Glossary of Risk-Related 

Terms in Support of RiskInformed Decisionmaking 

Manuscript Completed: May 2013 

Date Published: 

Prepared by: 

M.Drouin1

, M.Gonzalez1, 

S.Herrick1

, J.S. Hyslop1, D.Stroup1

J.Lehner2

, T.Pratt2

M.Dennis3

, J. LaChance3, T.Wheeler3

2

Brookhaven National Laboratory 

Upton, NY 11973 

3

Sandia National Laboratories 

Alburquerque, NM 87185 

M.Drouin, NRC Project Manager 

1

Office of Nuclear Regulatory Research 

NUREG-2122 
 
ABSTRACT

 iii 

The policy statement on the “Use of Probabilistic Risk Assessment (PRA) Methods in Nuclear 

Regulatory Activities” (Ref. 31) expressed the Commission’s belief that the use of PRA 

technology in U.S. Nuclear Regulatory Commission (NRC) regulatory activities should be 

increased. Consequently, the NRC carried out numerous risk-informed activities in all areas of 

NRC regulation. With increased risk-informed activities came the recognition that the agency 

could enhance regulatory stability and efficiency if it implemented the many potential 

applications of risk information in a consistent and predictable manner. An essential part of 

consistent and predictable implementation is the use of consistent terminology to ensure 

accurate communication and transfer of information. Further, the NRC recognizes that some 

risk-related terms have been used in ambiguous ways by practitioners. The increased 

development of guidance documents, regulations, and procedures related to risk-informed 

activities makes the fundamental understanding of these risk-related terms more imperative. 

Consistent terminology is essential to the appropriate implementation of risk-informed activities 

and the communication between the NRC and its stakeholders. It allows practitioners to 

eliminate communication issues and avoid unnecessary discussions that may have been 

erroneously perceived as technical issues. Therefore, a glossary with agreed-upon definitions 

of risk-informed related terms is an essential tool for future risk-informed activities. This 

glossary addresses risk-related terms used in the context of risk associated with a reactor of a 

nuclear power plant. 

 

TABLE OF CONTENTS 

 v 

Page

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

EXECUTIVE SUMMARY ............................................................................................................ vii

ACKNOWLEDGEMENTS ............................................................................................................ ix

ACRONYMS AND ABBREVIATIONS .......................................................................................... xi

1. INTRODUCTION ................................................................................................................ 1-1

1.1 Background ................................................................................................................. 1-1

1.2 Objective ...................................................................................................................... 1-2

1.3 Scope and Limitations ................................................................................................. 1-3

1.4 Approach Summary ..................................................................................................... 1-4

1.4.1 Identification and Selection of Terms ................................................................... 1-4

1.4.2 Development of Definitions .................................................................................. 1-6

2.0 APPROACH .................................................................................................................... 2-1

2.1 Introduction .................................................................................................................. 2-1

2.2 Task 1: Identification and Selection of Terms ............................................................. 2-1

2.2.1 Step 1-1: Initial Potential Candidates .................................................................. 2-2

2.2.2 Step 1-2: Important for Risk Communication ...................................................... 2-5

2.2.3 Step 1-3: A Risk-Context Specific Definition ....................................................... 2-7

2.2.4 Step 1-4: Availability of Definitions ...................................................................... 2-8

2.2.5 Step 1-5: Multiple Term Definitions ..................................................................... 2-8

2.2.6 Step 1-6: Consensually Established Definitions .................................................. 2-9

2.2.7 Step 1-7: Term Fundamental to Risk Communication....................................... 2-10

2.2.8 Step 1-8: Policy Implications ............................................................................. 2-12

2.3 Task 2: Development of Definitions .......................................................................... 2-13

2.3.1 Step 2-1: Develop Initial Glossary ..................................................................... 2-14

2.3.2 Step 2-2: Identify Related Terms ....................................................................... 2-14

2.3.3 Step 2-3: Finalize Definitions in Glossary .......................................................... 2-15

3.0 ABBREVIATIONS AND ACRONYMS ............................................................................. 3-1

4.0 GLOSSARY .................................................................................................................... 4-1

4.1 Understanding the Format and Structure .................................................................... 4-1

4.2 Terms and Definitions .................................................................................................. 4-5

5.0 REFERENCES ................................................................................................................ 5-1

APPENDIX A INTERNAL FIRE GLOSSARY ............................................................................ A-1

APPENDIX B PRA TECHNICAL ELEMENTS .......................................................................... B-1
 
EXECUTIVE SUMMARY 

 vii 

The final policy statement on the “Use of Probabilistic Risk Assessment Methods in Nuclear 

Regulatory Activities” (Ref. 33) expressed the U.S. Nuclear Regulatory Commission’s (NRC’s) 

belief that the use of probabilistic risk assessment (PRA) technology in NRC regulatory activities 

should be increased. Since the PRA policy statement, the staff has issued several PRA or 

risk-informed plans detailing various risk-informed activities. 

With increased risk-informed activities comes the recognition that regulatory stability and 

efficiency would be enhanced if the various risk-information activities are implemented 

consistently and predictably. An essential part of implementation is the use of consistent 

terminology to ensure a common understanding of information. A common understanding of 

information provides increased assurance that the analyses being performed are technically 

adequate to facilitate better risk-informed decisionmaking. 

A glossary with definitions of risk-informed-related terms is an essential tool for risk-informed 

activities. A glossary provides clarity on the meaning of many terms. For terms that are context 

or scope dependent, a single definition may not be appropriate, but a discussion on the use of 

these terms in different contexts will be helpful. 

This NUREG report identifies and defines terms used in risk-informed activities related to 

commercial nuclear power plants. It provides a single source in which these terms can be 

found. A major goal of the glossary is to reduce ambiguity in the definition of terms as much as 

possible, so that a common understanding can be achieved that will facilitate communication on 

risk-informed activities. 
 
ACKNOWLEDGMENTS

 ix 

Numerous individuals contributed to the development of this glossary. The majority of the 

definitions are based on definitions from numerous sources (see Section 5 for a list of 

references); as such, the authors do not claim sole authorship. However, it was an immense 

and challenging effort to perform the necessary research, identify the terms to be included in the 

glossary, develop definitions understandable to individuals regardless of their level of risk 

expertise and experience, and provide the necessary discussion on the usage of terms. 

Therefore, the following acknowledgements are made: 

NRC Project Manager and Technical Lead: Mary Drouin 

Main contributors or authors by organizations include: 

U.S. Nuclear Regulatory Commission 

Mary Drouin 

Michelle Gonzalez 

J.S. Hyslop 

Sandra Herrick 

David Stroup 

Brookhaven National Laboratory 

John Lehner 

Trevor Pratt 

Sandia National Laboratories 

Matthew Dennis 

Jeffery LaChance 

Timothy Wheeler 

Energy Math Solutions 

Zoran Musicki 

OMICRON Safety and Risk Technologies 

Pier Carrara 

Willard Thomas 

 
 
ACRONYMS AND ABBREVIATIONS 

 xi 

AOOs anticipated operational occurrences 

ATD atmospheric transport and diffusion 

BDBAs beyond-design-basis accidents 

BDBEs beyond-design-basis events 

ECCS emergency core cooling system 

FTR fails to run 

FTS fails to start 

IAEA International Atomic Energy Agency 

IPEEE individual plant examinations for external events 

IPEs individual plant examinations 

LOCCW loss of the component cooling water 

LWR light-water reactor 

NRC U.S. Nuclear Regulatory Commission 

POS plant operating states 

PRA IP Probabilistic Risk Assessment Implementation Plan 

PRA probabilistic risk assessment 

RCPs reactor coolant pumps 

RG regulatory guide 

RIRIP Risk-Informed Regulatory Implementation Plan 

RPP Risk-Informed Performance-Based Plan 

RPS reactor protection system 

SSC structures, systems, and components 

SSE safe-shutdown earthquake 

1. INTRODUCTION 

 1-1 

1.1 Background

The final policy statement on the “Use of Probabilistic Risk Assessment Methods in Nuclear 

Regulatory Activities” (Ref. 33) expressed the U.S. Nuclear Regulatory Commission’s (NRC’s) 

belief that the use of probabilistic risk assessment (PRA) technology in NRC regulatory activities 

should be increased. Since the PRA policy statement, the staff has issued several PRA or 

risk-informed plans detailing various risk-informed activities. The NRC used the first plan, the 

PRA Implementation Plan (PRA IP), until 1999. This plan identified the initial risk-informed 

activities undertaken as a result of the PRA policy statement. As the use of risk information in 

regulation increased further, the NRC replaced the PRA IP with the Risk-Informed Regulation 

Implementation Plan (RIRIP) in 2000. This plan reflected the increased sophistication and 

experience in the use of risk assessment methods that included not just PRA, but also 

integrated safety assessments and other risk-related techniques. The RIRIP was improved and 

became the Risk-Informed and Performance-Based Plan (RPP), submitted with SECY-06-0217, 

“Improvement to and Update of the Risk-Informed Regulation Implementation Plan,” dated 

October 25, 2006 (Ref. 99). The NRC implemented the RPP in response to a June 2006 

Commission-issued staff requirements memorandum, “Briefing on Status of Risk-Informed and 

Performance-Based Reactor Regulation,” dated June 1, 2006, (Ref. 101), which directed the 

staff to (1) improve the RIRIP so that it is an integrated master plan for activities designed to 

help the agency achieve the Commission’s goal of a holistic, risk-informed and 

performance-based regulatory structure, and (2) seek ways to communicate more transparently 

the purpose and use of PRAs in the NRC’s reactor regulatory program to the public and 

stakeholders. The RPP is updated annually. 

As these plans indicate, risk information is used in every aspect of the NRC’s work 

(e.g., regulation and guidance, licensing and certification, oversight, and operational 

experience). Examples of these include the following: 

 

• Regulation and guidance—Recent risk-informed rules include Title 10 of the Code of 

Federal Regulations (10 CFR) 50.44, “Combustible Gas Control for Nuclear Power 

Reactors” (Ref. 16); 10 CFR 50.48(c), “Fire Protection” (Ref. 17); and 10 CFR 50.69, 

“Risk-Informed Categorization and Treatment of Structures, Systems, and Component 

for Nuclear Power Reactors” (Ref. 21).

• Licensing and certification—Regulatory Guide (RG) 1.174, “An Approach for Using 

Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to 

the Licensing Basis,” Revision 1, issued November 2002 (Ref. 84) provides general 

guidance on what is needed for risk-informed applications for licensing -basis changes, 

while specific risk-informed guidance is offered in RG 1.177, “An Approach for 

Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications,” issued 

August 1998 (Ref. 86); RG 1.175, “An Approach for Plant-Specific, Risk-Informed 

Decisionmaking: Inservice Testing,” issued August 1998 (Ref. 85); and in RG 1.178, “An 

Approach for Plant-Specific, Risk-Informed Decisionmaking for Inservice Inspection of 

Piping,” issued September 2003 (Ref. 87). 

• Plant oversight—the Reactor Oversight Process uses risk-informed performance 

indicators and inspections, as well as a risk-informed significance determination process, 

for inspection findings. 
1. INTRODUCTION 

 1-2 

• Operational experience—Management Directive 8.3, “NRC Incident Investigation 

Program,” dated March 27, 2001, (Ref. 43), provides risk-informed incident investigation 

direction. The Office of Nuclear Reactor Regulation instruction LIC-504 focuses on 

risk-informed decisionmaking for event assessment, and the Accident Sequence 

Precursor Program (Ref. 96) calculates a conditional core damage probability to assess 

event significance. 

With increased risk-informed activities comes the recognition that regulatory stability and 

efficiency would be enhanced if the various applications of risk-information are implemented 

consistently and predictably. An essential part of implementation is the use of consistent 

terminology to ensure a common understanding of information. A common understanding of 

information provides increased assurance that the analyses being performed are technically 

adequate and, therefore, produce correct results. This assurance, in turn, leads to better 

risk-informed decisionmaking. 

Historically, some risk-related terms have been used somewhat differently by different 

practitioners, but the increased development of guidance documents, regulations, and 

procedures makes a common understanding of fundamental terms in these areas more 

imperative. A common understanding is fundamental to communication in the risk-informed 

arena for the consistent and appropriate treatment of risk-informed applications by industry as 

well as risk-informed regulatory actions by the NRC, and for communication between the NRC 

and its stakeholders. There are a variety of reasons why consistent use of terminology is not 

always found in the area of risk-informed activities: multiple definitions exist for the same term, 

terms are used interchangeably when they are not synonymous, or the definition is scope or 

context dependent. 

A glossary of risk terms is needed for appropriate interpretation of terminology to support a 

risk-informed regulatory structure (e.g., licensee applications, NRC regulations). A good 

illustration of this need is the use of the term “internal events.” The term is used to refer to 

potential events resulting from equipment failures and human errors that can result in a plant 

disturbance. In some instances, “internal events” has been defined as events occurring within 

the plant and includes internal fires or internal floods or both; however, in other instances it has 

been defined as events occurring within the component boundary and does not include internal 

fires and internal floods. Further, there are instances in which the term has been used without a 

definition and it is not clear which definition is intended. In the treatment of an application or 

implementation of a regulation, it is important to know if internal events include internal fires 

and/or floods, or not. 

Therefore, a glossary with definitions of risk-informed-related terms is an essential tool for 

risk-informed activities. A glossary provides clarity on the meaning of many terms. For terms 

that are context or scope dependent, a single definition may not be appropriate, but a 

discussion on the use of these terms in different contexts will be helpful. 

1.2 Objective

The objective of this glossary is to identify and define terms used in risk-informed activities 

related to commercial nuclear power plants. This glossary provides a single source in which 

these terms can be found. A major goal of the glossary is also to reduce ambiguity in the 

definition of terms as much as possible, so that a common understanding can be achieved that 

will facilitate communication on risk-informed activities. Among other things, this glossary will 
1. INTRODUCTION 

 1-3 

allow individuals to distinguish communication issues—erroneously perceived as technical 

issues—from actual technical discussions. Where terms are found to have a justifiable variety 

of definitions, depending on the context in which they are used, the objective of this glossary is 

to explain the individual definitions, along with the context, to ensure proper context-specific use 

of the term. Whenever possible, existing definitions are used and redefining terms is avoided. 

This NUREG glossary is a supporting document for documents and procedures that include risk 

terminology. As such, it is a reference for NRC staff, as well as other stakeholders in the 

risk-informed arena. 

1.3 Scope and Limitations

In developing the list of terms to be included in the glossary, it was not possible to identify every 

applicable source nor was it practical to review every available source. For this glossary, the 

sources are limited to internal NRC sources (e.g., regulations, NUREG reports, technical 

reports, regulatory guides, standard review plans, Commission documents) that are risk-related. 

Except for certain PRA standards and International Atomic Energy Agency (IAEA) documents, 

industry sources were not included. 

This glossary is intended to provide a common understanding of the identified terms for users 

with different backgrounds, skills, and experience. Therefore, this glossary contains terms that 

are likely to be familiar even to users with only limited exposure to risk-informed activities 

(although they may not have a common understanding of the term), as well as terms that only 

experienced practitioners would use in their work. Consistent with the objective to make this 

glossary also of use to analysts and regulators inexperienced in the area of risk-informed 

activities, certain “core” terms that are fundamental to the basic understanding of risk concepts 

and risk analyses are included. 

This glossary covers terms used in all three levels of a light-water reactor (LWR) PRA: core 

damage frequency analysis (Level 1 PRA), radionuclide release frequency analysis (Level 2 

PRA), and consequence analysis and risk integration (Level 3 PRA). It also includes terms 

used in other quantitative analyses (i.e., terms such as “seismic margin”) and in qualitative risk 

considerations (i.e., terms such as “dependency”). 

Terms that are common across all hazard groups—internal hazards (e.g., internal events, 

internal floods, and internal fires) and external hazards (e.g., earthquakes, external floods, high 

winds)—are included. For specific hazards (i.e., internal fires), a hazard-specific glossary is 

provided. At this time, the only hazard-specific glossary included in this NUREG is one for 

internal fires. 

The terminology in this glossary is meant to cover risk-related terms used for an at-power PRA, 

as well as for the other plant states (i.e., low power and shutdown). This glossary also 

addresses risk terms that may come up in the different life stages of a power reactor: design, 

licensing, operation, and decommissioning. 

There are general scientific terms that do not take on an additional meaning for risk-related 

activities; in general, these terms are not included in this glossary. However, in some cases, 

some of these terms are included because they are fundamental to understanding the results or 

insights used in risk-related activities. 
1. INTRODUCTION 

 1-4 

The terms defined in this glossary are those used in risk-informed activities related to 

commercial nuclear power reactors. As such, risk terms related to nonnuclear industries, such 

as the chemical industry, are not included. In addition, nuclear technology-related risk terms 

specific to nonpower reactor parts of the industry (e.g., nuclear material and waste, research 

reactors) also are excluded. 

This glossary contains many generic terms used throughout the risk field whose definitions are 

broadly applicable to areas outside of the area of power reactors. However, some of these 

generic terms may have different meanings when used in nonreactor or nonnuclear fields; 

therefore, these other definitions are not included in the glossary. 

In the power reactor area, this glossary is meant to be broad and wide-ranging. The terms 

included are meant to support all risk-informed activities related to LWRs (including advanced 

LWRs) that produce electric power. However, some risk-related reactor terms may not be 

included (e.g., terms that are non-LWR specific). 

In summary, this NUREG provides a single broad source for terms used in risk-informed 

activities related to LWR power reactors. 

1.4 Approach Summary 

A major challenge in developing such a glossary is selecting the terms to include in the 

glossary. The defined scope provides general boundaries for inclusion in the glossary. Although 

the focus of the glossary is on terms for risk communication, the list still could be exhaustive. 

Consequently, guidance was developed to identify and select the terms. The second major 

challenge is developing the definitions. Providing just a definition for each term would not 

completely accomplish the glossary’s objective. To meet the objective, both a definition of each 

term and insights on its meaning and use were needed. Consequently, guidance also was 

developed for developing the definitions. The approach is summarized below and discussed in 

detail in Section 2. 

1.4.1 Identification and Selection of Terms 

A major challenge in developing a glossary is in how to identify and select specific terms to 

include. An initial list was developed that was meant to be as broad as possible to help ensure 

a term was not prematurely excluded from consideration. The general guidance used to identify 

this list was that the term should be related to risk communication. For example, if the term is 

used to communicate what is meant by the term “risk,” then the authors considered such issues 

as what is a risk analysis, what are risk results and insights, and how risk results and insights 

are used in decisionmaking. Based on this guidance, an initial list of terms was compiled in a 

two-step process. 

Terms were identified by reviewing documents related to or that support risk-informed activities. 

The types of documents selected for review included PRA standards, NUREGs and technical 

methodology documents, regulatory guides and standard review plans for risk-informed 

applications, risk-informed regulations, and Commission documents on risk-informed activities. 

Section 5 provides the list of sources. 

The NRC staff and management also were asked to augment the initial list. Participants in this 

step included individuals with and without risk expertise and both junior and senior staff. 
1. INTRODUCTION 

 1-5 

Although the initial list was meant to be broad, it resulted in a list of more than 1,000 terms. 

Table 1-1 gives a small sample of this initial list of terms. 

Table 1-1 Sample List of Initial Terms for Inclusion in Glossary

Accident Consequences 

Accident Mitigation 

Assumption 

At-Power 

ATHEANA 

Atmospheric Transport and 

 Dispersion 

Authority Having 

 Jurisdiction 

Backstop 

Basic Event 

Bayesian 

Certified Seismic Design 

Circuit Failure Mode 

Classical Statistics 

Common Cause Failure 

Common Mode Failure 

Containment Event Tree 

Containment Failure Mode 

Core Cooling Water System 

Core Damage 

Defense-in-Depth 

Decision Model 

Deterministic 

Diagnosis 

Dominant Accident Sequence 

Dominant Contributor 

Dose Conversion Factor 

Dose Response Model 

Double Contingency Principle 

Dynamic Probabilistic Risk 

 Assessment 

Dynamic Risk Level 

Early Containment Failure 

Event Scenario 

Event Sequence 

External Hazards 

Failure Probability 

Frequentist 

Fussell-Vesely Importance 

 Measure 

Human Error 

Infrequent Event 

Initiating Event 

Key Assumption 

Living PRA 

Mean Value 

Operational Risk 

Performance Shaping Factor 

Recovery Action 

Significant Accident Sequence 

Technical Adequacy 

Uncertainty Distribution 

Zone Of Influence 

With such a large list, it was necessary to prioritize the selection of the final list. The authors 

developed guidelines consistent with the objectives of this project for ensuring that terms were 

appropriately selected for the glossary. 

Terms not relevant to risk-informed activities were excluded. For example, self-evident terms, 

such as names of organizations, names of structures, systems, and components (SSCs), and 

units of measure were excluded from the glossary (e.g., auxiliary feedwater system, becquerel). 

Terms necessary to understand risk analysis initially were retained as potential candidates for 

the glossary (i.e., terms needed to understand the term “risk,” to understand what constitutes a 

risk analysis, to understand the different kinds of risk analyses, including their associated 

terminology, were not excluded). Further, terms that play a role in risk-informed 

decisionmaking were not necessarily excluded from the list. While a term may not relate to the 

understanding of risk analysis, the term may often be used or associated with risk-informed 

decisionmaking. These terms also were retained as potential candidates for inclusion in the 

glossary. 

After the initial screening, although some terms were identified as candidates to be screened 

from the list, they were, nevertheless, retained. These terms were appropriately identified to be 

screened from the list because they are not used in the broad definition of risk communication; 

however, they are related to risk communication for a specific hazard. It was decided that these 

previously screened terms which were identified as being related to risk communication for a 

specific hazard should be retained in the glossary. Consequently, it was determined that there 

should be two types of glossaries: 

1. A main glossary that focuses on terms used in communicating risk at a high level. 

2. Hazard-specific glossaries that focus on terms used in communicating the details in 

performing the hazard-specific risk analysis. 
1. INTRODUCTION 

 1-6 

The main glossary is documented in Section 4, and the hazard-specific glossaries are 

documented in the appendices. At this time, only an internal fire hazard-specific glossary has 

been developed. 

Although a term is important to risk communication (whether at a high level or hazard-specific 

level), there still may be other reasons for excluding the term from the glossary. The term may 

not have a meaning that is unique in a risk context (i.e., a term’s definition may be well 

established throughout the community without a basis for possible miscommunication). These 

terms are potential candidates for exclusion from the glossary. However, the term may be 

fundamental to a basic understanding of risk. Such a fundamental term is not excluded and is 

retained as a potential candidate for the glossary. For example, “probability” is a term whose 

definition is well established and does not change when used in a risk context. As such, the 

term becomes a candidate for exclusion. However, the term “probability” is fundamental in 

understanding risk, and it is also frequently used in discussing both inputs to a PRA model and 

results from the PRA. Consequently, “probability” is a term that should not be excluded, but 

should be retained and included in the glossary. 

The last guideline involves determining if the term has policy implications and requires approval 

from the Commission to issue an officially documented definition. A good example of such a 

term is “defense-in-depth.” This term can be important in risk communication. There are many 

different understandings of the term “defense-in-depth” and there is no consensus on its 

meaning. Development of a definition, however, does have policy implications and, therefore, 

would require Commission approval. These types of terms are included in the glossary; 

however, a single “official” definition is not developed. The glossary discusses the various 

definitions that have been used and notes the policy implications. 

1.4.2 Development of Definitions 

This glossary does not recreate definitions. Consequently, where the definition of the term 

already exists, and there is consistency among the various sources, that definition is used as 

the basis for the definition provided in the glossary. However, there are terms with multiple 

definitions among different sources that are not in agreement. For terms that have multiple 

(legitimate) definitions, each definition is included in the glossary, and an explanation describing 

the differences and bases is provided. For terms that have multiple, conflicting definitions, an 

appropriate definition is selected, and an explanation is provided for the basis for the definition 

provided. 

As noted above, it is not the intent of this glossary to recreate definitions. In some cases, it was 

determined that (1) there may not be disagreement about the definition, (2) the definition was 

appropriate, or (3) for multiple definitions, there is consistency. However, it was determined that 

for some terms, an experienced risk analyst may be needed to understand the definition. In 

these cases, although the definition from the sources is included, a definition is developed in 

“plain language” (i.e., a definition is provided that does not rely on technical jargon). The reason 

the definition is written in plain language is to minimize any misunderstanding of the definitions. 

Furthermore, plain language helps PRA practitioners, including those who are not native English 

speakers, to understand the definitions with minimum language barriers. The use of plain 

language is an NRC policy to improve communication with the general public and other 

stakeholders. 

To help the user, numerous terms are cross-referenced in the glossary. The authors used 

these cross-references when they thought that to completely understand the term, related terms 
1. INTRODUCTION 

 1-7 

also needed to be understood. These cross-references are for terms, for example, that meet 

the following criteria: 

• They are similar but have different meanings and there are instances in which they are 

used incorrectly (e.g., “probability” versus “frequency,” “core damage” versus “core 

melt”) 

• They are the opposite of another term (e.g., “deterministic” and “probabilistic”). 

• They are closely related in meaning and may be a subset of the related term or an 

example of the related term (e.g., “risk analysis” and “probabilistic risk analysis”, 

“accident consequence” and “health effects”). 

• They are closely related in that understanding one term depends on understanding the 

related term (e.g., “aleatory uncertainty” and “uncertainty,” “core damage” and “core 

damage frequency” and “core damage probability”). 

 

The glossary also combines terms. Instead of appearing as individual terms, they are defined 

together in the glossary as a single term or a group of terms. These groups include terms, for 

example, which are: 

• composed of multiple words with the same “adjective” (e.g., “significant,” “significant 

contributor,” “significant basic event”; “significant accident sequence”) 

• similar and convey the same meaning (e.g., “general transient” and “transient”) 

• complementary of each other (e.g., “availability” and “unavailability”) 

For each term in the glossary, the definition(s) stated in the source documents are collected. 

Examples of the source documents are the PRA standards, NUREGs associated with 

risk-informed activities or risk methodologies, tools or data, regulatory guides and standard 

review plans for risk-informed applications, risk-informed regulations, and Commission 

documents on risk-informed activities (e.g., PRA policy statement). Section 5 provides a 

complete list of the source documents. 

2. APPROACH 

 2-1 

2.1 Introduction 

Two major tasks were performed while developing this glossary. The first task was identifying 

and selecting terms. The second task was developing the actual definitions of the selected 

terms. 

The approach used for these two tasks are discussed in detail in Sections 2.2 and 2.3. 

Both the identification and development of the definitions is an 

iterative process. Additional terms may need to be included in the 

glossary as terms are identified, selected, and defined, as shown 

in the illustration at right. However, the steps associated with the 

guidance are discussed sequentially, although each step in the 

two tasks also may result in identifying a new term to be included 

in the glossary. 

2.2 Task 1: Identification and Selection of Terms 

The list of terms was meant to be as extensive as possible. 

However, for practical purposes, guidelines were necessary to keep the list to a manageable 

size. Therefore, the list was developed by identifying potential candidate terms and then 

selecting the actual terms for the glossary. Guidance consistent with the objectives of this 

project was developed to ensure that terms were appropriately selected for the glossary. 

A set of eight steps was developed with the sole purpose of determining if a term should 1) be 

considered as a candidate term, and 2) if the candidate term should remain as part of the 

glossary. The process for identifying and selecting terms for the glossary consisted of the 

following eight steps: 

1-1 Initial Potential Candidate – A list of candidate terms is first developed by reviewing a 

set of risk-related documents and interviewing staff. A high-level screening is performed 

to exclude such terms as “auxiliary feed water.” 

1-2 Used in Risk Communication or Specific Hazard Risk – Not all candidate terms from 

Step 1 are essential to risk communication and perhaps should be excluded from the 

glossary. However, although a term is not considered essential in communicating risk at 

a high level, it may be essential in communicating the details of performing a 

hazard-specific risk analysis (e.g., internal fire). These terms should not be excluded. 

1-3 Risk Context Specific Definition – Terms that have a risk context specific meaning are 

terms whose meaning may be consistent with the actual definition in a dictionary, but 

their meaning in risk communication has a risk connotation (e.g., internal events). These 

terms should be retained as potential candidates. 

1-4 Availability of Definitions – Terms with readily available definitions are terms whose 

definition can be easily found in technical documents. These terms should be 

considered for potential exclusion. 

1-5 Multiple Definitions – Terms may have multiple definitions. For some terms, these 

definitions may be unrelated, some may be legitimate, and others may not be. 

Identification 

of terms

Selection of 

terms

Definition of 

terms
2. APPROACH 

 2-2 

1-6 Consensually Established Definitions – Not all sources necessarily have the same 

(or similar) definition for a term. However, if there is agreement among the different 

sources on a term’s meaning, then it can be assumed it has been consensually 

established. These terms should be considered for potential exclusion. 

1-7 Term Fundamental to Risk Communication – Some terms have been identified for 

potential exclusion, yet they may be essential to a basic understanding of risk-informed 

activities. These terms should be retained and included in the glossary. 

1-8 Policy Implications – If it is determined that a term has policy implications, its definition 

may require Commission approval. These terms are identified and a “formal” definition 

is not developed. 

This eight-step process is illustrated in Figure 2-1 and discussed in detail below. 

2.2.1 Step 1-1: Initial Potential Candidates 

This step identifies the initial list of potential candidates for inclusion in the glossary. This initial 

list of candidate terms is as broad as possible to ensure that a term is not prematurely excluded. 

The guidelines to develop an initial list of terms include: 

1. identifying initial list of candidate terms 

2. performing a high-level screening 
2. APPROACH 

 2-3 

No 

No 

Step 1-8 

Identify any policy 

implications 

Step 1-4 

Is the definition of the 

term readily available? 

Yes 

Step 1-1 

Is the term a 

potential candidate? 

Step 1-2 

Is the term important for 

communication of risk or 

specific hazard risk? 

Step 1-3 

Does the term have a 

risk context specific 

definition? 

No 

Yes No 

Step 1-5 

Identify terms with 

multiple definitions 

Step 1-7 

Is the term 

fundamental to risk 

communication? 

Yes 

Term Excluded 

from Main 

Yes -- risk Glossary 

Yes 

Step 1-6 

Is there a consensually 

established definition of 

the term? 

Yes 

No 

Term Included 

in Main 

Glossary 

No 

Term included in 

Hazard Specific 

Glossary 

Yes -- hazard

Reference 

Sources 

Figure 2-1 Process to identify terms for glossary 
2. APPROACH 

 2-4 

Guideline 1-1.1 – Identification of Initial List 

This guideline is used to develop an initial list of candidate terms. Terms were identified by 

reviewing documents related to risk communication. For example, the term may be used to 

communicate what is meant by the term “risk,” what is a risk analysis, what are risk results and 

insights, or how are risk results and insights used in decisionmaking. The types of documents 

selected for review included the following: 

• PRA standards 

• risk-informed related NUREGs 

• risk-informed regulatory guides and standard review plans 

• risk-informed regulations 

• NRC inspection manuals 

• Commission documents on risk-informed activities 

• selected IAEA documents 

The sources reviewed are listed in Section 5. 

NRC staff and management also were asked to augment the initial list. Participants included 

individuals with and without risk expertise and both junior and senior staff. The purpose and 

objective of the glossary was provided to these participants, along with the guidelines used to 

identify the initial list. Terms identified by the staff were added to the initial list of terms. 

The process discussed above for identifying potential terms resulted in an initial list of more than 

1,000 terms. A high-level review indicated many terms that simply do not belong in a “risk 

communication glossary.” Consequently, the second guideline of Step 1-1 involved a high-level 

screening to identify terms that do not fit the scope of the glossary. 

Guideline 1-1.2 – High-Level Screening 

This guideline is used to perform a high-level screening. The high-level screening identifies 

terms that do not have a risk context. These terms are well known and do not need to be 

defined in the glossary. These types of terms include names of SSCs and operational 

procedures, units of measure, organizations, chemicals, nuclear safety analyses, and names of 

computer codes (including acronyms). Examples of terms excluded from the glossary include 

the following: 

• Examples of SSCs and operational procedure: 

⎯ auxiliary feedwater 

⎯ cable 

⎯ heating, ventilation, and air conditioning 

• Examples of units of measure: 

⎯ becquerel 

⎯ rem 

⎯ curie 

• Examples of organizations: 

⎯ Gesellschaft fuer Anlagen-und Reaktorsicherheit (Germany) 

⎯ IAEA 
2. APPROACH 

 2-5 

• Examples of chemical terms: 

⎯ chlorinated polyethylene 

• Examples of nuclear safety analyses terms: 

⎯ cladding 

⎯ water hammer 

⎯ departure from nucleate boiling 

• Examples of computer code acronyms: 

⎯ CAFTA 

⎯ RISKMAN 

These types of terms were excluded from the list of candidate terms. 

OUTPUT

Once the guidance in this step has been applied, an initial list of potential candidate terms for 

the glossary was identified. 

Because a term is a potential candidate for inclusion in the glossary does not necessarily mean 

that it should remain in the glossary. For example, the term may not be important to risk 

communication; or while it may be important, it is a well-known term whose meaning is widely 

accepted. In the next steps, each term was further reviewed to determine if it should remain as 

a candidate for the glossary. 

2.2.2 Step 1-2: Important for Risk Communication 

This step determines if any of the candidate terms from Step 1-1 should be excluded from the 

glossary because they are not important to risk communication. If a term is considered not 

essential to risk communication, it can be excluded from the glossary; conversely, if a term is 

considered essential, it should remain as a potential candidate for the glossary. 

To determine if a term is essential for risk communication, guidelines need to be established to 

judge a term’s importance. The candidate terms are then reviewed against these guidelines. 

However, developing the guidelines requires an understanding of what is meant by “risk 

communication,” and therefore, what is meant by “term is needed (or essential) to risk 

communication.” 

Risk communication can have various meanings, from a very strict interpretation to a wide 

interpretation. For example, risk communication can mean communicating: 

• What is meant by the term “risk”? 

• What is a risk analysis? 

• What are risk results and insights? 

• How are risk results and insights used in decisionmaking? 

As discussed in Section 1, for the purposes of this glossary, risk communication includes all of 

the above. In addition, understanding the audience is equally important. The audience may 

range from senior executives to subject experts to lay personnel. Consequently, guidelines are 

developed consistent with these objectives and expectations. 
2. APPROACH 

 2-6 

The guidelines developed to retain or screen candidate terms included: 

1. The term is related to the science of risk analysis. 

2. The term plays a role in risk-informed decisionmaking. 

Guideline 1-2.1 – Term Related to Science of Risk Analysis 

This guideline is used to identify terms related to the science of risk analysis (i.e., terms needed 

to understand risk analysis). Understanding risk analysis involves more than providing a 

definition of the term “risk analysis.” Terms used in communicating a risk analysis and those 

used in understanding the details of performing a risk analysis need to be included. 

The former (i.e., communicating a risk analysis) involves understanding the following: 

• What constitutes a risk analysis? (i.e., what is a risk analysis?) 

• What are the different kinds of risk analyses? 

• What are the different terms used to explain risk analyses? 

• What are the objectives of a risk analysis? 

• What are the inputs and outputs of the analysis? 

Terms needed to understand risk analysis (using the above guidance) should be retained as 

potential candidates for inclusion in the glossary. Examples of terms related to the high-level 

understanding include: 

• consequence 

• probability 

• core damage frequency 

• health effects 

• initiating event 

Understanding the details of performing a risk analysis involves comprehending the different 

technical elements of the specific hazards for which a risk analysis is performed (e.g., internal 

fire risk analysis) and the associated terminology for each technical element. These terms are 

included and defined in the hazard-specific glossaries. At this time, only one hazard-specific 

glossary has been developed (internal fires). 

Determining if the term should be in the main glossary or the hazard-specific glossary can be 

subjective. For example, “hot short” is a term generally not used in risk communication, but it is 

part of the lexicon for internal fire. Consequently, it is defined in the internal fire-hazard 

glossary. 

Guideline 1-2.2 – Term Plays a Role in Risk-Informed Decisionmaking 

This guideline is used to identify terms that may not be related to the science of risk analysis, 

but often are used or associated with risk-informed decisionmaking. These terms are used in 

nuclear safety activities that are not necessarily risk specific, but are often used when 

risk-informed issues are discussed and communicated. There are also terms that may have risk 

aspects included under the umbrella of their broader meaning and should be retained as 

potential candidates for inclusion in the main glossary. 
2. APPROACH 

 2-7 

Examples of terms that play a role in risk-informed decisionmaking include: 

• safety margins 

• severe accidents 

• public health effects 

OUTPUT

Once the guidance in this step has been applied, terms are screened as either (1) important at a 

high level to risk communication for potential inclusion in the main glossary, (2) important for 

hazard-specific risk communication for potential inclusion in the hazard-specific glossary, or 

(3) not important to risk communication at any level and excluded from the glossary. 

Because a term is important to risk communication, it does not necessarily mean that the term 

should remain in the main glossary. For example, the term may have no risk context or the 

term’s definition may be well established throughout the community without any potential 

miscommunication. In the next steps, each candidate term for the main glossary is further 

reviewed to determine if it should remain as a candidate. 

The remaining steps and guidance are discussed relative to the main glossary. However, they 

are applicable to the hazard-specific glossary and used where appropriate. 

2.2.3 Step 1-3: A Risk-Context Specific Definition 

This step determines if any of the terms from Step 1-2 have a risk-context specific meaning. 

Specifically, although a term’s meaning may be consistent with the actual definition in a 

dictionary, its meaning in risk communication has a risk connotation. There are terms, however, 

whose meaning is the same regardless of their use. 

The sources are reviewed to determine if a term’s meaning has a risk connotation. At this point, 

there may be multiple risk meanings for a single term, which may or may not be consistent. 

However, the purpose of this step is only to identify those terms that do not have a risk 

connotation and, therefore, may be excluded from the main glossary. 

For example, consider the terms “probability” and “internal hazards.” The meaning of the term 

“probability” does not change when used in a risk context. Probability is defined as the “relative 

possibility that an event will occur as expressed by the ratio of the number of actual occurrences 

to the total number of possible occurrences.” The term “internal hazards,” however, does have 

a risk-context meaning. Outside of a risk context, an internal hazard would be “something 

dangerous that happens in the interior of something.” However, in a risk context, an internal 

hazard has a very specific meaning. An “internal hazard” can be “an event originating within a 

nuclear power plant that directly or indirectly disrupts the steady state operation of the plant.” 

For the main glossary, terms such as “probability” should be candidates for exclusion whereas 

terms such as “internal hazard” should be retained as a possible candidate. 

OUTPUT

Once the guidance in this step has been applied, terms that have a risk-context specific 

meaning have been identified, along with those whose meanings are not risk-context specific. 
2. APPROACH 

 2-8 

Because a term has risk-context specific meaning does not necessarily mean it should be 

included in the main glossary. The term may have a risk-context specific meaning that is 

consensually established, uniformly known, and consistently used to the extent that it does not 

need to be explained to potential audiences. Conversely, because a term has no risk-context 

specific meaning does not necessarily mean it should be excluded from the main glossary. 

