-reactor respondents identified the need to involve the Agreement States, 

given their role in regulating 85% of the materials licensees in the US as well as all of the 

LLW disposal sites. Specifically, the Organization of Agreement States and the Conference of 

Radiation Control Program Directors were cited as key stakeholders.

Several comments mentioned the senior reactor analyst and general PRA and probabilistic 

safety analysis community. One response suggested engaging prescreened academicians 

to engage in a workshop to identify implementation ideas. Most suggested the public at 

different stages of development, including critics. Some also suggested like-minded agencies 

(International Atomic Energy Agency, U.S. National Aeronautics and Space Administration, 

U.S. Department of Energy, U.S. Environmental Protection Agency, U.S. Department of 

Homeland Security), national labs, and professional standards groups (e.g., American Society of 

Mechanical Engineers, American Nuclear Society).
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A Proposed Risk Management Regulatory Framework

15. What do you feel are the two or three key obstacles to making further progress 

on moving toward more risk-informed or performance-based regulatory 

approaches? Are the current capabilities and limitations of probabilistic risk 

assessment, as used in your area, one of the major obstacles? (Also see 

Question 11.)

Many responses focused on internal and external communications. Some suggested a more 

broad-based risk education of staff, especially leadership and management. This training 

would move beyond the use of PRA to the use of risk information to make decisions in all 

NRC regulatory programs. Additionally, one response suggested comparing the limitations of 

probabilistic approaches to those of deterministic analyses. The response suggests that when 

the focus is only on the limitations of probabilistic approaches, there is an assumption that 

the deterministic approaches have fewer limitations. Some respondents identified the lack of 

risk information and appropriate tools and the cost of obtaining such information as obstacles. 

Others identified inertia in moving away from regulatory approaches that are known and 

predictable, while another identified the Agreement States as a challenge.

Many also discussed the limitations of PRAs, both at the NRC (addressing all internal and 

external initiators, all modes of operation, human factors, etc.) and within the industry. Also, one 

suggested the safety goals should better reflect public risk tolerance (e.g., Fukushima may not 

have violated the safety goals but would not be tolerated by the public).

I.3 External Comments

The RMTF published a request for public comment regarding an adoption of risk management 

concepts in the Federal Register (published on November 22, 2011 (76 FR 72220)). The notice 

asked for stakeholder input on the following questions:

1. Do you believe there is a common understanding and usage of the terms risk-informed, 

performance-based, and defense-in-depth within the NRC, industry, and other 

stakeholders? Which terms are especially unclear?

2. What are the relevant lessons learned from the previous successful and unsuccessful 

risk-informed and performance-based initiatives?

3. What are the relevant lessons learned from the previous successful and unsuccessful 

deterministic regulatory actions?

4. What are the key characteristics for a holistic risk management regulatory structure for 

reactors, materials, waste, fuel cycle, and security?

5. Should the traditional deterministic approaches be integrated into a risk management 

regulatory structure? If so, how?

6. What are the challenges in accomplishing the goal of a holistic risk management 

regulatory structure? How could these challenges be overcome?

7. What is a reasonable time period for a transition to a risk management regulatory 

structure?
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I-10 | Risk Management Task Force Appendix I

8. From your perspective, what particular areas or issues might benefit the most by 

transitioning to a risk management regulatory approach?

The responses to the Federal Register notice are available in the NRC’s Agencywide 

Documents Access and Management System (ADAMS) (Case/Reference No. NRC20110269). 

A summary of the responses is provided below:

Table I-1 Summary of Comments from External Stakeholders

Organization

(ADAMS 

Accession No.)

Relevant 

Report

Section(s)

Comments

Energy 

Solutions

(ML12023A038)

Decommissioning, 

waste disposal, 

transportation

NRC should hold a series of workshops for all program 

areas

Defense-in-depth (DID) well defined; not Risk-Informed 

(RI) or Performance-based (PB)

MARSSIM [Multi-agency radiation survey and site 

investigation manual] is an example of a good RIPB 

regulation

Bad deterministic example – Regulatory Guide (RG) 1.86 

uses standards based on 1974 technology limits

Avoid a one-size-fits-all approach

Difficulties include available resources, criteria for 

success, consistency in interpretation and application, 

data availability and sufficiency, and qualitative vs. 

quantitative standards

This should be implemented piecemeal after workshops 

with stakeholders

10 CFR 61 would benefit from risk management regulatory 

approach (some aspects don’t make sense at certain 

facilities)
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A Proposed Risk Management Regulatory Framework

Council on 

Radionuclides 

and 

Radiopharmaceuticals

(CORAR)

(ML12018A230)

Materials

Suggest workshops and training in risk management 

before implementation.