There may be other reasons for including it in the glossary, such as it is fundamental in risk 

communication. In the next steps, each term is further reviewed to determine if it should remain 

as a candidate for the main glossary. 

2.2.4 Step 1-4: Availability of Definitions 

This step determines if the terms identified in Step 1-3 have readily available definitions. The 

question becomes: Is there a readily available definition for terms whose usage (or definition) is 

the same whether it is used in a risk context or some other context? 

If the term is consistently defined in well-established and authoritative references, such as 

published governmental regulations, text books, dictionaries, or consensus standards, then its 

inclusion in the main glossary may not be necessary. Examples of terms used in risk 

communication that have well-defined definitions, which apply in a risk context as well as in 

other contexts, include: 

• probability 

• aleatory/epistemic 

• seismic 

OUTPUT

Once the guidance in this step has been applied, those terms for which a definition is readily 

available have been identified along with those that do not have documented sources. 

Because a term has consensually established sources for its definition, this does not mean it 

should be excluded from the main glossary. The term may be essential to understanding risk 

communication, for example, and on that basis should be included in the main glossary. In step 

1-7, these terms are reviewed for their importance to determine if they should remain in the 

main glossary. 

2.2.5 Step 1-5: Multiple Term Definitions 

This step determines if the terms identified in Step 1-3 as having a risk-context specific 

definition have more than one risk-context definition. Using a non-risk term for illustration, the 

term “fathom” has two meanings. One, a fathom is a nautical term meaning 6 feet; however, 

second, “to fathom” means to understand. The purpose of this step is to identify such terms. 

As such, this step does not serve any screening purpose. It is simply a checkpoint to identify if 

a candidate term has multiple definitions. 

The guidelines developed to determine if a term has multiple definitions include: 

1. reviewing the definitions 

2. performing a peer review 
2. APPROACH 

 2-9 

Guideline 1-5.1 – Definition Review

This guideline is used to review the definitions for the candidate terms. For each source, the 

definitions are reviewed to determine how each term is used. In some documents, this review is 

straightforward because the document provides a list of definitions or a glossary. In other 

cases, an actual definition may not be provided and a definition (or possible definitions) must be 

inferred. In these situations, the usage of the term may be very clear; however, there may be 

times when the usage is not clear and the intent of the term is ambiguous. For each term, the 

definitions are reviewed to determine if there are multiple definitions. 

Guideline 1-5.2 – Peer Review 

This guideline is used to perform a peer review. The objective of this review is to perform a 

sanity check. Reference documents may have been missed or not reviewed during the 

definition review resulting in terms with multiple definitions. Consequently, a peer review is 

performed to identify this gap. For each term, the reviewer identifies if there are multiple 

definitions based on his or her experience. If multiple definitions are identified, these additional 

definitions are added to the glossary (unless the term is eventually excluded from the glossary). 

In addition, the reviewer looks at the associated sources and, based on his or her experience, 

identifies any additional relevant documents that need to be included as a source. If additional 

sources are identified, the definitions are reviewed per Guideline1-5.1. 

OUTPUT

Once the guidance in this step has been applied, terms with multiple definitions have been 

identified (terms are not screened in this step). Regardless of whether a term has one or more 

definitions, it still needs to be reviewed to determine if it should remain in the main glossary. 

This step ensures the completeness of the definitions of the candidate terms for the main 

glossary. 

2.2.6 Step 1-6: Consensually Established Definitions 

This step determines if there is agreement on meaning in a risk context for the terms reviewed 

in Step 1-5. If there is agreement about a term’s meaning, then it can be assumed it has been 

consensually established. It cannot be assumed a term has been consensually established 

because it is defined in a standard. The definition has only been agreed upon for how the term 

is used in that standard. There may be disagreement or controversy in other uses 

(e.g., different uses in different standards, technical reports, guidance documents, or 

regulations). The challenge of this step is identifying the different uses of each term. 

The guidelines developed to determine if there is agreement include: 

1. reviewing the definitions 

2. performing a peer review 

The two guidelines of Step 1-6 are the same as Step 1-5. In general, determining if a term has 

a consensually established definition is performed at the same time as determining if a term has 

multiple definitions. 
2. APPROACH 

 2-10 

Guideline 1-6.1 – Definition Review 

This guideline is used to review the definitions for the candidate terms. For each source, the 

definitions are reviewed to determine how each term is used. In some documents, this review is 

straightforward because the document provides a list of definitions or a glossary. In other 

cases, an actual definition may not be provided and a definition (or possible definitions) must be 

inferred. In these situations, the usage of the term may be very clear; however, there may be 

times when the usage is not clear and the intent of the term is ambiguous. For each term, the 

definitions are reviewed for consistency. Terms whose usage has been consistent are those 

whose meanings have been consensually established. 

In this step, there also may be terms with multiple legitimate definitions. These multiple 

definitions are reviewed for consistency. Using the example of “fathom” from Step 1-5, both 

definitions would be reviewed to determine if both have been consistently used. 

Guideline 1-6.2 – Peer Review 

This guideline is used to perform a peer review. The objective of this review is to perform a 

sanity check. Reference documents may have been missed or not reviewed that could result in 

terms with definitions in disagreement or inconsistent usage. Based on experience, the peer 

reviewer is aware of disagreements or inconsistencies. Consequently, a peer review is 

performed to identify this gap. For each term, the reviewer looks at the associated sources, and 

based on his or her experience and expertise, identifies any additional relevant documents that 

need to be included as a source. If additional sources are identified, the definitions are 

reviewed under Guideline 1-6.1. 

OUTPUT

Once the guidance in this step has been applied, those terms whose meaning has been 

consensually established have been identified along with those in disagreement, 

misunderstanding, or controversy. 

If there is no disagreement, misunderstanding, or controversy about a term’s meaning, it should 

not necessarily be rejected from the main glossary. The term’s meaning may be essential to 

understanding risk communication, and on that basis it should remain as a candidate for the 

main glossary. In the next steps, these terms are further reviewed to determine if they should 

remain in the main glossary. 

2.2.7 Step 1-7: Term Fundamental to Risk Communication 

This step determines if any of the candidate terms from Steps 1-4 and 1-6, which would 

otherwise be excluded from the main glossary, should be retained because they are essential to 

a basic understanding of risk-informed activities. Terms that are potential candidates for 

exclusion from the main glossary include terms that have a well-documented definition source 

and that may or may not have a unique risk-context meaning. 

Similar to Step 1-2, this step somewhat depends on the intended audience for the glossary. 

The audience may range from senior executives to subject experts to lay personnel. 

Nonetheless, some terms are fundamental to the basic understanding of risk-informed activities, 

and for the main glossary to be most effective, these terms should be included. However, while 
2. APPROACH 

 2-11 

no special risk knowledge is presumed by the audience, a basic understanding of nuclear safety 

is assumed. 

The guidance developed to identify fundamental terms include: 

1. terms frequently used to communicate results of risk analyses 

2. terms used in decisionmaking and communicating risk-informed decisions 

3. terms sometimes misused or used in confusing ways 

Guideline 1-7.1 – Terms Frequently Used to Communicate Risk Results

This guideline is used to identify terms that are frequently used and have a well-documented 

and consensual definition. Because of their frequent usage, these terms are useful to almost 

any audience for communicating the results of a risk analysis. A correct understanding of these 

terms is essential for accurately presenting results, as well as an accurate appreciation of what 

is being presented. Examples of terms used to communicate results include: 

• health effects 

• core damage frequency 

• large early release frequency 

• latent fatality 

Also included are terms used to describe the analysis that produced the results. Examples of 

terms used to describe the analysis include: 

• minimal cutset 

• dose response model 

Guideline 1-7.2 – Terms Used in Decisionmaking 

This guideline is similar to Guideline 1-7.1, but it addresses terms used more often in 

decisionmaking and communication of decisions than in PRA results. A correct understanding 

of these terms is essential for accurately communicating and understanding a risk-informed 

decision. 

Examples of terms useful for communicating and understanding decisions include: 

• deterministic acceptance criteria 

• high-level requirement 

• consequences 

• acute health effects 

Guideline 1-7.3 – Terms Misused 

This guideline is used to identify terms that are sometimes misused or used in confusing ways. 

Some terms may have established definitions in the published literature that are consistent with 

their risk-specific definition and are also fundamental to risk communication. However, their 

definitions may be complex or ambiguous (e.g., their usage is inconsistent with the intent of their 

meaning). 
2. APPROACH 

 2-12 

Examples of terms sometimes misused or used in confusing ways include: 

• probability 

• frequency 

• model uncertainty 

• internal event 

OUTPUT

Once the guidance in this step has been applied, the final list of terms for the glossary has been 

identified. 

Because a term has been identified as fundamental to a basic understanding of risk-informed 

activities does not mean that it should be included in the main glossary. The term may have 

policy implications and require a policy decision by the Commission. In the next step, these 

terms are reviewed for policy implications. 

2.2.8 Step 1-8: Policy Implications 

This step identifies if the terms from Steps 1-6 and 1-7 (which comprise the final list of terms) 

have policy implications. These terms may have risk-context definitions, definitions that are not 

consensually established, or may be terms fundamental to a basic understanding of risk 

communication. However, some of the terms may have policy implications and, therefore, its 

definitions could require Commission approval. These terms are identified and the policy 

implications discussed. 

A definition for a term is considered to have policy implications if it: 

• sets a precedent with broad ramifications 

• states new Commission expectations 

• deviates from current policy 

• is fundamental to other decisions 

Examples of terms that have potential policy issues include: 

• defense-in-depth 

• large release frequency 

OUTPUT

Once the guidance in this step has been applied, terms that have a definition with policy 

implications have been identified, along with those that do not. 

Because a term’s definition has policy implications does not mean it should be excluded from 

the glossary. These terms remain in the glossary; however, a formal definition of the term is not 

developed. At this step in the process, a final list of terms has been identified and the next task 

is developing the definitions. As definitions are developed, however, new terms may be added 

or others may be deleted. 
2. APPROACH 

 2-13 

2.3 Task 2: Development of Definitions 

Guidance was established for developing definitions. For example, for each term the glossary 

could provide a single definition or multiple definitions or the glossary could just document the 

various definitions found in the different sources. To meet the objective established for the 

glossary, such as reducing ambiguity and to be helpful to the user (regardless of the individual’s 

level of risk experience or expertise), the glossary should provide more than just a simple 

definition. Understanding the meaning of a term may require some explanation (e.g., the bases 

for different definitions, the relationship to a related term). As such, guidance was developed to 

optimize the usability of the glossary. This guidance involves three major steps. These steps 

are not necessarily performed in a sequential manner, but more in an iterative and integrated 

fashion. 

2-1 Develop Initial Definitions – For the terms identified from Task 1, the various 

definitions from the relevant sources are documented. 

2-2 Identify Related Terms – For each term, the definitions are reviewed and terms are 

cross-referenced (e.g., related definitions) or grouped (similar definitions), where 

appropriate. 

2-3 Finalize Definitions – For each term, although there may be published definitions, there 

may not be agreement; the bases for these disagreements are discussed. Furthermore, 

the definitions may not be easily understandable; therefore, a definition in plain English 

without the use of technical jargon is developed. 

This process is illustrated in Figure 2-2 and discussed in detail below. 

 

Figure 2-2 Process to develop definitions 

Output from 

Task 1 

List of Terms

Main Glossary 

and 

Hazard-Specific 

Glossaries

Step 2-1 

Develop Initial 

Glossary 

Step 2-2 

Identify Related 

Terms 

Step 2-3 

Finalize 

Definitions 
2. APPROACH 

 2-14 

2.3.1 Step 2-1: Develop Initial Glossary 

This step collects the different definitions that exist for the final list of terms and documents both 

the definitions and their sources. For some terms, multiple or different definitions may exist; 

these are included in the glossary as potential definitions. 

Most of the effort performed for this step was completed in Task 1. The main focus of this step 

is to document the definition (or definitions) associated with each term and its associated 

sources. 

OUTPUT

Once the guidance in this step has been applied, the reference sources and definitions for the 

final list of terms have been collected and documented. The result is an initial glossary; 

however, there may be discrepancies in definitions. There also may be terms with similar 

definitions that are used interchangeably. Furthermore, some terms may be closely related, and 

understanding both terms is necessary. Identifying these situations and including explanations, 

where appropriate, helps in understanding the terms. 

2.3.2 Step 2-2: Identify Related Terms 

This step identifies terms related in some manner. For some related terms, an understanding of 

the relationship between the terms is needed to fully understand each one. These terms need 

to be cross-referenced to compare their differences or similarities with regard to their meanings 

and applications. In addition, there are terms that convey the same or similar meaning. These 

terms should be grouped together to avoid redundancy in definitions. 

The guidance developed to identify related terms includes: 

1. cross-referenced terms 

2. grouped terms 

Guideline 2-2.1 – Cross-Reference of Terms 

This guideline is used to identify those terms that are related and should be cross-referenced in 

the glossary. These terms are: 

• similar, but they have different meanings, and there are instances in which they are used 

incorrectly (e.g., “probability” versus “frequency,” “core damage” versus “core melt”) 

• related, in that one is the opposite of the other (e.g., “deterministic” and “probabilistic”) 

• closely related in meaning, and may be a subset of the related term(e.g., “harzard,” 

“external hazard” and “internal hazard”) or an example of the related term (e.g., “risk 

analysis” and “probabilistic risk assessment”, “accident consequence” and “health 

effects”). 

• closely related, in that understanding one term depends on understanding the related 

term (e.g., “risk analysis” and “probabilistic risk assessment”, “accident consequence” 

and “health effects”). 
2. APPROACH 

 2-15 

These terms are identified and cross-referenced. In some cases, a definition is provided for 

each term, and the reason for the cross-reference is discussed. In other cases, the term is 

cross-referenced to another term and the definition is provided by the cross-referenced term 

For example: 

• “Risk-informed approach” is a term in the glossary. A definition is not provided for this 

term. However, in the Discussion Column, it is noted that this term “is related to the term 

risk-informed and is defined under “Risk-Informed”.” 

• “Probability” is a term in the glossary. A definition is provided and there is a 

cross-reference to frequency. 

Guideline 2-2.2 – Grouping of Terms 

This guideline is used to identify related terms that should be grouped in the glossary. This 

grouping assists the reader so that it is clear which terms should be defined together. These 

are terms that are: 

• composed of multiple words with the same adjective (e.g., “significant,” “significant 

contributor,” “significant basic event”; “accident sequence class,” “accident sequence 

type,” “accident sequence group”) 

• similar and convey the same meaning (e.g., “general transient” and “transient”) 

• complements of each other (e.g., “availability” and “unavailability”) 

These terms are identified and grouped. A single definition is provided in the glossary. When 

the group is a result of a common adjective, then the terms being modified are discussed where 

appropriate. 

OUTPUT

Once the guidance in this step has been applied, terms that should be cross-referenced or 

grouped have been identified and the bases understood. With this accomplished, the actual 

definitions for each term need to be finalized and associated discussion included. 

2.3.3 Step 2-3: Finalize Definitions in Glossary 

This step finalizes the definitions and associated discussion to complete the glossary. 

The guidance developed to complete the glossary includes: 

1. modify definition 

2. develop discussion 

 
2. APPROACH 

 2-16 

Guideline 2-3.1 – Modify Definition 

This guideline is used to modify definition(s) of terms where necessary. As noted previously, 

the intent of the glossary is not to recreate definitions; consequently, if an appropriate definition 

exists, it is used as the starting point. However, it is the objective of the glossary that the 

definition provided is stated in plain English with little-to-no reliance on technical jargon. 

The definitions collected from the various sources generally use technical jargon and are often 

written in a complex manner. For these terms, while the definition may be accurate, it can be 

difficult to understand and sometimes requires a specific expertise to understand. In these 

instances, the definition is modified and written in plain language (e.g., use of common, 

everyday words in short sentences) so that it can be clearly understood. The more complex 

definition (or explanation) is provided as part of the discussion. 

Guideline 2-3.2 – Develop Discussion 

This guideline is used to add the necessary discussion to each term to assist the user in 

understanding the meaning of the term as used in risk communication, where appropriate. 

Discussion might include: 

• The definition provided is written in plain English, but does not necessarily provide a 

complete risk context. In these cases, the risk (in particular, PRA) context is explained 

in the discussion. 

• There may be multiple legitimate definitions, each of which is included in the glossary. 

While the appropriate definition(s) are provided, a discussion is also included to explain 

why certain definitions are not appropriate. 

• There may be discrepancies and inconsistencies among the definitions. An explanation 

is provided for the discrepancies or differences in use of the term. 

• Where terms are cross-referenced, the reason for the cross-reference is provided where 

necessary (i.e., for some cross-references, the basis does not require an explanation, 

but for others, an explanation may be necessary to understand the relationship). 

Further, a discussion may be needed on the use of the terms. Examples include: 

⎯ “Probability” and “frequency” are similar terms, but they have different meanings 

and it is not surprising that these terms should be cross-referenced. However, one 

of the main reasons for the cross-reference is to remind the reader that these terms 

are not the same and have been incorrectly used in the past. 

⎯ “External events” and “external hazards” have the same meaning and should be 

cross-referenced. In current usage, external hazard has replaced the term external 

event, therefore a cross-reference was essential to emphasize this matter. 

• There may be some terms whose definitions have policy implications(e.g., “defense-indepth”). For these terms, the various definitions are provided; however, a single 

definition is not defined. Further, a discussion of why the definition is considered to have 

policy implications is provided. 
2. APPROACH 

 2-17 

OUTPUT

Once the guidance in this step has been applied, a complete glossary, given the scope and 

limitations of the document, has been developed. A definition(s) has been developed for each 

term. Discussion has been included to provide explanation and clarity to assist in 

understanding the meaning of each term. Terms whose definitions have policy implications also 

have been identified. 

3. ABREVIATIONS AND ACRONYMS 

 3-1 

In risk communication, there are abbreviations and acronyms that have become common and in 

some cases, often are not defined. Table 3-1 is not meant to be all inclusive; it is meant to 

identify the more commonly used abbreviations or acronyms. 

Table 3-1 Commonly Used Abbreviations and Acronyms 

ACRS Advisory Committee on Reactor Safeguards 

ANS American Nuclear Society 

APET accident progression event tree 

ASME (formerly) American Society of Mechanical Engineers 

ATWS anticipated transient without scram 

BE basic event 

BWR boiling-water reactor

CCDF complementary cumulative distribution function 

CCDP conditional core damage probability 

CCF common-cause failure 

CD core damage 

CDF core damage frequency 

CDP core damage probability 

CET containment event tree 

CLERP conditional large early release probability 

CM core melt 

CMF common-mode failure 

core-melt frequency 

CRM configuration risk management 

CY calendar year 

DBA design-basis accident 

DBE design-basis earthquake 

design-basis event 

DCF dose conversion factor 

DCH direct containment heating 

EAB exclusion area boundary 

EP emergency preparedness 

EPRI Electric Power Research Institute 
3. ABREVIATIONS AND ACRONYMS 

 3-2 

Table 3-1 Commonly Used Abbreviations and Acronyms 

ET event tree 

F&B feed and bleed (bleed and feed) 

FM failure mode

FMEA failure modes and effects analysis 

FT fault tree 

HCLPF high confidence in low probability of failure 

HEP human error probability

HFE human failure event 

HLR high-level requirement 

HPME high-pressure melt ejection 

HRA human reliability analysis 

IE initiating event

IM importance measure

ISLOCA interfacing-systems loss-of-coolant accident 

LBE licensing-basis event 

LERF large early release frequency 

LOCA loss-of coolant accident 

LOOP, 

LOSP 

loss of offsite power; loss of station power 

LP/SD low power/shut down 

LWR light-water reactor 

NEI Nuclear Energy Institute 

NPP nuclear power plant 

NRC U.S. Nuclear Regulatory Commission 

OG owners group 

PDS plant damage state 

POS plant operational state 

PRA probabilistic risk assessment (base, baseline) 

PWR pressurized-water reactor 

QA quality assurance 

QHO quantitative health objective 
3. ABREVIATIONS AND ACRONYMS 

 3-3 

Table 3-1 Commonly Used Abbreviations and Acronyms 

RAW risk achievement worth 

RG regulatory guide 

RIDM risk-informed decisionmaking 

RY reactor-year 

SA systems analysis 

SB, SBO station blackout 

SGTR steam generator tube rupture 

SM seismic margin 

SOKC state-of-knowledge correlation 

SR supporting requirement 

ST source term 

4. GLOSSARY

 4-1 

4.1 Understanding the Format and Structure 

This section describes the format and structure used in listing and defining the terms in the 

glossary. The understanding of this process is essential to being able to use the glossary 

effectively and efficiently. 

For each term, a definition is provided in plain English with little-to-no technical jargon so that 

understanding does not depend on an individual’s level of risk experience or expertise. In 

addition, a discussion is provided to include explanations to help the user understand the 

meaning of the term. This discussion, where appropriate, generally includes: 

• the definition of the term in a risk context 

• the different definitions of the term 

• how the term has been and should be used 

• how the term relates to other terms for a complete understanding of the definition 

The table contains two columns. The first column provides the term and its definition and any 

appropriate cross-references; the second column contains the discussion of the term. 

Where appropriate, terms are grouped, related, and cross-referenced as follows: 

• When a term is related to another term(s), these related terms are referenced at the end 

of the definition of the main term. These related terms are not subsets of the main term 

or synonyms. The main term and related term each has a distinctly different definition; 

consequently, each term is listed separately in the glossary with its own definition. 

Furthermore, the reason for referencing any related term(s) is discussed in the 

discussion column. 

For example, common-cause failure and common-mode failure are related terms, but 

they have very different meanings: 

 TERM AND 

DEFINITION 

DISCUSSION 

Common-Cause Failure 

Main term 

Main term 

definition 

Related terms 

Discussion on 

relationship for 

the two terms

A failure of two or 

more structure, 

system or 

components as a 

result of a single 

shared cause. 

(see CommonMode Failure, 

Failure Mode)

In a PRA, common-cause failure (CCF) is a special form of 

dependent failure in which the failure of the structure, system, or 

component (SSC) has occurred from the same fault. CCF faults 

generally reflect errors occurring as a result of a common 

manufacturer, environment, maintenance, etc. 

The CCF term is often incorrectly used interchangeably with 

common-mode failure (CMF). CCF only accounts for the SSCs 

failing because of the same, single cause, not if they ultimately 

fail in the same manner (or in the same mode), which is CMF. In 

data provided to quantify CCF events, the failure mode is usually 

presented (i.e., failure to start, fail to run), and the cause is not 

always provided about why the failure mode occurs. There could 

be multiple causes lumped into the data presentation for a given 

failure mode. Thus, the available failure data dictate whether the 

PRA model is modeling CCF or CMF. 
4. GLOSSARY

 4-2 

Moreover, for all the related terms, there is a cross-reference for each related term. For 

example: 

TERM AND DEFINITION TERM AND DEFINITION 

Common-Cause Failure 

A failure of two or more structure, 

system or components as a result 

of a single shared cause. (see 

Common-Mode Failure, Failure 

Mode)

 Common-Mode Failure 

A failure of two or more structures, 

systems, or components in the 

same manner or mode as the result 

of a single shared cause. (see 

Common-Cause Failure, Failure 

Mode)

 

• When a term is a subset of another term (i.e., main term), this subset term is listed with 

the main term. It is also listed separately in the glossary and cross-referenced to the 

main term. In these cases, no definition appears with the subset term; the subset term is 

defined with the main term. The definition provided is broad enough to encompass the 

main term and all the subset terms. 

For example, early containment failure (ECF) is a subset of containment failure, which is 

the main term; therefore, ECF is cross-referenced to the main term, containment failure, 

where it is defined. 

Subset term 

Crossreference to 

main term 

TERM AND 

DEFINITION 

 TERM AND DEFINITION Main

term

Subset 

terms 

Definition 

of main

term 

Early Containment 

Failure 

(see Containment 

Failure) 

 Containment Failure 

(Early, Late) 

Loss of integrity of the 

containment from a core 

damage accident that is 

expected to result in an 

unacceptable release of 

radioactive materials. (see 

Containment, Containment 

Bypass, Containment 

Pressure Boundary)

Late Containment 

Failure 

(see Containment 

Failure) 

No definition provided 

For situations in which terms are a subset of another term, the specific words causing 

the term to be a subset are enclosed in parentheses after the main term. The definitions 

for the subset terms are provided in the discussion column. For example, early

containment failure is defined under containment failure. However, the definition 

provided is for containment failure, and the distinction and definition of early containment 

failure is provided in the discussion column. 
4. GLOSSARY

 4-3 

Main term 

Subset terms 

Definition of 

main term 

Discussion and 

definition of 

subset terms 

TERM AND DEFINITION DISCUSSION 

Containment Failure (Early, Late) 

Loss of integrity of the 

containment from a core 

damage accident that is 

expected to result in an 

unacceptable release of 

radioactive materials. (see 

Containment, Containment 

Bypass, Containment 

Pressure Boundary)

In a PRA, determining when and if the containment fails 

or is bypassed during a severe accident is very 

important from a risk perspective. If the containment 

pressure boundary remains leak-tight, the offsite 

consequence will be low. Conversely, if the 

containment fails or is bypassed, then the consequence 

to the surrounding population can be potentially high. 

For specific containments there can be selected severe 

accident scenarios in which the containment fails before 

fission products have penetrated the primary system. If 

the accident is successfully arrested at this point, no 

release will occur. However, usually containment failure 

represents the failure of the final barrier preventing a 

radioactive material release. 

 

Containment failure is often categorized as early or late. 

Early containment failure occurs in a timeframe before 

the surrounding population within 1 mile of the site 

boundary can be evacuated. Late containment failure 

occurs in a timeframe that allows the surrounding 

population from 1 to 10 miles to be evacuated.

• When a term has the same meaning as another term, the terms are grouped. The terms 

generally are listed in the group alphabetically, unless one of the terms is prevalently 

used over the other terms. The prevalently used term is then listed first in the group. 

The first term listed in the group is the main term. The latter terms of the group are 

separately listed in the glossary where they are cross-referenced to the main term. No 

definition is provided for these latter similar terms; they are defined with the main term. 

For example, “accident sequence class,” “accident sequence group,” and “accident 

sequence type” are similar terms and are grouped together, with “accident class” 

identified as the main term. The terms are grouped alphabetically, and the first term is 

the main term. 

TERM AND 

DEFINITION

 TERM AND DEFINITION Main

 term

Similar 

terms 

Definition 

of main

term 

Similar term 

Cross-reference 

to main term 

Accident 

Sequence 

Group 

See Accident 

Sequence Class 

Accident Sequence Class, 

Accident Sequence Group, 

Accident Sequence Type, 

Event Sequence Class, 

Event Sequence Group, 

Event Sequence Type

A grouping of accident 

sequences with similar 

characteristics or end states. 

(see Accident Sequence)

Accident 

Sequence 

Type 

See Accident 

Sequence Class 

 

 No definition provided 
4. GLOSSARY

 4-4 

For situations in which terms are similar or synonymous and are grouped, the terms are 

separated by commas, which indicates that they are terms with the same meaning and 

are not related terms. Moreover, the discussion will explain the reason for the grouping. 

TERM AND 

DEFINITION

DISCUSSION 

Accident Sequence Class, Accident Sequence Group, Accident 

Sequence Type, Event Sequence Class, Event Sequence Group, Event 

Sequence Type 

Term 

Group 

Definition 

for group 

A grouping of 

accident sequences 

with similar 

characteristics or 

end states. (see 

Accident Sequence)

In a PRA, the accident sequences typically are combined into 

accident sequence classes (groups or types). For example, an 

accident sequence class might represent a set of accident 

sequences with similar initiating events (e.g., loss-of-coolant 

accidents (LOCAs), loss of offsite power (LOOP), loss of heat 

removal or similar safety function responses. The purpose for 

combining like sequences is generally done to understand the 

type of sequences contributing to the risk. 

The terms accident sequence class, accident sequence group, 

and accident sequence type are similar in meaning and often 

correctly used interchangeably. Moreover, accident sequence is 

also used interchangeably with event sequence. Consequently, 

the terms event sequence class, event sequence group, and 

event sequence type also are similar in meaning and used 

interchangeably. 

• There also may be instances of multiple subsets. For example, hazard type, hazard 

group, and hazard event are all subsets of hazard; internal and external hazards are 

both subsets of hazard type and hazard and are each listed in the glossary. There is still 

a single main term and the definition is provided for the main term. Each subset term is 

still separately listed, cross-referenced to the main term, and no definition is provided. 

The specific words causing the term to be a subset are still enclosed in parentheses 

after the main term. The definitions for the subset terms are provided in the discussion. 
4. GLOSSARY

 4-5 

For example: 

Subset terms for 

hazard 

Subset terms 

for hazard 

type 

Term Term Main

term 

Subset

terms 

Definition

of main

term 

Hazard Type 

(see Hazard) 

Hazard (Type (Internal, 

External), Group, Event) 

Anything that has the potential to 

cause an undesired event or 

condition that leads to equipment 

damage.

Hazard Group 

(see Hazard) 

No definitions provided 

Hazard Event 

(see Hazard) 

External Hazard

(see Hazard) 

Internal Hazard

(see Hazard) 

Discussion

In a PRA, there are three different uses of the term hazard as an adjective (the terms hazard and plant hazard tend to 

be correctly used interchangeably): types, groups, and events. The first, hazard type, classifies hazards as either 

internal or external to the plant. Within each hazard type, internal and external, there are subcategories, which are 

referred to as hazard groups. For internal hazards, this hazard group includes internal events, internal floods, and 

internal fires. For external hazards, this includes seismic events, high winds, external floods, and other external 

hazards. Finally, a hazard event represents the events brought about by the occurrence of the specified hazard. For 

example, those of interest in a PRA are ones that directly or indirectly cause an initiating event and may further cause 

safety system failures or operator errors that may lead to core damage or radioactive material release. 

As defined in Regulatory Guide 1.200 (Ref.91), a hazard group “is a group of similar causes of initiating events that 

are assessed in a PRA using a common approach, methods, and likelihood data for characterizing the effect on the 

plant.” 

A hazard event is described in terms of the specific levels of severity of impact that a hazard can have on the plant. 

The hazard event is an occurrence of the phenomenon that can result in a plant trip and possibly other damage when 

the plant is at-power or result in the loss of a key safety function during non-power operations. The ASME/ANS PRA 

Standard (Ref. 2) states that there “is a range of hazard events associated with any given hazard, and, for analysis 

purposes, the range can be divided into bins characterized by their severity.” An example of the overall concept of 

hazard, hazard event, and initiating event is as follows: 

• Earthquakes are a hazard;

• 0.1g, 0.3g, 0.5g earthquakes and their associated spectral shapes and time histories may be defined as 

hazard events;

• A manual plant trip is typically the initiating event for the 0.1g earthquake, and a loss of offsite power is 

typically assumed as the initiating event for the 0.3g and 0.5g earthquakes.

The ASME/ANS PRA Standard (Ref. 2) defines a hazard as “an event or a natural phenomenon that poses some risk 

to a facility. Internal hazards include events such as equipment failures, human failures, and flooding and fires 

internal to the plant. External hazards include events such as flooding and fires external to the plant, tornadoes, 

earthquakes, and aircraft crashes.” 
4. GLOSSARY

 4-6 

4.2 Terms and Definitions 

Table 4-1 provides the terms and their definitions with the associated discussion. The terms are 

listed alphabetically. Hazard-specific terms are listed, but their definitions are provided in the 

noted appendix. 

Table 4-1 Term and Definition(s)

TERM AND 

DEFINITION DISCUSSION 

Accident Consequence 

The health effects 

or the economic 

costs resulting 

from a nuclear 

power plant 

accident. (see 

Health Effects, 

Accident 

Consequence 

Analysis)

In a Level 3 PRA, the consequences can be measured by health effects and economic costs 

resulting from a nuclear accident. 

The accident consequences analyzed in a risk analysis generally involve evaluating the extent 

to which the health of the surrounding population or the condition of the surrounding 

environment is affected. The health effects and economic costs of a nuclear accident can be 

incurred both on the plant site as well as in the surrounding community. In most cases, the 

focus is on offsite consequences (i.e., (1) radiation doses from various exposure pathways and 

consequent health effects to the public, and (2) the economic costs associated with protective 

measures, such as evacuation and relocation of the public, destruction of contaminated 

foodstuffs, and decontamination or interdiction of contaminated land and property). 

Accident Consequence Analysis 

The calculation of 

the extent of 

health effects or 

the economic 

costs resulting 

from a nuclear 

power plant 

accident. (see 

Accident 

Consequence)

In a PRA, the accident consequence analysis is the actual quantification of the potential 

magnitude of health effects and/or economic costs that can result from a nuclear accident. 

Accident consequence analysis attempts to answer the third of the three questions used to 

define risk: (1) What can go wrong? (2) How likely is it? (3) What might be its consequences?

Accident Event Sequence 

 (see Accident 

Sequence)

The term accident event sequence has the same meaning as accident sequence and is defined 

under “Accident Sequence.” 

Accident Mitigation 

Actions taken to 

reduce the severity 

of an accident. 

(see Accident 

Prevention, 

Emergency 

Preparedness, 

Emergency 

Response) 

In a PRA, accident mitigation typically refers to actions taken to reduce the severity of an 

accident once core damage has started, as opposed to actions to prevent a core damage event 

from occurring. Successful accident mitigation implies that a core damage event occurred, but 

its consequences were minimized. 

Some strategies used for accident mitigation include preventing fission product releases by 

maintaining barrier integrity, or reducing fission product releases by filtration. 

Also, accident mitigation measures typically refer to plans or actions taken on the plant site, 

while emergency preparedness measures and emergency response (e.g., evacuation, 

sheltering) refer to plans or actions taken to reduce exposure of onsite workers, as well as the 

surrounding population offsite.

Accident Precursor, Precursor Event 

A change in plant 

status that could 

lead to core 

damage accidents. 

A PRA is used to evaluate an event to determine if it will be considered an accident precursor. 

A conditional core damage probability (CCDP) is calculated for the event. The event is 

considered a precursor event, according to the NRC’s Performance and Accountability Report 

(Ref. 55), if the event “has a probability of greater than 1 in 1 million of leading to substantial 
4. GLOSSARY

 4-7 

TERM AND 

DEFINITION DISCUSSION 

damage to the reactor fuel.” An event is considered to be a “significant precursor” when the 

event “has a probability of 1 in 1,000 (or greater) of leading to substantial damage to the 

reactor fuel.” 

The terms accident precursor and precursor event generally have the same meaning. In some 

documents, the definition of accident precursor or precursor event includes quantitative criteria 

(e.g., as in the definition above), whereas some other definitions do not include quantitative 

criteria. 

Accident Prevention 

Actions taken to 

reduce the 

likelihood of an 

accident. (see 

Accident 

Mitigation)

In a PRA, accident prevention typically refers to actions taken to prevent a core damage event 

from occurring, as opposed to reducing the severity once core damage has started. Successful 

accident prevention implies that a core damage event does not occur. 

Some strategies used for accident prevention include: physical protection, maintaining plant 

stable operation, reactor protective systems, and maintaining barrier integrity. 

Accident Progression Event Tree 

A logic diagram 

that begins with 

the onset of core 

damage and 

identifies the 

potential 

responses of the 

containment and 

associated 

equipment, as well 

as operator 

actions, to the 

severe accident 

loads. (see Bridge 

Tree, Containment 

Event Tree, Event 

Tree)

In the PRAs documented in the NUREG-1150 (Ref. 51) series of reports, an accident 

progression event tree (APET) was used to analyze containment response to severe accident 

loads. An APET is a detailed representation of the containment response to severe accident 

loads, including the interaction of phenomena, the availability of equipment, and the 

performance of operators. For most modern PRAs, a containment event tree (CET), which is a 

less complex representation, is used to emphasize the status of the containment and 

containment equipment during a severe accident. The end states of both the APET and the 

CET are no containment failure, various containment failure modes, or containment bypass. 

Accident Scenario 

(see Accident 

Sequence)

The term accident scenario has the same meaning as accident sequence and is defined under 

“Accident Sequence.”

Accident Sequence Analysis, Event Sequence Analysis 

The process used 

to determine the 

series of events 

that can lead to 

undesired 

consequences. 

(see Accident 

Sequence)

In a PRA, accident sequence analysis is the process used to determine the combination of 

events that can lead to the undesired end state (e.g., core damage or large early release). The 

results of the accident sequence analysis are expressed in terms of individual accident 

sequences, each of which includes an initiating event followed by the necessary set of failures 

or successes of additional events (such as system, function, or operator performance) that will 

cause the undesired event. 

The terms accident sequence analysis and event sequence analysis are similar in meaning and 

often correctly used interchangeably. However, generally the terminology “accident” refers to 

leading to core damage, and the terminology “event” does not necessarily reflect a negative 

outcome such as core damage. 

The ASME/ANS PRA Standard (Ref. 2) defines accident sequence analysis as “the process to 

determine the combinations of initiating events, safety functions, and system failures and 

successes that may lead to core damage or large early release.” 
4. GLOSSARY

 4-8 

TERM AND 

DEFINITION DISCUSSION 

Accident Sequence Class, Accident Sequence Group, Accident Sequence Type , Event 

Sequence Class, Event Sequence Group, Event Sequence Type 

A grouping of 

accident 

sequences with 

similar 

characteristics or 

end states. (see 

Accident 

Sequence)

In a PRA, the accident sequences typically are combined into accident sequence classes 

(groups or types). For example, an accident sequence class might represent a set of accident 

sequences with similar initiating events (e.g., loss-of-coolant accidents (LOCAs), loss of offsite 

power (LOOP), loss of heat removal or similar safety function responses. The purpose for 

combining like sequences is generally done to understand the type of sequences contributing 

to the risk. 

The terms accident sequence class, accident sequence group, and accident sequence type are 

similar in meaning and often correctly used interchangeably. Moreover, accident sequence is 

also used interchangeably with event sequence. Consequently, the terms event sequence 

class, event sequence group, and event sequence type also are similar in meaning and used 

interchangeably. 

Accident Sequence Frequency 

(see Frequency) Accident sequence frequency is a type of frequency used in PRA and is defined in the 

discussion under “Frequency.” 

Accident Sequence Group 

(see Accident 

Sequence 

Class)

The term accident sequence group has the same meaning as accident sequence class and is 

defined under “Accident Sequence Class.”

Accident Sequence Type 

(see Accident 

Sequence 

Class)

The term accident sequence type has the same meaning as accident sequence class and is 

defined under “Accident Sequence Class.”

Accident Sequence, Accident Event Sequence, Accident Scenario, Event Sequence, 

Event Scenario, Event Tree Sequence 

A series of events 

that can lead to 

undesired 

consequences. 