Some regulations that claim to be risk-informed put a 

burden on all licensees when only some are of concern

Risk information could help incentivize self-identification of 

deficiencies

The framework should only apply to those considerations 

that affect all licensees

Risk management framework could help to exempt 

licensees from regulations that are not appropriate for 

them

Organization 

of Agreement 

States (OAS)

(ML12013A123)

Materials

Performance-based is well understood in terms of 

inspections, but not sure about licensing. Risk-informed 

and DID do not seem to make sense for materials as 

outlined in the FRN

Can’t answer Qs 3-8 because risk management concepts 

are not used as outlined in FRN

Could use a performance-based approach to reduce 

frequencies of inspections for good licensees

Society for 

Nuclear 

Medicine (SNM)

(ML12079A135)

Materials

Current definitions and examples of RI, PB, and DID are 

for reactors, need examples for nuclear medicine

Dose-based release criteria have been very effective in 

nuclear medicine community

Holistic risk management approach would require rewrite 

of 10 CFR 35.75 and would take 5 years

SNM believes a RIPB approach would be extremely useful 

for nuclear medicine.
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I-12 | Risk Management Task Force Appendix I

NEI 

Attachment 2

(ML12027A141) Materials and 

Fuel Cycle

Each sector developed its own meaning of RI, PB, and 

DID that is acceptable to NRC. Believe NRC understands 

different risks and should build upon current flexible 

framework

Decision to stop Cesium-137/Chloride rule was based on 

RIPB insights; several others

Many requests for additional information don’t seem to 

consider risk

Need to have engagement and workshops like for Safety 

Culture Policy Statement

Could use risk information to determine licensing fees 

(Level 1 facility vs. power reactor)

Fuel cycle facilities could potentially benefit from RIPB 

approach depending on implementation

Next 

Generation 

Nuclear Plant 

(NGNP)

(ML12010A098)

Power Reactors

Need to better define DID; Idaho National Laboratory (INL) 

has suggestions 

INL has reported on how risk concepts affect NGNP in 

several white papers referenced

Need a blend of deterministic and probabilistic

Defining licensing-basis events (LBEs) are a good place to 

start in looking at risk-informed approach to NGNP

The white papers could provide a basis to develop the 

entire process.
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Pressurized 

Water Reactor 

(PWR) Owners 

Group

(ML12032A029)

Power Reactors

NRC needs to better define DID

There is a current overreliance on numbers coming from 

PRA

Analyses are too risk-based (especially Significance 

Determination Process (SDP)), rather than risk-informed

Sometimes conservatisms in model are not considered 

Maintenance rule and Mitigating System Performance 

Indication (MSPI) work well

Generally, NRC too focused on refining methods and not 

on context of decisions 

Important to have early buy-in, pilot cases, and a common 

vision of regulatory shift

Risk-informed approach needs to percolate through staff 

level and would be helpful in GSI-191 and Fukushima 

lessons learned

PWROG 

– WCAP 

Attachment

(ML12032A029)

Power Reactors An attached Risk-Informed safety analysis developed in 

2003

NEI 

Attachment 1

(ML12027A141) Power Reactors

RI is not well defined, too much of a focus on numbers. 

DID is sometimes used to avoid RI applications

Maintenance rule has been successful PB approach

ATWS and station blackout rule were based on PRA 

insights and an example of a positive use of risk 

information

NFPA 805 and MSPI have been challenging to implement; 

General structure of MSPI has been good, but too much 

focus on making it a “true” PRA

The SDP can lead plants to spend resources refining 

assumptions in PRA rather than fix the problems

Decades to fully implement risk management regulatory 

structure. Can use risk insights immediately

GSI-191 would be a good place to use an RI approach
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I-14 | Risk Management Task Force Appendix I

National Fire 

Protection 

Association 

(NFPA)

(ML12010A026)

Power Reactors

Agree with the need for RIPB. NFPA 805 is a good 

example of PB. NFPA 806 will become an important part 

of oversight. NFPA 804 is prescriptive but should be part 

of NRC’s oversight and regulatory process

Farshid 

Shahrokhi

(ML12013A125)

Overall

Should establish high-level goals for all sectors; these 

would then naturally lead to individual sector criteria

Need a collaborative environment with open-ended ways 

to meet safety criteria

Precedent has played a role in past decisions, but this 

does not necessarily lead to better safety regulations

NEI Main

(ML12027A141) Overall

Current risk-informed approaches should be encouraged 

and sustained

Don’t layer PRA on top of deterministic regulations

Currently, focus too much on numerical thresholds

Prasad 

Kadambi

(ML12005A212)

Power Reactors

Worried about focusing too much on numbers coming out 

of a PRA

Likes reactor oversight process (ROP) and seismic 

margins analysis

Believes we should implement risk management in a 

looser, case-by-case decision analysis context

NRC needs to be more explicit in performance-based 

approaches 

Put nuclear consequences in context of total 

consequences (Fukushima: reactors vs. earthquake, 

tsunami)
Appendix I Risk Management Task Force | I-15

A Proposed Risk Management Regulatory Framework

STP

(ML12079A134) Power Reactors

The RIPB and DID definitions are fuzzy; need criteria, 

details for implementation and should apply beyond 

reactors to materials, security, etc.

Need to add oversight and documented expectations 

for RIPB approaches. The Risk Managed Technical 

Specifications is an example of successful risk-informed 

initiative

Deterministic regulations lead to resource misallocation. 

Risk should be in regulatory policy statements (work 

hour rule, cybersecurity) and risk methods to determine 

significance criteria for regulatory response

Successful implementation requires Commission buy-in 

and pilot tests; full implementation $5 million to $10 million/

year and 50 FTEs

This approach would lead to improved safety, better 

operational flexibility and decisionmaking; improved 

regulatory decisionmaking; better societal quality of life; 

better worker safety and quality of life

Organization of 

Test, Research 

and Training 

Reactors 

(TRTR)

(ML12013A122)

Nonpower 

Reactors

The ratio of regulation to risk from NPRs is higher than for 

the other regulatory licensees

NPRs should not be required to do PRAs

Biggest issue to implementation is public risk perception

Switching to a risk management framework may benefit 

RTRs by lessening their regulatory burden

Penn State

(ML12010A027)

Nonpower 

Reactors

Prefer less regulation and more consultation 

NRC seems to be less focused on risk to public than on 

political considerations and other difficult-to-quantify risks

Would like to move to a more performance-based 

approach
A Proposed Risk Management Regulatory Framework

I-16 | Risk Management Task Force Appendix I

Kennecott 

Uranium 

Company

(ML12013A154)

Uranium Recovery

Supports RIPB because of low risk in handling radioactive 

materials

RI, PB, DID are well understood

Rad risks should be ranked against other risks at reactors, 

materials, waste, fuel cycle, and security sites

Main difficulty is perceived vs. real risk, especially with 

the public

Benefits of RIPB include review ground water restoration 

and remediation, soil remediation, standards for 

Radon222, etc.

Many favorable supporting documents

National Mining 

Association 

(NMA)

(ML12019A111)

Uranium Recovery

RIPB approach promotes efficient use of limited resources

RPB approach should apply to licensing actions, policy 

development, and inspection/enforcement actions

The Regulatory Issue Statement is a good example of a 

RIPB success

Currently too much of a focus on numerical (baseline) 

recovery when in certain sites that doesn’t make sense

Wyoming 

Mining 

Association

(ML12013A124)

Uranium Recovery Similar to Kennecott comments...
Appendix J Risk Management Task Force | J-1

A Proposed Risk Management Regulatory Framework

APPENDIX J

 FINDINGS AND RECOMMENDATIONS

The various sections of the report provide the key RMTF findings and recommendations related 

to proposed Risk Management Regulatory Framework and its application to specific NRC 

program areas. A summary of the RMTF findings and recommendations is provided in 

Table J-1.

Table J-1 Risk Management Task Force Findings and Recommendations

Chapter 2 – A Proposed Risk Management Regulatory Framework

Finding 2.1

Whether used explicitly, as for power reactors, or implicitly, as for 

materials programs, the concept of defense-in-depth has served 

the NRC and the regulated industries well and continues to be 

valuable today. However, it is not used consistently, and there is 

no guidance on how much defense-in-depth is sufficient.

Finding 2.2

Risk assessments provide valuable and realistic insights 

into potential exposure scenarios. In combination with other 

technical analyses, risk assessments can inform decisions about 

appropriate defense-in-depth measures.