(see Accident 

Sequence 

Analysis, Severe 

Accident, End 

State, Event Tree)

In a PRA, this series of events (e.g., an accident sequence, scenario, or event sequence) 

refers to an event tree pathway that follows from a particular initiating event, through system 

and operator responses, and ultimately to a well-defined end state, such as core damage. If 

the end state involves extensive core damage and radioactive material release into the reactor 

vessel and containment, with potential release to the environment, the accident sequence 

would represent a severe accident sequence. The system and operator responses may 

involve success, failure, or both. 

The terms accident sequence, accident event sequence, accident scenario, event scenario, 

event sequence, and event tree sequence are similar in meaning and are often correctly used 

interchangeably. 

The ASME/ANS PRA Standard (Ref. 2) defines an accident sequence as “a representation in 

terms of an initiating event followed by a sequence of failures or successes, of events (such as 

system, function or operator performance) that can lead to undesired consequences with a 

specified end state (e.g., core damage or large early release).” 

The following figure is an example of an accident sequence: 
4. GLOSSARY

 4-9 

TERM AND 

DEFINITION DISCUSSION 

Active Component 

A component 

whose operation or 

function depends 

on an external 

source of power 

(e.g., air, electrical, 

hydraulic). (see 

Passive 

Component)

In a PRA, important elements of the model include both active and passive components. 

NUREG/CR-5695 (Ref. 74) defines active component as: “A component which normally is 

operating or can and should change state under normal operating conditions or in response to 

accident conditions (e.g., pumps, valves, switches).” 

Some examples of active components include pumps, fans, relays, and transistors. These are 

identified as active components because they rely on an external driving mechanism to perform 

their function. 

The IAEA Safety Glossary (Ref. 7) mentions “certain components, such as rupture discs, check 

valves, safety valves, injectors, and some solid state electronic devices, have characteristics 

that require special consideration before designation as an active or passive component.” This 

special consideration implies that some components are not easily labeled as either active or 

passive because they may have characteristics of both. 

 

The ability to change state is sometimes considered as the defining characteristic of whether a 

component is active or passive. For example, a check valve normally has a passive function, 

but in a safety injection system it could be considered active since it needs to open and then 

reclose to prevent backflow. 

Acute Exposure 

(see Exposure) The term acute exposure is a type of exposure and is defined in the discussion under 

“Exposure.” 

Acute Health Effects 

(see Health 

Effects)

The term acute health effect refers to a type of health effect and is defined in the discussion 

under “Health Effects.” 

Aging 

General process in 

which 

characteristics of a 

structure or 

component 

gradually change 

(e.g., degrade) 

with time or use. 

(see Bathtub 

Curve)

In a PRA, the aging of a component is generally not explicitly modeled but is sometimes 

assumed to be reflected in the failure probability used to represent the performance of the 

component. 

The performance of structures or components may degrade with time (e.g., increasing failure 

rates, new failure modes) because of wearout and exposure to environmental conditions. 

Aging can lead to increasing failure rates in the later stages of life of a component. During the 

early life (burn-in) of a component, failure rates can decrease until a plateau is reached, as 

seen in the bathtub curve. 

The definition provided is based on the definition in the IAEA Safety Glossary (Ref. 7).

IE Main 

Chute

Reserve 

Chute

This path is an example of an accident 

sequence: 

Initiating Event (Jump from Airplane)Main 

Chute fails  Reserve Chute fails  End 

State (Both chutes fail, jumper casualty)

End 

State

Main chute works, float to ground

Initiating Event:

Jump from 

Airplane

Reserve chute works, float to ground

Both chutes fail, jumper casualty

System succeeds

System fails
4. GLOSSARY

 4-10 

TERM AND 

DEFINITION DISCUSSION 

Air Submersion 

(see 

Cloudshine)

Air submersion has the same meaning as cloudshine and is defined under “Cloudshine.”

Aleatory Uncertainty

(see 

Uncertainty) 

The term aleatory uncertainty is a specific type of uncertainty and is defined under the term 

“Uncertainty.” 

Anticipated Transient Without Scram 

An event that 

requires a plant 

trip and challenges 

safety systems but 

is followed by 

failure of control 

rod insertion to 

terminate the 

fission process. 

(see Transient)

In a PRA, anticipated transient without scram (ATWS) is referred to as both the initiating event 

and an accident sequence class. When referring to ATWS as the initiating event, this includes 

the initiating event (e.g., failure of the feedwater system) and failure of the reactor protection 

system (RPS). When referring to ATWS as an accident sequence, this includes the initiating 

event, failure of the RPS, and failure of other methods for terminating the fission process (e.g., 

emergency boron injection for a boiling-water reactor). 

A few examples of definitions for ATWS include: 

• “An ATWS is one of the "worst-case" accidents, consideration of which frequently 

motivates the NRC to take regulatory action. Such an accident could happen if the 

scram system (which provides a highly reliable means of shutting down the reactor) 

fails to work during a reactor event (anticipated transient). The types of events 

considered are those used for designing the plant.” (NRC Web site Glossary, 

Ref. 36) 

• The Code of Federal Regulations formally defines ATWS as “an anticipated 

operational occurrence followed by the failure of the reactor trip portion of the 

protection system specified in General Design Criterion 20.” (10 CFR Part 50, 

Appendix A, Ref. 22) 

• “The event is a perturbation in the state of some system or component at full reactor 

power that initiates a deviation from the full-power, steady-state operating conditions 

that have been previously considered and analyzed, which would normally result in a 

reactor scram. However, in this case, the reactor does not scram, either 

automatically or manually.” (NUREG-1742, Ref. 59) 

 

As-Built As-Operated (As-Designed) 

The accurate and 

current design and 

operation of the 

plant. (see PRA 

Configuration 

Control, Living 

PRA, Plant 

Configuration 

Control)

When applied to a PRA, as-built as-operated refers to the fidelity of the PRA model matching 

the current plant design, configuration, procedures, and performance data (e.g., component 

failure rates). Similarly, as-designed refers to the PRA matching the plant configuration in the 

design certification or combined operating license stage, in which the plant is not yet built or 

operated. 

Because the plant’s configuration and operating procedures are continuously upgraded and 

modified and operating experience is accrued, the PRA model needs to be updated from time 

to time to reflect the as-built, as-operated plant. In that case, the model is said to be up-to-date 

(i.e., current). A PRA that is continuously updated to incorporate plant changes is called a 

living PRA. 

In the ASME/ANS PRA Standard (Ref. 2), as-built as-operated is defined as “a conceptual term 

that reflects the degree to which the PRA matches the current plant design, plant procedures, 

and plant performance data, relative to a specific point in time.” 
4. GLOSSARY

 4-11 

TERM AND 

DEFINITION DISCUSSION 

As-Designed 

(see As-Built AsOperated) 

The term as-designed is defined in the discussion of the term “As-Built As-Operated.” 

Assumption (Key) 

A decision or 

judgment that is 

made in the 

development of a 

model or analysis. 

(see Model 

Uncertainty)

In a PRA, an assumption is either related to a source of model uncertainty or to scope or level 

of detail. An assumption related to a model uncertainty is made about the choice of the data, 

approach, or model used to address an issue because there is no consensus. A credible 

assumption is one that has a sound technical basis, such that the basis would receive broad 

acceptance within the relevant technical community. An assumption related to scope or level 

of detail is one that is made for modeling convenience. 

An assumption is considered to be key to a risk-informed decision when it could affect the PRA 

results that are being used in a decision and, consequently, may influence the decision being 

made. An effect on the PRA results could include the introduction of a new functional accident 

sequence or other changes to the risk profile (e.g., overall core damage frequency or large 

early release frequency, event importance measures). Key sources of model uncertainty are 

identified in the context of an application. 

The definition provided is based on the definition in the ASME/ANS PRA Standard (Ref. 2). 

The NRC Web site Glossary (Ref. 36) states, “in the context of individual plant examinations 

(IPEs), individual plant examinations for external events (IPEEE), and probabilistic risk 

assessments (PRAs), assumptions are those parts of the mathematical models that the analyst 

expects will hold true for the range of solutions used for making decisions.” 

Atmospheric Transport and Diffusion 

The movement 

and variation in 

concentration of a 

radioactive plume 

after release to the 

environment. (see 

Atmospheric 

Transport and 

Diffusion Analysis, 

Level 1,2,3 PRA)

In a PRA, assumptions about atmospheric transport and diffusion of the radioactive plume are 

used in the calculation of the health effects or economic consequences of a severe accident. A 

Level 3 PRA takes the result of a Level 2 PRA (frequencies, amounts, timing durations, and 

energies of radioactivity releases) and produces offsite consequences (health effects, 

economic consequences) as output. 

To calculate the offsite consequences, the movement and concentration of the radioactive 

plume under various weather conditions (e.g., high winds, rain) has to be determined. The 

plume characteristics can then be combined with the population information to calculate the 

health effects. The plume characteristics also can be used to determine land contamination 

and economic consequences of a severe accident. 

Atmospheric Transport and Diffusion Analysis 

An analysis to 

determine the 

movement and 

concentration of a 

radioactive plume. 

(see Atmospheric 

Transport and 

Diffusion)

In a Level 3 PRA, atmospheric transport and diffusion (ATD) models are used in the 

consequence calculations. ATD models range from simple straight-line, steady-state Gaussian 

dispersion models, which calculate ground-level instantaneous and time-integrated airborne 

concentrations in the plume, to more sophisticated models that allow terrain-dependent effects 

and temporal variations in wind speed and atmospheric stability. Probabilistic consequence 

modeling codes typically include sampling of meteorological data from a site-specific annual 

database of hourly weather data to determine appropriately weighted scenarios of plume 

transport under different weather conditions to provide probabilistic results. 

At-Power 

The state of 

operation in which 

the reactor is 

critical and 

A PRA models the different plant operating states (POSs), generally defined as at-power, 

low-power, and shutdown. These POSs are distinguished in the PRA model because the plant 

responses (e.g., accident sequences) are different. 
4. GLOSSARY

 4-12 

TERM AND 

DEFINITION DISCUSSION 

producing power 

from a range of 

states between full 

and low power. 

(see Full Power, 

Low Power/ 

Shutdown, Plant 

Operational State)

At-power plant status includes all power levels above low-power. In this instance, the reactor is 

producing a significant amount of power from fission in the core fuel, above and beyond the 

decay heat levels. The safety systems are on automatic actuation and not blocked or defeated 

(as they might be in low-power and shutdown states). The support systems are aligned in their 

normal configuration (e.g., electric power is being drawn from the grid). These are all important 

initial conditions for PRA modeling. 

The borderline between at-power and low-power and shutdown depends on plant evolution (the 

changes in configuration used to bring the plant down from full power or up from low-power and 

shutdown) and is typically on the order of 15%-25% of full power. 

Historically, the term “full power” was used for all power levels between low-power and 100% 

power. This has been modified such that at-power now refers to intermediate power levels 

ranging from low-power and up to and including 100% power, while “full power” is reserved for 

just 100% reactor power. The figure below is a pictorial representation of the different plant 

operating states. 

Note: The overlap shows that PRAs have used different denominations for At-power and Lowpower. 

The ASME/ANS PRA Standard (Ref. 2) defines at-power as “those plant operating states 

characterized by the reactor being critical and producing power, with automatic actuation of 

critical safety systems not blocked and with essential support systems aligned in their normal 

power operation configuration.” 

Availability (Unavailability) 

The probability that 

a system, 

structure, or 

component of 

interest is 

functional at a 

given point in time. 

(see Reliability) 

In a PRA, unavailability is one of the attributes of a system, structure, or component that may 

affect the plant’s response to an initiating event. 

Unavailability is the complement of availability (i.e., shortfall between availability and unity). 

In the ASME/ANS PRA Standard (Ref. 2), unavailability is defined as “the probability that a 

system or component is not capable of supporting its function including, but not limited to, the 

time it is disabled for test or maintenance.” 

The definition provided is based on the definition in National Fire Protection Association 

(NFPA)-805 (Ref.11). 

Base PRA, Baseline PRA 

(see PRA) The terms base PRA and baseline PRA represent a specific type of PRA and are defined under 

“PRA.” 

Shutdown

(0%)

At Power

(generally from

~15% to 100%)

Low Power

(generally from

0% to ~25%)

Full Power

(100%)
4. GLOSSARY

 4-13 

TERM AND 

DEFINITION DISCUSSION 

Basic Event 

An element of the 

PRA model for 

which no further 

decomposition is 

performed 

because it is at the 

limit of resolution 

consistent with 

available data. 

(see Component, 

Fault Tree)

In a PRA, in developing the fault trees, the basic events represent those failures for which there 

is available data, and as such, represent the termination of a branch of the fault tree. There are 

typically two types of failures (or basic events): equipment unavailability and human errors. 

The term basic event can have other (more specific) definitions, as stated below: 

• “An event in a fault tree model that requires no further development, because the 

appropriate limit of resolution has been reached.” (Ref.2) 

• The individual events that collectively form a cutset, which is a combination of failures 

needed to result in the occurrence of a condition of interest (e.g., accident sequence, 

system failure). 

In the quantification process of the PRA, the model uses or manipulates the basic events to 

model the core damage frequency. At this point, the initiating event is part of the quantification 

process; consequently, an initiating event is sometimes referred to as a basic event. 

The following figure is an example of a basic event: 

Basic Event Failure Probability

(see Probability) The term basic event failure probability is a specific type of failure probability and is defined 

under “Probability.” 

Bathtub Curve

Graphical 

representation of 

failure rate time 

dependency in the 

life of a typical 

component. (see 

Aging)

In a PRA, the mid-life or constant failure rate stage in the life of a component is the one 

typically modeled. However, the life of certain types of components is often considered to have 

three stages of failure rate behavior: I) burn-in (or infant mortality) stage, characterized by 

failure rates decreasing with time, II) mid-life or constant failure rate stage, and III) wearout 

stage in which failure rates increase with time. These three stages together form a curve that 

looks like the cross-section of a bathtub. The following figure represents a bathtub curve: 

Pump A

failed

Pump C 

failed

Pump B 

failed

Pump D 

failed

Pump E 

failed

Pump Systems 

failed

Pump System I 

failed

Pump System II 

failed

A B C D E

Transfer to 

AC Power Failed 

FT

These are basic events in the fault tree.
4. GLOSSARY

 4-14 

TERM AND 

DEFINITION DISCUSSION 

• Region I – The failure rate is usually high at the beginning of a component’s life 

because of defects. It decreases if the component survives. 

• Region II – The failure rate becomes stable and remains constant in the middle of the 

component’s life. 

• Region III – The failure rate increases toward the end of the component’s life. 

Bayesian Analysis, Bayesian Estimation, Bayesian Statistics 

Type of data 

analysis in which 

an initial estimate 

about a parameter 

value is combined 

with evidence to 

arrive at a more 

informed estimate. 

(see Frequentist, 

Bayesian Update)

In a PRA, Bayesian analysis is commonly used in the computation of the frequencies and 

failure probabilities in which an initial estimation about a parameter value (e.g., event 

probability) is modified based on actual occurrences of the event. The initial parameter value 

may have a probability distribution associated with it. Thus, the event probability to be 

determined is based on a belief, rather than on occurrence ratios. Any actual occurrence or 

lack of occurrence of the event is used to measure consistency with the original hypothesis, 

which is then modified to reflect this evidence. The modified or updated hypothesis is the most 

meaningful estimate of the parameter. 

The initial hypothesis is called the “prior”. The prior should be as relevant as possible to the 

parameter value in question. The final parameter estimate will depend on the prior chosen to a 

certain extent. For example, industry average (generic) data may be used as the prior. 

Noninformative priors can be used if no basis for making an educated guess exists. The prior 

is modified by actual observations of the event occurrences (e.g., plant-specific data) to 

calculate the “posterior” or best estimate of the parameter. The process is called “Bayesian 

update.” 

Bayesian analysis is used when occurrences of an event are sparse or nonexistent, such that 

probability estimates using the proportion of actual event occurrences (frequentist approach) 

are not reliable. It also can be used to produce a probability distribution for the parameter in 

question. 

In risk analysis, both frequentist and Bayesian analysis may be used. Frequentist analysis is 

used when the occurrence data is sufficiently abundant, Bayesian analysis is used otherwise. 

The terms Bayesian analysis, Bayesian estimation, and Bayesian statistics are used 

interchangeably. 

Bayesian Estimation 

(see Bayesian 

Analysis) 

The term Bayesian estimation has the same meaning as Bayesian analysis and is defined the 

same as the term “Bayesian Analysis.” 

Bayesian Statistics 

(see Bayesian 

Analysis)

The term Bayesian statistics has the same meaning as Bayesian analysis and is defined the 

same as the term “Bayesian Analysis.” 

Failure Rate

Time

I II III
4. GLOSSARY

 4-15 

TERM AND 

DEFINITION DISCUSSION 

Bayesian Update 

Modification of a 

probability 

(frequency) of an 

event by 

incorporating 

additional 

observations of 

event occurrence. 

(see Bayesian 

Analysis)

In a PRA, Bayesian update is the process of using the Bayesian approach to incorporate new 

information and combine it with existing information to come up with a new characterization of 

the state-of-knowledge about a parameter. It is used to incorporate new information as it 

becomes available or to account for plant-specific information when primarily relying on generic 

data (or some other initial guess) to generate event failure probabilities or frequencies. For 

example, an initial guess of a pump failure rate is based on industry generic data. 

Observations of a certain number of failures (or no failures) of that type of pump over a certain 

time period in the plant are used in the Bayesian update to obtain a better estimate of the pump 

failure rate in that particular plant. 

Industry generic failure rates might be used as the starting estimate (called the prior). These 

would be combined with the observed occurrences of failure of such components to calculate 

the updated failure rates. A similar process may be used to obtain plant-specific initiating event 

frequencies, by starting from generic data and updating with plant-experienced occurrences to 

arrive at the updated initiating event frequencies. 

Best Estimate 

Approximation of a 

quantity based on 

the best available 

information. (see 

Mean, Point 

Estimate)

In a PRA, the term best estimate is not generally used. The term is sometimes mistakenly 

used in place of point estimate or mean value to characterize a parameter value estimate used 

in a PRA. 

The term is used for deterministic calculations, in which best estimate designates inputs or 

results obtained by using the most realistic assumptions available to the analyst (i.e., not 

biased by conservatism or optimism). For example, best estimate codes may be used to 

deterministically predict the pressure rise in containment from a hydrogen burn. 

Beyond-Design-Basis Accident 

A postulated 

accident that is 

more severe than 

those accidents 

used to establish 

the design of a 

nuclear facility. 

(see Design-Basis 

Accident, Severe 

Accident)

In a PRA, beyond-design-basis accidents (BDBAs) are a major focus of the analysis. For 

example, PRAs for currently operating light-water reactors (LWRs) have focused almost 

exclusively on BDBAs. Recent PRAs for proposed high-temperature graphite reactors have 

included design-basis accidents and anticipated occurrences in the analysis.

A nuclear facility must be designed and built to withstand a design-basis accident (DBA) 

without threatening public health and safety. However, the nuclear facility is not necessarily 

designed to withstand BDBAs. Therefore, an important role of PRA is to determine how a 

nuclear facility will behave in a BDBA and analyze the adequacy of the systems, structures, 

and components that are included to ensure public health and safety are maintained. Although 

BDBAs might exceed the design envelope, they do not necessarily result in significant core 

damage. Those BDBAs that do result in significant core damage are termed severe accidents. 

All severe accidents are by definition BDBAs since their challenges exceed the design 

envelope of the plant. 

The NRC Web site Glossary (Ref. 36) defines the term beyond-design-basis accident as “a 

technical way to discuss accident sequences that are possible but were not fully considered in 

the design process because they were judged to be too unlikely. (In that sense, they are 

considered beyond the scope of design-basis accidents that a nuclear facility must be designed 

and built to withstand.) As the regulatory process strives to be as thorough as possible, 

beyond-design-basis accident sequences are analyzed to fully understand the capability of a 

design.” 

Beyond-Design-Basis Event 

An event more 

severe than the 

In a PRA, beyond-design-basis events (BDBEs) represent conditions beyond the plant design 

envelope and, therefore, exceed the already considered anticipated transients (e.g., tripping of 
4. GLOSSARY

 4-16 

TERM AND 

DEFINITION DISCUSSION 

events for which 

the nuclear power 

plant was 

designed to 

withstand and 

specified in the 

safety analysis. 

(see Design-Basis 

Event, Severe 

Accident)

turbine generator), anticipated operational occurrences (AOOs), design-basis accidents 

(DBAs), and design-basis natural phenomena. 

A BDBE challenges the systems, structures, and components that are included in the design to 

ensure public health and safety. Generally, BDBEs have been excluded from the design-basis 

because they were considered to have a low probability of occurrence. Extremely unlikely 

earthquakes or aircraft impacts would be considered beyond-design-basis events which, while 

not considered in the nuclear plant design, can be analyzed in the PRA to determine how the 

plant would respond given such an event. 

Bin, Binning 

A group of 

initiating events or 

accident 

sequences with 

similar 

characteristics.

In a PRA, binning is a process used to group similar types of initiating events, accident 

scenarios, or sequences together to simplify the analysis. The term bin generally is associated 

with binning event tree sequences into groups that have similar characteristics and lead to 

similar end states called plant damage states. Initiating events also are grouped by similar 

characteristics (e.g., failure of a main steam isolation valve and failure of a feedwater pump are 

generally grouped (or binned) into a loss of feedwater initiator group). 

Bin is the actual group and binning is the process. 

Birnbaum Importance 

 (see Importance 

Measure) 

The term Birnbaum importance is one type of importance measure and is defined under 

“Importance Measure.” 

Bounding Analysis 

An analysis that 

uses assumptions 

such that the 

assessed outcome 

will meet or 

exceed the 

maximum severity 

of all credible 

outcomes, both in 

magnitude as well 

as frequency. (see 

Conservative 

Analysis)

In a PRA, a bounding analysis of a contributor or parameter may be performed to bound the 

risk or to screen the PRA item as a potential contributor to risk. When used for screening, the 

bounding analysis demonstrates that the item can be omitted from the PRA model because, 

even in the worst case, the impact on calculated risk is insignificant. 

As discussed in NUREG-1855 (Ref. 62), in the context of a specific PRA scope or level of detail 

item, a bounding analysis includes the worst credible outcome of all known possible outcomes 

that result from the risk assessment of that item. The worst credible outcome is the one that 

has the greatest impact on the defined risk metric(s). Thus, a bounding probabilistic analysis 

must be bounding both in terms of the potential outcome and the likelihood of that outcome. 

Consequently, a bounding analysis considers both the frequency of the event and the outcome 

of the event. 

NUREG-1855 states that if a bounding analysis is being used to bound the risk (i.e., determine 

the magnitude of the risk impact from an event), then both its frequency and outcome must be 

considered. However, if a bounding analysis is being used to screen the event (i.e., 

demonstrate that the risk from the event does not contribute to the defined risk metric(s)), then 

the event can be screened based on frequency, outcome, or both, depending on the specific 

event. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 2). 

Bridge Event Tree

(see Bridge 

Tree)

The term bridge event tree has the same meaning as bridge tree and is defined under “Bridge 

Tree.” 
4. GLOSSARY

 4-17 

TERM AND 

DEFINITION DISCUSSION 

Bridge Tree, Bridge Event Tree 

An event tree used 

to transfer 

information from 

one analysis stage 

to another in a 

manner that 

ensures the critical 

information is 

preserved. (see 

Containment Event 

Tree, Event Tree, 

Accident 

Progression Event 

Tree)

In a PRA, the most common use of bridge trees is in linking the core damage states, which are 

the end points of the Level 1 PRA analysis, with the plant damage states. The plant damage 

states often are used as the starting point of the accident progression event tree or the 

containment event tree (i.e., Level 2 analysis). In this case, the bridge trees provide the 

information on the status of systems that were not relevant for determining core damage, but 

that can influence further accident progression. The terms bridge tree and bridge event tree 

are similar in meaning and often correctly used interchangeably. 

The figure below is an example of a bridge tree: 

Capability Categories 

Categories used to 

indicate different 

levels of detail, 

plant-specificity, 

and realism in 

defining technical 

requirements for 

an acceptable 

PRA. 

For a PRA used with a risk-informed application, the level of detail, plant specificity, and 

realism needs to be commensurate with the scope of the specific application under 

consideration, as recognized in NRC Regulatory Guide 1.200 (Ref. 91). 

Capability categories are used in the ASME/ANS PRA Standard (Ref. 2) to recognize that the 

various elements in the PRA model can be constructed to different levels of detail, levels of 

plant-specificity, and levels of realism. The PRA standard defines three categories of the 

acceptable level of detail, plant-specificity and realism, starting at the minimal for capability 

Category I, and increasing through Category II, and Category III. The use of capability 

categories supports the concept that a PRA needs only to have the scope and level of detail 

necessary to support the application for which it is being used, but it always needs to be 

technically acceptable. 

As stated in the ASME/ANS PRA Standard (Ref. 2), “as the capability category increases, the 

depth of the analysis required also increases.” As further stated in the ASME/ANS PRA 

Standard, “the level of conservatism may decrease as the capability category increases and 

more detail and more realism are introduced into the analysis. However, this is not true for all 

requirements and should not be assumed.” 

Chemical Element Group 

A group of 

radioactive 

materials with 

similar physical 

and chemical 

properties used to 

simplify the 

estimate for offsite 

health effects. 

(see Source Term)

In a PRA, the source term used to characterize the radioactive material release is based on the 

defined chemical element groups. 

During a core damage accident, the number of different radioactive materials released from the 

fuel, reactor vessel, and containment to the environment can be quite large. The number of 

radioactive materials considered can be reduced to a manageable size by grouping those with 

similar physical and chemical properties. For example, in NUREG-1150 (Ref. 51) the 

60 radionuclides considered in the consequence calculation were not dealt with individually in 

the source term calculation. Since some different elements behave similarly enough both 

Level -1 Core 

Damage 

Accident

Sequences 

(CDAS)

Bridge Event Tree 

(containment 

systems)

Plant 

Damage 

State

Level-2 

Containment Event 

Tree 

Source 

Terms

CDAS cutsets binned by 

reactor core coolant and 

containment status
4. GLOSSARY

 4-18 

TERM AND 

DEFINITION DISCUSSION 

chemically and physically that they can be considered together, the 60 isotopes were placed in 

nine radionuclide groups. These nine groups were treated individually in the source term 

analysis. 

Chronic Exposure 

 (see Exposure) The term chronic exposure is a type of exposure and is defined in the discussion under 

“Exposure.” 

Cloudshine 

Direct external 

exposure from 

radioactive 

material in the 

atmosphere. (see 

Exposure 

Pathways, Water 

Immersion, 

Groundshine, 

Inhalation, 

Ingestion, Skin 

Deposition) 

In a Level 3 PRA, cloudshine, also referred to as air immersion, is one of the assumed 

pathways by which an individual can receive doses in the consequence calculation. The 

pathways of exposure include: (1) direct external exposure from radioactive material in a 

plume, principally due to gamma radiation (air immersion or cloudshine), (2) direct exposure 

from radioactive material in contaminated water given to an individual immersed in the water, 

(3) exposure from inhalation of radioactive materials in the cloud and resuspended material 

deposited on the ground, (4) exposure to radioactive material deposited on the ground 

(groundshine), (5) radioactive material deposited onto the body surfaces (skin deposition), and 

(6) ingestion from deposited radioactive materials that make their way into the food and water 

pathway. 

Cohort 

A group of 

individuals that is 

defined by some 

statistical or 

demographic 

factor. (see 

Emergency 

Response)

In the emergency response modeling of a Level 3 PRA, a cohort is a subset of the offsite 

population that mobilizes or moves differently from others. The planning and analysis of the 

offsite response to a severe accident is driven by the demographics of the surrounding 

population (i.e., the attributes (e.g., age, location) of the various cohorts (e.g., school children, 

hospital patients, prisoners) and their potential for being exposed to severe health effects). 

Collective Dose

(see Dose) The collective dose is a summation of dose that is defined under “Dose.” 

Committed Dose Equivalent 

(see Dose 

Equivalent) 

The committed dose equivalent is one measure of dose that can be used to calculate the effect 

of radiation received by an individual and is defined under “Dose Equivalent.” 

Committed Effective Dose Equivalent 

(see Dose 

Equivalent)

The committed effective dose equivalent is one measure of dose that can be used to calculate 

the effect of radiation received by an individual and is defined under “Dose Equivalent.” 

Common Cause Component Group

Similar 

components that 

are modeled as a 

group because 

they are subject to 

failure by a 

common cause. 

(see CommonCause Failure)

In a PRA, one failure mechanism of a component may be from a common cause that also fails 

other components. 

A common cause component group is a collection of like components considered to have the 

potential to fail by the same cause. For example, redundant diesel generators in a nuclear 

power plant are modeled as having the potential to fail by common cause (as well as 

independently) and form a common cause component group. Turbine-driven and motor-driven 

pumps in a secondary cooling system may form a common cause component group (failures 

because of a common environment), while at the same time the motor-driven pumps may form 

a separate common cause group because of separate common cause failures. 
4. GLOSSARY

 4-19 

TERM AND 

DEFINITION DISCUSSION 

Common cause failure among like components usually is not modeled to occur across system 

boundaries. This is because the operating regime may be different and thus failure rates may 

be different. An exception may be in external events, such as seismic events, in which 

components may be subject to similar stresses.

Common-Cause Failure 

A failure of two or 

more structures, 

systems, or 

components as a 

result of a single 

shared cause. 

(see CommonMode Failure, 

Failure Mode)

In a PRA, common-cause failure (CCF) is a special form of dependent failure in which the 

failure of the structure, system, or component (SSC) has occurred from the same fault. CCF 

faults generally reflect errors occurring as a result of a common manufacturer, environment, 

maintenance, etc. 

The CCF term is often incorrectly used interchangeably with common-mode failure (CMF). 

CCF only accounts for the SSCs failing because of the same, single cause, not if they 

ultimately fail in the same manner (or in the same mode), which is CMF. In data provided to 

quantify CCF events, the failure mode is usually presented (i.e., failure to start, fail to run), and 

the cause is not always provided about why the failure mode occurs. There could be multiple 

causes lumped into the data presentation for a given failure mode. Thus, the available failure 

data dictate whether the PRA model is modeling CCF or CMF. 

To illustrate the relationship between CCF and CMF, consider potential causes of failure for 

emergency diesel generators (EDGs) as shown in the figure below. Potential failure causes 

include a plugged radiator, a failed load sequencer, bad fuel oil, or faulty bearings. As 

indicated in the figure below, each of these causes can result in failure of multiple diesel 

generators in either the same failure mode or in different failure modes. Diesel failure modes 

included in this example are fails to start (FTS) and fails to run (FTR). 

Failure 

Cause 

Failure Mode

Basic Event Comments CCF Types EDG 

A 

EDG

B 

Plugged 

radiator 

FTS FTR CCF-DG-ABFTS/R-1 Same cause results in a 

different failure mode of 

each DG 

CCF without 

CMF 

Failed load 

sequencer 

FTR FRT CCF-DG-AB-FTR Same cause results in the 

same failure mode of both 

EDGs

CCF with 

CMF 

Bad fuel oil FTS FTS CCF-DG-AB-FTS Same cause results in the 

same failure mode of both 

EDGs 

CCF with 

CMF 

Faulty 

Bearings 

FTS FTR CCF-DG-AB-FTSR2 Same cause results in a 

different failure mode of 

each DG

CCF without 

CMF 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 2).

Common-Mode Failure 

A failure of two or 

more structures, 

systems, or 

components in the 

same manner or 

mode as the result 

of a single shared 

cause. (see 

Common-Cause 

Failure, Failure 

Mode)

In a PRA, common-mode failure (CMF) is a special form of dependent failure that reflects (1) a 

common manner of failure (e.g., failure to start, failure to run) and (2) failure from a common 

cause. Consequently, CMF is actually a type of common-cause failure (CCF) in which the 

SSCs fail in the same way and from the same cause. CMF and CCF are often incorrectly used 

interchangeably. However, CCF only addresses the cause of the failure, while CMF addresses 

both the cause and the manner. 

 

In data provided to quantify CCF or CMF events, the failure mode is usually presented 

(i.e., fails to start (FTS), fails to run (FTR)), and the cause is not always provided about why the 

failure mode occurs. There could be multiple causes lumped into the data presentation for a 

given failure mode. Thus, the available failure data dictate if the PRA model is modeling CCF 

or CMF. 
4. GLOSSARY

 4-20 

TERM AND 

DEFINITION DISCUSSION 

Consider the figure displayed in the discussion section for CCF. Potential failure modes for 

emergency diesel generators are FTS and FTR. Potential failure causes include a plugged 

radiator, a failed load sequencer, bad fuel oil, or faulty bearings. As indicated in the figure for 

CCF, each of these causes can result in failure of multiple diesel generators in either the same 

failure mode or in different failure modes. Examples of CMF are shown in the comment column 

under the term “Common-Cause Failure.” 

The definition provided was based on the definition in the IAEA Safety Glossary (Ref. 7). 

Complementary Cumulative Distribution Function 

(see Cumulative 

Distribution 

Function)

The term complementary cumulative distribution function is a type of cumulative distribution 

function and is defined under “Cumulative Distribution Function.” 

Completeness Uncertainty 

 (see 

Uncertainty)

The term completeness uncertainty is related to epistemic uncertainty and defined under 

“Uncertainty.”

Component 

A part of a system 

in a nuclear power 

plant. (see Basic 

Event) 

In a PRA, the plant is usually modeled at the component level. The ASME/ANS PRA Standard 

(Ref. 2) defines a component as “an item in a nuclear power plant, such as a vessel, pump, 

valve, or circuit breaker.” 

Basic events are associated with individual components, such that different basic events will be 

associated with different failure modes of a particular component. 

Conditional Containment Failure Probability 

(see Conditional 

Probability)

The term conditional containment failure probability is a type of conditional probability and is 

defined under “Conditional Probability.”

Conditional Core Damage Probability 

(see Conditional 

Probability) 

The term conditional core damage probability is a type of conditional probability and is defined 

under “conditional probability.”

Conditional Large Early Release Probability 

(see Conditional 

Probability)

The term conditional large early release probability is a type of conditional probability and is 

defined under “Conditional Probability.” 

Conditional Probability (Containment Failure, Core Damage, Large Early Release) 

Probability of 

occurrence of an 

event, given that a 

prior event has 

occurred. (see 

Probability)

In a PRA, a conditional probability can be calculated for containment failure, core damage, and 

large early release given the knowledge of a variety of prior events have occurred. Examples 

include: 

• Conditional containment failure probability can be calculated given that a particular 

accident type (large loss-of-coolant accident, transient) has occurred. 

• Conditional core damage probability can be calculated given an initiating event (a 

plant upset causing a demand for shutdown) has occurred, or given that a certain 

plant system has been taken out of service. 
4. GLOSSARY

 4-21 

TERM AND 

DEFINITION DISCUSSION 

• Conditional large early release probability can be calculated given that a core 

damage event has occurred, or given that a bypass sequence has occurred. 

Conditional probability exists in other contexts. For example, seismic fragility is the conditional 

probability of a component, structure, or system failure given a seismic motion of a certain 

magnitude. 

Confidence Interval

A range of values 

that has a 

specified likelihood 

of including the 

true value of a 

random variable. 

(see Uncertainty 

Interval)

In a PRA, a confidence interval is sometimes used to describe the uncertainty of a parameter 

input. However, confidence intervals cannot be propagated through the PRA model. A 

confidence interval with a confidence level p is defined such that the probability that the true 

value of a random variable contained within that interval p can be stated with a specified 

likelihood. The confidence level can take a specified value, with the most common being 95% 

or 99%. The following figure shows a 95% confidence interval. In this case, 2.5% of the 

probability distribution is greater than the 95% confidence interval (shaded area under the 

probability distribution function curve), while 2.5% of the probability distribution is less than the 

95% confidence interval. 

Configuration Risk Profile 

(see PRA 

Configuration 

Control)

The configuration risk profile is related to configuration control and is defined under “PRA 

Configuration Control.” 

Consequence 

(see Accident 

Consequence)

In the context of a PRA, the term consequence has the same meaning as accident 

consequence, which is defined under “Accident Consequence.”

Consequence Analysis 

(see Accident 

Consequence 

Analysis)

In the context of a PRA, the term consequence analysis has the same meaning as accident 

consequence analysis, which is defined under “Accident Consequence Analysis.”

Consequential Steam Generator Tube Rupture, Induced Steam Generator Tube Rupture

A break or breach 

in a steam 

generator tube 

caused by the 

consequences of 

an accident. (see 

Steam Generator 

Tube Rupture, 

Containment 

Bypass )

In a PRA for a pressurized-water reactor, steam generator tube ruptures (SGTRs) are modeled 

either as an initiating event or a subsequent failure as part of an accident sequence. If the 

SGTR occurs randomly while the plant is operating, it is an initiating event modeled in the PRA. 

However, if the SGTR occurs because of excessive conditions produced as a result of the 

accident, it is considered to be a consequential or induced SGTR and is modeled in the PRA as 

an event in an accident sequence. These excessive conditions generally involve high 

pressures or high temperatures that could rupture a steam generator tube. For example, this 

might occur if the steam generator were to boil dry (steam generator dryout). 

Confidence Interval

Probability 

Distribution 

Function

(2.5%) (2.5%) (95%)

Mean
4. GLOSSARY

 4-22 

TERM AND 

DEFINITION DISCUSSION 

Accidents involving SGTRs are modeled in PRAs because it allows reactor coolant to flow from 

the reactor vessel to the secondary side of the steam generator. As such, an SGTR can 

become a significant contributor to risk because it can serve as a possible mechanism for 

radioactive material transport to the environment. There is the potential that if a tube bursts 

while a plant is operating, radioactivity from the primary coolant system could escape directly to 

the atmosphere through the safety valves on the secondary side. This scenario is referred to 

as containment bypass. 

Conservative Analysis (Demonstrably) 

An analysis that 

uses assumptions 

such that the 

assessed outcome 

is meant to be less 

favorable than the 

expected outcome. 

(see Bounding 

Analysis) 

In a PRA, conservative analysis may be performed to show that a certain contributor is not 

significant to risk, and thus, resources do not need to be spent on more accurate modeling. A 

conservative analysis provides a result that may not be the worst result of a set of outcomes, 

but produces a quantified estimate of a risk metric that is significantly greater than the risk 

metric estimate obtained by using the most realistic information obtainable (i.e., a realistic 

analysis). Therefore, in a PRA, if there is not much change in risk with the contributor in 

question set at an unfavorable value (as opposed to its most favorable value), then the 

contributor can be omitted from the analysis. For example, a licensee’s request for change in 

technical specifications may show that the requested change will result in acceptable risk 

increases, even with pessimistic assumptions associated with the proposed change. If that is 

the case, then it may be acceptable not to perform a realistic assessment of the proposed 

change since it may involve detailed and time-consuming modeling. Conservative analysis 

also may be used to demonstrate that an item that is not modeled in the PRA has negligible 

impact on risk and therefore can be justifiably neglected. A conservative analysis provides a 

result that may not be the worst result of a set of outcomes, but produces a quantified estimate 

of a risk metric that is significantly greater than the risk metric estimate obtained by using a 

best-estimate evaluation. 