Recommendation 2.1

The goal to adopt risk-informed and performance-based 

approaches, where practical, should continue and should be 

incorporated into the revised regulatory framework.

Recommendation 2.2

The general regulatory approach of the NRC should be defined 

in terms of “managing the risks” posed to workers and the 

public from the various uses of byproduct, source, and special 

nuclear materials.

Recommendation 2.3

In defining requirements for the protection of workers and the 

public, the NRC should recognize and address uncertainties 

associated with the hazards and the events, including human 

errors, which could challenge or degrade barriers and controls. 

A balanced approach that considers traditional and risk 

assessment techniques should be used to identify barriers and 

controls so that appropriate requirements are defined to prevent, 

contain, and mitigate exposures to radioactive materials.

Recommendation 2.4 The NRC should formally adopt the proposed Risk Management 

Regulatory Framework through a Commission Policy Statement.
A Proposed Risk Management Regulatory Framework

J-2 | Risk Management Task Force Appendix J

Chapter 4 – Implementation Options

Section 4.2.1 Nuclear Power Reactors

Finding PR-F-1

The concept of design-basis events and accidents continues 

to be a sound licensing approach, but the set of design-basis 

events and accidents has not been updated to reflect insights 

from power reactor operating history and more modern 

methods, such as PRA.

Recommendation PR-R-1

The set of design-basis events and accidents should be 

reviewed and revised, as appropriate, to integrate insights from 

power reactor operating history and more modern methods, 

such as PRA.

Finding PR-F-2:

Requirements for beyond-design-basis accident scenarios 

(e.g., station blackout) were established at different times and 

in different ways. Differences in implementation approaches 

have reduced the efficiency and consistency of the NRC’s 

regulatory and oversight activities.

Finding PR-F-3:

The extent to which licensee activities undertaken as part 

of voluntary industry initiatives can be credited has been a 

source of contention in the Reactor Oversight Process and has 

reduced the efficiency of that process.

Recommendation PR-R-2

The NRC should establish through rulemaking a designenhancement category of regulatory treatment for beyonddesign-basis accidents. This category should use risk as a 

safety measure, be performance-based (including the provision 

for periodic updates), include consideration of costs, and be 

implemented on a site-specific basis.

Finding PR-F-4

The processes for establishing the external hazard design 

bases do not use consistent event frequency and 

magnitude methods.

Finding PR-F-5

New information that would provide the basis for external 

hazard frequency updates is not systematically collected, 

evaluated, and communicated.

Finding PR-F-6

PRA methods for assessing external hazard risks are available, 

but expertise in performing such studies is very limited. 

Uncertainty analyses and the recognition of the limitations of 

available scientific knowledge are a key element of 

these methods.
Appendix J Risk Management Task Force | J-3

A Proposed Risk Management Regulatory Framework

Recommendation PR-R-3

The NRC should reassess methods used to estimate the 

frequency and magnitude of external hazards and implement 

a consistent process that includes both deterministic and PRA 

methods. Consideration of the risks from beyond-designbasis external hazards should be included in the designenhancement category described in Recommendation PR-R-2.

Recommendation PR-R-4 The NRC should establish a program to systematically collect, 

evaluate, and communicate external hazard information. 

Finding PR-F-7

The availability and broad-scale use of quantitative risk 

assessment methods (PRA) for power reactors provide an 

opportunity for a more quantitative characterization of defensein-depth.

Recommendation PR-R-5

The NRC should apply the risk-informed and performancebased defense-in-depth concept to power reactors in a more 

quantitative manner.

Finding PR-F-8

Vulnerability assessments performed to assess security have 

important similarities in scope (e.g., facility equipment and 

radioactive hazards considered) and methods to 

risk assessments.

Finding PR-F-9

Differences in regulatory language and approaches between 

power reactor security and safety regulation may have reduced 

the efficiency and effectiveness of the NRC’s work.

Finding PR-F-10

In the past decade, considerable research has been performed 

on estimating security risks, much of it sponsored by the 

U.S. Department of Homeland Security.

Recommendation PR-R-6

The NRC should develop and implement guidance for use in its 

security regulatory activities that uses a common language with 

safety activities and harmonizes methods with risk assessment 

and the proposed risk-informed and performance-based 

defense-in-depth framework.

Section 4.2.1.1 - Operating Power Reactors

Recommendation OR-R-1

For operating reactors, the establishment of the design 

extension category can be followed by a review of design basis 

events/accidents and related revisions to AOOs and DBAs 

to integrate insights from operating history and more modern 

methods. The NRC need not impose such a requirement 

but should be amenable to related industry initiatives should 

they pursue revisions to design basis events based on the 

introduction of the design extension category. 
A Proposed Risk Management Regulatory Framework

J-4 | Risk Management Task Force Appendix J

Recommendation OR-R-2

For operating reactors, the RMTF recommends that the 

NRC should establish through rulemaking a designenhancement category of regulatory treatment for beyonddesign-basis accidents.

Recommendation OR-R-3

The NRC should reassess the methods used to estimate the 

frequency and magnitude of external hazards and implement 

a consistent process that includes both deterministic and PRA 

methods. For operating reactors, the RMTF recommends 

that the design-enhancement category rulemaking include 

consideration of external hazards.

Recommendation OR-R-4

For operating reactors, the RMTF recommends that the 

NRC develop and implement guidance for the collection and 

dissemination of external hazard information.

Recommendation OR-R-5

The NRC should apply the risk-informed and performancebased defense-in-depth concept to power reactors in a more 

quantitative manner. For operating reactors, the RMTF 

recommends that this recommendation be implemented in the 

form of guidance to the NRC staff and in future requirements 

established for operating reactor licensees.

Recommendation OR-R-6

For operating reactors, the RMTF recommends that guidance 

be developed and implemented to better harmonize 

terminology and methods used for reactor safety and security.

Section 4.2.1.2 - New Reactors

Recommendation NR-R-1

For new reactors, the RMTF recommends that the NRC be 

amenable to and promote, where practical, the adoption of 

more risk-informed and performance-based approaches for the 

selection of more relevant scenarios for design-basis events. 

Changes pursued for operating reactors (OR-R-1) should also 

consider applicability to new reactors.

Recommendation NR-R-2 Apply Recommendation PR-R-2 (design-enhancement 

category) to new reactors.

Recommendation NR-R-3 Apply Recommendation PR-R-3 (include external events in 

design-enhancement category) to new reactors.

Recommendation NR-R-4 Apply Recommendation PR-R-4 (periodically evaluate new 

information regarding external hazards) to new reactors.
Appendix J Risk Management Task Force | J-5

A Proposed Risk Management Regulatory Framework

Recommendation NR-R-5

Apply Recommendation PR-R-5 (issue guidance to adopt riskinformed and performance-based defense-in-depth) to new 

reactors.

Recommendation NR-R-6

Apply Recommendation PR-R-6 (develop guidance and 

consistent approach between safety and security) to 

new reactors.

Section 4.2.1.3 - Generation IV Reactors

Recommendation GIV-R-1

For Generation IV reactors, the RMTF recommends that the 

concept of design-basis accidents be maintained, but the NRC 

should be amenable to and promote, where practical, the 

adoption of more risk-informed approaches for the selection 

of relevant scenarios (e.g., alternatives to the single failure 

criterion) for design-basis accidents.

Recommendation GIV-R-2 Apply Recommendation PR-R-2 (design-enhancement 

category) to Generation IV reactors.

Recommendation GIV-R-3 Apply Recommendation PR-R-3 (include external events in 

design-enhancement category) to Generation IV reactors.

Recommendation GIV-R-4

Apply Recommendation PR-R-4 (periodically evaluate new 

information regarding external hazards) to Generation 

IV reactors.

Recommendation GIV-R-5

Apply Recommendation PR-R-5 (issue guidance to adopt 

risk-informed and performance-based defense-in-depth) to 

Generation IV reactors.

Recommendation GIV-R-6

Apply Recommendation PR-R-6 (develop guidance and 

consistent approach between safety and security) to 

Generation IV reactors.

Section 4.2.2 - Nonpower Reactors

Finding NPR-F-1

The concept of defense-in-depth for NPRs remains relevant. 

Knowledge gaps and uncertainties continue to be effectively 

addressed by the defense-in-depth approach.

Finding NPR-F-2

The analysis of design basis and the maximum hypothetical 

accidents based on conservative design limits, acceptance 

criteria, safety margins, and assumptions in conjunction with 

the application of a defense-in-depth philosophy continues 

to be a sound but highly conservative licensing approach to 

ensuring adequate safety of NPRs.
A Proposed Risk Management Regulatory Framework

J-6 | Risk Management Task Force Appendix J

Recommendation NPR-R-1

The proposed defense-in-depth framework should be applied 

to the NPR licensing process to ensure that the current 

amount of defense-in-depth is appropriate given the relatively 

small radioactive hazard. This application should include 

safety and security licensing matters.