A conservative analysis should be distinguished from a bounding analysis in which 

assumptions and parameters are chosen such that the impact on risk is as detrimental as 

possible; therefore, bounding analysis is a special case of conservative analysis. For example, 

for a conservative analysis a human error probability event can be set to a value that is unlikely 

to be exceeded, whereas for a bounding analysis, the error probability would be set to 1.0. 

Conservative analyses, then, include a spectrum of assessments with results less favorable 

than those of realistic analysis all the way to bounding assessments with the most unfavorable 

results. 

Examples of areas in which conservative analyses can be used in Level 1 risk assessments 

are initiating events, success criteria, thermal-hydraulics, and human error probabilities. 

The terms conservative and demonstrably conservative are used interchangeably. 

The definition is based on the ASME/ANS PRA Standard (Ref. 2), which defines demonstrably 

conservative analysis as one “that uses assumptions such that the assessed outcome will be 

conservative relative to the expected outcome.” 

Containment Building 

(see 

Containment) 

The term containment building has the same meaning as containment and is defined under 

“Containment.” 

Containment Bypass 

A flow path that 

allows the 

unintended 

release of 

radioactive 

material directly to 

In a PRA, the potential for containment bypass is modeled and such a bypass often is 

determined to be a significant risk contributor. A containment bypass circumvents the 

containment’s design function, which is to confine and reduce a release of radioactive material. 

Therefore, a containment bypass can lead to a significant release of fission products in the 

event of a core damage accident. A containment bypass can result from the failure of various 

containment components so that a direct path to the environment is opened. For example, a 
4. GLOSSARY

 4-23 

TERM AND 

DEFINITION DISCUSSION 

the environment, 

bypassing the 

containment. (see 

Containment 

Failure, 

Containment 

Isolation Failure, 

Interfacing 

Systems Loss-ofCoolant Accident) 

containment bypass can result from an interfacing-system loss-of-coolant accident (i.e., an 

accident in which a high-pressure system containing fission products leaks into a lowerpressure system, part of which is outside of containment). For example, a steam generator 

tube rupture in a core damage accident provides a pathway for the fission products in the highpressure primary system to enter the low-pressure side of the steam generator, which has relief 

valves outside of containment. 

Containment bypass is distinct from containment isolation failure in which the containment is 

not acceptably leak-tight. 

The definition provided is based on the definition found in the ASME/ANS PRA Standard (Ref. 

2). 

Containment Capacity 

The ability of the 

containment to 

withstand the 

challenges that 

result from 

accidents. (see 

Containment, 

Containment 

Capacity Analysis, 

Containment 

Pressure 

Boundary) 

In a Level 2 PRA, the containment capacity is evaluated so that it can be compared against the 

postulated challenges to the containment that could result from a severe accident, both preand post-core damage. As such, the containment performance in response to severe accident 

conditions can be assessed. 

The containment capacity is the ability of the structures, systems, and components that make 

up the containment pressure boundary to withstand postulated loads and challenges. 

Containment Capacity Analysis 

A calculation that 

estimates the 

ability of the 

containment to 

withstand the 

challenges that 

result from 

accidents. (see 

Containment 

Capacity) 

In a Level 2 PRA, the containment capacity analysis involves selecting a method or methods to 

evaluate the structural capacity to withstand challenges (e.g, high pressure, temperature, etc.) 

of the structures, systems, and components (SSCs) that make up the containment pressure 

boundary. A plant-specific containment capacity analysis usually involves developing and 

solving a computer model of the relevant SSCs using finite element analysis or similar 

techniques. In the simplest case, the containment capacity can be inferred from that of a 

previously analyzed similar containment of a reference plant. 

Containment Event Tree 

A logic diagram that 

graphically 

represents the 

status of the 

containment and 

containment 

equipment when 

subjected to severe 

accident loads. (see 

Accident 

Progression Event 

Tree, Event Tree) 

In a PRA, a containment event tree (CET) begins with the onset of core damage and 

progresses through a limited number of branches that depict the various scenarios of the 

containment and containment equipment performance when subjected to severe accident 

loads (e.g., high temperatures, pressures). 

As noted in NUREG-1150 (Ref.51), an accident progression event tree (APET) is a more 

detailed representation of the containment response to severe accident loads. The APET 

includes the interaction of phenomena, the availability of equipment, and the performance of 

operators. 

The end states of both the CET and the APET are: no containment failure, various 

containment failure modes, or containment bypass. 
4. GLOSSARY

 4-24 

TERM AND 

DEFINITION DISCUSSION 

The figure below represents a containment event tree with the following acronyms: Core 

Damage (CD), Reactor Coolant System depressurization (RCS Depress), Vessel Breach (VB), 

Steam Generator Tube Rupture (SGTR). 

Containment Failure Mode 

The various ways 

in which the ability 

of the containment 

to prevent 

radioactive 

material release is 

compromised. 

(see Containment 

Failure, 

Containment 

Bypass, 

Containment 

Isolation Failure)

In a PRA, the modes of containment failure define the manner in which containment integrity is 

lost (i.e., the way a radioactive material release pathway from inside the containment to the 

environment is created). Containment failure mode encompasses both structural failures of 

containment induced by containment challenges when they exceed containment capability, as 

well as the failure modes of containment induced by human failure events, isolation failures, or 

bypass events such as interfacing-systems loss-of-coolant accidents. 

The definition provided is based on the definition in the ASME/ANS PRA Standard (Ref. 2). 

Containment Failure Probability 

(see Probability) The term containment failure probability is a type of failure probability that is computed based 

on the likelihood of containment failure and is discussed under the discussion for the term 

“Probability.” 

Containment Failure (Early, Late) 

Loss of integrity of 

the containment 

from a core 

damage accident 

that is expected to 

result in an 

unacceptable 

release of 

radioactive 

materials. (see 

Containment, 

Containment 

Bypass, 

In a PRA, determining when and if the containment fails or is bypassed during a severe 

accident is very important from a risk perspective. If the containment pressure boundary 

remains leak-tight, the offsite consequence will be low. Conversely, if the containment fails or 

is bypassed, then the consequence to the surrounding population can be potentially high. For 

specific containments there can be selected severe accident scenarios in which the 

containment fails before fission products have penetrated the primary system. If the accident is 

successfully arrested at this point, no release will occur. However, usually containment failure 

represents the failure of the final barrier preventing a radioactive material release. 

 

Containment failure is often categorized as early or late. Early containment failure occurs in a 

timeframe before the surrounding population within 1 mile of the site boundary can be 

evacuated. Late containment failure occurs in a timeframe that allows the surrounding 

population from 1 to 10 miles to be evacuated. 

Succeeds

Fails

Core

Damage

(CD)

Containment

Isolation or no

Bypass

RCS

Depress

CD 

Arrested 

w/o VB

No 

Induced 

STGR

No 

Containment

Failure at VB

No Potential

for Early

Fatalities

Large 

Early 

Release

No

No

No

No

No

No

No

No

Yes

Yes

Yes

Yes
4. GLOSSARY

 4-25 

TERM AND 

DEFINITION DISCUSSION 

Containment 

Pressure 

Boundary)

Containment bypass failures (e.g., interfacing-system loss-of-coolant accidents) occur in the 

early timeframe but usually are categorized separately from early structural failures of the 

containment. 

The definition is derived from the ASME/ANS PRA Standard (Ref. 2). 

Containment Integrity 

The ability of the 

containment to 

function as a 

barrier to prevent 

release of 

radioactive 

materials as a 

result of an 

accident. (see 

Containment 

Failure Mode)

In a Level 2 PRA, an important concern is the potential loss of containment integrity. 

Containment integrity depends on the structures, systems, and components of the reactor 

containment pressure boundary that perform the containment function. Maintaining 

containment integrity largely depends on the individual containment design and the particular 

phenomena or load that challenges the integrity of the containment. Examples of particular 

severe accident challenges to the containment integrity include overpressure, internal missiles, 

external missiles, melt-through, and bypass. 

Containment Isolation Failure 

A failure in the 

piping, valves, or 

actuators that 

isolate the 

containment. (see 

Containment 

Bypass, 

Containment 

Failure Mode)

In a PRA, containment isolation failures are one of the containment failure modes considered in 

a Level 2 analysis. Containment isolation is provided to prevent or limit the escape of fission 

products that may result from postulated accidents. In a containment isolation failure, fission 

products can pass to the environment through the containment because the containment is not 

properly isolated (i.e., not acceptably leak-tight). 

In some severe accident scenarios, an accident management strategy, referred to as 

containment venting, may be used. Containment venting involves a deliberate breach of 

containment isolation by the plant operators who open a controlled, filtered or unfiltered, 

pathway from the containment to the environment to prevent an uncontrolled overpressure 

failure of the containment. 

The containment isolation system consists of the piping, valves, and actuators that are 

designed so that fluid lines penetrating the containment boundary are isolated in the event of 

an accident. 

Containment Pressure Boundary 

Those parts of the 

reactor 

containment that 

sustain loading 

and provide a 

pressure boundary 

in the performance 

of the containment 

function. (see 

Containment)

In a Level 2 PRA, the evaluation of containment integrity is an evaluation of the structures, 

systems, and components of the reactor containment pressure boundary that perform the 

containment function (i.e., that form the containment system). As stated in NUREG-0800 (Ref. 

44), the reactor containment system design must include the functional capability of enclosing 

the reactor system and of providing a final barrier (boundary) against the release of radioactive 

fission products in case of postulated accidents. 

Leak-tightness of the containment is ensured by a continuous pressure boundary consisting of 

nonmetallic seals and gaskets and metallic components that are either welded or bolted 

together. Each containment also includes numerous access and process penetrations that 

complete the pressure boundary. 

The definition provided is derived from Chapter 6 of NUREG-0800 (Ref. 46). 

Containment Structure 

 (see 

Containment)

The term containment structure has the same meaning as containment and is defined under 

“Containment.” 
4. GLOSSARY

 4-26 

TERM AND 

DEFINITION DISCUSSION 

Containment, Containment Building, Containment Structure 

A physical 

structure 

surrounding a 

reactor that is 

designed to 

prevent or control 

the release of 

radioactive 

material. (see 

Containment 

Capacity, 

Containment 

Failure, 

Containment 

Failure Mode, 

Containment 

Integrity, 

Containment 

Pressure 

Boundary) 

In a Level 2 PRA, the ability of the containment (containment building or containment structure) 

to contain fission products that have escaped from the reactor is analyzed to estimate the limits 

of the containment’s capacity. 

A containment, containment building, or containment structure, in its most common usage, is a 

steel or reinforced concrete structure enclosing a nuclear reactor designed to contain the 

escape of radiation to the environment. The containment is the final barrier to radioactive 

material release. 

Containments are designed to remain intact when subject to the pressure and temperature 

loads from design-basis accidents (DBAs). Moreover, because of safety factors built into 

containment designs, they are predicted to fail at pressures and temperatures (from core melt 

accidents) that are significantly higher than those of DBAs. 

The NRC Web site Glossary (Ref. 36) defines the term containment building as an “air-tight 

building, which houses a nuclear reactor and its pressurizer, reactor coolant pumps, steam 

generator, and other equipment or piping that might otherwise release fission products to the 

atmosphere in the event of an accident. Such buildings usually are made of steel-reinforced 

concrete.” 

The NRC Web site Glossary (Ref. 36) also defines the term containment structure as “a 

gas-tight shell or other enclosure around a nuclear reactor to confine fission products that 

otherwise might be released to the atmosphere in the event of an accident. Such enclosures 

are usually dome-shaped and made of steel-reinforced concrete.” 

Core Damage 

Sufficient damage 

that could lead to a 

release of 

radioactive 

material from the 

core that could 

affect public 

health. (see Core 

Melt, Core 

Damage 

Frequency, Core 

Damage 

Probability) 

In a PRA, the potential for core damage is evaluated in the Level 1 part of the analysis. 

Specifically, a Level 1 PRA calculates the core damage frequency given the design and 

operation of the plant. In this context, core damage in a Level 1 PRA is actually the onset of 

core damage; that is, being the onset of sufficient damage to the core that (1) if not immediately 

arrested could potentially result in a release of radioactive material from the core, and (2) if 

released from the vessel and containment, could result in offsite public health effects. 

In deterministic analyses, quantitative criteria often are used to define the onset of core 

damage (e.g. a peak clad temperature of 2,200 degrees Fahrenheit). 

The ASME/ANS PRA Standard (Ref. 2) defines core damage as “uncovery and heatup of the 

reactor core to the point at which prolonged oxidation and severe fuel damage are anticipated 

and involving enough of the core, if released, to result in offsite public health effects.” 

The terms core damage and core melt are sometimes incorrectly used as synonyms. 

However, core melt occurs after the onset of core damage. Core damage does not necessarily 

indicate that the reactor fuel has melted, only that radioactive material could be released from 

the core into the reactor vessel. An illustration differentiating the concepts of core damage, 

core melt, and their timing is provided below. 
4. GLOSSARY

 4-27 

TERM AND 

DEFINITION DISCUSSION 

Core Damage Frequency 

 (see 

Frequency) 

The term core damage frequency is a type of frequency used in PRA and is defined under 

“Frequency.” 

Core Damage Probability 

 (see Probability) The term core damage probability is a type of probability used in PRA and is defined under 

“Probability.” 

Core Melt

Damage beyond 

the onset of core 

damage that could 

progress to a 

complete melting 

of the core. (see 

Core Damage, 

High- Pressure 

Melt Ejection, 

Reactor Core) 

In a PRA, the potential for core melt is evaluated in the Level 2 part of the analysis. A Level 1 

PRA calculates the onset of core damage, while the Level 2 evaluates the effects starting with 

the onset of core damage and then progressing to a complete melting of the core. The 

evaluation considers the different releases that can occur during the core melt progression and 

the frequencies of the associated accident progressions. 

The terms core melt and core damage are sometimes incorrectly used as synonyms. 

However, core damage entails only the potential release of radioactive material from the core 

into the reactor vessel, not necessarily the melting of any portion of the reactor core. An 

illustration differentiating the concepts of core damage, core melt, and their timing is provided 

under the discussion for the term “Core Damage.” 

The NRC Web site Glossary (Ref. 36) defines a core melt accident as “an event or sequence of 

events that result in the melting of part of the fuel in the reactor core.” 

Cumulative Distribution Function (Complementary) 

A function that 

provides the 

probability that a 

parameter is less 

than or equal to a 

given value. (see 

Probability 

Distribution)

In a PRA, the cumulative distribution function is often used to present the results of the 

analysis. 

The cumulative distribution function gives the probability that the random variable does not 

exceed a specified value. The cumulative distribution function is the integral of the probability 

distribution functions. The cumulative distribution function adds up the probabilities of 

occurrence of all possible parameter values less than the specified value, as represented by 

the probability distribution function of the parameter. The following graphs illustrate the 

cumulative distribution function and the probability distribution function. 

Coolant inventory decreasing

core melt progressing

Coolant above core

decreasing

Cooling 

systems 

lost

No unacceptable 

releases

Start of core melt

Level 1 Analysis Level 2 Analysis

Amount of release dependent on 

whether and when coolant is 

restored

Full melting of

the core

Sufficient coolant has been lost such that, if

recovered, the unmitigated release of radioactive 

material from the core would be sufficient to result in 

public health effects

Onset of core damage
4. GLOSSARY

 4-28 

TERM AND 

DEFINITION DISCUSSION 

The cumulative distribution function may be used to calculate the quantiles or the probability of 

not exceeding the mean of a risk metric. 

Other examples of using the cumulative distribution function are calculation of the seismic 

fragility of a component, or the calculation of probability of recovery of offsite power within a 

certain time period. 

NUREG/CR-6823 (Ref. 78) defines cumulative distribution function as one that “gives the 

probability that the random variable does not exceed a given value.” 

The complementary cumulative distribution function is the complement of the cumulative 

distribution function (i.e., the result of subtracting the cumulative distribution function from 

unity). Therefore, the complementary cumulative distribution function can be defined as a 

function that provides the probability that a parameter value is greater than a given value. The 

following graphs illustrate the complementary cumulative distribution function and its 

corresponding cumulative distribution function. Cumulative Probability

Cumulative Distribution Function

Risk Metric

Event Probability

Risk Metric

Probability Distribution Function

Probability

Complementary Cumulative Distribution Function (CCDF)

Vs.

Cumulative Distribution Function (CDF)

Risk Metric

Probability

CCDF CDF
4. GLOSSARY

 4-29 

TERM AND 

DEFINITION DISCUSSION 

Some examples of using the complementary cumulative distribution function are calculating the 

probability of exceeding a certain release fraction of radioactive material in core melt accidents, 

calculating the frequency of exceeding a certain intensity of external hazard occurrence, 

calculating the frequency of loss of offsite power events exceeding a certain duration, or 

calculating the probability of emergency diesel generator repair lasting longer than a certain 

time period. 

The definition provided was based on the definition in NUREG/CR-6823 (Ref. 78). 

Cumulative Dose

(see Dose) The cumulative dose is a total dose that is defined under “Dose.” 

Cutset (Minimal Cutset) 

A combination of 

failures that result 

in a particular 

outcome. (see 

Truncation Limit) 

In a PRA, a cutset (sometimes also written as “cut set”) is the product (i.e., result) of the 

analysis and identifies a combination of failures that would result in core damage or 

containment failure. However, the cutsets produced by the PRA are minimal cutsets in which 

each minimal cutset is the smallest combination of failures needed to cause core damage or 

containment failure. 

Cutsets are expressed in the form of combinations of basic events. Basic events represent 

elements of the PRA model for which no further decomposition is performed because they are 

at the limit of resolution consistent with available failure data. Basic events can represent 

equipment unavailability, human errors, and initiating events. 

NUREG-1560 (Ref. 56) defines cutset as a “combination of a set of events (e.g., initiating event 

and component failures) that, if they occur, will result in an undesirable condition (such as the 

onset of core damage or containment failure).” In addition, NUREG-1560 defines the term 

“minimal cutset” as “the minimum combination of the set of events that would result in the 

undesirable condition.” 

The Fault Tree Handbook (Ref. 49) defines minimal cutset in the context of a fault tree as “a 

smallest combination of component failures which, if they all occur, will cause the top event to 

occur.” 

To illustrate the concept of a minimal cutset, consider an accident involving the combination of 

loss of offsite power, emergency diesel generator (EDG) failure, and electrically-driven 

emergency cooling pump failure: 

• For this postulated accident, a “cutset” may include separate events that represent 

(1) failure of offsite power, (2) failure of all EDGs, and (3) independent failure of the 

electrically-driven emergency cooling pumps; however, this would represent a 

nonminimal cutset because the electrically-driven emergency cooling pumps rely on 

the EDGs. If the EDGs fail, the electrically-driven emergency cooling pumps will not 

function, regardless if they independently fail. 

• For this accident, a “minimal cutset” would represent (1) failure of offsite power and 

(2) failure of all EDGs. These are the minimal failures required to cause failure of 

emergency cooling regardless if the electrically-driven emergency cooling pumps fail. 
4. GLOSSARY

 4-30 

TERM AND 

DEFINITION DISCUSSION 

Deep Dose Equivalent 

(see Dose 

Equivalent)

The deep dose equivalent is one measure of dose that can be used to calculate the effect of 

radiation received by an individual and is defined under “Dose Equivalent.” 

Defense-in-Depth 

Formal definition 

requires 

Commission 

approval. (see 

Safety Margin, 

Uncertainty, 

Rationalist, 

Structuralist) 

In a PRA, defense-in-depth is not an explicitly modeled element. Rather, the results of the 

PRA provide insights into defense-in-depth. 

Over time, various definitions have been used for defense-in-depth, including: 

• three barriers to contain radioactive material: fuel cladding, primary system boundary, 

and the containment 

• the use of successive measures to prevent an accident or to mitigate the 

consequences of an accident 

• the use of redundancy and diversity 

• implementation of the single failure criterion 

Regardless of its definition, defense-in-depth is an integral part of the NRC’s safety philosophy. 

The NRC Web site Glossary (Ref. 36) defines defense-in-depth as: “An approach to designing 

and operating nuclear facilities that prevents and mitigates accidents that release radiation or 

hazardous materials. The key is creating multiple independent and redundant layers of 

defense to compensate for potential human and mechanical failures so that no single layer, no 

matter how robust, is exclusively relied upon. Defense-in-depth includes the use of access 

controls, physical barriers, redundant and diverse key safety functions, and emergency 

response measures.” 

The NRC Commission has referred to defense-in-depth as a concept that: 

Has always been and will continue to be a fundamental tenet of regulatory practice in the 

Pump A

failed

Pump C 

failed

Pump B 

failed

Pump D 

failed

Pump E 

failed

Pump Systems

failed

Pump System I 

failed

Pump System II 

failed

Cutset Example forPump Systems:

A B C D E

Possible Cutsets:

A*D

A*E

A*B*D

A*B*E

A*C*D

A*C*E

A*B*C*D

A*B*C*E

A*B*C*D*E

A*C*D*E

A*B*D*E

B*D

B*E

B*C*D

B*C*E

B*D*E

B*C*D*E

C*D

C*E

C*D*E

Minimal Cutsets: 

A*D

A*E

B*D

B*E

C*D 

C*E
4. GLOSSARY

 4-31 

TERM AND 

DEFINITION DISCUSSION 

nuclear field, particularly regarding nuclear facilities. Risk insights can make the elements of 

defense-in-depth clearer by quantifying them to the extent practicable. Although the 

uncertainties associated with the importance of some elements of defense may be 

substantial, the fact that these elements and uncertainties have been quantified can aid in 

determining how much defense makes regulatory sense. Decisions on the adequacy of, or 

the necessity for, elements of defense should reflect risk insights gained through 

identification of the individual performance of each defense system in relation to overall 

performance. 

The Commission further states: 

Defense-in-depth is an element of the NRC’s Safety Philosophy that employs successive 

compensatory measures to prevent accidents or mitigate damage if a malfunction, 

accident, or naturally caused event occurs at a nuclear facility. 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. 

Demonstrably Conservative Analysis 

(see 

Conservative 

Analysis) 

A demonstrably conservative analysis has the same meaning as a conservative analysis and is 

defined under “Conservative Analysis.” 

Dependency 

Reliance of a 

function, system, 

component, or 

human action on 

another part of the 

system or another 

human action to 

accomplish its 

function. 

Dependency is significant to the fidelity of a PRA model to capture the interrelationship 

between the modeled systems and human actions. 

As an example of systems dependency, many core cooling systems depend on electric power 

or cooling water systems. Also, operator actions closely spaced in time may have dependency 

in that a failure to perform a certain action may negatively affect successful performance of a 

subsequent action. 

Dependency has also been defined as: 

• “Requirement external to an item and upon which its function depends and is 

associated with dependent events that are determined by, influenced by, or correlated 

to other events or occurrences.” (Ref. 2) 

• “Requirement external to a structure, system, or component (SSC), and upon which 

the SSC’s function depends.” (Ref. 59) 

Design-Basis Accident 

A postulated 

accident that a 

nuclear facility 

must be designed 

and built to 

withstand without 

loss to the 

systems, 

structures, and 

components 

necessary to 

ensure public 

In a PRA, the accidents traditionally modeled are not design-basis accidents (DBAs). Instead, 

the PRA typically models accidents that are more severe than DBAs, which are referred to as 

beyond-design-basis accidents (BDBAs) or severe accidents. It is important, though, to 

distinguish that the term “severe accident” indicates that core damage occurred; however, the 

term “beyond-design-basis accident” merely indicates that the accident exceeded the design 

limits of the plant. 

When developing a nuclear power plant, DBAs are selected to bound credible accident 

conditions and to ensure that the nuclear power plant can withstand and recover from these 

accidents. An example of a DBA is a major rupture of a pipe containing reactor coolant up to 

and including the double-ended rupture of the largest pipe containing reactor coolant. 
4. GLOSSARY

 4-32 

TERM AND 

DEFINITION DISCUSSION 

health and safety. 

(see BeyondDesign-Basis 

Accident, Severe 

Accident, DesignBasis Event)

Another term, design-basis event (DBE), is used to broadly describe any event, internal or 

external to the plant, which could challenge safety functions. Therefore, DBAs are a subset of 

DBEs, and other examples of DBEs are anticipated transients (e.g., tripping of turbine 

generator), external events, and natural phenomena. 

NUREG-0800, Standard Review Plan 15.0 (Ref. 47), defines design-basis accidents as 

“postulated accidents that are used to set design criteria and limits for the design and sizing of 

safety-related systems and components.” 

The definition provided was based on the definition in the NRC Web site Glossary (Ref. 36). 

Design-Basis Event 

Any of the events 

specified in the 

nuclear power 

plant’s safety 

analysis that are 

used to establish 

acceptable 

performance for 

safety-related 

functions. (see 

Design-Basis 

Accident, Severe 

Accident) 

In a PRA, the outcome of concern is whether or not a particular accident leads to core damage. 

Therefore, beyond-design-basis accidents (BDBAs) that exceed the design envelope and lead 

to core damage are typically modeled. In this instance, these BDBAs that lead to core damage 

are referred to as severe accidents. Because a plant is designed and engineered to contend 

with design-basis accidents (DBAs), they typically are not the focus of current PRAs. However, 

DBAs represent only a portion of a broader category, design-basis events (DBEs). DBEs 

represent conditions within the plant design envelope and include anticipated transients (e.g., 

tripping of turbine generator), anticipated operational occurrences (AOOs), DBAs, external 

events, and natural phenomena. 

AOOs, an example of a DBE mentioned above, are a type of DBE described in NUREG-0800, 

Standard Review Plan 15.0 (Ref. 47), as “conditions of normal operation that are expected to 

occur one or more times during the life of the nuclear plant unit,” (e.g., example loss of all 

offsite power). 

DBAs are a subset of DBEs, as noted above. An example of a DBA is a major rupture of a 

pipe containing reactor coolant up to and including the double-ended rupture of the largest pipe 

containing reactor coolant. 

The definition provided was based on the definition in NUREG-1560 (Ref. 56). 

Deterministic (Analysis, Approach, Regulation) 

A characteristic of 

decisionmaking in 

which results from 

engineering 

analyses, not 

involving 

probabilistic 

considerations, are 

used to support a 

decision. (see 

Risk-Informed, 

Probabilistic) 

A PRA represents an approach for assessing the likelihood of accidents and their potential 

consequences. However, the PRA model cannot be separated from and depends on 

deterministic analyses. For example, success criteria for various systems used in PRA to 

prevent and mitigate core damage are based on deterministic analyses. Another example of a 

deterministic analysis would be the calculation of peak cladding temperatures after emergency 

core cooling system actuation in a loss-of-coolant accident, or the timing of vessel breach in a 

core melt accident. 

As discussed in SECY-98-144 (Ref. 96), a deterministic regulation assumes that adverse 

conditions can exist and establishes a specific set of design-basis events (i.e., what can go 

wrong?). The deterministic approach involves implied, but unquantified, elements of probability 

in the selection of the specific accidents to be analyzed as design-basis events. It then 

requires that the design include safety systems capable of preventing or mitigating the 

consequences (i.e., what are the consequences?) of those design-basis events to protect 

public health and safety. 

The NRC Web site Glossary (Ref. 36) defines the term deterministic as “consistent with the 

principles of ‘determinism,’ which hold that specific causes completely and certainly determine 

effects of all sorts. As applied in nuclear technology, it generally deals with evaluating the 

safety of a nuclear power plant in terms of the consequences of a predetermined bounding 

subset of accident sequences.” A deterministic approach or regulation is the opposite of a riskinformed approach or regulation in which the likelihood of potential accidents is integrated. 

Deterministic approaches or regulations do not account for likelihood, and thus do not 

incorporate risk results obtained from a PRA.
4. GLOSSARY

 4-33 

TERM AND 

DEFINITION DISCUSSION 

Deterministic Analysis 

(see 

Deterministic)

The term deterministic analysis is defined under “Deterministic.”

Deterministic Approach 

(see 

Deterministic)

The term deterministic approach is defined under “Deterministic.”

Deterministic Regulation 

 (see 

Deterministic)

The term deterministic regulation is defined under “Deterministic.”

Direct Containment Heating 

(see HighPressure Melt 

Ejection) 

The term direct containment heating is a mechanism for challenging containment integrity and 

is defined under “High-Pressure Melt Ejection.” 

Dose

A measure of the 

amount of 

radiation absorbed 

by a person. (see 

Dose Equivalent)

In a Level 3 PRA, dose is calculated to assess offsite health effects. The NRC Web site 

Glossary (Ref. 36) defines dose as “a general term, which may be used to refer to the amount 

of energy absorbed by an object or person per unit mass. Known as the ‘absorbed dose,’ this 

reflects the amount of energy that ionizing radiation sources deposit in materials through which 

they pass, and is measured in units of radiation-absorbed dose (rad). The related international 

system unit is the gray (Gy), where 1 Gy is equivalent to 100 rad. By contrast, the biological 

dose or dose equivalent, given in rems or sieverts (Sv), is a measure of the biological damage 

to living tissue as a result of radiation exposure.” 

The collective dose (i.e., total dose obtained by summing over individual exposures of the 

affected population) is also used as a risk measure in value-impact analyses carried out in 

conjunction with PRAs. NUREG-0713, Vol. 28 (Ref. 45), states that the concept of collective 

dose is used by the NRC to denote the summation of the total effective dose equivalent 

received by all monitored workers at a nuclear facility, usually over the course of a year, and is 

reported in units of person-rem per year. 

The cumulative dose is the total dose that an individual receives as a result of repeated 

exposures to ionizing radiation to the same portion of the body, or to the whole body, over time. 

Cumulative dose usually is used for measuring occupational exposures of workers in the 

nuclear industry. 

When defining dose and the way it is used in PRAs to estimate health effects the following 

considerations are relevant: 

Under ‘radiation dose’ two concepts commonly used are: deterministic or non-stochastic 

dose and stochastic dose. The former implies that a health effect will occur within a short 

period following exposure with near certainty; the latter that a health effect may occur at 

some later time with some probability. In a PRA, the former is used with a threshold 

(depending on organ) to estimate early health effects. The latter is used, usually with a 

linear no-threshold model, to estimate latent cancers. 
4. GLOSSARY

 4-34 

TERM AND 

DEFINITION DISCUSSION 

Dose Coefficient 

Dose coefficients 

relate the dose to 

organs and tissues 

of the body from 

concentrations of 

radionuclides. 

(see Dose, Dose 

Conversion 

Factor) 

In a Level 3 PRA, dose coefficients are incorporated into the consequence model. Dose 

coefficients relate the dose to organs and tissues of the body from concentrations of 

radionuclides. Dose coefficients for external exposure relate the organ and tissue doses to the 

concentrations of radionuclides in environmental media. Since the radiation arises outside the 

body, this is referred to as external exposure, while dose coefficients for internal exposure 

relate the organ and tissue doses to the intake of radionuclides by inhalation or ingestion, 

where the radiation is emitted inside the body. 

Dose Conversion Factor 

A factor used to 

determine the 

biological effect of 

different types of 

radiation on an 

individual’s organs. 

(see Dose)

In a Level 3 PRA, dose conversion factors are incorporated into the consequence model and 

used to calculate the effect of radiation received by an individual on different organs. 

As discussed in WASH-1400 (Ref. 44), dose conversion factors for the incorporation of 

radioactive material in the body give the dose received by individual organs over a time interval 

per curie intake by inhalation or ingestion. For external exposure, the dose conversion factors 

give the dose received by each organ per curie of radioactive material in a cubic meter of air or 

per curie of radioactive material deposited uniformly on a square meter of horizontal surface. 

The calculation of these dose conversion factors requires elaborate computer models with 

appropriate physiological parameters for a human body. These calculations need only be 

performed once for each type of radioactive material, organ, exposure mode, and time interval. 

From these calculations, a table can be prepared for use in the consequence model.

Dose Equivalent 

A measure of the 

biological damage 

to living tissue as a 

result of radiation 

exposure. (see 

Dose)

In a Level 3 PRA, a measure of biological damage because of radiation exposure is needed to 

estimate health effects. The dose equivalent is calculated as the product of absorbed dose in 

tissue multiplied by a quality factor and then sometimes multiplied by other necessary 

modifying factors at the location of interest. The dose equivalent is expressed numerically in 

units of rems or sieverts. 

The NRC Web site Glossary (Ref. 36) states that as defined in Title 10 of the Code of Federal 

Regulations (10 CFR) 20.1003, “Definitions” (Ref. 13), the committed dose equivalent (CDE) is 

the dose to some specific organ or tissue of reference that will be received from an intake of 

radioactive material by an individual during the 50-year period following the intake. In the event 

that an individual inhales or ingests radioactive material, the individual will continue to receive a 

dose from this event for the rest of his or her life. 

The NRC Web site Glossary (Ref. 36) also states that as defined in 10 CFR 20.1003 (Ref. 12), 

the committed effective dose equivalent (CEDE) is the sum of the products of the committed 

dose equivalents for each of the body organs or tissues that are irradiated, multiplied by the 

weighting factors applicable to each of those organs or tissues. The CEDE reflects the fact that 

different organs in the body are affected differently by radiation. 

The total effective dose equivalent (TEDE) is the sum of the external and the internal doses to 

an individual exposed to radiation. In a PRA, the total effective dose equivalent is needed to 

calculate offsite health effects. According to the NRC Web site Glossary (Ref. 36), the TEDE is 

the sum of the deep-dose equivalent (for external exposures) and the CEDE (for internal 

exposures). The deep-dose equivalent is the external whole-body exposure dose equivalent at 

a tissue depth of 1 cm. Whole body exposure includes at least the external exposure, head, 

trunk, arms above the elbow, or legs above the knee. Where a radioisotope is uniformly 

distributed throughout the body tissues, rather than being concentrated in certain parts, the 

irradiation can be considered as whole-body exposure. 
4. GLOSSARY

 4-35 

TERM AND 

DEFINITION DISCUSSION 

Dose Rate

The amount of 

absorbed dose 

delivered per unit 

time. (see Dose, 

Exposure, 

Exposure Time)

In a Level 3 PRA a dose rate is needed to calculate the health effects. The units in which the 

dose rate is expressed are usually rems or sieverts per hour. Dose rate is the same as 

exposure rate. A PRA considers two types of exposures: acute and chronic. An acute 

exposure involves a large exposure received over a short period of time, i.e., a high exposure 

rate. Chronic exposures involve exposure at a low rate received over a long period of time, 

such as during a lifetime. 

Dose Response Model 

A model that 

reflects the 

relationship 

between low doses 

of ionizing 

radiation and the 

potential for 

cancer. (see 

Dose, Linear NoThreshold Model) 

In a Level 3 PRA, a dose response model is used to calculate frequency of latent cancers in 

the affected population, based on the dose received from the postulated accidents. 

There is some debate about the appropriate dose-response relationship for cancer risk 

following exposure to ionizing radiation. For example, in most PRAs, a linear relationship is 

assumed in which the cancer risk increases in direct proportion to the dose and there is no 

lower dose limit below which there is no risk. Others believe there is a nonlinear relationship, in 

which cancer risk increases in a more complex manner relative to dose. 

Dosimetry 

The measurement 

and calculation of 

the absorbed dose 

in matter and 

tissue resulting 

from the exposure 

to ionizing 

radiation. (see 

Dose) 

In a Level 3 PRA, dose is calculated to estimate health effects on the population affected by a 

severe accident. Dosimetry is the process of determining dose from exposure to radiation. 

To determine the dose received by exposed individuals, dosimetry attempts to estimate the 

dose received directly or indirectly via the various dose pathways, including cloudshine, water 

immersion, groundshine, skin deposition, inhalation, and ingestion. 

Dynamic PRA 

A PRA that 

accounts for 

time-dependent 

effects by 

integrating them 

directly into the 

computer model. 

(see PRA, Living 

PRA) 

In a traditional PRA, the coupling of deterministic analyses into the PRA model is achieved by 

manually constructing the linkage between the probabilistic and deterministic models. Thus, 

the manner in which an accident evolves with time (i.e., time-dependent effects) is based on a 

set of system and operator response characteristics that are manually entered into the PRA 

model. This is done by constructing event sequences in a discrete way such that they bound 

the contribution from all the scenarios that differ in the timing of the contributing events. 

In contrast, a dynamic PRA models accident sequences by automatically constructing the 

linkage between the probabilistic and deterministic models such that system and operator 

response characteristics are automatically accounted for in the PRA model. 

A dynamic PRA is not the same as a living PRA. In a living PRA, the PRA is updated as 

necessary to reflect changes in plant characteristics (e.g., design, operations) so that it 

represents the as-built as-operated plant. 

Early Containment Failure 

(see 

Containment 

Failure) 

The term early containment failure is discussed under the discussion for the term “Containment 

Failure.”
4. GLOSSARY

 4-36 

TERM AND 

DEFINITION DISCUSSION 

Early Fatality 

 (see Fatality) The term early fatality is discussed under the discussion for the term “Fatality.” 

Early Fatality Risk 

(see Fatality) The term early fatality risk is a type of risk-involved fatality caused by exposure to radioactive 

materials and is defined under “Fatality.” 

Economic Factors 

The considerations 

taken into account 

when assessing 

costs related to a 

release of 

radioactive 

material to the 

environment. (see 

Economic Impact)

The Level 3 portion of a PRA assesses the injuries and economic losses that might result if 

radioactivity escaped from containment. The economic factors in assessing risk include the 

costs of various actions taken to protect the public from short-term and long-term exposure 

through different exposure pathways (e.g., evacuation, relocation, decontamination), the costs 

of health effects and health care following exposure, and secondary economic effects. 

An illustrative list of required cost inputs from NUREG/CR-2300 (Ref. 69) includes: 

• evacuation cost per person 

• value of residential, business, and public areas per person 

• relocation cost per person 

• decontamination cost per acre for farm areas 

• decontamination cost per person for residential, business, and public areas 

• compensation rate per year for residential, business, and public areas (i.e., fraction of 

value) 

• average value of farmland per acre for state, county, or smaller areas 

• average annual value of farm sales per acre for state, county, or smaller areas 

• miscellaneous information, such as seeding and harvesting month, fraction of land 

devoted to farming, and fraction of farm sales due to dairy production. 

Economic Impact 

The incurred costs 

of evacuation and 

relocation of the 

population, the 

costs of land 

condemnation, 

and the cost of 

condemned crops 

and other farm 

products as a 

result of an 

accident. (see 

Economic Factors)

In a Level 3 PRA, in addition to the health effects on the surrounding population, the impact of 

the severe accident on the surrounding economy is often estimated. Therefore, the economic 

impact risk is one of the risk categories calculated in a Level 3 PRA. 