Finding NPR-F-3

The application of modern risk assessment methods at 

NPRs could provide valuable insights into accident scenarios 

not previously identified by the earlier deterministic safety 

assessment and could be valuable in focusing the application 

of licensing and oversight resources on areas of risk 

importance. Risk assessment insights, in conjunction with 

a formal risk management decisionmaking process, could 

significantly contribute to the development of a more efficient 

and effective NPR regulatory framework. NPR PRA models 

developed by others could be used as a starting point for 

facility-specific PRA models at NRC-licensed NPRs. Even 

with this background, however, funding such assessments 

could be problematic for NPRs.

Recommendation NPR-R-2

The NRC should evaluate the utility of performing a pilot risk 

assessment, including consideration of external hazards, 

using modern risk assessment methods at an NPR. This 

evaluation would assess the value of the risk insights gained 

from the risk assessment on the basis of possible safety 

enhancements and possible contributions to a more efficient 

and effective risk-informed and performance-based regulatory 

framework for NPRs.

Finding NPR-F-4

The traditional NPR licensing approach shares the same 

limitations as the power reactor approach for methods used to 

estimate frequencies and magnitudes of external events.

Recommendation NPR-R-3

NPRs should be considered, to the extent practical, in the 

implementation of power reactor Recommendation PR-R-3 

and Recommendation PR-R-4.

Finding NPR-F-5 NPR security requirements are not risk-informed beyond the 

use of a graded approach based on NPR power levels.

Finding NPR-F-6

The development of risk-informed security regulations or 

guidance would enhance efficiency by focusing resources on 

identified areas important to facility security. 

Recommendation NPR-R-4

If the NRC decides to develop and implement a risk-informed 

and performance-based defense-in-depth regulatory 

framework to ensure the safety of NPRs, then the agency 

should also develop guidance for use in its NPR security 

regulatory activities that uses a common language with safety 

activities and harmonizes methods with the risk-informed and 

performance-based defense-in-depth framework.
Appendix J Risk Management Task Force | J-7

A Proposed Risk Management Regulatory Framework

Section 4.3 - Materials Uses

Finding-M-F-1

The materials program has successfully developed and 

incorporated risk insights and performance considerations into 

its rulemaking, policy development, and routine licensing and 

inspection activities.

Finding-M-F-2

Deterministic approaches such as dose limits and possession 

limits are useful concepts across the wide range of materials 

uses that should be retained, but broader use of deterministic 

approaches can lead to ineffective use of limited licensee, 

NRC, and Agreement State resources.

Finding-M-F-3

Buy-in of the 37 Agreement States is essential to the success 

of risk management process implementation given their role in 

regulating more than 85 percent of the materials licensees in 

the United States.

Recommendation-M-R-1

The NRC materials program should continue to apply 

risk insights and performance-based considerations, as 

appropriate, in rulemaking, guidance and policy development, 

and implementation in accordance with the proposed risk 

management framework.

Recommendation-M-R-2

The development and rollout of the recommended 

Risk Management Policy Statement should be closely 

coordinated with the leadership of the Agreement States 

and a joint NRCAgreement State Working Group should be 

established to guide risk management implementation in the 

materials area.

Finding M-F-4

Differences in regulatory language and approaches between 

safety and security in the materials area may have reduced 

the efficiency and effectiveness of the NRC and Agreement 

State regulatory programs.

Recommendation M-R-3

The NRC should apply common risk approaches to safety 

and security based on the proposed risk management and 

defense-in-depth regulatory framework. 

Finding M-F-5 The terminology of defense-in-depth is not used consistently 

across the agency’s materials regulatory programs.

Recommendation M-R-4

As part of the implementation of the proposed risk 

management regulatory framework, the RMTF recommends 

that the materials program should more explicitly consider the 

defense-in-depth philosophy in rulemaking, guidance, and 

program implementation, and modify appropriate parts of staff 

training to make these concepts a central part of 

such training.
A Proposed Risk Management Regulatory Framework

J-8 | Risk Management Task Force Appendix J

Finding M-F-6

The materials program could benefit from a more structured 

application of the risk management process in resource 

allocation. This process would allow program managers to 

more systematically apply resources to those areas where the 

safety or security risk warrants it.

Recommendation M-R-5

Headquarters and Region managers should undertake a more 

formal review of operational data and risk considerations to 

help determine if and where budgeted resources need to be 

adjusted. This option takes into account the fact that budgets 

are developed well in advance of budget implementation and 

that operational realities may change in the intervening time.

Section 4.4 – Waste

Section 4.4.1 - Low Level Waste

Finding LLW-F-1

The regulatory framework for LLW disposal is, for the most 

part, implicitly followed a risk-informed and performancebased approach. However, changes in the LLW environment 

and the maturing of the performance assessment method 

over the past 30 years have underscored the need to provide 

a stronger risk basis to the program. The staff is currently 

implementing Commission direction to make it more so and 

will be sending various products to the Commission over the 

next year or two. However, as noted in earlier sections of this 

report, the terms risk-informed and performance-based are 

not well understood and have not been consistently applied 

across NRC regulatory programs. 

Finding LLW-F-2

Certain aspects of the LLW regulatory framework readily lend 

themselves better to the risk management approach, such 

as waste classification or concentration averaging. Applying 

the proposed risk management approach to comprehensive 

LLW licensing decisions, however, may be more challenging 

because it involves estimating facility performance due to 

events potentially far into the future.

Recommendation LLW-R-1

The NRC should adopt the concept of risk management 

to the LLW program, including any revisions proposed 

to 10 CFR Part 61 (including performance assessment 

requirements) and related guidance documents.
Appendix J Risk Management Task Force | J-9

A Proposed Risk Management Regulatory Framework

Finding LLW-F-3

The interlocking and reinforcing systems approach in 

10 CFR Part 61 (site suitability, waste form and classification, 

intruder barrier, institutional controls, etc.) represents an 

implicit consideration of defense-in-depth features, based on 

the risk posed by various classes of waste. 

Finding LLW-F-4 The NRC has not developed an explicit characterization of 

defense-in-depth considerations for the LLW program. 

Recommendation LLW-R-2

The NRC should develop an explicit characterization of how 

defense-in-depth within the proposed risk management 

framework applies to the LLW program and build this into 

current and future staff guidance documents and into training 

and development activities for the staff.

Finding LLW-F-5 Consideration of environmental risks as well as safety risks is 

a central part of the LLW regulatory program.

Recommendation LLW-R-3 The NRC should include environmental reviews within the 

scope of its risk management framework.

Section 4.4.2.2 - High Level Waste

Finding-HLW-F-1

The development of regulations for geologic disposal of 

HLW at Yucca Mountain was based significantly on risk 

information developed from performance assessments and 

closely followed the proposed Risk Management Regulatory 

Framework.

Finding-HLW-F-2

The NRC’s regulatory philosophy of defense-in-depth is 

reflected in the multiple-barrier requirement for post-closure 

in 10 CFR Part 63. Compliance with the multiple barrier 

requirements is demonstrated through the performance 

assessment. 

Finding-HLW-F-3

As performance assessment capabilities and experience 

increased at the NRC during the past 30 years, so did the 

use of risk insights to help guide the HLW program. Risk 

insights and performance assessment capabilities have been 

used to improve the efficiency and effectiveness of guidance 

documents; inform pre-licensing interactions with DOE; and 

help identify and direct data needs and experimental activities.

RecommendationHLW-R-1

Any future revisions to the regulatory framework for geologic 

disposal of HLW should be done in accordance with the 

proposed risk management framework to ensure that risk 

information continues to be appropriately considered in the 

development of requirements and appropriately reflect any 

future HLW disposal paradigm. 
A Proposed Risk Management Regulatory Framework

J-10 | Risk Management Task Force Appendix J

Finding-HLW-F-4

The risk-informed, performance-based approach for regulating 

geologic disposal (YMRP and regulations) was intended 

to provide flexibility to the applicant in developing the best 

approach in meeting the regulatory requirements, and allow 

the use the risk information to focus regulatory review on 

safety-significant issues.

Section 4.5 – Uranium Recovery

Finding-UR-F-1

Uranium recovery facilities are of low radiological risk to 

workers and members of the public under normal operational 

conditions and most accident scenarios.

Finding-UR-F-2

Although the NRC staff has made inroads to risk-informed, 

performance-based licensing of uranium recovery facilities, 

the regulatory framework is largely a deterministic one. NRC 

regulations in 10 CFR Part 40 and Appendix A to 10 CFR 

Part 40 reflect the requirements of UMTRCA and EPA’s 

rules in 40 CFR Part 192, which pre-date the Commission’s 

move to a risk-informed, performance-based framework in 

the mid-1990s. Similarly, program guidance, especially for 

conventional mills, could benefit from a greater risk basis.