The economic model in a Level 3 PRA includes the direct costs associated with protective 

actions taken after the accident, such as evacuation and relocation of the population, 

temporary or permanent interdiction of contaminated land and property, destruction of crops 

and foodstuffs. The model also may include other direct costs of actions, such as 

decontamination. Therefore, costs are a function of the stringency of post-accident radiation 

protection measures. Other direct costs may include costs of treatment of individuals exposed 

to radiation. Some models may include indirect economic impacts (e.g., litigation costs, 

government spending for disaster relief, regional economic activity impacts). 

Economic Impact Risk 

(see Economic 

Impact)

The economic impact risk is the risk resulting from the economic impact of the accident and is 

defined in the discussion under “Economic Impact.” 
4. GLOSSARY

 4-37 

TERM AND 

DEFINITION DISCUSSION 

Emergency Preparedness 

The actions put 

into place to 

prepare personnel 

to rapidly identify, 

evaluate, and 

react to 

emergencies. 

(see Emergency 

Response,

Accident 

Mitigation)

In a Level 3 PRA, to credit an effective emergency response when calculating the 

consequences of postulated accidents, adequate emergency preparedness (EP) is assumed. 

EP includes the programs, plans, training, exercises, and resources necessary to prepare 

emergency personnel to respond to emergencies, including those arising from terrorism or 

natural events such as hurricanes. EP strives to ensure that nuclear power plant operators can 

implement measures to protect public health and safety in the event of a radiological 

emergency. 

The definition provided is based on the definition in the NRC Web site Glossary (Ref. 36). 

Emergency Response 

The actions 

initiated by the 

plant to mitigate 

the consequences 

of an accident that 

could potentially 

result in 

radioactive 

material release. 

(see Emergency 

Preparedness, 

Accident 

Mitigation, Cohort)

In a Level 3 PRA, the emergency response is taken into account when calculating the 

consequences of the postulated accidents. 

The emergency response encompasses the actions used to mitigate the consequences of an 

emergency, such as a severe nuclear accident, to human health and safety, quality of life, 

property, and the environment. The feasibility of some emergency actions may be limited by 

the hazard type (e.g., seismic events). 

The definition provided is based on the definition in the IAEA Safety Glossary (Ref. 7). 

End State 

A set of conditions 

selected to 

characterize the 

plant states at the 

end of a chain of 

events. (see

Accident 

Sequence) 

In most PRAs, end states associated with Level 1 accident sequences typically include: 

success states (i.e., those states with negligible impact), and core damage or plant damage 

states. End states associated with Level 2 sequences usually are containment failure modes 

or release categories. 

The following figure illustrates different end states of an event tree: 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 2). 

Environmental Qualification 

A process for 

demonstrating that 

equipment will be 

capable of 

withstanding the 

accident ambient 

In most PRAs, the focus is on severe accidents. The environment during a severe accident 

can be quite harsh and affect equipment performance. Safety equipment may experience high 

temperatures, pressures, humidity, radiation levels, and aerosol and particulate levels. The 

equipment may or may not be credited in the PRA as continuing to function under these 

conditions for many hours. One issue is that the environmental qualification carried out for 

equipment in currently operating reactors is carried out for the ambient conditions expected for 

Reserve chute works, float to ground

Event Tree (ET)

IE Main 

Chute

Reserve

Chute End State

Main chute works, float to ground

Both chutes fail, jumper casualty

Initiating Event: 

Jump from airplane

System succeeds

System fails
4. GLOSSARY

 4-38 

TERM AND 

DEFINITION DISCUSSION 

conditions that 

could exist when 

functionality is 

required. 

design-basis accidents, and these conditions are likely to differ from those encountered in a 

severe accident. 10 CFR 50.49 (Ref.18) establishes requirements for environmental 

qualification for safety electric equipment important to safety for nuclear power plants. 

The definition provided was based on the definition in the NRC Web site Glossary (Ref. 36). 

Epistemic Uncertainty 

(see 

Uncertainty)

Epistemic uncertainty is a type of uncertainty and is defined under “Uncertainty.” 

Error Factor (Human)

A measure of 

uncertainty 

associated with 

probability 

estimates. 

In a PRA, error factors are used to account for the uncertainty of the various parameters in the 

PRA model, such as the probability associated with a component failure or human error event. 

The error factor is a measure of the spread of the distribution of a parameter in the calculation 

of these types of failure. 

The term human error factor refers to the uncertainty in the probability of a human error. The 

probability of a human error event is often referred to as the human error probability. 

From a mathematical perspective, when the uncertainty distribution for an event failure 

probability is characterized by the log-normal distribution, uncertainties on these probability 

estimates are expressed as error factors. The lognormal error factor is defined as the 95th

percentile divided by the median (i.e., the 50th percentile).

Event Scenario 

(see Accident 

Sequence)

The term event scenario has the same meaning as accident sequence and is defined under 

“Accident Sequence.”

Event Sequence 

(see Accident 

Sequence)

The term event sequence has the same meaning as accident sequence and is defined under 

“Accident Sequence.”

Event Sequence Analysis 

(see Accident 

Sequence 

Analysis)

The term event sequence analysis is another way of describing an accident sequence and is 

defined under “Accident Sequence Analysis.”

Event Sequence Class 

(see Accident 

Sequence 

Class)

The term event sequence class has the same meaning as accident sequence class and is 

defined under “Accident Sequence Class.”

Event Sequence Diagram 

A flowchart that 

represents various 

accident scenarios 

that can occur as a 

result of a plant 

upset condition. 

(see Event Tree, 

Top Event) 

In a PRA, event sequence diagrams (ESDs) sometimes have been used to represent the 

progression of an initiating event by asking questions about successes and failures of plant 

responses to that initiating event. Each leg of the ESD ends with a successful or undesired 

end state for individual sequences. Once an ESD is developed, it can be mapped into an event 

tree, which relates more directly to a practical quantification of accident scenarios in a PRA. 

However, in comparison to event trees, ESDs tend to include additional supporting details on 

plant design and operational information that illustrates why a branch in the event tree 

proceeds down a particular success path. In this regard, ESDs are related to event trees in 
4. GLOSSARY

 4-39 

TERM AND 

DEFINITION DISCUSSION 

that they can help document the assumptions used in constructing an event tree. 

The following figure illustrates a simple ESD. The oval to the left corresponds to top events in 

the “jump from airplane” event tree. 

Event Sequence Group 

(see Accident 

Sequence 

Class)

The term event sequence group has the same meaning as accident sequence group and is 

defined under “Accident Sequence Class.”

Event Sequence Type 

(see Accident 

Sequence 

Class)

The term event sequence type has the same meaning as accident sequence type and is 

defined under “Accident Sequence Class.”

Event Tree 

A logic diagram 

that graphically 

represents the 

various scenarios 

that can occur as a 

result of an upset 

condition. (see 

Accident 

Sequence, 

Containment Event 

Tree, Top Event, 

Accident 

Progression Event 

Tree, Bridge Tree)

In a PRA, event trees are used in various parts of the analysis: 

• Level 1 event trees provide the plant response logic from the initiating event to the 

successful prevention of core damage or core damage end states. 

• Bridge event trees often are used as the interface between the Level 1 event trees 

and Level 2 event trees, in that they define the initial conditions for the Level 2 

analysis (i.e., plant damage states), based on the plant conditions when core damage 

occurs. 

• Level 2 event trees provide the plant response logic from the plant damage states to 

the successful prevention of containment failure or containment failure and release 

end states. In Level 2, these event trees are referred to as a containment event tree 

or accident progression event tree. 

Event trees start with an initiating event and progress through questions about successes and 

failures of plant responses to that initiating event, ending with a successful or undesired end 

state for individual sequences. Individual sequences are pathways through the event tree. 

An example of a simple event tree is shown below: 

Evaluate the Status 

of Both Chutes 

(Main and Reserve)

Float to 

Ground

Initiating 

Event—

Jump from 

Airplane

Main 

Chute 

Works?

Reserve 

Chute 

Works?

Jumper

Casualty

Float to 

Ground

Yes Yes

Yes

No

No
4. GLOSSARY

 4-40 

TERM AND 

DEFINITION DISCUSSION 

An event tree has also been defined as: 

• “A logic diagram that begins with an initiating event or condition and progresses 

through a series of branches that represent expected system or operator performance 

that either succeeds or fails. The progression arrives at either a successful or failed 

end state.” (Ref. 2)

• “An event tree graphically represents the various accident scenarios that can occur as 

a result of an initiating event (i.e., a challenge to plant operation). Toward that end, an 

event tree starts with an initiating event and develops scenarios, or sequences, based 

on whether a plant system succeeds or fails in performing its function. The event tree 

then considers all of the related systems that could respond to an initiating event, until 

the sequence ends in either a safe recovery or reactor core damage.” (Ref. 36) 

Event Tree Sequence 

(see Accident 

Sequence)

The term event tree sequence is a specific description of an accident sequence and is defined 

under “Accident Sequence.”

Event Tree Top Event 

(see Top Event) The term event tree top event is discussed under the discussion for the term “Top Event.” An 

illustration of an event tree top event is shown under the discussion for the term “Event Tree.”

Exclusion Area Boundary 

The boundary of 

the area 

surrounding the 

plant where the 

plant owner has 

the authority to 

determine all 

PRA consequence calculations usually are concerned with the consequences outside of the 

exclusion area boundary. The exclusion area is that area around the plant where public 

residence is not normally permitted. The exclusion area boundary is the inner edge of the low 

population zone. 

 

The exclusion area and its boundary are important for reactor siting considerations as a 

location where acceptable dose limits following a release must be met. For example, Title 10 

Reserve

Chute

Fails

Event Tree (ET)

Fault Tree (FT)

ET Top Events

Initiating Event: 

Jump from airplane

FT Top Events

Chute not

deployed

Chute 

tangled

Rip cord

breaks

Auto activation 

device fails

Alternator

malfunction

Battery is 

dead

Reserve chute works, float to 

IE Mai

Chut

Reserve

Chute End 

Main chute works, float to ground

Both chutes fail, jumper 

System 

System 
4. GLOSSARY

 4-41 

TERM AND 

DEFINITION DISCUSSION 

activities, including 

exclusion or 

removal of 

personnel and 

property.

of the Code of Federal Regulations (10 CFR) 100.11, “Determination of Exclusion Area, Low 

Population Zone, and Population Center Distance” (Ref. 26), states that the applicant (of a 

siting permit) should determine the following: an exclusion area of such size that an individual 

located at any point on its boundary for 2 hours immediately following onset of the postulated 

fission product release would not receive a total radiation dose to the whole body in excess of 

25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure. 

The definition provided is based on the definition in the NRC Web site Glossary (Ref. 36).

Expert Elicitation 

A formal, 

structured, and 

documented 

process in which 

judgments from 

expert(s) are 

obtained. (see 

Expert Judgment) 

In a PRA, expert elicitation may be used to obtain information from technical experts on topics 

that are uncertain. An expert elicitation is a process in which experts are assembled and their 

judgment is sought and aggregated in a formal way. 

NUREG-1563 (Ref. 57) states, “Typically an elicitation is conducted to evaluate uncertainty. 

The uncertainty could be associated with: the value of a parameter to be used in a model; the 

likelihood and frequency of various future events; or the relative merits of alternative conceptual 

models. In each of these cases, the information regarding uncertainty would be represented by 

encoding the subjective probabilities from each subject-matter expert.” 

An expert elicitation is a more formal process than expert judgment. Expert judgment may be 

the opinion of one or more experts, whereas expert elicitation is a highly structured process in 

which the opinions of several experts are sought, collected, and aggregated in a very formal 

way. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 2).

Expert Judgment 

Information (or 

opinion) provided 

by one or more 

technical experts 

that is based on 

their experience 

and knowledge. 

(see Expert 

Elicitation)

In a PRA, expert judgment is used when there is a lack of information. For example, if certain 

parameter values are unknown, or there are questions about phenomenology in accident 

progression, then expert judgment may be used. Expert judgment may be part of a structured 

approach, such as expert elicitation. 

Obtaining expert judgment is not necessarily as formal as invoking an expert elicitation 

process. Expert judgment may be the opinion of one or more experts, whereas expert 

elicitation is a highly structured process in which the opinions of several experts are sought, 

collected, and aggregated in a very formal way. 

NUREG-1563 (Ref. 57) states, “expert judgments may also be opinions that can be analyzed 

and interpreted, and used in subsequent technical assessments. Expert judgments can be 

either qualitative or quantitative. Expert judgments also can be judgments about uncertain 

quantities or judgments about value preferences.” 

The ASME/ANS PRA Standard (Ref. 2) defines expert judgment as “information provided by a 

technical expert, in the expert’s area of expertise, based on opinion, or on an interpretation 

based on reasoning that includes evaluations of theories, models, or experiments.” 

Exposure 

The state of being 

subjected to 

ionizing radiation. 

(see Exposure 

Time, Cloudshine, 

Groundshine, 

Inhalation, 

Ingestion, Skin 

Deposition, Health 

Effects)

In a Level 3 PRA, the offsite health effects resulting from exposure to ionizing radiation is 

considered. As stated in the NRC Web site Glossary (Ref.36), exposure occurs through 

absorption of ionizing radiation because of an external source or an internal exposure caused 

by inhalation or ingestion of a radioisotope. Acute exposure is a large exposure received over 

a short period of time. Chronic exposure is exposure received over a long period of time, such 

as during a lifetime. 
4. GLOSSARY

 4-42 

TERM AND 

DEFINITION DISCUSSION 

Exposure Pathways

The various 

means by which 

exposure to 

radiation occurs 

and dose to 

recipients is 

delivered. (See 

Exposure, 

Exposure Time, 

Cloudshine, Water 

Immersion, 

Groundshine, 

Inhalation, 

Ingestion, Skin 

Deposition, Health 

Effects)

In a Level 3 PRA, exposure pathways to an individual are assumed for the consequence 

calculations. Cloudshine, sometimes referred to as air submersion, is the pathway by which 

external dose is given to an individual exposed to contaminated air; water immersion is a 

pathway by which external dose is given to an individual immersed in contaminated water (e.g., 

by bathing or swimming); inhalation is the pathway by which internal dose is given by breathing 

in contaminated air (resuspension inhalation is the pathway by which internal dose is given to 

an individual from breathing resuspended material previously deposited on the ground); 

ingestion is the pathway by which internal dose is given from consuming contaminated food or 

water; groundshine is the pathway by which external dose is given to an individual standing on 

contaminated ground; and skin deposition is exposure resulting from radioactive material 

deposited directly onto the surface of the body. 

Exposure Rate 

(see Dose Rate) The exposure has the same meaning as dose rate and is defined under “Dose Rate”.

Exposure Time 

Duration of 

radiation exposure 

used to estimate 

the dose received 

by an individual.

(see Health 

Effects, Exposure)

In a Level 3 PRA, the exposure time is needed to calculate the dose and subsequent health 

consequences to affected individuals. 

The PRA considers two types of exposures: acute and chronic. An acute exposure involves a 

large exposure received over a short period of time. Chronic exposures involve exposure 

received over a long period of time, such as during a lifetime. 

External Event

The term external 

event is no longer 

used and has 

been replaced by 

the term external 

hazard. (see 

Hazard) 

A full scope PRA includes accidents resulting from both internal and external hazards. Internal 

hazards could include internal events, internal floods, and internal fires. External hazards could 

include seismic events, high winds, external floods, and other external hazards. 

The no-longer-used term, external event, is defined in the ASME/ANS PRA Standard (Ref. 2) 

as “an event originating outside a nuclear power plant that directly or indirectly causes an 

initiating event and may cause safety system failures or operator errors that may lead to core 

damage or large early release. Events such as earthquakes, tornadoes, and floods from 

sources outside the plant and fires from sources inside or outside the plant are considered 

external events. By historical convention, loss of offsite power not caused by another external 

event is considered to be an internal event.” 

Historically, the difference between an internal event and an external event was the equipment 

boundary. The internal event represented something that occurred “internal” to the boundary of 

the piece of equipment. Conversely, occurrences external to the equipment boundary but 

within the plant boundary were classified as external events. With time, the definition for 

internal hazards has come to encompass all the hazards within the plant boundary, not just 

within the equipment. Thus, the external events have changed to currently represent events 

that occur outside the plant boundary but can cause undesired outcomes or conditions leading 

to plant equipment damage. Loss of offsite power is still considered an internal event. 

The term external event and external hazard have been used incorrectly interchangeably. The 

term external event is no longer used and has been subsumed by the term external hazard. 

 
4. GLOSSARY

 4-43 

TERM AND 

DEFINITION DISCUSSION 

External Flood 

A flood initiated 

outside the plant 

boundary that can 

affect the 

operability of the 

plant. (see 

Hazard, External 

Flood Analysis, 

Internal Flood)

In a PRA, external floods are a specific hazard group in which the flood occurs outside the 

plant boundary. The PRA considers floods because they have the potential to cause 

equipment failure by the intrusion of water into plant equipment through submergence, spray, 

dripping, or splashing. 

The definition provided was based on the definition in NUREG-1742 (Ref. 59). 

External Flood Analysis 

A process used to 

assess potential 

risk from external 

floods. (see 

Hazard Analysis, 

External Flood)

In a PRA, an external flood analysis quantifies the risk contribution (e.g., core damage 

frequency and large release frequency) as a result of an external flood. The analysis models 

the potential failures of plant systems and components from external floods, as well as random 

failures. Floods have the potential to cause equipment failure by the intrusion of water into 

plant equipment through submergence, spray, dripping, or splashing. The likelihood of an 

external flood is determined through an external flood hazard analysis, which evaluates the 

frequency of occurrence of different external flood severities. The frequency of the external 

flood is used as input to the model used to assess external flood risk. 

External Flood Fragility Analysis 

(see Fragility 

Analysis)

The term external flood fragility analysis is a type of fragility analysis and is included in the 

discussion to the term “Fragility Analysis.” 

External Flood Hazard Analysis 

(see Hazard 

Analysis)

The term external flood hazard analysis is a specific type of hazard analysis and is defined 

under “Hazard Analysis.” 

External Flood Plant Response Analysis/Model

(see Plant 

Response 

Analysis/Model)

The term external flood plant response analysis is a type of plant response analysis and is 

included under “Plant Response Analysis/Model.” 

External Hazard 

(see Hazard) The term external hazard is related to the term hazard and is defined under “Hazard.”

External Hazard Analysis

(see Hazard 

Analysis)

The term external hazard analysis is a type of hazard analysis and is defined under “Hazard 

Analysis.” 

Failure Mechanism 

The fault 

associated with a 

component that 

causes it to 

malfunction. (see 

Failure Mode)

In a PRA, the concept of failure mechanism is used to explain the immediate cause of 

component failure. The fault that causes failure could be electrical, mechanical, chemical, 

physical, thermal, or human error. An example of a failure mechanism would be an electrical 

short in the electric motor winding that causes failure of a pump to start. 

The ASME/ANS PRA Standard (Ref. 2) defines failure mechanism as “any of the processes 

that results in failure modes, including chemical, electrical, mechanical, physical, thermal, and 

human error.” 

While failure mechanism is a cause of failure, failure mode is the functional manifestation of 

failure (e.g., failure to start, failure to run).
4. GLOSSARY

 4-44 

TERM AND 

DEFINITION DISCUSSION 

Failure Mode 

The manner in 

which a 

component fails to 

perform its 

function. (see 

Failure 

Mechanism, 

Failure Modes and 

Effects Analysis)

In a PRA, the failure modes of a component are represented as basic events, and while it is a 

visible manifestation of failure, it is distinguished from failure mechanism, which is a cause of 

failure. Failure of a component is distinguished by its failure mode. Each failure mode is 

modeled separately, with its own failure probability. Failure mode is failure in a distinct 

functionality of a component that is necessary for it to successfully operate (e.g., failure modes 

of a valve might be failure to open, failure to close, or inadvertent opening). Failure of a pump 

may be distinguished into two separate failure modes, namely failure to run or failure to start. 

In a fire PRA, spurious (unintended) operation is also defined as a failure mode. 

The ASME/ANS PRA Standard (Ref. 2) defines failure mode as “a specific functional 

manifestation of a failure (i.e., the means by which an observer can determine that a failure has 

occurred) by precluding the successful operation of a piece of equipment, a component, or a 

system (e.g., fails to start, fails to run, leaks).” 

A failure modes and effects analysis can be used to identify component failure modes and 

evaluate their effects on other components, subsystems, and systems. 

Failure Modes and Effects Analysis 

A process for 

identifying failure 

modes of specific 

components and 

evaluating their 

effects on other 

components, 

subsystems, and 

systems. (see 

Failure Mode)

In a PRA, a failure modes and effects analysis (FMEA) generally is not used except to identify 

initiating events for a new plant design with no operational history or failure data. A FMEA is 

aimed at analyzing the effects of a single component or function failure on other components, 

systems, and subsystems. A FMEA can be useful in identifying initiating events that involve 

support system failures and the expected effects on the plant (especially on mitigating 

systems). 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 2). 

Failure Probability 

 (see Probability) The term failure probability is a specific type of probability and is defined under “Probability.” 

Fatality (Early, Latent, Prompt, Latent Cancer) 

Death occurring as 

a result of 

exposure to 

radioactive 

material. (see 

Exposure, 

Quantitative Health 

Objectives)

In a Level 3 PRA, one of the objectives is to calculate the dose received by the population 

surrounding the plant as a result of a potential release of radioactive material. Depending on 

the amount of dose and the duration over which it is received, early and latent fatalities can 

occur. The risk of incurring fatalities, both early and latent fatalities, is one of the most 

important outputs of a Level 3 PRA. 

Early fatalities, synonymous with prompt fatalities, are defined as deaths from the acute effects 

of radiation that may occur within a few months of the exposure. Latent cancer fatalities are 

defined as deaths from cancer caused by chronic effects of radiation exposure; latent cancer 

fatalities may occur years after the exposure. 

Prompt or early fatalities are usually the result of acute exposures (large exposure received 

over a short period of time). Latent fatalities resulting from cancer that became active after a 

latent period can result from exposure from early pathways (e.g., groundshine, cloudshine, and 

skin deposition), as well as long-term pathways (e.g., resuspension inhalation and ingestion). 

Fatality Risk (Early, Latent, Prompt)

(see Fatality) The fatality risk (early or prompt fatality risk, latent fatality risk) is the risk involving fatalities 

caused by exposure to radioactive materials and is defined in the discussion under “Fatality.”
4. GLOSSARY

 4-45 

TERM AND 

DEFINITION DISCUSSION 

Fault Tree 

A deductive logic 

diagram that 

graphically 

represents the 

various failures 

that can lead to a 

predefined 

undesired event. 

(see Top Event, 

Event Tree)

In a PRA, fault trees are used to depict the various pathways that lead to a system failure. 

Fault trees describe how failures of top events occur because of various failure modes of 

components, human errors, initiator effects, and failures of support systems that combine to 

cause a failure of a top event in the event trees. 

A fault tree also has been defined as: 

• “A deductive logic diagram that depicts how a particular undesired event can occur as 

a logical combination of other undesired events.” (Ref. 2) 

• “A fault tree identifies all of the pathways that lead to a system failure. Toward that 

end, the fault tree starts with the top event, as defined by the event tree, and 

identifies ...what equipment and operator actions, if failed, would prevent successful 

operation of the system. All components and operator actions that are necessary for 

system function are considered. Thus, the fault tree is developed to a point where 

data are available for the failure rate of the modeled component or operator action.” 

(Ref. 36) 

The following is an example of a fault tree diagram: 

Fault Tree Top Event 

 (see Top Event) The term fault tree top event is a type of top event in a PRA model and is defined under “Top 

Event.” An illustration of a fault tree top event is shown under the discussion for the term 

“Event Tree.” 

Feed and Bleed, Bleed and Feed

A method of core 

cooling in a 

pressurized-water 

reactor by 

providing cooling 

water to the 

reactor while 

removing heated 

coolant through 

open reactor 

vessel relief 

valves.

In a PRA, feed and bleed is often included as a core heat removal option for pressurized-water 

reactors when secondary cooling (e.g., auxiliary feedwater) is unavailable. To remove the core 

(i.e., decay) heat from the reactor vessel, water from a storage tank or recirculated from the 

containment sump is injected into the reactor vessel through safety or nonsafety grade 

pumping systems (feed), and the pressurizer power-operated relief valves (PORVs) or safety 

valves are opened to discharge the heated coolant from the reactor vessel (bleed). 

The terms feed and bleed and bleed and feed are similar in meaning and often used 

interchangeably. However, in certain instances, these terms may be used to distinguish the 

manner in which this decay heat removal option is accomplished. In some plants, the injection 

pumps may be capable of injecting coolant at full reactor coolant system pressure while 

discharging reactor coolant through the safety valves. In this design, the injection of water 

Pump A

failed

Pump C 

failed

Pump B 

failed

Pump D 

failed

Pump E 

failed

Pump Systems 

failed

Pump System I

failed

Pump System II 

failed

A B C D E

Transfer to 

AC Power Failed 

FT
4. GLOSSARY

 4-46 

TERM AND 

DEFINITION DISCUSSION 

(feed) can occur before opening the safety valves (bleed), such that this decay heat option may 

be referred to as feed and bleed. In other plants, the injection pumps are not capable of 

injecting coolant at full system pressure, but instead must rely upon operator actions to open 

one or more PORVs in a timely matter. In this situation, the reactor vessel pressure is first 

reduced by the release of coolant (bleed), with subsequent injection of coolant from the 

injection pumps (feed). This decay heat option may be referred to as bleed and feed. 

Fire Probabilistic Risk Assessment Plant Response Model (Analysis)

(see Plant 

Response 

Analysis)

The term fire probabilistic risk assessment plant response analysis is a type of plant response 

analysis and is defined under “Plant Response Analysis/Model.”

The term fire probabilistic risk assessment plant response model is also a technical element for 

internal fires in the ASME/ANS PRA Standard (Ref. 2) whose objective is to identify the 

initiating events that can be caused by a fire event and develop a related accident sequence 

model, and to depict the logical relationships among equipment failures (both random and fire 

induced) and human failure events for core damage frequency and large early release 

frequency assessment when combined with the initiating event frequencies. 

Fission Product (Release) 

The byproduct of 

the nuclear fission 

process. (See 

Radioactive 

Material, 

Radionuclide)

In a PRA, the terms radionuclide, radioactive material, and fission product are used 

interchangeably. These terms are meant to refer to the substance that is the source of the risk 

being evaluated. A fission product release, therefore, refers to the release of the radioactive 

material from the reactor and from the containment that could adversely affect public health 

and safety. 

The NRC Web site Glossary (Ref. 36) defines fission product as, “The nuclei (fission 

fragments) formed by the fission of heavy elements, plus the nuclide formed by the fission 

fragment’s radioactive decay.”

Fission Product Release 

(see Radioactive 

Material Release)

For purposes of a Level 2 and Level 3 PRA, the term fission product release is used 

interchangeably with radioactive material release. 

Fragility 

The likelihood that 

a component, 

system, or 

structure will 

cease to function 

given the 

occurrence of a 

hazard event of a 

certain intensity. 

(see Fragility 

Analysis, High 

Confidence of Low 

Probability of 

Failure, Fragility 

Curve) 

In a PRA, fragility is a concept used in the evaluation of external hazards. The fragility of a 

component, system, or structure is generally calculated for seismic events, high wind events, 

and external flood events 

Since a given component may fail because of various mechanisms (e.g., seismic motion may 

cause anchor failure, structural failure, systems interactions), fragility can be calculated for 

each of these failure mechanisms, or the results can be presented for the dominant 

mechanism. 

The ASME/ANS PRA Standard (Ref. 2) states, “fragility of a structure, system or component 

(SSC) is the conditional probability of its failure at a given hazard input level. The input could 

be earthquake motion, wind speed, or flood level.” 

Fragility Analysis (External Flood, High Winds, Other External Hazards, Seismic)

Estimation of the 

likelihood that a 

given component, 

system, or 

In a PRA, fragility analysis identifies the components, systems, and structures susceptible to 

the effects of an external hazard and estimates their fragility parameters. Those parameters 

are then used to calculate fragility (conditional probability of failure) of the component, system, 

or structure at a certain intensity level of the hazard event. Fragility analysis considers all 
4. GLOSSARY

 4-47 

TERM AND 

DEFINITION DISCUSSION 

structure will 

cease to function 

given the 

occurrence of a 

hazard event of a 

certain intensity. 

(see Fragility, 

Fragility Curve) 

failure mechanisms due to the occurrence of an external hazard event and calculates fragility 

parameters for each mechanism. This is true whether the fragility analysis is used for an 

external flood hazard, fire hazard, high wind hazard, seismic hazard, or other external hazards. 

For example, for seismic events, anchor failure, structural failure, and systems interactions are 

some of the failure mechanisms that would be considered. 

Fragility Curve

A graph that plots 

the likelihood that 

a structure, system 

or component will 

fail versus the 

increasing 

intensity of a 

hazard event. 

(see Fragility, 

Fragility Analysis)

In a PRA, fragility curves generally are used in seismic analyses and provide the conditional 

frequency of failure for structures, systems, or components as a function of an 

earthquake-intensity parameter, such as peak ground acceleration. Fragility curves also can 

be used in PRAs examining other hazards, such as high winds or external floods. 

Frequency (Accident Sequence, Core Damage, Initiating Event, Large Early Release, 

Large Release, Radioactive Material Release) 

The expected 

number of 

occurrences of an 

event or accident 

condition 

expressed per unit 

of time. (see 

Probability)

In a PRA, a frequency is calculated for various events. For a Level 1 PRA, frequencies are 

calculated for the initiating events and for the core damage accident sequences; the latter 

frequencies are summed to provide an overall core damage frequency. For a Level 2 PRA, 

frequencies are calculated for the plant damage states and for the release of radioactive 

material (e.g., large early release frequency, large release frequency, and the overall 

radioactive material release frequency). For a Level 3 PRA, frequencies are calculated for 

accident consequences (i.e.; early and latent fatalities) and, sometimes, economic 

consequences. 

Frequency is normally expressed in events per plant (or reactor) operating year or events per 

plant (or reactor) calendar year. 

The subset terms of frequency can be defined as follows: 

• Accident Sequence Frequency: The frequency associated with a series of events 

that follow from a particular initiating event, through system and operator responses, 

and ultimately to a well-defined end state, such as core damage. (see Accident 

Sequence) 

• Core Damage Frequency: The sum of the accident sequence frequencies of those 

accident sequences whose end state is core damage. 

• Initiating Event Frequency: The frequency of an event originating from an internal or 

external hazard that both challenges normal plant operation and requires successful 

mitigation. 

• Large Early Release Frequency: The frequency of a rapid, unmitigated release of 

airborne fission products from the containment to the environment that occurs before 

effective implementation of offsite emergency response, and protective actions, such 

that there is a potential for early health effects. 

• Large Release Frequency: The Commission has not approved a formal definition of 

a large release or a large release frequency. One informal definition for large release 

frequency is the frequency of an unmitigated release of airborne fission products from 

the containment to the environment that is of sufficient magnitude to cause severe 
4. GLOSSARY

 4-48 

TERM AND 

DEFINITION DISCUSSION 

health effects, regardless of its timing. The history of the use of the term “Large 

Release Frequency” is provided in SECY-13-0029 (Ref.100). (see Large Release)

• Radioactive Material Release Frequency: The frequency of the release of radioactive 

material from the containment to the environment. This may refer to the total 

frequency of all releases regardless of size or timing. The radioactive material 

release frequency may also be subdivided depending on the size and timing of the 

release. Large early release frequency and large release frequency are defined 

above. A small early release frequency can be defined as the frequency of early 

releases of low enough magnitude to have minimum potential for early health effects. 

A small late release frequency can be defined as the frequency of late releases of low 

enough magnitude and with a long enough delay to have minimum potential for early 

health effects. A large late release frequency can be defined as the frequency of late 

releases that have sufficient magnitude to cause severe health effects, but which 

occur in a timeframe that allows effective emergency response and protective actions 

so that the offsite health effects will be significantly reduced compared to those of a 

large early release. (see Radioactive Material Release) 

In some instances, the terms frequency and probability are used interchangeably, but 

incorrectly. Unlike frequency, probability represents a unitless quantity. 

Frequentist Analysis, Frequentist Estimation, Frequentist Statistics 

A type of data 

analysis that relies 

solely on actual 

occurrences of the 

event under 

consideration. 

(see Bayesian 

Analysis)

In a PRA, frequentist analysis is only used when occurrences of an event are sufficiently 

abundant such that a reliable estimate of event probability can be expressed as the ratio of 

number of event occurrences to total number of occurrences in which the event could occur. In 

frequentist statistics, error probability can be calculated as the number of errors experienced 

over some number of tries divided by the number of tries. 

 

In the frequentist approach, the probability of a random event is interpreted as the fraction of 

times that the event would occur, in a large number of trials. 

In risk analysis, both frequentist and Bayesian analysis may be used, depending on whether 

occurrence data is sufficiently abundant. 

The terms frequentist analysis, frequentist estimation, and frequentist statistics are used 

interchangeably. 

Frequentist Estimation 

(see Frequentist 

Analysis) 

The term frequentist estimation has the same meaning as frequentist analysis and is defined 

the same as the term “Frequentist Analysis.” 

Frequentist Statistics 

(see Frequentist 

Analysis)

The term frequentist statistics has the same meaning as frequentist analysis and is defined the 

same as the term “Frequentist Analysis.” 

Front-Line System 

A system used to 

directly provide a 

safety function. 

(see Support 

System)

In a PRA, front-line systems are modeled to help represent the ways in which a plant can 

prevent core damage or prevent containment failure. The ASME/ANS PRA Standard (Ref. 2) 

defines a front-line system as “a system (safety or non-safety) that is capable of directly 

performing one of the accident mitigating functions (e.g., core or containment cooling, coolant 

makeup, reactivity control, or reactor vessel pressure control) modeled in the PRA.” 

In some references, the definition of a front-line system only includes safety-related systems. 

However, other definitions are more generalized to include the possibility that a front-line 

system can be a nonsafety system, such as the ASME/PRA Standard definition cited above. 
4. GLOSSARY

 4-49 

TERM AND 

DEFINITION DISCUSSION 

Full Power 

The state of 

operation in which 

the reactor is 

critical and 

producing 100-

percent power. 

(see At-Power, 

Low Power and 

Shutdown)

A PRA models the different plant operating states (POSs) of the plant. Operation at full power 

is one POS, while several POSs are needed to characterize the plant during the various stages 

of low-power and shutdown. These POSs are distinguished in the PRA model because the 

plant response (e.g., accident sequences) differs during different POSs. 

Historically, the term full power was used to denote any power level between low power and 

100-percent power. This definition has been recently modified so that full power currently 

refers just to 100-percent power of the reactor core, while at-power covers the range of powers 

from low power up to and including 100-percent power. 

Full-Scope PRA 

A PRA that 

considers all the 

various challenges 

that could 

contribute to the 

risk posed by the 

plant to the health 

and safety of the 

public. (see PRA, 

Risk Metric) 

A full-scope PRA generally only considers the reactor and associated systems and is 

comprised of three distinct parts, referred to as Levels. The full-scope PRA includes a Level 1 

(core damage), Level 2 (radioactive material release) and Level 3 (consequences) PRA that 

addresses both internal and external hazards at all power modes (at-power, low-power, and 

shutdown). These power modes commonly are referred to as plant operating states (POSs). 

A full-scope site PRA may also consider risks from the spent fuel pool and any other fuel 

storage facility on site. Offsite risk metrics in the Level 3 portion may include both health 

effects and economic considerations brought about by the release of radioactive material. 

Fussell-Vesely Importance 

(see Importance 

Measure) 

The term Fussell-Vesely importance is one type of importance measure and is defined under 

“Importance Measure.” 

General Transient 

(see Transient) The term general transient has the same meaning as transient and is defined under 

“Transient.”

Groundshine 

Exposure from 

radioactive 

material deposited 

on the ground. 

(see Exposure 

Pathways, 

Cloudshine, Water 

Immersion, 

Inhalation, 

Ingestion, Skin 

Deposition)

In a Level 3 PRA, for the consequence calculation groundshine is one of the assumed 

pathways by which an individual can receive doses. The pathways of exposure include: (1) 

direct external exposure from radioactive material in a plume, principally due to gamma 

radiation (air immersion or cloudshine), (2) direct exposure from radioactive material in 

contaminated water given to an individual immersed in the water, (3) exposure from inhalation 

of radioactive materials in the cloud and resuspended material deposited on the ground, (4) 

exposure to radioactive material deposited on the ground (groundshine), (5) radioactive 

material deposited onto the body surfaces (skin deposition), and (6) ingestion from deposited 

radioactive materials that make their way into the food and water pathway. 

Hazard (Type (Internal, External), Group, Event) 

Anything that has 

the potential to 

cause an 

undesired event or 

condition that 

leads to equipment 

damage. (see 

Hazard Analysis, 

Initiating Event)

In a PRA, there are three different uses of the term hazard as an adjective (the terms hazard 

and plant hazard tend to be correctly used interchangeably): types, groups, and events. The 

first, hazard type, classifies hazards as either internal or external to the plant. Within each 

hazard type, internal and external, there are subcategories, which are referred to as hazard 

groups. For internal hazards, this hazard group includes internal events, internal floods, and 

internal fires. For external hazards, this includes seismic events, high winds, external floods, 

and other external hazards. Finally, a hazard event represents the events brought about by the 

occurrence of the specified hazard. For example, those of interest in a PRA are ones that 

directly or indirectly cause an initiating event and may further cause safety system failures or 

operator errors that may lead to core damage or radioactive material release. 
4. GLOSSARY

 4-50 

TERM AND 

DEFINITION DISCUSSION 

As defined in Regulatory Guide 1.200 (Ref.91), a hazard group “is a group of similar causes of 

initiating events that are assessed in a PRA using a common approach, methods, and 

likelihood data for characterizing the effect on the plant.” 

A hazard event is described in terms of the specific levels of severity of impact that a hazard 

can have on the plant. The hazard event is an occurrence of the phenomenon that can result 

in a plant trip and possibly other damage when the plant is at-power or result in the loss of a 

key safety function during non-power operations. The ASME/ANS PRA Standard (Ref. 2) 

states that there “is a range of hazard events associated with any given hazard, and, for 

analysis purposes, the range can be divided into bins characterized by their severity.” An 

example of the overall concept of hazard, hazard event, and initiating event is as follows: 

• Earthquakes are a hazard;

• 0.1g, 0.3g, 0.5g earthquakes and their associated spectral shapes and time histories 

may be defined as hazard events;

• A manual plant trip is typically the initiating event for the 0.1g earthquake, and a loss 

of offsite power is typically assumed as the initiating event for the 0.3g and 0.5g 

earthquakes.

The ASME/ANS PRA Standard (Ref. 2) defines a hazard as “an event or a natural 

phenomenon that poses some risk to a facility. Internal hazards include events such as 

equipment failures, human failures, and flooding and fires internal to the plant. External 

hazards include events such as flooding and fires external to the plant, tornadoes, 

earthquakes, and aircraft crashes.”