Finding-UR-F-3

The exact nature of the ISR rule under development by the 

EPA is not clear at this time; therefore, presents an uncertainty 

to adoption of the proposed risk management regulatory 

framework.

Recommendation-UR-R-1

Notwithstanding the current uncertainty associated with 

the EPA rulemaking, the NRC should adopt the proposed 

risk management regulatory framework to the uranium 

recovery program to provide greater efficiency, effectiveness, 

and predictability in policy development and regulatory 

decisionmaking.

Recommendation-UR-R-2

The NRC should work closely with the Agreement States 

and the regulated community to guide implementation of risk 

management in the uranium recovery program.

Finding-UR-F-4 Consideration of environmental risks is a central part of the 

uranium recovery regulatory program. 

Recommendation-UR-R-3 The NRC should include environmental reviews within the 

scope of its risk management framework.
Appendix J Risk Management Task Force | J-11

A Proposed Risk Management Regulatory Framework

Section 4.6 – Fuel Cycle

Finding F-F-1

The current fuel cycle regulatory approach incorporates 

several elements of the proposed risk management regulatory 

framework, such as the use of ISAs to identify safety 

significant items, and the implementation of a revised fuel 

cycle oversight program as directed by the Commission.

Finding F-F-2

The concept of defense-in-depth, as embedded in fuel cycle 

regulatory requirements and practices, is consistent with 

Commission guidance. Its implementation changes as the 

processes change at the fuel cycle facilities.

Recommendation F-R-1

The fuel cycle regulatory program should continue to evaluate 

the risk and the associated defense-in-depth protection by 

using insights gained from ISAs. ISAs should continue to 

evolve to support regulatory decisionmaking.

Section 4.7 – Interim Spent Fuel Storage

Finding S-F-1

The regulatory approach for SNF storage is largely based 

on meeting applicable industry consensus standards 

and conservative guidance to ensure adequate safety 

margins in the facility and cask designs and operations.

More recently, insights from a limited number of risk studies 

have been gradually factored into this regulatory approach. 

Furthermore, though qualitative, a systematic approach that 

parallels answering the risk triplet was used in the latest 

revision of the Standard Review Plan.

Recommendation S-R-1

While elements of the proposed risk management approach 

have been used in the SNF storage regulatory approach 

to evaluate the acceptable level of risk and the sufficiency 

of defense-in-depth (physical barriers, controls or margins) 

more consistently, the NRC should develop the necessary 

risk information, the corresponding decision metrics, and 

numerical guidelines. This is important in guiding further 

changes to the existing SNF storage regulatory approach 

and the evaluation of strategies for extended SNF 

storage activities.

Finding S-F-2 The concept of defense-in-depth is not explicitly or 

consistently applied in the SNF storage regulatory program.
A Proposed Risk Management Regulatory Framework

J-12 | Risk Management Task Force Appendix J

Recommendation S-R-2

As part of the implementation of the proposed risk 

management regulatory framework, the NRC should 

more consistently consider the concept of defense-indepth explicitly and evaluate its proper use in the SNF 

storage regulatory program. The NRC should also improve 

appropriate parts of staff training to make this concept a 

central part of such training.

Section 4.8 – Transportation

Finding T-F-1

While the U.S. transportation regulatory approach is governed 

by the IAEA transportation regulations, the current NRC 

transportation regulatory approach uses several elements of 

the proposed risk management framework.

Finding T-F-2

Risk assessments have been conducted on the safety of 

transportation of spent fuel. However, there is a lack of risk 

information on the transportation of other 

radioactive materials.

Recommendation T-R-1

Considering the strong international regulatory basis for 

transportation and the need to conform U.S. standards to 

those of the IAEA and other member states, application of 

the proposed risk management regulatory framework should 

focus on implementation guidance.

Recommendation T-R-2

The risk management process should be used to influence 

the future outcome of IAEA deliberations on proposed 

changes in international transportation regulations.

Recommendation T-R-3

The NRC should explore the value of using risk insights 

to justify regulations different from the IAEA’s for domestic 

use only, such as regulations dealing with domestic storage 

and transportation of high burnup fuel. Risk information 

could be used to develop a more flexible approach toward 

implementing and making gradual changes to current 

transportation regulations.



NUREG-2150

April 2012