Hazard Analysis (External, External Flood, High Wind, (Probabilistic) Seismic, Other 

Hazards) 

A process used to 

assess potential 

plant challenges, 

including natural 

phenomena, and 

to assess their 

likelihood, typically 

as a function of 

severity. 

In a PRA, it is important to identify and characterize the nature and causes of specific types of 

hazards. A hazard represents an event or a natural phenomenon that poses some challenge 

to a facility. Examples of external hazards typically evaluated in a PRA include external floods, 

high winds, seismic events, and external fires. A hazard analysis is used to evaluate the 

frequency of occurrence of different severities for the hazard being analyzed. Results from the 

hazard analysis are used as input to the PRA, which subsequently examines the hazards with 

respect to risk. 

Listed below are specific types of hazard analyses: 

• External hazard analysis: The objective is to evaluate the frequency of occurrence of 

different severities or intensities of external events or natural phenomena (e.g., 

external floods or high winds). 

• External flood hazard analysis: The objective is to evaluate the frequency of 

occurrence of different external flood severities. 

• High wind hazard analysis: The objective is to evaluate the frequency of occurrence 

of different intensities of high winds. 

• (Probabilistic) seismic hazard analysis: A seismic hazard analysis expresses “the 

seismic hazard in terms of the frequency of exceedance for selected ground motion 

parameters during a specified time interval. The analysis involves identification of 

earthquake sources, evaluation of the regional earthquake history, and an estimate of 

the intensity of the earthquake-induced ground motion at the site. As stated in 

Regulatory Guide 1.200 (Ref. 86): “at most sites, the objective is to estimate the 

probability or frequency of exceeding different levels of vibratory ground motion” The 

term probabilistic seismic hazard analysis is similar in meaning to the definition of 

seismic hazard analysis as stated above. 
4. GLOSSARY

 4-51 

TERM AND 

DEFINITION DISCUSSION 

• Other hazards analysis: Evaluates the frequency of occurrence of different intensities 

of other internal or external hazards (e.g., external fires). 

The ASME/ANS PRA Standard (Ref. 2) defines hazard analysis as “the process to determine 

an estimate of the expected frequency of exceedance (over some specified time interval) of 

various levels of some characteristic measure of the intensity of a hazard (e.g., peak ground 

acceleration to characterize ground shaking from an earthquake). The time period of interest is 

often taken as 1 year, in which case the estimate is called the annual frequency of 

exceedance.” 

An example of a hazard curve is shown below. 

Hazard Event 

(see Hazard) The term hazard event is related to the term hazard and is defined under “Hazard.” 

Hazard Group 

(see Hazard) The term hazard group is related to the term hazard and is defined under “Hazard.” 

Hazard Type

(see Hazard) The term hazard type is related to the term hazard and is defined under “Hazard.”

Health Effects 

The effects of 

radioactive material 

on the health and 

safety of exposed 

individuals. (see 

Quantitative Health 

Objectives, 

Accident 

Consequence, 

Exposure Time, 

Land 

Contamination)

In a Level 3 PRA, the health effects represent the main component of the calculated risk. 

Health effects from radioactive material (i.e., ionizing radiation) usually are distinguished as 

acute or latent. 

Acute health effects are adverse health symptoms (e.g., fatalities) occurring within a short time 

(days or months rather than years) of an exposure to large radiation doses. Acute fatalities and 

injuries are expected to occur within 1 year of an accident or sooner. 

Latent health effects refer to cancer deaths that may occur with a considerable latency period, 

from approximately 2 to 25 years, depending on the type of cancer involved. 

Public health effects refer to illnesses or fatalities to the population beyond the site boundary 

resulting from the release of radiation. 
4. GLOSSARY

 4-52 

TERM AND 

DEFINITION DISCUSSION 

High Confidence of Low Probability of Failure 

A measure of 

seismic capacity of 

a structure, 

system, or 

component, 

expressed in terms 

of a threshold 

earthquake 

intensity, below 

which failure of the 

structure, system, 

or component is 

highly unlikely. 

(see Seismic 

Margin, Fragility)

In a seismic PRA, the high confidence in low probability of failure (HCLPF) measure is 

generally not used, but it is a key parameter primarily in a seismic margin analysis. 

The HCLPF capacity is a measure of the seismic capacity of a structure, system, or component 

(SSC) or of the whole plant. It indicates an earthquake intensity level at which there is high 

(95%) confidence the conditional probability of failure of the SSC is low (5% or less). At the 

plant level, HCLPF can refer to the peak ground acceleration level at which there is a high 

(95%) confidence of low (5%) conditional probability of core damage. It is used extensively in a 

seismic margin analysis. 

The ASME/ANS PRA Standard (Ref. 2) states that “HCLPF capacity: refers to the High 

Confidence of Low Probability of Failure capacity, which is a measure of seismic margin.” 

High-Level Requirements 

The minimum 

requirements for a 

technically 

acceptable 

baseline PRA, 

independent of 

application. (see 

Supporting 

Requirements)

For a base PRA, NRC Regulatory Guide (RG) 1.200 (Ref. 91) defines a set of technical 

characteristics and associated attributes that make it technically acceptable. One approach to 

demonstrate a PRA is acceptable is to use a national consensus PRA standard, supplemented 

to account for the NRC staff’s regulatory positions. The ASME/ANS PRA Standard (Ref. 2) is 

one example of a national consensus PRA standard. The ASME/ANS PRA Standard uses 

high-level requirements and supporting requirements. 

RG 1.200 states, “Technical requirements may be defined at two different levels: (1) high-level 

requirements and (2) supporting requirements. High-level requirements are defined for each 

technical element and capture the objective of the technical element. These high-level 

requirements are defined in general terms, need to be met regardless of the level of analysis 

resolution and specificity (capability category), and accommodate different approaches. 

Supporting requirements are defined for each high-level requirement. These supporting 

requirements are those minimal requirements needed to satisfy the high-level requirement.” 

The ASME/ANS PRA Standard (Ref. 2) states, “The high level requirements are defined in 

general terms and present the top level logic for the derivation of more detailed supporting 

requirements. The high level requirements reflect not only the diversity of approaches that 

have been used to develop the existing PRAs, but also the need to accommodate future 

technological innovations.” 

The definition provided was based on the definition in the introduction section of ASME/ANS 

PRA Standard (Ref. 2).

High-Pressure Melt Ejection 

A phenomenon in 

which molten core 

material 

penetrates the 

reactor vessel and 

is forcibly ejected 

under high 

pressure. (see 

Core Melt)

In a PRA, high-pressure melt ejection (HPME) is a phenomenon that could lead to containment 

failure and release of radioactive material to the environment before evacuation of the 

surrounding population. 

If the core melts and penetrates the reactor pressure vessel while the reactor coolant system is 

at high pressure (>400psi), the core debris would be ejected into the reactor cavity. This 

phenomenon is called HPME. 

A phenomenon often associated with HPME is direct containment heating (DCH). DCH can 

occur in the following manner: As the core debris is being ejected from the reactor vessel 

(depending on the configuration of the reactor cavity), it is possible that it will be transported 

into the containment atmosphere and directly heat the atmosphere. This heating can 

substantially increase the pressures in containment. It is also possible that combustible gases 

in the containment atmosphere could ignite and burn as a result of the transported core debris, 

adding to the containment heating and therefore the pressure in containment. 
4. GLOSSARY

 4-53 

TERM AND 

DEFINITION DISCUSSION 

High-Wind Fragility Analysis 

(see Fragility 

Analysis)

High-wind fragility analysis is a type of fragility analysis and is included in the discussion under 

“Fragility Analysis.”

High-Wind Hazard Analysis 

(see Hazard 

Analysis)

The term high-wind hazard analysis is a specific type of hazard analysis and is defined under 

“Hazard Analysis.”

High-Wind Plant Response Analysis/Model

(see Plant 

Response 

Analysis/Model)

The high-wind plant response analysis is a type of plant response analysis and is included in 

the discussion under “Plant Response Analysis/Model.” 

High Winds 

Winds of a certain 

size that could 

potentially damage 

or affect the 

operability of a 

nuclear power 

plant. (see 

Hazard) 

In a PRA, the typical high winds analyzed as a hazard include the following: tornadoes, 

hurricanes (or cyclones or typhoons as they are known outside of the United States), 

extratropical (thunderstorm) winds, and other wind phenomena depending on the site location 

(Ref. 2). High winds are a hazard group and, more specifically, a type of external hazard. 

Human Action (Operator Action) 

An action 

performed by plant 

personnel. (see 

Human Failure 

Event, Human 

Reliability 

Analysis) 

In a PRA, the human actions that are modeled include those actions that plant personnel might 

fail to perform or might fail to perform correctly. Plant personnel interact with the plant in a 

number of ways. For example, maintenance personnel perform surveillance tests, calibrate 

equipment, and repair failed equipment. Control room operators control the plant and, after an 

initiating event, bring the plant to a safe stable state using as guidance written or memorized 

procedures. These actions are of concern for the PRA because failure to perform any of the 

actions correctly can lead to a reduced capability of responding to a transient or accident. For 

example, failure to restore a system following maintenance can lead to its unavailability to 

perform its function when called upon. Failure of the control room crew to correctly follow their 

procedures might lead to a loss of a critical safety function. 

A human action and an operator action do not necessarily mean the same thing. A human 

action can be performed by different types of nuclear power plant personnel, while an operator 

action is an action performed by a licensed individual in the control room. 

Human actions are an important component in conducting a human reliability analysis (HRA). 

HRA is used to support the development of a PRA by identifying relevant human actions and 

the associated human errors that might occur. Human errors modeled in the PRA are referred 

to as human failure events. 

Human Error (Operator Error) 

Any human action, 

including inaction, 

which exceeds 

some limit of 

acceptability, 

excluding 

In a PRA, human (operator) errors are modeled in the PRA as human failure events if they are 

unrecovered and lead to the failure or unavailability of a component, system, or function. 

Human errors of interest are those that result in the unavailability of a component, system, or 

function, or a failure to initiate, terminate, or control a system or function that can affect an 

accident sequence. 
4. GLOSSARY

 4-54 

TERM AND 

DEFINITION DISCUSSION 

malevolent 

behavior. (see 

Human Failure 

Event, Human 

Reliability 

Analysis)

A human error and an operator error do not necessarily mean the same thing. A human error 

can be attributed to different types of nuclear power plant personnel, while an operator error is 

specifically attributed to a licensed individual (i.e., operator) in the control room. 

Human reliability analysis (HRA) is used to identify the possible human errors that might occur. 

The term human failure event is synonymous with and has replaced the term human error in 

the PRA lexicon. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 2). 

Human Error Event 

(see Human 

Failure Event)

A human error event is a type of human error modeled in a PRA and is defined under “Human 

Failure Event.”

Human Error Factor 

(see Error 

Factor)

A human error factor is a specific type of error factor applicable to human reliability analysis 

and is defined under “Error Factor.”

Human Error Probability 

(see Probability) A human error probability is a specific type of probability applicable to human reliability analysis 

and is defined under “Probability.”

Human Failure Event, Human Error Event 

A basic event that 

represents a 

failure or 

unavailability of a 

component, 

system, or function 

that is caused by 

human inaction, or 

inappropriate 

action. (see 

Human Action, 

Human Error) 

In a PRA, potential human errors (i.e., human actions or inappropriate human actions) are 

modeled as basic events. The term human failure event is synonymous with and has replaced 

the term human error in the PRA lexicon. 

Human failure events can be classified as either errors of omission or errors of commission. 

An error of omission would be failure to perform a system-required task or action. An error of 

commission would be incorrectly performing a system-required task or action, or performing an 

extraneous task that is not required and could contribute to component, system, or function 

failure or unavailability. In the PRA, failures to restore a function, referred to as recovery, are 

also modeled as human failure events. 

The terms human failure event and human error event have the same meaning in a PRA 

context and it is correct and appropriate to use them interchangeably. 

 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 2). 

Human Reliability Analysis 

A structured 

approach used to 

identify potential 

human failure 

events and to 

systematically 

estimate the 

probability of those 

events using data, 

models, or expert 

judgment. (see 

Human Action, 

Human Error) 

In a PRA, a human reliability analysis is used to identify relevant human actions and possible 

human errors that might occur. Human actions considered in the human reliability analysis 

include those actions that plant personnel might fail to perform or might fail to perform correctly. 

Failure to correctly perform certain human actions can lead to a reduced capability of 

responding to a transient or accident, including the loss of one or more critical safety functions. 

The failure to correctly perform a human action is referred to as a human error. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 2). 
4. GLOSSARY

 4-55 

TERM AND 

DEFINITION DISCUSSION 

Importance Measure (Risk Reduction Worth, Risk Achievement Worth, Fussell-Vesely, 

Birnbaum Importance, Uncertainty Importance) 

A metric that 

provides either the 

absolute or relative 

contribution of a 

component, 

system, structure, 

or human action to 

the defined risk.

In a PRA, importance measures are used to determine the contribution of the basic events to a 

number of risk metrics, such as core damage frequency. By using importance measures, the 

PRA analyst can determine the risk-significance of structures, systems, and components 

(SSCs) or human actions. Different importance measures provide different perspectives. For 

example, importance measures can evaluate the risk-reduction potential of improving SSC 

performance or human action, or they can show the significance of an SSC or human failure 

event for maintaining the current risk level. There are five importance measures typically used 

in a PRA: 

• Risk Reduction Worth : As defined in NUREG/CR-3385 (Ref.71), risk reduction worth 

is: “The decrease in risk if a plant feature (e.g., system or component) were assumed 

to be optimized or were assumed to be made perfectly reliable. Depending on how 

the decrease in risk is measured, the risk reduction worth can either be defined as a 

ratio or an interval.” 

• Risk Achievement Worth: The increase in risk if a plant feature (e.g., system or 

component) was assumed to be failed or was assumed to be always unavailable. 

Depending on how the increase in risk is measured, the risk achievement worth can 

either be defined as a ratio or an interval. Sometimes risk achievement worth is 

referred to as “risk increase.”

• Fussell-Vesely: For a specified basic event, Fussell-Vesely importance is the relative 

contribution of a basic event to the calculated risk. This relative or fractional 

contribution is obtained by determining the reduction of the risk if the probability of the 

basic event to zero. 

• Birnbaum Importance (Bi): NUREG-1489 (Ref.54) defines birnbaum importance as: 

“An indication of the sensitivity of the accident sequence frequency to a particular 

basic event.” Bi measures the change in total risk as a result of changes to the 

probability of an individual basic event. 

• Uncertainty Importance: The uncertainty in each input parameter, as expressed 

through its probability distribution, contributes to the uncertainty in the output 

parameter of interest (e.g., core damage frequency). The uncertainty importance 

measure attempts to quantify the contribution of each individual basic event’s 

uncertainty to this total output uncertainty. The uncertainty importance is the 

Birnbaum importance multiplied by the standard deviation of the input probability 

distribution (Ref.83).

Important to Safety 

(see Safety 

Significant) 

The term important to safety has a safety connotation and is defined under “Safety Significant.” 

Incremental Conditional Probability (Core Damage, Large Early Release) 

A measure of the 

impact of a 

temporary plant 

modification on the 

probability of an 

undesired end 

state. (see 

Conditional 

Probability, 

As applied to PRA and plant risk evaluations, the term incremental conditional probability refers 

to the change in the probability of an undesired plant end state attributable to (conditional on) a 

temporary modification in plant configuration or operations, over the time that the modification 

is in place. Usually, this incremental change in conditional probability is reflected as an 

increase in the probability of an undesired end state such as core damage when compared to 

the baseline core damage probability. Because the probability of core damage depends on the 

temporary modification or change at the plant, it is therefore a conditional probability. 

Incremental conditional probability also is calculated in a PRA for large early release. 
4. GLOSSARY

 4-56 

TERM AND 

DEFINITION DISCUSSION 

Instantaneous 

Conditional 

Probability). 

Incremental conditional probability differs from instantaneous conditional probability in that 

instantaneous conditional probability represents the probability that an undesired plant end 

state is reached given an initiating event and the actual (instantaneous) plant configuration. 

The incremental conditional probability is integrated over the duration of the temporary 

condition, while the instantaneous conditional probability represents a point-in-time measure. 

Induced Steam Generator Tube Rupture 

(see 

Consequential 

Steam 

Generator Tube 

Rupture) 

The term induced steam generator tube rupture is similar in definition to, and is grouped with, 

the term “Consequential Steam Generator Tube Rupture.” 

Ingestion

Exposure from 

intake of food and 

water contaminated 

with radioactive 

material. (see 

Exposure 

Pathways, 

Exposure, 

Exposure Time, 

Cloudshine, Water 

Immersion, 

Groundshine, 

Inhalation, Skin 

Deposition, Health 

Effects)

In a Level 3 PRA, for the consequence calculation ingestion is one of the assumed pathways 

by which an individual can receive doses. The pathways of exposure include: (1) direct 

external exposure from radioactive material in a plume, principally due to gamma radiation (air 

immersion or cloudshine), (2) direct exposure from radioactive material in contaminated water 

given to an individual immersed in the water, (3) exposure from inhalation of radioactive 

materials in the cloud and resuspended material deposited on the ground, (4) exposure to 

radioactive material deposited on the ground (groundshine), (5) radioactive material deposited 

onto the body surfaces (skin deposition), and (6) ingestion from deposited radioactive materials 

that make their way into the food and water pathway. 

Inhalation 

Exposure from 

breathing 

radioactive 

material. (see 

Exposure 

Pathways, 

Cloudshine, Water 

Immersion, 

Groundshine, 

Ingestion, Skin 

Deposition)

In a Level 3 PRA, for the consequence calculation inhalation is one of the assumed pathways 

by which an individual can receive doses. The pathways of exposure include: (1) direct 

external exposure from radioactive material in a plume, principally due to gamma radiation (air 

immersion or cloudshine), (2) direct exposure from radioactive material in contaminated water 

given to an individual immersed in the water, (3) exposure from inhalation of radioactive 

materials in the cloud and resuspended material deposited on the ground, (4) exposure to 

radioactive material deposited on the ground (groundshine), (5) radioactive material deposited 

onto the body surfaces (skin deposition), and (6) ingestion from deposited radioactive materials 

that make their way into the food and water pathway. 

Initiating Event, Initiator 

An event that 

perturbs the 

steady-state 

operation of the 

plant and could 

lead to an 

undesired plant 

condition. 

In a PRA, an initiating event is an event originating from an internal or external hazard that both 

challenges normal plant operation and requires successful mitigation. As such, these events 

represent the beginning of accident sequences modeled in the PRA. Having a reasonably 

complete set of initiating events is crucial in determining what events could propagate to core 

damage. 

Initiating events can arise from the following: 

• Internal Hazards, which include:

⎯ Internal event (see Internal Event) 
4. GLOSSARY

 4-57 

TERM AND 

DEFINITION DISCUSSION 

⎯ Floods (see Internal Flood) 

⎯ Fires (see Appendix A for fire terms) 

• External Hazards, which include:

⎯ Floods (see External Flood) 

⎯ High winds (see High Winds) 

⎯ Seismic events (see Hazard Analysis) 

⎯ Other external hazards 

These hazards result in different types of initiating events. Examples of initiating events are 

transients (see Transient) and loss-of-coolant accidents (see Loss-of-Coolant Accident). 

The terms initiating event and initiator are both used in a PRA context and generally have the 

same meaning. In some cases, the term initiator may refer to a class of initiators 

(e.g., transient), while the term initiating event may refer to the actual event (e.g., loss of 

a feedwater pump resulting in a transient). 

The ASME/ANS PRA Standard (Ref. 2) defines an initiating event as “an event either internal 

or external to that which perturbs the steady state operation of the plant by challenging plant 

control and safety systems whose failure could potentially lead to core damage or release of 

airborne fission products. These events include human-caused perturbations and failure of 

equipment from either internal plant causes (such as hardware faults, floods, or fires) or 

external plant causes (such as earthquakes or high winds).” 

Initiating Event Analysis 

The process used 

to identify events 

that perturb the 

steady- state 

operation of the 

plant and could 

lead to an 

undesired plant 

condition. (see 

Initiating Event, 

Master Logic 

Diagram)

In a PRA, the initiating event analysis considers how accidents can start by identifying and 

quantifying those events that challenge plant operation and require successful mitigation to 

prevent core damage from occurring. To facilitate the efficient modeling of potential accidents, 

initiating events typically are identified using a systematic process (e.g., master logic diagram) 

and grouped according to their mitigation requirements. The frequencies of these initiating 

event groups are then quantified. 

 

NRC Regulatory Guide 1.200 (Ref. 91) states that initiating event analysis “identifies and 

characterizes the events that both challenge normal plant operation during power or shutdown 

conditions and require successful mitigation by plant equipment and personnel to prevent core 

damage from occurring. Events that have occurred at the plant and those that have a 

reasonable probability of occurring are identified and characterized. An understanding of the 

nature of the events is performed such that a grouping of the events, with the groups defined 

by similarity of system and plant responses (based on the success criteria), may be performed 

to manage the large number of potential events that can challenge the plant.” 

Initiating Event Frequency 

 (see Frequency) The term initiating event frequency is a type of frequency that is defined under “Frequency.”

Initiator 

(see Initiating 

Event)

The term initiator is similar in meaning to initiating event and is defined under “Initiating Event.”

Instantaneous Conditional Probability (Core Damage, Large Early Release) 

Event probability at 

the specific time 

the plant is 

analyzed, given 

that a prior event 

Using a PRA, instantaneous conditional probability can be calculated for core damage and 

large early release. The probability of either of those undesired outcomes occurring depends 

on the occurrence of an initiating event while the plant is in a given configuration. Thus, core 

damage or large early release is “conditional” on the probability of a prior event occurring. 
4. GLOSSARY

 4-58 

TERM AND 

DEFINITION DISCUSSION 

has occurred. 

(see Conditional 

Probability, 

Incremental 

Conditional 

Probability)

The following are other definitions that could describe instantaneous conditional probability: 

• The probability that an undesired plant end state is reached given an initiating event 

and the actual (instantaneous) plant configuration. 

• The average probability that an undesired plant end state is reached, weighted over 

all credible initiating events, for the actual (instantaneous) plant configuration. 

Instantaneous conditional probability differs from incremental conditional probability in that 

incremental conditional probability represents the impact of a temporary plant modification on 

the probability of an undesired end state. The incremental conditional probability is integrated 

over the duration of the temporary condition, while the instantaneous conditional probability 

represents a point-in-time measure. 

Interfacing-Systems Loss-of-Coolant Accident 

A loss-of-coolant 

accident 

characterized by 

high-pressure 

reactor coolant 

being released into 

a low-pressure 

system. (see 

Loss-of-Coolant 

Accident)

In a PRA, accidents involving an interfacing-systems loss-of-coolant accident (ISLOCA) are 

modeled because they represent a loss of isolation between an ancillary system and the 

reactor coolant system, which contains radioactive material. This type of accident is important 

in the PRA because it may lead to radioactive material bypassing containment and loss of 

reactor coolant inventory. 

 

The ASME/ANS PRA Standard (Ref. 2) defines ISLOCA as “a loss of coolant accident (LOCA) 

when a breach occurs in a system that interfaces with the reactor coolant system, where 

isolation between the breached system and the reactor coolant system fails. An ISLOCA is 

usually characterized by the over-pressurization of a low-pressure system when subjected to 

reactor coolant system pressure and can result in containment bypass.” 

ISLOCAs of most concern are those accidents during which the break flow is discharged 

outside the reactor containment building. The two main reasons for this concern are: 

(1) potential high offsite radiological consequences caused by radioactive material bypassing 

the containment and (2) potential loss of long-term core cooling resulting from loss of reactor 

coolant system inventory that would otherwise be available for recirculation from the 

containment sumps. 

Internal Event 

Failure of 

equipment as a 

result of either an 

internal random 

cause or a human 

event which 

perturbs the 

steady-state 

operation of the 

plant and could 

lead to an 

undesired plant 

condition. (see 

Hazard) 

In a PRA, internal events result from or involve random mechanical, electrical, structural, or 

human failures within the plant boundary and are a specific hazard group. An example of an 

internal event modeled in a PRA would be the random structural failure of a reactor coolant 

system pipe resulting in a loss-of-coolant accident (LOCA) initiating event. Until the 2009 

ASME/ANS PRA Standard revision (Ref. 2), this term did not have a consistent definition. In 

some cases, a fire or flood or both occurring within the plant were considered an internal event. 

The ASME/ANS PRA Standard has been revised and internal flood and internal fire are not 

considered internal events. 

The ASME/ANS PRA Standard (Ref.2) defines an internal event as “an event resulting from or 

involving random mechanical, electrical, structural, or human failures from causes originating 

within a nuclear power plant that directly or indirectly causes an initiating event and may cause 

safety system failures or operator errors that may lead to core damage or large early release. 

By historical convention, loss of offsite power is considered to be an internal event, and internal 

fire is considered to be an external event, except when the loss is caused by an external 

hazard that is treated separately (e.g., seismic-induced loss of offsite power). Internal floods 

sometimes have been included with internal events and sometimes considered as external 

events. For this standard, internal floods are considered to be internal hazards separate from 

internal events.” 
4. GLOSSARY

 4-59 

TERM AND 

DEFINITION DISCUSSION 

Internal Fire 

A fire initiated 

within the plant 

that can affect the 

operability of the 

plant. (see Hazard 

and Appendix A)

In a PRA, internal fires are a specific hazard group in which the fire occurs within the plant 

boundary. The PRA considers fires because they have the potential to cause equipment failure 

by direct flame impact or high thermal radiation. 

Internal Flood, Internal Flooding Event 

A flood initiated 

within the plant 

that can affect the 

operability of the 

plant. (see 

Hazard, External 

Flood) 

In a PRA, internal floods are a specific hazard group in which the flood occurs within the plant 

boundary. The PRA considers floods because they have the potential to cause equipment 

failure by the intrusion of water into plant equipment through submergence, spray, dripping, or 

splashing. 

The term internal flooding event represents the occurrence of an internal flood.

Internal Flooding Event 

(see Internal 

Flood)

The term internal flooding event is the occurrence of an internal flood and is defined under 

“Internal Flood.” 

Internal Hazard 

(see Hazard) The term internal hazard is a specific type of hazard and is defined under “Hazard.” 

Key Assumption 

(see 

Assumption)

The term key assumption is a specific type of assumption and is defined under “Assumption.” 

Key Model Uncertainty

 (see 

Uncertainty)

The term key model uncertainty is a type of uncertainty and is defined under “Uncertainty.” 

Key Source of Model Uncertainty

 (see 

Uncertainty)

The term key source of model uncertainty is defined under “Uncertainty.” 

Key Source of Uncertainty

 (see 

Uncertainty)

The term key source of uncertainty is defined under “Uncertainty.” 

Land Contamination 

Contamination of 

land outside of the 

nuclear power 

plant site boundary 

with radioactive 

material released 

in an accident. 

(see Health 

Effects) 

In a Level 3 PRA, land contamination often is evaluated along with health effects. 

Land contamination refers to the radioactive material deposited on the ground by gravitational 

settling or the impact during plume passage. Land contamination depends on the 

characteristics of the radioactivity release and how the land surrounding the plant is used. 

Land contamination risk involves the frequency and amount of land contamination and its 

associated cost. 
4. GLOSSARY

 4-60 

TERM AND 

DEFINITION DISCUSSION 

Land Contamination Risk 

(see Land 

Contamination) 

Land contamination risk is sometimes calculated in a Level 3 PRA and is defined in the 

discussion under “Land Contamination.” 

Large Early Release 

(see Radioactive 

Material 

Release) 

The term large early release is a type of radioactive material release and is defined in the 

discussion under “Radioactive Material Release.” 

Large Early Release Frequency 

(see Frequency) The term large early release frequency is a type of frequency used in PRA calculation and is 

defined in the discussion under “Frequency.” 

Large Early Release Frequency Analysis 

(see Radioactive 

Material Release 

Frequency 

Analysis)

The term large early release frequency analysis is a type of radioactive material release 

frequency analysis and is defined under “Radioactive Material Release Frequency Analysis.” 

Large Late Release 

(see Radioactive 

Material 

Release) 

The term large late release is a type of radioactive material release and is defined in the 

discussion under “Radioactive Material Release.” 

Large Late Release Frequency 

 (see Frequency) The term large late release frequency is a type of frequency used in PRA calculation and is 

defined in the discussion under “Frequency.”

Large Late Release Frequency Analysis 

(see Radioactive 

Material Release 

Frequency 

Analysis) 

The term large late release frequency analysis is a type of radioactive material release 

frequency analysis and is defined under “Radioactive Material Release Frequency Analysis.” 

Large Release 

Formal definition 

requires 

Commission 

approval. (see 

Radioactive 

Material Release)

The notion of a large release implies that in the range of possible releases there exists a 

threshold value that distinguishes large releases from not large releases. Many PRAs include 

their own specific definitions of a large release, but no universally accepted definition has been 

established. Attempts have been made to define a large release magnitude based on offsite 

health effects. There is an inherent arbitrariness in definitions since offsite health effects 

depend not only on release magnitude but also on site-specific parameters, such as population. 

Therefore, what would be a large release at one site would not necessarily be one at another 

site. Weather and wind variability are other site-specific factors. 

In the past, the NRC staff has considered several alternate definitions of a large release. These 

include: 

• A release that would result in one or more early fatalities;
4. GLOSSARY

 4-61 

TERM AND 

DEFINITION DISCUSSION 

• A release that has the potential to result in one early offsite fatality within 1 mile of the 

plant boundary;

• A definition of a large release source term in the traditional form of a fractional 

release of the core inventory of various radionuclide groups to the environment, the 

timing of the release, etc.

• Any release from an event that involves severe core damage, primary system 

pressure boundary failure, and early containment failure.

The Commission has not approved a formal definition for the term large release. 

Large Release Frequency 

(see Frequency) The term large release frequency is a type of frequency used in PRA calculation and is defined 

in the discussion under “Frequency.”

Late Containment Failure 

(see 

Containment 

Failure)

The term late containment failure is a type of containment failure and is defined under 

“Containment Failure.”

Latent Cancer Fatality

(see Fatality) The term latent cancer fatality is a type of fatality caused by exposure to radioactive materials 

and is defined under “Fatality.”

Latent Fatality 

 (see Fatality) The term latent fatality is a type of fatality caused by exposure to radioactive materials and is 

defined under “Fatality.”

Latent Fatality Risk 

 (see Fatality) The term latent fatality risk is a type of risk-involved fatality caused by exposure to radioactive 

materials and is defined under “Fatality.”

Latent Health Effects 

(see Health 

Effects) 

The term latent health effect refers to a type of health effect and is defined in the discussion 

under “Health Effects.” 

Level 1, 2, 3 PRA 

A characterization 

of the scope of a 

PRA in terms of 

increasing 

specification of 

consequences. 

(see PRA)

The three types of PRA are distinguished by the risk metric calculated, and when all three are 

calculated for a particular plant, it is referred to as a full-scope PRA. Level 1 refers to core 

damage frequency as the risk measure, Level 2 refers to radioactivity releases as the risk 

measure, and Level 3 refers to offsite consequences as the risk measure. 

A Level 2 PRA takes the results of the Level 1 PRA (accident sequences resulting in core 

damage) as input and produces frequencies of radioactivity releases as output. A Level 3 PRA 

takes the results of the Level 2 PRA as input and produces offsite consequences (health 

effects, economic consequences) as output. In some usages, a Level 2 PRA includes the 

Level 1 analysis, and the Level 3 PRA includes both the Level 1 and Level 2 analyses. The 

figure below illustrates the different PRA “Levels” and what each calculates. 
4. GLOSSARY

 4-62 

TERM AND 

DEFINITION DISCUSSION 

Level of Detail 

The degree of 

resolution or 

specificity in the 

analyses 

performed in the 

PRA. (see Model, 

Capability 

Categories)

In a PRA, the level of detail generally refers to the level to which a system is modeled 

(e.g., function level, train level, component level), the extent to which systems are included in 

the success criteria (e.g., safety systems and nonsafety systems), the extent to which 

phenomena are included in the challenges to the plant in the Level 2 analysis, and the extent to 

which operator actions are considered (e.g., accident management strategies).

Level of detail generally is dictated by four factors: (1) the level of detail to which information is 

available, (2) the level of detail required so that dependencies are included, (3) the level of 

detail so that the risk contributors are included, and (4) the level of detail sufficient to support 

the application. 

In the ASME/ANS PRA Standard (Ref. 2), the degree to which the level of detail (and scope) of 

the plant design, operation, and maintenance are modeled forms one of the bases for the 

capability categories defined in the Standard.

Licensing Basis

The collection of 

documents or 

technical criteria 

that provides the 

basis upon which 

the NRC issues a 

license to 

construct or 

operate a nuclear 

facility. 

A PRA is part of the licensing basis for plants licensed under 10 CFR Part 52, “Licenses, 

Certifications, and Approvals for Nuclear Power Plants.” A PRA also is used to support 

changes to the licensing basis carried out using regulatory guidance documents such as 

Regulatory Guide (RG) 1.174 (Ref. 84), RG 1.175 (Ref. 85), or RG 1.177 (Ref. 86). 

The NRC Web site Glossary (Ref. 36) defines licensing basis as “the collection of documents 

or technical criteria that provides the basis upon which the NRC issues a license to construct or 

operate a nuclear facility; to conduct operations involving the emission of radiation; or to 

receive, possess, use, transfer, or dispose of source material, byproduct material, or special 

nuclear material.” 

10 CFR Part 54 (Ref. 27) defines current licensing basis (CLB) as “the set of NRC 

requirements applicable to a specific plant and a licensee’s written commitments for ensuring 

compliance with and operation within applicable NRC requirements and the plant-specific 

design basis (including all modifications and additions to such commitments over the life of the 

license) that are docketed and in effect.” The CLB includes NRC regulations, orders, license 

conditions, exemptions, technical specifications, final safety analysis reports, and licensee 

commitments to NRC bulletins, generic letters, enforcement actions, and licensee event 

reports. 

The definition provided was based on the definition in the NRC Web site Glossary (Ref. 36). 

Licensing-Basis Event 

A postulated 

accident that a 

nuclear facility 

must be designed 

and built to 

withstand.

The term licensing-basis event (LBE) is not used in current PRAs or the current nuclear power 

plant regulatory licensing structure. It is a term being used for a potentially new regulatory 

process. Further information on this regulatory framework can be found in NUREG-1860 (Ref. 

63). 

This potential future licensing structure is a process that uses both deterministic and 

probabilistic criteria for selecting the postulated accidents, called LBEs, which a nuclear facility 

Level 1 Level 2 Level 3

Computation of

core damage 

frequency

Computation of

radioactive 

material release 

frequency

Analysis of

early and latent 

fatality 
4. GLOSSARY

 4-63 

TERM AND 

DEFINITION DISCUSSION 

must demonstrate it can withstand (i.e., the facility design and operation must be able to 

withstand the impact of LBEs without loss to the structures, systems, and components (SSCs) 

needed to ensure public health and safety). 

Linear No-Threshold Model

A dose response 

model that 

assumes cancer 

risk is proportional 

to the dose 

received no matter 

how small the 

dose. (see Dose, 

Dose Response 

Model) 

In a Level 3 PRA, a dose response model is used to calculate the cancer risk for given levels of 

a dose to individuals after a severe accident. 

There is some debate on the appropriate dose response relationship for cancer risk following 

exposure to ionizing radiation. A linear relationship in which the cancer risk increases in direct 

proportion to the dose is one view. Another view advocates a nonlinear relationship, in which 

cancer risk increases in a more complex manner relative to dose. There is also a question 

about whether a minimum dose exists, below which no increased risk of cancer is found 

(threshold model), or whether any dose, no matter how small, increases cancer risk (nothreshold model).

Living PRA 

A probabilistic risk 

assessment that is 

maintained so that 

it reflects the 

current plant 

design and 

operational 

features. (see 

Dynamic PRA, 

PRA Configuration 

Control, As-Built 

As-Operated) 

The term living PRA designates a PRA that is updated as necessary to reflect any changes in 

the plant (e.g., design, operating procedures, data) to continue to represent the as-built 

as-operated plant. Therefore, the living PRA can be used in risk-informed decisionmaking 

processes, such as plant-specific changes to the licensing basis discussed in NRC Regulatory 

Guide 1.174 (Ref. 84). PRA configuration control is part of the process used to support a living 

PRA. 

A living PRA is not the same as a dynamic PRA. A dynamic PRA refers to a PRA that 

accounts for time-dependent effects by integrating these effects directly into the computer 

model. 

Loss-of-Coolant Accident (Small, Medium, Large) 

An accident that 

results in a loss of 

coolant from the 

reactor. (see 

InterfacingSystems Loss-ofCoolant Accident)

In a PRA, two major categories of initiating events are evaluated; namely, transients and 

loss-of-coolant accidents (LOCAs). LOCAs represent a particularly challenging accident 

because reactor coolant, usually water, cannot be replaced at a sufficient rate to prevent 

uncovering the reactor core leading to core damage and potential fueling melting. Once 

considered to be the most severe design-basis accident, PRAs have revealed that other 

accident initiators, such as long-term station blackout, are far more frequent and can lead to 

equally undesired consequences. 

LOCA initiating event frequencies used in the PRA are dependent on the size of LOCA. These 

sizes are typically referred to as small, medium, or large LOCAs. The break sizes which define 

small, medium, and large LOCAs are also dependent on the type of reactor, either PWR or 

BWR, and whether the lost coolant is liquid or steam. NUREG/CR-6928 (Ref. 82) provides the 

following description for BWRs: 

• Small LOCA (SLOCA): A break size less than 0.004 square feet (1-inch inside 

diameter pipe equivalent) for liquid and less than 0.05 square feet (approximately 4-

inch inside diameter pipe equivalent) for steam in a primary system pipe with leakage 

rate greater than100 gallons per minute.

• Medium LOCA (MLOCA): A break size between 0.004 to 0.1 square feet 

(approximately 1- to 5-inch inside diameter pipe equivalent) for liquid and between 

0.05 to 0.1 square feet (approximately 4- to 5-inch inside diameter pipe equivalent) 

for steam in a primary system pipe. 
4. GLOSSARY

 4-64 

TERM AND 

DEFINITION DISCUSSION 

• Large LOCA (LLOCA): A break size greater than 0.1 square feet (approximately 5-

inch inside diameter pipe equivalent) for liquid or steam in a primary system pipe. 

NUREG/CR-6928 (Ref. 82) also provides the following description for PWRs:

• Small LOCA (SLOCA): A pipe break in the primary system boundary with an inside 

diameter between 0.5- and 2-inches

• Medium LOCA (MLOCA): A pipe break in the primary system boundary with an 

inside diameter between 2- and 6- inches. 

• Large LOCA (LLOCA): A pipe break in the primary system boundary with an 

equivalent inside diameter greater than 6-inches. 

Historically, NUREG-1150 (Ref. 51) defines SLOCA as < 1 inch, MLOCA as 1 to 5 inches, and 

LLOCA as > 5 inches for BWRs and SLOCA as 0.5 to 2 inches, MLOCA as 2 to 6 inches, and 

LLOCA as > 6 inches for PWRS. 

 

Appendix A to 10 CFR Part 50 (Ref. 22) and the NRC Web site Glossary (Ref. 34) define the 

term LOCAs as “those postulated accidents that result in a loss of reactor coolant at a rate in 

excess of the capability of the reactor makeup system from breaks in the reactor coolant 

pressure boundary, up to and including a break equivalent in size to the double-ended rupture 

of the largest pipe of the reactor coolant system.” 

Loss of Offsite Power 

The loss of all AC 

power from the 

electrical grid to 

the plant. (see 

Transient)

In a PRA, loss of offsite power (LOOP) is referred to as both an initiating event and an accident 

sequence class. As an initiating event, LOOP to the plant can be a result of a weather-related 

fault, a grid-centered fault, or a plant-centered fault. During an accident sequence, LOOP can 

be a random failure. Generally, LOOP is considered to be a transient initiating event. 

NUREG/CR-6890 (Ref.80) defines a LOOP as “the simultaneous loss of electrical power to all 

plant safety buses, requiring all emergency power generators to start and supply power to the 

safety buses.” 

Low-Power and Shutdown 

The states of 

nuclear power 

plant operation 

when the reactor is 

producing power in 

a range below a 

specified level or is 

shutdown. (see 

Full Power, 

At-Power)

A PRA models the different plant operating states (POSs) of the plant. Operation at-power is 

one POS, while several POSs are needed to characterize the plant during the various stages of 

low-power and shutdown. These POSs are distinguished in the PRA model because the plant 

response (e.g., accident sequences) differs during different POSs. 

Low power and shutdown is the term applicable for other than at-power conditions (i.e., the 

reactor is typically producing less than 15-25% of its rated power). Low-power and shutdown 

analysis is further separated into consideration of low power and shutdown states. 

In a low-power initial condition, the core is producing power from fissioning of fuel, over and 

above the decay heat levels, although at lower amounts than at-power. Most safety systems 

are on automatic actuation but some may be disabled or blocked (e.g., main feedwater trip in 

boiling-water reactors). The support systems are aligned in their normal configuration 

(e.g., electrical power is being drawn from the grid). In these POSs, the power level may be 

changing as the reactor is shutting down or starting up, or the power level may be constant at a 

reduced level. The power level that distinguishes nominal full power from low power is the 

power level below which major plant evolutions are required to reduce or increase power that 

significantly increase the likelihood of a plant trip (e.g., taking manual control of feedwater 

level). 
4. GLOSSARY

 4-65 

TERM AND 

DEFINITION DISCUSSION 

In shutdown conditions, the core is not producing power (i.e., the reactor is subcritical). The 

reactor temperature and pressure are lower than at-power, coolant inventory may be lower or 

higher, the reactor may be relying on alternate cooling systems, some safety systems may be 

defeated, or containment may be open. 

A representation of the different plant operating states (i.e., low power and shutdown) is shown 

under the discussion for the term At-Power.

Master Logic Diagram 

A graphical model 

that can be 

constructed to 

guide the selection 

of initiating events. 

(see Fault Tree)

In a PRA, a master logic diagram (MLD) is often used to identify the specific events that are 

potential initiating events and to group them according to the challenges they pose to plant 

safety. An MLD is developed using fault tree logic to show general categories of initiating 

events proceeding to increasingly detailed information at lower levels, with specific initiating 

events presented at the bottom level. In a more general sense, an MLD is a fault tree 

identifying all the hazards that affect a system or mission. 

An MLD generally uses a fault tree logic approach to identify the logic or relationship between 

events. However, the difference between an MLD and a fault tree is that a fault tree focuses on 

accounting for the specific causes leading to failure of a system or group of systems, whereas 

the MLD focuses on listing the hazards that can affect a top event. An example of an MLD is 

provided below. 

The ASME/ANS PRA Standard (Ref. 2) defines an MLD as a “summary fault tree constructed 

to guide the identification and grouping of initiating events and their associated sequences to 

ensure completeness.” 

Mean 

The expected 

value of a random 

variable. (see

Median, Best 

Estimate, Point 

Estimate, 

Probability 

Distribution)

In a PRA, the metrics (e.g., core damage frequency, large early release frequency) generally 

are evaluated and presented as mean values to reflect the uncertainties in the parameter 

values used as input to the evaluation of the metrics. The mean values and the distributions 

from which they are calculated can be used to address the parameter uncertainties. 

The mean is the average value from a probability distribution. It is the expected value one 

would get from many samples taken of the random variable. The random variable in question 

could be a risk parameter, such as a component failure probability, or a risk measure, such as 

core damage frequency. 

The mean and median provide different information and cannot be used interchangeably. An 

illustration of the difference between mean and median is shown below. 

Initiating

Event

Insufficient

Reactivity 

Control

Insufficient 

Core Heat 

Removal

Pipe 

Rupture

Safety/Relief

Valve 

Opens

Transients LOCAs
4. GLOSSARY

 4-66 

TERM AND 

DEFINITION DISCUSSION 

Mechanistic Source Term 

A source term that 

is calculated 

considering the 

characteristics of 

specific accidents. 

(see Source Term)

In a Level 2 PRA, the source term calculated is usually a mechanistic source term. A 

mechanistic source term is calculated using validated models and supporting scientific data 

that simulate the physical and chemical processes that describe the radioactive material 

inventories and the time-dependent radioactive material transport mechanisms necessary and 

sufficient to predict the source term. 

For licensing calculations not involving a PRA, current light-water reactors (LWRs) use a 

deterministic predetermined source term into containment for different accidents, instead of a 

mechanistic source term, to analyze the effectiveness of the containment and site suitability for 

licensing purposes. 

Median 

That value of a 

random variable 

for which the 

occurrence of 

larger values is 

just as likely as 

occurrence of 

smaller values. 

(see Mean, 

Probability 

Distribution)

In a PRA, median values are not usually calculated. In some cases, median values of the risk 

metric are calculated in addition to the mean to provide a perspective on the distribution of the 

risk metric. Conclusions can be made about the spread and shape of a probability distribution 

of a risk metric or a parameter by comparing the median to the mean and to the other 

quantiles. 

The median is the middle value in a probability distribution. It is a reference point in which half 

the data values in a probability distribution (e.g., uncertainty distribution) lie below it and half lie 

above it. For example, if the median of a failure rate of a particular type of electric motor is 

2x10-4/hr then half of all electric motors of that type would have failure rates below 2x10-4/hr 

and half would have failure rates above 2x10-4/hr. 

An illustration of the difference between mean and median is under the discussion of the term 

“Mean.” 

Minimal Cutset 

(see Cutset) The term minimal cutset is a type of cutset used in PRA and is defined under “Cutset.” 

Mission Time 

The time period 

that a system or 

component is 

required to operate 

In a PRA, the failure probability of a component to operate is directly related to its mission time. 

By convention, in a Level 1 internal events PRA, mission time usually is specified as 24 hours. 

After that initial time period, multiple options for dealing with the accident would become 

available so that the residual risk results, beyond the 24-hour timeframe, would be negligibly 

6’ 6.5’ 7’ 7.5’ 8’ 4’ 4’ 7’ 7’ 7’

Team B

The median height of this team is 7’, 

but the mean height of this team is 

only 5’8” (173cm).

Team A

The mean height and the median height 

of this team are both 7’ (213cm).
4. GLOSSARY

 4-67 

TERM AND 

DEFINITION DISCUSSION 

to successfully 

perform its 

function.

small. For Level 1 PRAs that examine external hazards, the mission times usually are longer 

(e.g., 72 hours) because of area wide effects of such events. 

The definition provided is based on the definition in the ASME/ANS PRA Standard (Ref. 2). 

Mitigating System 

A plant system 

designed to 

minimize the 

effects of initiating 

events. (see 

Initiating Event, 

Front-Line System, 

Support System)

In a PRA, the accident mitigating functions and mitigating systems modeled are based on the 

initiating event(s) being analyzed. Mitigating systems can prevent an accident or reduce the 

consequences of a potential accident by directly performing or supporting one or more accident 

mitigating functions (e.g., core or containment cooling, coolant makeup, reactivity control, or 

reactor vessel pressure control). 

Front-line systems are mitigating systems that directly perform an accident mitigating function. 

Typically, support systems (e.g., electric power, control power, or cooling) are required to 

enable the operation of systems that directly perform an accident mitigating function. In this 

regard, support systems also may be considered mitigating systems. 

Model (PRA) 

A representation of 

a physical process 

or system that 

allows one to 

predict the 

system’s behavior. 

(see Uncertainty)

The term “model” is used in a variety of ways in a PRA: 

• The entire PRA is sometimes referred to as a PRA model or risk model.

• Different submodels are used inside the PRA in the performance of the various 

technical elements (system model, human reliability analysis model). 

• Other submodels may be phenomenological models (e.g., direct containment heating 

or core-concrete interaction). 

All of these types of models may be sources of model uncertainty in the PRA. 

Model Uncertainty

(see Uncertainty) The term model uncertainty is related to epistemic uncertainty and is defined under 

“Uncertainty.” 

Nonsafety Related 

(see Safety 

Significant)

The term nonsafety related indicates the safety category of a structure, system, or component 

and is defined under “Safety Significant.” 

Operating-Basis Earthquake 

An earthquake that 

could be expected 

to affect the site of 

a nuclear reactor, 

but for which the 

plant's power 

production 

equipment is 

designed to remain 

functional without 

undue risk to public 

health and safety. 

(see SafeShutdown 

Earthquake)

In a seismic PRA, the operating-basis earthquake (OBE) is sometimes used in the initiating 

event selection process to develop a hierarchy to ensure that every earthquake greater than a 

certain defined size produces a plant shutdown within the systems model. As noted in the 

ASME/ANS PRA Standard (Ref. 2), it is generally a requirement at all nuclear power plants that 

any earthquake larger than a certain size—usually defined as the OBE—will require the plant to 

shut down to reduce energies that may cause loss-of-coolant accidents and to enable 

inspection for possible earthquake-caused damage. 

The ASME/ANS PRA Standard (Ref. 2) defines an OBE as “that earthquake for which those 

features of the nuclear power plant necessary for continued operation without undue risk to 

health and safety are designed to remain functional. In the past, the OBE was commonly 

chosen to be one-half of the safe shutdown earthquake (SSE).” 

The definition provided is based on the definition in the NRC Web site Glossary (Ref. 36). 
4. GLOSSARY

 4-68 

TERM AND 

DEFINITION DISCUSSION 

Operator Action 

(see Human 

Action)

The term operator action is a specific type of human action that is defined under the term 

“Human Action.” 

Operator Error

(see Human 

Error) 

The term operator error is a specific type of human error that is defined under the term “Human 

Error.” 

Other External Hazard Fragility Evaluation/ Analysis

(see Fragility 

Analysis)

The term other external hazard fragility analysis is a type of fragility analysis and is included in 

the discussion under “Fragility Analysis.”

Other External Hazard Plant Response Analysis/Model

(see Plant 

Response 

Analysis)

The term other external hazard plant response analysis is a type of plant response analysis 

and is included the discussion under “Plant Response Analysis/Model.”

Other Hazards Analysis

(see Hazard 

Analysis)

The term other hazards analysis is a specific type of hazard analysis and is defined under the 

term “Hazard Analysis.” 

Parameter 

The variables used 

to calculate and 

describe 

frequencies and 

probabilities. (see 

Uncertainty, Point 

Estimate)

In a PRA, parameters are used directly in supporting PRA models. Initiating event frequencies, 

component failure rates and probabilities, and human error probabilities are several parameters 

used in quantifying the accident sequence frequencies. 

Generally accepted probability models exist for many of the basic events modeled in the PRA 

model. These “basic event” models typically are simple mathematical models with only one or 

two parameters. An example is the simple constant failure rate reliability model, which 

assumes that the failures of components in a standby state occur at a constant rate. The 

parameter(s) of such models may be estimated using appropriate data, which, in the example 

above, may come from the number of failures observed in a population of like components in a 

given period of time. Statistical uncertainties are associated with the estimates of the model’s 

parameters. Because most of the events that constitute the building blocks of the risk model 

(e.g., some initiating events, operator errors, and equipment failures) are relatively rare, the 

data are scarce and the uncertainties can be relatively significant. 

Parameter Uncertainty

 (see 

Uncertainty) 

The term parameter uncertainty is related to epistemic uncertainty and is defined under 

“Uncertainty.” 

Passive Component 

A component 

whose operation or 

function does not 

depend on an 

external source of 

motive power. 

(see Active 

Component)

In a PRA, both passive and active components are modeled. A passive component has no 

moving parts, and it can experience changes in pressure, temperature, or fluid flow in 

performing its functions. Some examples of passive components include heat exchangers, 

pipes, vessels, and electrical cables and structures. 

The IAEA Safety Glossary (Ref. 7) defines passive components as “a component whose 

functioning does not depend on an external input such as actuation, mechanical movement, or 

supply of power.”
4. GLOSSARY

 4-69 

TERM AND 

DEFINITION DISCUSSION 

Performance-Based (Approach, Regulation, Regulatory Action)

Focusing on 

measurable 

outcomes, rather 

than prescriptive 

processes, 

techniques, or 

procedures. (see 

Risk-Based)

In a PRA, a quantitative evaluation is made about the performance of the plant in response to 

potential accident conditions. The results of this evaluation can be used to support a 

performance-based approach to plant operations in which measureable outcomes are used to 

show compliance with regulation. 

NUREG/BR-0318 (Ref. 66) defines the term performance-based as “an approach to regulatory 

practice that establishes performance and results as the primary bases for decision-making. 

Performance-based regulations have four common attributes: (1) Measurable, calculable, or 

objectively observable parameters exist or can be developed to monitor performance. (2) 

Objective criteria exist or can be developed to assess performance. (3) Licensees have 

flexibility to determine how to meet the established performance criteria in ways that encourage 

and reward improved outcomes. (4) A framework exists or can be developed in which the 

failure to meet a performance criterion, while undesirable, will not constitute or result in an 

immediate safety concern.” 

The terms performance-based regulation and performance-based regulatory action are defined 

below based on the NRC Web site Glossary (Ref. 36): 

• Performance-Based Regulation: “A regulatory approach that focuses on desired, 

measurable outcomes, rather than prescriptive processes, techniques, or procedures. 

Performance-based regulation leads to defined results without specific direction 

regarding how those results are to be obtained. At the NRC, performance-based 

regulatory actions focus on identifying performance measures that ensure an 

adequate safety margin and offer incentives for licensees to improve safety without 

formal regulatory intervention by the agency.” 

• Performance-Based Regulatory Action: “Licensee attainment of defined objectives 

and results without detailed direction from the NRC on how these results are to be 

obtained.” 

Performance-Based Approach 

(see 

PerformanceBased)

The term performance-based approach indicates an evaluation that is based on measureable 

outcomes and is defined under “Performance-Based.” 

Plant Configuration Control 

The process of 

maintaining 

consistency 

between the 

physical condition 

of a nuclear plant 

and its associated 

design and 

engineering 

records. 

A PRA relies on plant configuration control to ensure that the as-built as-operated plant is 

accurately modeled. Without plant configuration control, uncertainty can be introduced about 

the extent to which the PRA accurately reflects important characteristics of the plant (e.g., the 

design of plant structures, systems, and components (SSCs)). 

Plant configuration control represents the process of identifying and documenting the 

characteristics (e.g., design or operating conditions) of plant SSCs, and of ensuring that 

changes to these characteristics are properly developed, assessed, approved, issued, 

implemented, verified, recorded, and incorporated into the facility documentation. 

Plant Damage State 

A group of 

accident sequence 

end states that 

share similar 

characteristics with 

accident 

progression, and 

containment or 

engineered safety 

feature operability. 

(see Bin)

In a Level 2 PRA, the critical first step is developing a structured process for defining the 

specific accident conditions to be examined. Attributes have to be determined for binning the 

large number of accident sequences developed for Level 1 PRA analysis into a practical 

number for detailed Level 2 analysis. Combinations of attributes of similar accident conditions 

define the plant damage states. 

The definition provided is based on the definition in the ASME/ANS PRA Standard (Ref. 2).
4. GLOSSARY

 4-70 

TERM AND 

DEFINITION DISCUSSION 

Plant Hazard 

(see Hazard) The term plant hazard has the same meaning as hazard and is defined under “Hazard.” 

Plant Operational Mode 

(see Plant 

Operational 

State) 

The term plant operational mode has the same meaning as plant operational state and is 

defined with “Plant Operational State.” 

Plant Operational State, Plant Operational Mode 

A particular plant 

configuration with 

specified 

operational 

characteristics. 

The scope of the PRA determines the various individual plant operational states (POSs) that 

the PRA model must include for the risk estimation results (i.e., if a PRA is being conducted for 

at-power operations, the plant configuration in that state or mode will be considered to obtain 

the risk results). The term plant operational state has the same meaning as plant operational 

mode. 

The plant conditions that define a POS usually include core decay heat level, primary water 

level, primary temperature, primary vent status, containment status, and decay heat removal 

mechanisms. A POS can be a steady state or represent a transition between steady-state 

POSs. For example, full power and cold shutdown while on residual heat removal cooling may 

be two steady-state POSs. To transition from full power to cold shutdown, there may be one or 

more transition POSs to cover the range of temperatures and pressures the plant goes through 

in shutting down to cold shutdown. 

Note that the impacts of unavailability of individual systems or components because of test or 

maintenance typically are not included as part of the definition of a POS. The complete set of 

POSs for a specific outage type shows a discrete representation of the outage from a risk 

perspective. 

Plant Partitioning 

The defining of the 

plant physical 

boundary affected 

by the flood and 

fire hazard and the 

segmenting of the 

physical boundary 

into smaller spatial 

units. 

In a PRA, plant partitioning is used in flood and fire evaluations to define the physical analysis 

units in terms of flood or fire areas and flood or fire compartments. In the ASME/ANS PRA 

Standard (Ref. 2), the objective of plant partitioning for internal floods (referred to as internal 

flood plant partitioning) is to account for plant-specific physical layouts and separations in such 

a way as to identify in the PRA plant areas where internal floods could lead to core damage. 

Plant Response Analysis, Plant Response Model (External Floods, Internal Fire, High 

Winds, Other External Hazard, Seismic) 

The logic 

framework for 

identification and 

analysis of 

accident scenarios 

resulting from the 

effects of a hazard 

on the plant.

In a PRA conducted to evaluate the effect of an external hazard group on the plant, or the 

effect of internal fires on the plant, plant response analysis usually involves modification of the 

internal events PRA model. This modification includes the event trees and fault trees and the 

initiating event set. It involves identifying and selecting important initiating events, deleting 

unlikely events from event trees, deleting unimportant internal failures and human errors (from 

fault trees or event trees), modifying event tree logic to conform to event-specific procedures, 

and adding hazard event induced failure events and human errors (to fault trees and event 

trees). These modifications are performed when the plant response model is used in 

conducting an external flood, internal fire, high wind, seismic, or other external hazards 

analysis. 
4. GLOSSARY

 4-71 

TERM AND 

DEFINITION DISCUSSION 

For example, in a seismic analysis, the initiating event is assumed to be a loss of offsite power. 

Recovery of offsite power is trimmed from the event trees. Seismic failures of structures and 

equipment are added and comparatively unimportant internal failures are trimmed. Human 

errors and their probabilities are adjusted. Mission time is extended, usually to 72 hours. 

A simplified plant response model also can be constructed “from scratch” (ad hoc model), 

without starting with the internal events model. 

Note that in an internal flood PRA the plant response also is determined in a manner similar to 

that described above. The ASME/ANS PRA Standard (Ref. 2) states that the expected plant 

response(s) to the selected set of flood scenarios is determined, and an accident sequence, 

from the internal events at power PRA that is reasonably representative of this response is 

selected for each scenario. 

Plant Response Model 

(see Plant 

Response 

Analysis) 

The term plant response model has the same meaning as plant response analysis and is 

defined under “Plant Response Analysis.” 

Plant Risk Profile 

(see Risk 

Profile)

The term plant risk profile has the same meaning as risk profile and is defined under “Risk 

Profile.”

Point Estimate 

An estimate of a 

parameter in the 

form of a single 

value. (see Mean)

In a PRA, the preferred parameter point estimate is the mean of the value obtained from a 

probability distribution for the parameter. 

 

NUREG-1855 (Ref. 62) states, “a point estimate is a single value estimate for a parameter 

population. For example, the mean of a sample of values of a random variable X 

(i.e., expected value) is a commonly used point estimate of the mean of the distribution. When 

parameter distributions are not available, a maximum likelihood estimate or a value obtained 

from expert elicitation can serve as a point estimate.” 

 

For a point estimate of a risk metric (e.g., core damage frequency) mean values of various 

parameters are used. The mean value of the risk metric usually is very close to this point 

estimate. 

The definition provided was based on the definition in NUREG/CR-6823 (Ref.78). 

PRA, Probabilistic Safety Assessment (Base, Baseline)

A systematic 

method for 

assessing the 

likelihood of 

accidents and their 

potential 

consequences. 

(see Probability, 

Dynamic PRA, 

Full-Scope PRA, 

Level 1, 2, 3 PRA)

The term probabilistic risk assessment (PRA) has numerous, similar definitions. Some of the

formal definitions used are presented below: 

• “A qualitative and quantitative assessment of the risk associated with plant operation 

and maintenance that is measured in terms of frequency of occurrence of risk 

metrics, such as core damage or a radioactive material release and its effects on the 

health of the public (also referred to as a probabilistic safety assessment (PSA)).” 

(Ref. 2)

• “For a method or approach to be considered a PRA, the method or approach 

provides (1) a quantitative assessment of the identified risk in terms of scenarios that 

result in undesired consequence (e.g., core damage or large early release) and their 

frequencies, and (2) is comprised of specific technical elements in performing the 

quantification.” (Ref. 91)
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

• “A systematic method for assessing three questions used to define “risk.” These 

questions consider (1) what can go wrong, (2) how likely it is, and (3) what its 

consequences might be. These questions allow understanding of likely outcomes, 

sensitivities, areas of importance, system interactions, and areas of uncertainty, 

which can identify risk-significant scenarios. The PRA determines a numeric estimate 

of risk to provide insights into the strengths and weaknesses of the design and 

operation of a nuclear power plant.” (Ref. 36)

A specific type of PRA is the base or baseline PRA, which represents the as-built as-operated 

plant to the extent needed to support the application. For a nuclear power plant at the design 

certification or combined operating license stage, where the plant is not built or operated, the 

base(line) PRA model reflects the as-designed plant. This type of PRA is also used as a 

benchmark to estimate the change in risk from a proposed design change. A dynamic PRA is 

a special type of PRA that automatically accounts for time-dependent effects by integrating 

these effects directly into the computer model. In a traditional PRA, time-dependent effects are 

accounted for manually. A full-scope PRA addresses three specific levels of analysis; namely, 

Level 1 (core damage), Level 2 (radioactive material release), and Level 3 (consequences). 

The term probabilistic safety assessment is another term that is sometimes used 

interchangeably with PRA. Typically, the term probabilistic safety assessment is used outside 

of the U.S. 

PRA Configuration Control (Maintenance, Upgrade)

A process that 

maintains and 

updates the 

probabilistic risk 

assessment so 

that it reflects the 

as-built asoperated facility. 

(see Living PRA, 

Risk Management)

In a PRA, updates to the model may be needed to ensure that the PRA reflects the as-built asoperated facility. As described in the ASME/ANS PRA Standard (Ref. 2), a “PRA configuration 

control program shall include a process to monitor changes in the design, operation, 

maintenance, and industry-wide operational history that could affect the PRA. These changes 

shall include inputs that impact operating procedures, design configuration, initiating event 

frequencies, system or subsystem unavailability, and component failure rates. The program 

should include monitoring of changes to the PRA technology and industry experience that 

could change the results of the PRA model.” 

As further described in the ASME/ANS PRA Standard (Ref. 2), PRA maintenance involves 

“update of the PRA models to reflect plant changes such as modifications, procedure changes, 

or plant performance (data).” 

Additionally, the ASME/ANS PRA Standard (Ref. 2) states that a PRA upgrade involves “the 

incorporation into a PRA model of a new methodology or changes in scope or capability that 

impact the significant accident sequences or the significant accident progression sequences. 

This could include items such as new human error analysis methodology, new data update 

methods, new approaches to quantification or truncation, or new treatment of common cause 

failure.” 

PRA configuration control is part of the process used to support a living PRA (i.e., a PRA that is 

continuously updated to reflect current plant design, configuration, operating procedures, and 

plant-specific data). 

Listed below are definitions of related terms: 

• Configuration risk management: The term configuration risk management is the 

same as risk management and is defined under “Risk Management.”

• Configuration risk profile: A change in the overall nuclear power plant risk metric 

value as a result of a change from the initial plant configuration. Results from a PRA 

can be used as the basis for developing configuration risk profiles using various risk 

metrics (e.g., core damage frequency, large early release frequency). The 

configuration risk profile can depend on the plant operational status. For example, 

during certain shutdown operations when the containment function is not maintained, 
4. GLOSSARY

 4-73 

TERM AND 

DEFINITION DISCUSSION 

the risk metric represented by large early release fraction is not applicable; therefore, 

licensees may use more stringent baseline core damage frequency guidelines to 

maintain an equivalent risk profile. 

PRA Maintenance 

(see PRA 

Configuration 

Control)

The term PRA maintenance is part of PRA configuration control and is defined under “PRA 

Configuration Control.” 

PRA Model 

(see Model) The term PRA model is a specific type of model and is defined under the term “Model.” 

PRA Technical Acceptability 

(see Technical 

Acceptability)

The term PRA technical acceptability is discussed in the discussion for the term “Technical 

Acceptability.”

PRA Technical Adequacy 

(see Technical 

Adequacy)

The term PRA technical adequacy is discussed in the discussion for the term “Technical 

Adequacy.”

PRA Technical Elements 

The basic pieces 

(or analyses) 

required to 

produce the PRA 

model. (see 

Appendix B)

The individual analyses used in the development of a PRA model are organized according to a 

set of PRA technical elements. As described in the ASME/ANS PRA Standard (Ref. 2), a 

number of specific PRA technical elements are used to support the evaluation of contributors to 

risk (e.g., the evaluation of hazard groups). Examples of PRA technical elements include the 

following: initiating events analysis, accident sequence analysis, and high wind hazard 

analysis. 

PRA Upgrade 

(see PRA 

Configuration 

Control) 

The term PRA upgrade is part of PRA configuration control and is defined under “PRA 

Configuration Control.” 

Precursor Event 

(see Accident 

Precursor) 

The term precursor event is the same as accident precursor and is defined under “Accident 

Precursor.”

Probabilistic (Analysis, Approach) 

A characteristic of 

an evaluation that 

includes 

consideration of 

events with regard 

to their likelihood. 

(see Deterministic, 

PRA, Risk-Based, 

Risk-Informed)

A PRA is an example of a probabilistic analysis, which can be defined as a mathematical 

evaluation of random (stochastic) events or processes and their consequences. While a PRA 

uses probabilistic analysis, a PRA also depends on deterministic analyses. For example, 

success criteria for various systems modeled in a PRA to prevent and mitigate core damage 

are based on deterministic analyses. 

A probabilistic approach can be defined as a method that accounts for the likelihood of possible 

states that a physical entity or system can assume and predictions of models describing the 

entity or system. 

Both risk-based and risk-informed approaches to decisionmaking and regulation rely upon 

probabilistic analysis. A risk-based approach to decisionmaking or regulation means that the 

decision or regulation is based only on risk information generated from a probabilistic analysis 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

(e.g., from a PRA), whereas a risk-informed approach combines risk information generated 

from a probabilistic analysis with other factors to arrive at a decision or develop regulations. 

The NRC Web site Glossary (Ref. 36) states the following: “The term ‘probabilistic’ is 

associated with an evaluation that explicitly accounts for the likelihood and consequences of 

possible accident sequences in an integrated fashion.” Therefore, a probabilistic analysis or 

approach is unlike a deterministic analysis or approach, which does not include consideration 

of events with regard to their likelihood. 

Probabilistic Analysis 

(see 

Probabilistic)

The term probabilistic analysis is defined under “Probabilistic.”

Probabilistic Approach 

 (see 

Probabilistic)

The term probabilistic approach is defined under “Probabilistic.”

Probabilistic Safety Assessment 

(see PRA) The term probabilistic safety assessment is another term for PRA and is defined under “PRA.”

Probabilistic Seismic Hazard Analysis

(see Hazard 

Analysis)

The term probabilistic seismic hazard analysis is a specific type of hazard analysis and is 

defined under “Hazard Analysis.” 

Probability (Basic Event Failure, Containment Failure, Core Damage, Failure, Human 

Error)

The likelihood that 

an event will occur 

as expressed by 

the ratio of the 

number of actual 

occurrences to the 

total number of 

possible 

occurrences. (see 

Frequency)

In a PRA, probability is calculated for various types of PRA input and output parameters 

(e.g., failures of equipment associated with basic events, core damage, and containment 

failure). 

The probability assigned to a basic event is often referred to as the basic event failure 

probability. A basic event is an element of the PRA model for which no further decomposition 

is performed because it is at the limit of resolution consistent with available data. A failure 

probability is calculated for each failure mode of a component (e.g., failure to start and failure to 

run for a pump). In addition, a failure probability may be calculated for a system failing to 

perform its function or a structure failing (e.g., given a seismic event). For example, 

containment failure probability is the likelihood that the containment structure fails to perform its 

function of retaining fission products. 

The ASME/ANS PRA Standard (Ref. 2) defines failure probability as “the likelihood that a 

system or component will fail to operate upon demand or fail to operate for a specific mission 

time.” 

Failure probability is also calculated for human actions and is then called human error 

probability. The ASME/ANS Standard (Ref. 2) defines human error probability as a measure of 

the likelihood that plant personnel will fail to initiate the correct, required, or specified action or 

response in a given situation, or by commission performs the wrong action. 

Some PRA studies also calculate the probability of core damage, also referred to as core 

damage probability, given a particular initiating event or set of initiating events. 

There is a tendency in risk communication to use frequency and probability synonymously, but 

incorrectly. Probability only conveys the likelihood of an event; frequency conveys that 

likelihood per unit time. 

The definition provided was based on the definition in NUREG/CR-6823 (Ref. 78).
4. GLOSSARY

 4-75 

TERM AND 

DEFINITION DISCUSSION 

Probability Density Function 

(see Probability 

Distribution)

The term probability density function is an equivalent term for probability distribution and is 

defined under “Probability Distribution.” 

Probability Distribution (Probability Density Function) 

A curve that shows 

all the values that 

a random variable 

can take and the 

likelihood that 

each will occur. 

(see Cumulative 

Distribution 

Function, Mean, 

Median, 

Uncertainty 

Interval)

In a PRA, probability distributions are used to express uncertainties associated with the stateof-knowledge about the parameter values and models used in constructing the PRA. 

A probability distribution can represent either a discrete or continuous set of values for a 

random variable. It is usually represented as a probability density function. The probability 

density function is a function of a continuous random variable whose integral over an interval 

gives the probability that its value will fall within the interval. 

In comparison, the cumulative distribution function adds up the probabilities of occurrence of all 

possible parameter values in a probability distribution function that are less than a specified 

value. An illustration of a probability distribution function and its corresponding cumulative 

distribution function is shown under the discussion for the term “Cumulative Distribution.” 

Prompt Fatality 

(see Fatality) The term prompt fatality is a type of fatality caused by exposure to radioactive materials and is 

defined under “Fatality.”

Prompt Fatality Risk 

(see Fatality) The term prompt fatality risk is a type of fatality caused by exposure to radioactive materials 

and is defined under “Fatality.”

Public Health Effects 

(see Health 

Effects)

The term public health effect refers to a type of health effect and is defined in the discussion 

under “Health Effects.” 

Qualitative Risk Assessment 

(see Risk) A qualitative risk assessment is one type of risk assessment and is defined under “Risk.”

Qualitative Screening

(see Screening) A qualitative screening is one type of screening performed and is defined under “Screening.”

Quantitative Health Objectives 

Numerical criteria 

for the acceptable 

levels of risk to 

In some risk-informed decisions, the results of a PRA are used to compare the risk from the 

plant with the quantitative health objectives (QHOs) that support the NRC’s reactor safety 

goals.

Event Probability

Risk Metric

Probability Distribution Function
4. GLOSSARY

 4-76 

TERM AND 

DEFINITION DISCUSSION 

public health and 

safety in the 

population 

surrounding a 

nuclear power 

plant that satisfy 

the NRC's reactor 

safety goals. (see 

Fatality, Risk to 

Average individual)

The NRC safety goals (Ref. 30) are expressed by two QHOs: (1) the annual average individual 

probability of prompt fatality in the population within 1 mile of the site boundary of a nuclear 

power plant should not exceed one-tenth of 1 percent of the risk of prompt fatality due to all 

other risks (non-nuclear) that the U.S. population is generally exposed to, and (2) the annual 

average individual probability of latent cancer fatality in the population within 10 miles of the 

site boundary of a nuclear power plant should not exceed one-tenth of 1 percent of the U.S. 

cancer fatality rate due to all other (non-nuclear) causes. 

Quantitative Screening 

 (see Screening) A quantitative screening is one type of screening and is defined under “Screening.” 

Radioactive Material 

A substance that 

emits ionizing 

radiation. (see 

Radionuclide, 

Fission Product)

In a PRA, the terms radionuclide, radioactive material, and fission product are used 

interchangeably. These terms are meant to refer to the substance that is the source of the risk 

being evaluated. However, a release of this substance (i.e., radioactive material) from the 

reactor and from the containment, or from another source such as the spent fuel pool, could 

have an adverse impact on public health and safety is generally not referred to as radioactive 

material release. Generally, either radionuclide release or fission product release, or just 

‘release’ is used. 

Radioactive Material Release (Large Early, Small Early, Large Late, Small Late)

The release of 

radioactive 

material to the 

environment. (see 

Radioactive 

Material, 

Radioactive 

Material Release 

Frequency 

Analysis, Health 

Effects)

In a Level 2 PRA, the release of radioactive material from the reactor core to the environment is 

calculated. Usually this is referred to as the ‘release,’ ‘radionuclide release,’ or ‘fission product 

release.’ This release may occur early or late and may be large or small. 

In the ASME/ANS PRA Standard (Ref. 2), a large early release is defined as a rapid, 

unmitigated release of airborne fission products from the containment to the environment 

occurring before the effective implementation of offsite emergency response and protective 

actions so there is a potential for early health effects. 

A small early release is of low enough magnitude to have minimal potential for early health 

effects. 

A large late release can be defined as a release of airborne fission products from the 

containment to the environment of sufficient magnitude to cause severe health effects. 

However, the release occurs in a timeframe that allows the effective implementation of offsite 

emergency response and protective actions such that the offsite health effects can be 

significantly reduced compared to those of a large early release. 

A small late release is of low enough magnitude and is delayed long enough to have minimal 

potential for health effects. 

For both early and late large releases, significant land contamination and property damage is to 

be expected. The term large release is discussed as its own entry in this glossary. The 

Commission has not approved a formal definition for the term large release. 

Radioactive Material Release Frequency (Large Early, Small Early, Large Late, Small Late)

 (see 

Frequency)

The term radioactive material release frequency (large early, small early, large late, small late) 

is a type of frequency used in PRA and is defined in the discussion under “Frequency.” 
4. GLOSSARY

 4-77 

TERM AND 

DEFINITION DISCUSSION 

Radioactive Material Release Frequency Analysis (Large Early, Small Early, Large Late, 

Small Late) 

An estimation of 

the frequency of 

radioactive 

material releases 

by evaluating the 

core and 

containment 

behavior under 

severe accident 

conditions. (see 

Radioactive 

Material Release, 

Health Effects)

In a Level 2 PRA, the frequency of release of radioactive material from the reactor core to the 

environment is calculated. This release may occur early or late and may be large or small. For 

operating reactors, a large early release frequency is one of the risk metrics used for 

risk-informed decisions. For new reactors, a large release frequency is one of the risk metrics 

used for risk-informed decisions. 

Radiological Source Term

 (see Source 

Term) 

The term radiological source term has the same meaning as source term and is defined under 

“Source Term.” 

Radiological Source Term Analysis 

(see Source 

Term Analysis)

The term radiological source term analysis has the same meaning as source term analysis and 

is defined under “Source Term Analysis.” 

Radionuclide 

An atom with an 

unstable nucleus 

that emits radiation 

(see Radioactive 

Material, Fission 

Product)

In a PRA, the terms radionuclide, radioactive material, and fission product are used 

interchangeably. These terms are meant to refer to the substance that is the source of the risk 

being evaluated. A radionuclide release, therefore, refers to the release of the substance (i.e., 

radionuclides) from the reactor and from the containment that could have an adverse impact on 

public health and safety. 

The NRC Web site Glossary (Ref. 38) defines radionuclide as “an unstable isotope of an 

element that decays or disintegrates spontaneously, thereby emitting radiation. Approximately 

5,000 natural and artificial radioisotopes have been identified.” 

Random Failure 

A failure not 

anticipated to 

occur at a certain 

time (i.e., occurring 

with no specific 

pattern). 

In a PRA, potential failures of the modeled structures, systems, and components (SSCs) are 

treated as random events. This treatment is necessary because it is not possible to predict 

when an SSC will possibly fail. Instead, it is only possible to predict the likelihood that an SSC 

will fail. The likelihood that an SSC will fail is based on failure rate data, which represents the 

expected number of failures of the SSC per unit time. Failure rate data are developed for each 

SSC modeled in a PRA. 

Random Uncertainty 

 (see 

Uncertainty)

The term random uncertainty is related to aleatory uncertainty and defined under “Uncertainty.” 

Rare Initiator 

An initiating event 

that is extremely 

unlikely and not 

expected to occur 

In a PRA, rare initiators generally are screened because of their low frequencies. Examples of 

rare initiators include aircraft impact, meteor strikes, and very large earthquakes. These 

occurrences are also correctly referred to as rare events. 
4. GLOSSARY

 4-78 

TERM AND 

DEFINITION DISCUSSION 

in nuclear power 

plants. (see 

Initiating Event)

The ASME/ANS PRA Standard (Ref. 2) defines the term rare event as “one that might be 

expected to occur only a few times throughout the world nuclear industry over many years 

(e.g., <1E-4/r-yr).” However, the ASME/ANS PRA Standard only allows screening of initiating 

events if the frequency is much lower than 1E-4/yr (e.g., if the frequency <1E-7/yr and the 

event does not involve either an ISLOCA, containment bypass, or reactor, or reactor pressure 

vessel rupture). 

Rationalist 

An approach to 

defense-in-depth 

that uses 

probabilistic 

information to 

evaluate the 

uncertainties and 

to determine what 

steps should be 

taken to 

compensate for 

those 

uncertainties. (see 

Structuralist, 

Defense-in- Depth)

When used in a PRA context, the term rationalist is a relatively new term associated with 

defense-in-depth. The rationalist approach is made practical by the ability to quantify risk and 

estimate uncertainties using PRA techniques. In this approach, results from a PRA or other 

probabilistic analysis are used to assess the strengths and weaknesses of defense-in-depth, 

while accounting for analysis uncertainties. Ultimately, the rationalist approach provides a way 

to increase the degree of confidence in the conclusion that the defense-in-depth is sufficiently 

robust to achieve adequate safety. 

In contrast, the fundamental principle of the structuralist approach is that if a system is 

designed to withstand all the worst-case credible accidents, then it is by definition protected 

against any credible accident. It is a deterministic method of establishing how precautions can 

be placed into a system, just in case an existing barrier or system fails. 

The Advisory Committee on Reactor Safeguards (Ref. 39) describes that the rationalist will 

“(1) establish quantitative acceptance criteria, such as the quantitative health objectives, core 

damage frequency and large early release frequency, (2) analyze the system using PRA 

methods to establish that the acceptance criteria are met, and (3) evaluate the uncertainties in 

the analysis, especially those due to model incompleteness, and determine what steps should 

be taken to compensate for those uncertainties.” 

Reactor Core 

The location within 

a nuclear reactor 

where the fission 

process occurs.

In a PRA, the source of risk generally evaluated is the reactor core with an understanding that 

the actual risk is from the fuel. The reactor core includes the fuel assemblies, moderator, 

neutron poisons, control rods, and support structures. The other sources of risk at the plant 

site (e.g., spent fuel) generally are not included in the reactor core PRA; however, there are 

several PRAs, separate from the reactor core PRAs, which evaluate the risk of the spent fuel. 

The NRC Web site Glossary (Ref. 36) defines reactor core as “the central portion of a nuclear 

reactor, which contains the fuel assemblies, moderator, neutron poisons, control rods, and 

support structures. The reactor core is where fission takes place.”

Reactor-Operating-State-Year 

(see ReactorYear) 

The term reactor-operating-state-year is related to the term reactor-year and is defined under 

“Reactor-Year.” 

Reactor-Year (Reactor-Operating-State-Year) 

A unit of time by 

which risk 

parameters are 

measured in a 

PRA. (see Plant 

Operational State)

In a PRA, the terms reactor-year and reactor-operating-state-year refer to units of time by 

which risk parameters (e.g., core damage frequency, large early release frequency) are 

measured. The ASME/ANS PRA Standard (Ref. 2) defines the term reactor-year as 

“a calendar year in the operating life of one reactor, regardless of power level.” The term 

reactor-year assumes that more than one reactor can operate during a year (e.g., a calendar 

year during which five reactors operated would be the experience equivalent of 

5 reactor-years). 

For some applications, such as configuration risk management or analyses that compare 

specific risks during different modes of operation, it may be appropriate to develop risk metrics 

that consider the time period associated with a given plant operational state. For at-power 
4. GLOSSARY

 4-79 

TERM AND 

DEFINITION DISCUSSION 

operation, this basis is sometimes referred to as per reactor critical year (i.e., assuming that the 

reactor operated continuously for a year). On a more general basis, it could be considered to 

be per reactor-operating-state-year. The ASME/ANS PRA Standard (Ref. 2) defines the term 

reactor-operating state-year as “an equivalent calendar year of operation in a particular plant 

operating state.” 

Realistic Analysis 

(see 

Conservative 

Analysis)

The term realistic analysis is discussed in the discussion for “Conservative Analysis” and is 

defined there. 

Recovery 

Restoration of a 

failed function. 

(see Repair)

In a PRA, the term recovery usually refers to an action or series of actions performed by an 

operator or other plant personnel to restore a function in response to a failed component or 

system. This term is sometimes used incorrectly as a synonym for repair. However, repair is 

restoring a failed function by fixing the actual cause of the failure while recovery is restoring the 

function in some other way. 

The ASME/ANS PRA Standard (Ref. 2) defines the term recovery as “restoration of a function 

lost as a result of a failed structure, system or component (SSC) by overcoming or 

compensating for its failure. Generally modeled by using human reliability analysis (HRA) 

techniques.” 

Release 

(see Radioactive 

Material 

Release) 

For purposes of a Level 2 and Level 3 PRA, the term release is used interchangeably with 

“Radioactive Material Release.” 

Release Category 

A group of 

radioactive 

material releases 

expected to result 

in similar 

consequences. 

(see Source Term) 

In a Level 2 PRA, a release category is a grouping of accident sequences into an accident 

sequence class or family based on a common potential for release of radioactive material. 

The release categories are characterized by a bounding mechanistic source term. This 

grouping is based on the following common attributes: common initiating events, combination 

of successful and failed safety functions, release magnitude, release timing and location, and 

radioactive material species released from the plant as a result of an accident. 

Release Fraction 

The amount of 

radioactive 

material released 

from the reactor 

core expressed as 

a fraction of the 

original inventory 

of the radioactive 

material. (see 

Source Term)

In a Level 2 PRA, the release fraction specifies the amount of radioactive materials released to 

the environment and provides the basis for the subsequent dose calculations to the affected 

population. 

NUREG-1489 (Ref. 54) states that the release fraction is expressed as the amount of 

radioactive material released from the containment as a function of time given as a fraction of 

the fission product inventory in the core at the time of the start of the accident. 

Release Timing and Duration 

The time of 

release and the 

timeframe over 

In a Level 3 PRA, the time of release and its duration are used to calculate the health 

consequences to the affected population. Both the timing and duration of the release also form 

the basis for potential offsite protective action strategies. 
4. GLOSSARY

 4-80 

TERM AND 

DEFINITION DISCUSSION 

which the 

radioactive 

materials are 

released to the 

environment 

during an accident. 

(see Source Term) 

Reliability (Unreliability) 

The likelihood that 

a system, 

structure, or 

component 

performs its 

required 

function(s) for a 

specified period of 

time. (see 

Availability) 

In a PRA, the unreliability of systems, structures and components, as well as human actions, 

are used as input to the PRA model, as opposed to the reliability. Unreliability is the 

complement of reliability and is the likelihood that a structure, system, and component (SSC) 

does not operate for its mission time when required. 

The term reliability is often inappropriately used interchangeably with the term availability. 

Availability only represents the degree to which a SSC is operational and accessible when 

required for use, with no reference to a mission time. Availability is the likelihood that the SSC 

is in a state to perform its required function at a given moment in time. 

In the ASME/ANS PRA Standard (Ref. 2), unreliability is defined as “the probability that 

a system or component will not perform its specified function under given conditions upon 

demand or for a prescribed time.” 

Repair

The restoration of 

a failed function by 

correcting the 

cause of failure. 

(see Recovery)

In a PRA, the term repair usually refers to an action or series of actions performed by an 

operator or other plant personnel to restore the function of a failed structure, system, or 

component (SSC) by correcting the cause of failure and returning the failed SSC to service so 

that it can perform its intended function(s). 

This term is sometimes used incorrectly as a synonym for the term recovery. However, repair 

is restoring a failed function by fixing the actual cause of the failure while recovery is restoring 

the function in some other way. 

The ASME/ANS PRA Standard (Ref. 2) defines the term repair as “restoration of a failed SSC 

by correcting the cause of failure and returning the failed SSC to its modeled functionality. 

Generally modeled by using actuarial data.” 

Response Time

The period of time 

something takes to 

react to a given 

input.

In a PRA, the term response time has different connotations, depending on the situation. 

Some of these connotations are as follows: 

• When referring to plant components, response time is “the period of time necessary 

for a component to achieve a specified output state from the time that it receives a 

signal requiring it to assume that output state.” (Ref. 7)

• When referring to human reliability analysis, response time is the time required for 

“the actions carried out after the operator has received and processed information 

related to his tasks. These responses constitute the human outputs in a manmachine system and serve as inputs to the man-machine interfaces.” (Ref. 68)

• When referring to a Level 3 PRA emergency response, response time is the time 

required for offsite responders to arrive at a plant site during an emergency (as 

related to accident response and accident preparedness). 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

Risk (Assessment, Analysis) 

The combined 

answer to three 

questions that 

consider (1) what 

can go wrong, (2) 

how likely it is, and 

(3) what its 

consequences 

might be. (see 

PRA, Level 1, 2, 3 

PRA, Risk Metric)

Risk assessment or risk analysis and PRA are often incorrectly used as synonyms. A PRA is 

one type of risk assessment or risk analysis. The PRA has a structured format and quantifies 

the ultimate consequences. A risk assessment or risk analysis does not necessarily reflect all 

the technical elements. For example, a seismic margin risk analysis is not a PRA. A 

qualitative risk assessment or analysis is a risk evaluation that uses descriptions or distinctions 

based on some characteristic rather than on some quantity or measured value. 

In comparison to a risk assessment or analysis, a PRA generates different ways to measure 

risk, called risk metrics, which satisfy specified safety objectives or goals. The consequences 

are manifested in the onset of core damage and each level of the PRA uses different risk 

metrics, which can be found in the discussion of Level 1, 2, 3 PRA. 

The ASME/ANS PRA Standard (Ref. 2) defines the term risk as the “probability and 

consequences of an event, as expressed by the “risk triplet” that is the answer to the following 

three questions: (a) What can go wrong? (b) How likely is it? (c) What are the consequences if 

it occurs?” 

The definition provided was based on the definition in the NRC Web site Glossary (Ref. 36). 

Risk Achievement Worth 

(see Importance 

Measure)

The term risk achievement worth is one type of importance measure and is defined under 

“Importance Measure.”

Risk Characterization 

 (see Risk Metric) The term risk characterization is a process that uses risk metrics to determine risk and is 

defined under “Risk Metric.” 

Risk Insights

The understanding 

about a facility’s 

response to 

postulated 

accidents. (see 

Risk, Risk-Based, 

Risk-Informed)

One of the main objectives of a PRA is to gain insights about a facility’s response to initiating 

events and accident progression, including the expected interactions among facility structures, 

systems, and components (SSCs), and between the facility and its operating staff. Risk insights 

are derived by investigating in a systematic manner: (1) what can go wrong, (2) how likely it is, 

and (3) what the consequences are. A risk assessment is a systematic method for addressing 

these questions as they relate to understanding issues like: important hazards and initiators, 

important accident sequences and their associated SSC failures and human errors, system 

interactions, vulnerable plant areas, likely outcomes, sensitivities, and areas of uncertainty. 

Risk insights can be obtained via both quantitative and qualitative investigations. As noted in 

RG 1.174 (Ref.84), quantitative risk results from PRA calculations are typically the most useful 

and complete characterization of risk, but they are generally supplemented by qualitative risk 

insights and traditional engineering analysis. Qualitative risk insights include generic results, i.e., 

results that have been learned from numerous PRAs that have been performed in the past, and 

from operational experience, and that are applicable to a group of similar plants. 

Risk insights are an important part of risk-informed regulation, in which regulatory decisions are 

made by integrating risk insights with considerations of defense-in-depth and safety margins. 

Risk Management

A process used at 

a nuclear power 

plant to keep the 

risk at acceptable 

levels. 

A PRA is a tool used to evaluate a nuclear plant from a risk management perspective. The 

PRA quantifies the plant risk and also quantifies changes in plant risk because of modifications 

of the plant design or operation. Examples of risk management activities that are supported by 

PRA are listed below: 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

• A PRA represents an important risk management tool that, as stated in Regulatory 

Guide (RG) 1.177 (Ref. 86), “ensures that other potentially lower probability, but 

nonetheless risk-significant, configurations resulting from plant maintenance and 

other operational activities are identified and compensated for.” 

• Regarding the use of PRA findings and risk insights to support licensee requests for 

changes to a plant’s licensing basis, RG 1.174 (Ref. 84) states the following: “All 

safety impacts of the proposed change are evaluated in an integrated manner as part 

of an overall risk management approach in which the licensee is using risk analysis to 

improve operational and engineering decisions broadly by identifying and taking 

advantage of opportunities to reduce risk and not just to eliminate requirements the 

licensee sees as undesirable. For those cases in which risk increases are proposed, 

the benefits should be described and should be commensurate with the proposed risk 

increases. The approach used to identify changes in requirements should be used to 

identify areas in which requirements should be increased as well as those in which 

they can be reduced.”

• In reference to the Maintenance Rule, 10 CFR 50.65 (Ref. 20) states, “the licensee 

shall assess and manage the increase in risk that may result from the proposed 

maintenance activities. The scope of the assessment may be limited to structures, 

systems, and components that a risk-informed evaluation process has shown to be 

significant to public health and safety.” 

Risk Management is used in a broader context in NUREG-2150 (Ref.67), “A Proposed Risk 

Management Regulatory Framework,” to refer to an approach for achieving a more 

comprehensive, holistic, risk-informed, performance-based regulation for reactors, materials, 

waste, fuel cycle, and transportation that would continue to ensure the safe and secure use of 

nuclear material. The objective of such an approach is described NUREG-2150 as managing 

the risks from the use of byproduct, source and special nuclear materials through appropriate 

performance based regulatory controls and oversight. 

Risk Metric 

A measure that is 

used to express 

the risk quantity of 

interest. (see 

Risk, Level 1, 2, 3 

PRA, Risk Profile, 

Full-Scope PRA)

In a PRA, several risk metrics are evaluated. Examples of risk metrics are core damage 

frequency, developed as part of a Level 1 PRA and large early release frequency, developed as 

part of a Level 2 PRA. Health effects developed in a Level 3 PRA also can be used as a risk 

metric. In this instance, limiting to a threshold value the annual average individual probability of 

death due to acute radiation syndrome within 1 mile of the site boundary would be an example 

of a risk metric. A full-scope PRA develops risk metrics associated with Levels 1, 2, and 3. 

Risk metrics are used among other things, to illustrate compliance with safety goals. Risk 

metrics focus attention on those areas where risk is most likely (such as events that cause core 

damage) and how the risk metric value for that area is achieving the desired safety objective. 

Risk metrics can be used in performing risk characterization. Risk characterization combines 

the major components of risk (hazards, consequences, frequency, and probability), along with 

quantitative estimates of risk, to give a combined and integrated risk perspective (i.e., a risk 

profile). Additionally, it shows the key assumptions and rationale, expert elicitation, 

uncertainties associated with the analysis, and sensitivity analysis. 

Risk Monitor 

A plant-specific 

analysis tool used 

to determine the 

risk in real-time 

based on the 

current plant 

configuration. (see 

Living PRA)

The model the risk monitor uses is based on, and is consistent with, the living PRA for the 

facility. At any given time, the risk monitor reflects the current plant configuration in terms of 

the known status of the various systems or components (e.g., if any components are out of 

service for maintenance or tests). The risk monitor assists plant personnel in making decisions 

about plant configuration changes. 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

Risk Profile (Plant) 

The major results 

generated by a 

PRA that 

characterize plant 

risk.

A plant risk profile presents a concise synopsis of the major PRA results. This synopsis may 

consist of numerous characterizations of risk, including: 

• Core damage frequency and large early release frequency for internally and 

externally initiated events during various modes of operation.

• Percentage contributions to core damage frequency and large early release 

frequency by initiating event and accident sequence type.

• Ranking of the contribution of individual basic events and cutsets to core damage 

frequency and large early release frequency, based on various importance measures. 

• Comparison of PRA results to PRAs for other plants.

• Qualitative risk insights on plant design features.

Risk Reduction Worth 

(see Importance 

Measure)

The term risk reduction worth is one type of importance measure and is defined under 

“Importance Measure.” 

Risk Significant 

A level of risk 

associated with a 

facility’s system, 

structure, 

component, human 

action or modeled 

accident sequence 

that exceeds a 

predetermined level.

(see Safety 

Significant, 

Significant)

A principal focus of a PRA is to determine the risk significance of the various ‘features,’ i.e., the 

systems, structures, and components (SSCs), human actions or the accident sequences 

involving those SSCs, of the facility being analyzed. Usually, an item is considered risk 

significant when the risk associated with it exceeds a predetermined limit for contributing to the 

risk associated with the facility. Since the overall risk of a nuclear facility can be calculated in 

terms of core damage frequency (CDF) (Level 1 PRA), or releases (Level 2 PRA), or health 

effects (Level 3 PRA), risk significance can also be determined as related to these various risk 

measures. Note that risk significant does not have the same meaning as safety significant 

(defined elsewhere in this glossary) and safety significance is not evaluated in a PRA. 

The term also describes a level of risk exceeding a predetermined ‘significance’ level. (Ref. 36)

Risk Significant Equipment 

(see Significant) The term risk significant equipment is related to the term significant and is defined under 

“Significant.” 

Risk to Average Individual 

A measure of the 

risk to an individual 

that represents an 

average over the 

parameters 

characterizing the 

population at risk 

(see Fatality, 

Quantitative Health 

Objectives) 

In a Level 3 PRA, the risk to an average individual is calculated as the total fatalities in the 

surrounding population as a result of an accident divided by the total population. For example, 

the risk of prompt fatality to an average individual within 1 mile of the plant boundary can be 

calculated as the number of prompt fatalities per year to the total population within 1 mile of the 

plant boundary because of each accident sequence, summed over all accident sequences 

weighted by their frequency of concurrence, divided by the population within 1 mile. The 

average individual in the vicinity of the plant is defined as the average individual biologically (in 

terms of age and other risk factors) and who resides within 1 mile of the plant site boundary. 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

Risk-Based Approach 

(see RiskBased)

The term risk-based approach is related to the term risk-based and is defined under 

“Risk-Based.” 

Risk-Based (Approach, Decisionmaking, Regulation)

A characteristic of 

decisionmaking in 

which a decision is 

solely based on 

the results of a risk 

assessment. (see 

Risk-Informed)

The modifying term “risk-based” is applied to decisionmaking and regulation activities that rely 

solely on the use of risk information from PRA results. The terms risk-based approach, riskbased decisionmaking, and risk-based regulation are often used interchangeably and 

somewhat correctly to describe the same concept; therefore, these terms are grouped under 

the same definition. However, as indicated below, each of these terms has its own distinct 

meaning: 

• Risk-Based Approach: A philosophy on decisionmaking “in which a safety decision is 

solely based on the numerical results of a risk assessment.” (Ref. 96)

• Risk-Based Decisionmaking: “An approach to regulatory decisionmaking that 

considers only the results of a probabilistic risk assessment.” (Ref. 36)

• Risk-Based Regulation: An approach to regulation that uses the results of a risk 

assessment to develop applicable regulations. 

Risk-informed is a term that is often used incorrectly in place of risk-based. These terms are 

not synonyms. Unlike a risk-based approach, a risk-informed approach to decisionmaking or 

regulation combines risk information with other factors (e.g., engineering design features) to 

arrive at a decision or develop regulations. 

Since risk-based approaches, decisionmaking, and regulation put a greater emphasis on risk 

assessment results than is currently practical because of uncertainties in PRA, such as 

completeness, the Commission does not endorse a solely “risk-based” approach. 

Risk-Based Decisionmaking

 (see Risk-Based) The term risk-based decisionmaking is related to the term risk-based and is defined under 

“Risk-Based.”

Risk-Based Regulation

(see Risk-Based) The term risk-based regulation is related to the term risk-based and is defined under 

“Risk-Based.” 

Risk-Informed (Approach, Decisionmaking, Regulation) 

A characteristic of 

decisionmaking in 

which risk results 

or insights are 

used together with 

other factors to 

support a decision. 

(see Risk-Based, 

Deterministic, 

Probabilistic)

The modifying term “risk-informed” is applied to decisionmaking and regulation activities that 

combine risk information (e.g., PRA results) with other factors (e.g., engineering design 

features) to arrive at a decision. The terms risk-informed approach, risk-informed 

decisionmaking, and risk-informed regulation are often used interchangeably and somewhat 

correctly to describe the same concept; therefore, these terms are grouped under the same 

definition. However, as indicated below, each of these terms has its own distinct meaning: 

• Risk-Informed Approach: “A ‘risk-informed’ approach to regulatory decision-making 

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. A ‘risk-informed’ approach enhances the traditional approach by: (a) allowing 

explicit consideration of a broader set of potential challenges to safety, (b) providing a 

logical means for prioritizing these challenges based on risk significance, operating 

experience, and/or engineering judgment, (c) facilitating consideration of a broader 
4. GLOSSARY

 4-85 

TERM AND 

DEFINITION DISCUSSION 

set of resources to defend against these challenges, (d) explicitly identifying and 

quantifying sources of uncertainty in the analysis, and (e) leading to better decisionmaking by providing a means to test the sensitivity of the results to key assumptions. 

Where appropriate, a risk-informed regulatory approach can also be used to reduce 

unnecessary conservatism in deterministic approaches, or can be used to identify 

areas with insufficient conservatism and provide the bases for additional 

requirements or regulatory actions.” (Ref. 96) 

• Risk-Informed Decisionmaking: “An approach to regulatory decision making, in which 

insights from probabilistic risk assessment are considered with other engineering 

insights.” (Ref. 36) 

• Risk-Informed Regulation: “An approach to regulation taken by the NRC, which 

incorporates an assessment of safety significance or relative risk. This approach 

ensures that the regulatory burden imposed by an individual regulation or process is 

appropriate to its importance in protecting the health and safety of the public and the 

environment.” (Ref. 36)

A term often used incorrectly in place of risk-informed is risk-based; these terms are not 

synonyms. A risk-based approach to decisionmaking or regulation means that the decision or 

regulation is based only on risk information (e.g., risk results obtained from a PRA), whereas a 

risk-informed approach combines risk information with other factors to arrive at a decision or 

develop regulations. 

Risk-Informed Approach 

(see RiskInformed) 

The term risk-informed approach is related to the term risk-informed and is defined under “RiskInformed.” 

Risk-Informed Decisionmaking

(see RiskInformed) 

The term risk-informed decisionmaking is related to the term risk-informed and is defined under 

“Risk-Informed.” 

Risk-Informed Regulation

(see RiskInformed) 

The term risk-informed regulation is related to the term risk-informed and is defined under 

“Risk-Informed.” 

Safe-Shutdown Earthquake 

The maximum 

earthquake for 

which certain 

structures, 

systems, and 

components are 

designed to remain 

functional to shut 

down the reactor. 

(see Seismic 

Margin Analysis)

In a seismic PRA, the plant’s response to earthquakes of all magnitudes appropriate for the site 

are evaluated. In a seismic margin analysis, the capability of the plant to withstand an 

earthquake larger that the safe-shutdown earthquake (SSE) is often assessed. The 

ASME/ANS PRA Standard (Ref. 2) defines the SSE as “that earthquake for which certain 

structures, systems and components (SSCs) are designed to remain functional. In the past, 

the SSE has been commonly characterized by a standardized spectral shape anchored to 

a peak ground acceleration value.” 

Appendix S to 10 CFR 50 (Ref.25) states that the “safe-shutdown earthquake ground motion 

(SSE) is the vibratory ground motion for which certain structures, systems, and components 

must be designed to remain functional.” The SSCs required to withstand the effects of the 

safe-shutdown earthquake ground motion are those necessary to ensure: 

(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
4. GLOSSARY

 4-86 

TERM AND 

DEFINITION DISCUSSION 

(3) The capability to prevent or mitigate the consequences of accidents that could result 

in potential offsite exposures comparable to the guideline exposures of 10 CFR 

50.34(a)(1) (Ref. 14).

The definition provided is based on the definition in the NRC Web site Glossary (Ref. 36). 

Safe Stable State 

Condition of the 

reactor in which 

the necessary 

safety functions 

are achieved.

In a PRA, safe stable states are represented by success paths in modeling of accident 

sequences. A safe stable state implies that the plant conditions are controllable within the 

success criteria for maintenance of safety functions. 

The ASME/ANS PRA Standard (Ref. 2) defines the term safe stable state as “a plant condition, 

following an initiating event, in which reactor coolant system conditions are controllable at or 

near desired values.” 

Safety Function 

Those functions 

needed to shut 

down the reactor, 

remove the 

residual heat, and 

contain any 

radioactive 

material release. 

A PRA involves the analysis of the performance of the plant safety functions in response to 

accidents. The common general safety functions for a nuclear power plant as stated in the 

IAEA Safety Glossary (Ref. 7) are: 

• The capability to safely shut down the reactor and maintain it in a safe shutdown 

condition during and after appropriate operational states and accident conditions. 

• The capability to remove residual heat from the reactor core after shutdown, and 

during and after appropriate operational states and accident conditions. 

• The capability to reduce the potential for the release of radioactive material and to 

ensure that any releases are within prescribed limits during and after operational 

states and within acceptable limits during and after design-basis accidents. 

The ASME/ANS PRA Standard (Ref. 2) defines safety function as “function that must be 

performed to control the sources of energy in the plant and radiation hazards.” 

Safety Margin 

The extra capacity 

factored into the 

design of a 

structure, system, 

or component so 

that it can cope 

with conditions 

beyond the 

expected to 

compensate for 

uncertainty. (see 

Defense-in-Depth, 

Uncertainty)

In a PRA, the extra capacity of systems, structures, and components (SSC) provided by the 

safety margin is used in calculating the plant response to an accident. A safety margin is used 

to provide capacity for emergency situations, unexpected loads, misuse, or attrition. 

Many engineering codes and standards provide quantitative guidance on appropriate safety 

margin for a particular design application. However, the term safety margin also is often found 

in regulatory documents that contain phrases such as “maintain adequate safety margin,” or 

“provide sufficient safety margin,” without specification of a particular quantitative margin. 

Safety margins can be considered a part of, or complementary to, defense-in-depth in that they 

provide extra (redundant) capacity. Incorporation of safety margins is one of the ways 

designers deal with the uncertainty of the challenges that the designed SSCs face. 

The figure below illustrates several concepts on safety margins. A regulator may impose the 

requirement that a margin is maintained between a component’s allowable limit of operation, 

the regulatory limit, and the component’s ultimate capacity. The component designer may 

want to design or select the component so that during normal operation it operates below, 

rather than right at, the regulatory limit (i.e., he or she may want to add an additional margin). 

The total safety margin then encompasses both the designer and regulatory margins. 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

Safety-Related 

(see Safety 

Significant)

The term safety-related indicates the safety significance of a structure, system, or component 

and is defined under “Safety Significant.” 

Safety Significant (Important to Safety, Safety-Related, Nonsafety-Related) 

A qualifying term 

that indicates if 

something does 

not meet some 

predetermined 

criterion, it has the 

potential to affect 

safety. 

In a PRA, the risk significance of nuclear power plant structures, systems, and components 

(SSCs) are determined, not the safety significance. This risk significance is then used in a riskinformed regulatory framework to determine the safety significance of SSCs. The term safety 

significant is generally used to categorize nuclear power plant SSCs using the process outlined 

in 10 CFR 50.69 (Ref. 21). In this application, a plant-specific PRA is used to delineate and 

quantify severe accident scenarios resulting from internal initiating events at full-power 

operation. In 10 CFR 50.36, Technical Specifications, (Ref. 15) Criterion 4 requires that “a 

structure, system, or component which operating experience or probabilistic risk assessment 

has shown to be significant to public health and safety” must have a technical specification 

limiting condition for operation established for it. 

The term important to safety refers to both safety related and non-safety related SSCs that 

have been deemed important. In Regulatory Guide 1.201 (Ref. 92), the NRC has stated that it 

does not endorse the Nuclear Energy Institute (NEI) usage of important to safety as having the 

same connotation as safety significant. 

Another term, safety related, has a specific meaning in the regulatory arena. Part 50 of the 

Code of Federal Regulations (Ref. 13), as well as the NRC Web site Glossary (Ref. 36) state 

that the term “safety-related applies to systems, structures, components, procedures, and 

controls (of a facility or process) that are relied upon to remain functional during and following 

design basis events. Their functionality ensures that key regulatory criteria, such as levels of 

radioactivity released, are met. Examples of safety related functions include shutting down the 

nuclear reactor and maintaining it in a safe-shutdown condition.” Conversely, nonsafety-related 

indicates that the SSCs, procedures, and controls are not relied upon to remain functional 

during a design-basis event. 

The NRC Web site Glossary (Ref. 36) makes the following statement about the term safety 

significant: “When used to qualify an object, such as a system, structure, component, or 

accident sequence, this term identifies that object as having an impact on safety, whether 

determined through risk analysis or other means, which exceeds a predetermined significance 

criterion.” Safety significance is not evaluated in a PRA.

Screening (Analysis, Criteria, Qualitative, Quantitative) 

A process that 

distinguishes items 

that should be 

included or 

excluded from an 

analysis based on 

In a PRA, screening may be applied in a variety of ways (e.g., screening out (eliminating) 

component failure events from the PRA based on a low probability or frequency). Another form 

of screening is to identify the more significant events that should be analyzed in a detailed 

manner. Insignificant events may be addressed using less detailed and usually conservative 

methods. Screening is an integral step in most PRAs to reduce the complexity of the PRA 

model using sound judgment. The terms screening and screening analysis are similar in 

Safety Margin

Ultimate

Capacity

Design

Analysis

Increasing value of safety variable 

(e.g. Containment pressure)
4. GLOSSARY

 4-88 

TERM AND 

DEFINITION DISCUSSION 

defined criteria. meaning and often used interchangeably. 

The definitions of the grouped terms are presented below as they apply to screening: 

• Screening criteria: “The values and conditions used to determine whether an item is 

a negligible contributor to the probability of an accident sequence or its 

consequences.” (Ref. 2)

• Qualitative screening: The objective is to identify portions of the analysis whose 

potential risk contribution can be judged negligible without quantitative analysis. 

• Quantitative screening: The objective is to eliminate portions of the analysis from 

further consideration based on preliminary estimates of risk contribution through the 

use of established quantitative screening criteria. 

The ASME/ANS PRA Standard (Ref. 2) defines screening as “a process that eliminates items 

from further consideration based on their negligible contribution to the probability of an accident 

or its consequences.” 

Screening Analysis 

(see Screening) The term screening analysis is similar in meaning to screening and is discussed under 

“Screening.”

Screening Criteria 

 (see Screening) The term screening criteria is defined under “Screening.”

Seismic Fragility Analysis 

(see Fragility 

Analysis) 

Seismic fragility analysis is a type of fragility analysis and is included in the discussion under 

“Fragility Analysis.” 

Seismic Hazard Analysis 

(see Hazard 

Analysis)

The term seismic hazard analysis is a type of hazard analysis and is defined under “Hazard 

Analysis.” 

Seismic Margin 

A measure of the 

capacity of the 

plant to withstand 

an earthquake 

more severe than 

the design-basis 

earthquake. (see 

High Confidence of 

Low Probability of 

Failure, Safe 

Shutdown 

Earthquake, 

Seismic Margin 

Analysis) 

For some applications, seismic margin, rather than a PRA risk metric, has been used to 

estimate the ability of a plant to safely withstand seismic events. The ASME/ANS PRA 

Standard (Ref. 2) states that “seismic margin is expressed in terms of the earthquake motion 

level that compromises plant safety, specifically leading to severe core damage. The margin 

concept also can be extended to any particular structure, function, system, equipment item, or 

component for which ‘compromising safety’ means sufficient loss of safety function to 

contribute to core damage either independently or in combination with other failures.” 

NUREG-1742 (Ref. 59) defines seismic margin as “the ability of a plant, system, component or 

structure to safely withstand seismic demands or input ground-motion levels beyond those 

imposed by the design basis earthquake.” 

Seismic Margin Analysis 

The process to 

estimate the 

seismic margin of 

For some applications, seismic margin analysis is an alternative to a seismic PRA for 

identifying seismic vulnerabilities at a plant. The earthquake specified for assessing the 

seismic margin can depend on a number of factors, usually the plant’s location. In the 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

the plant and to 

identify any 

seismic 

vulnerabilities in 

the plant. (see 

High Confidence of 

Low Probability of 

Failure, Seismic 

Margin, SafeShutdown 

Earthquake) 

individual plant examination for external events (IPEEE), plants were assessed against a 

review-level earthquake whose intensity was higher than the design-basis earthquake and 

varied according to the plant location. 

Seismic margin analysis is performed to show high confidence of low probability of failure 

(HCLPF) at a certain earthquake level (peak ground acceleration) above the design-basis 

(safe-shutdown) earthquake. 

A number of methods can be used to calculate seismic margin: 

• In the IPEEEs, most licensees that carried out a seismic margin analysis used a 

method developed by the Electric Power Research Institute (EPRI). In the EPRI 

method, two success paths, addressing transients, are developed based on a group 

of safety functions capable of bringing the plant to a safe-shutdown condition after an 

earthquake. Each success path has to rely on different equipment and each path 

assumes a loss of offsite power. One path also has to be capable of mitigating a 

small loss-of-coolant accident (LOCA). HCLPFs are developed for the two success 

paths. 

The NRC also developed a seismic margin method for the IPEEEs, used by a few licensees. 

In the NRC IPEEE method, accident sequence models are developed for transients and small 

LOCAs and HCLPF values are evaluated for the accident sequences developed from these two 

initiators. Neither the EPRI nor the NRC method requires fragility curves to be developed and 

allow HCLPFs to be based on the conservative deterministic failure margin method. 

 

• More recently, the NRC has endorsed a seismic margin method in which fragility 

curves are developed. In this PRA-based method, accident sequence models are 

developed for all the initiators and HCLPF values are evaluated for the accident 

sequences developed from all the initiators.

The definition provided is based on the definition in the ASME/ANS PRA Standard (Ref. 2). 

Seismic Plant Response Analysis/Model

(see Plant 

Response 

Analysis/Model)

The term seismic plant response analysis is a type of plant response analysis and is included in 

the discussion under “Plant Response Analysis/Model.”

Sensitivity Analysis

An analysis in 

which one or more 

input parameters 

to a model are 

varied in order to 

observe their 

effects on the 

model results. 

In a PRA, sensitivity analyses often are performed to help assess the results. Sensitivity 

analyses often involve variations of quantitative parameters (e.g., component failure 

probabilities, initiating event frequencies, human error rates). 

The definition provided was based on the definition in NUREG-1560 (Ref. 56). 

Severe Accident (Sequence, Progression Sequence) 

A type of accident 

that involves core 

damage. (see 

Accident 

Sequence, 

Beyond-DesignBasis Accident, 

Design-Basis 

Accident)

In a PRA, beyond-design-basis accidents (BDBAs) are analyzed to determine which ones 

could lead to core damage. The BDBAs that have an end state resulting in core damage are 

termed severe accidents. All severe accidents are by definition beyond-design-basis accidents 

since their challenges exceed the design envelope of the plant. However, not all beyonddesign-basis accidents are severe accidents, since the design envelope can be exceeded 

without core damage occurring. 

The ASME/ANS PRA Standard (Ref. 2) defines a severe accident as “an accident that involves 

extensive core damage and fission product release into the reactor vessel and containment, 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

with potential release to the environment.”

In a Level 1 PRA, severe accident sequences are a subset of the accident sequences 

(i.e., many of the accident sequences in a Level 1 PRA do not result in core damage). In a 

Level 2 PRA, severe accident sequences are the only sequences considered because they 

involve core damage. The term severe accident progression sequence usually is used 

correctly as a synonym for the term severe accident sequence.

Severe Accident Progression Sequence 

(see Severe 

Accident) 

Severe accident progression sequence has the same meaning as severe accident sequence 

and is defined under “Severe Accident.”

Severe Accident Sequence 

(see Severe 

Accident) 

A severe accident sequence is an accident sequence that results in a severe accident and is 

defined under “Severe Accident.” 

Shutdown 

(see Low-Power 

and Shutdown)

The term shutdown is part of low power and shutdown operation and is defined under “LowPower and Shutdown.” 

Significant (Accident Sequence, Accident Progression Sequence, Basic Event, 

Containment Challenge, Contributor, Cutset, Equipment) 

A factor that can 

have a major or 

notable influence 

on the results of a 

risk analysis. 

In a PRA, the modifying term significant is applied to factors that have an important influence 

on causing a measurement of risk to exceed a predetermined level or limit. The terms 

significant and risk significant have the same meaning in a PRA context and are often used 

interchangeably, which is correct and appropriate in this context. 

As discussed in NRC Regulatory Guide 1.200 (Ref. 91), the determination of significance is 

a function of how the PRA is being, or is intended to be, used. When a PRA is being used to 

support an application, the significance of an accident sequence or contributor is measured 

with respect to whether its consideration has an effect on the decision being made. 

Quantitative thresholds (criteria) often are used to determine if a basic event, cutset, accident 

sequence, or accident progression sequence is considered significant from a risk perspective 

(e.g., based on importance measures, percentage contribution). The previously mentioned 

items (e.g., basic event, cutset) represent the different types of significant risk contributors that 

could influence the results of a risk analysis. These quantitative criteria may vary, depending 

on the source of the guidance. The following terms (excluding risk significant) and the 

subsequent definitions are based on the ASME/ANS PRA Standard (Ref. 2): 

• Significant Accident Sequence: “One of the sets of accident sequences resulting 

from the analysis of a specific hazard group, defined at the functional or systematic 

level, which, when rank-ordered by decreasing frequency, sum to a specified 

percentage of the core damage frequency for that hazard group, or that individually 

contribute more than a specified percentage of core damage frequency. For this 

version of the Standard [RA-Sa-2009], the summed percentage is 95% and the 

individual percentage is 1% of the applicable hazard.” (Ref. 2)

• Significant Accident Progression Sequence: “One of the sets of accident sequences 

contributing to large early release frequency resulting from the analysis of a specific 

hazard group that, when rank-ordered by decreasing frequency, sum to a specified 

percentage of the large early release frequency, or that individually contribute more 

than a specified percentage of large early release frequency for that hazard group. 

For this version of the Standard [RA-Sa-2009], the summed percentage is 95% and 

the individual percentage is 1% of the applicable hazard.” (Ref. 2) 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

• Significant Basic Event: “A basic event that contributes significantly to the computed 

risks for a specific hazard group. For internal events, this includes any basic event 

that has an FV importance greater than 0.005 or a RAW importance greater than 2.” 

(Ref. 2)

• Significant Containment Challenge: “A containment challenge that results in a 

containment failure mode that is represented in a significant accident progression 

sequence.” (Ref. 2)

• Significant Cutset: “One of the sets of cutsets resulting from the analysis of a specific 

hazard group that, when rank-ordered by decreasing frequency, sum to a specified 

percentage of the core damage frequency (or large early release frequency) for that 

hazard group, or that individually contribute more than a specified percentage of core 

damage frequency (or large early release frequency). For this version of the 

Standard [RA-Sa-2009], the summed percentage is 95% and the individual 

percentage is 1% of the applicable hazard.” (Ref. 2)

• Risk Significant Equipment: “Equipment associated with a significant basic event.” 

(Ref. 2)

A significant contributor can refer to an important factor associated with a significant accident 

sequence, such as a particular accident sequence cutset, a significant basic event, or an 

initiating event. As stated in the ASME/ANS PRA Standard (Ref. 2), a significant contributor 

also can be “an essential characteristic (e.g., containment failure mode, physical phenomena) 

of a significant accident progression sequence, and if not modeled would lead to the omission 

of the sequence.” 

Significant Accident Progression Sequence 

 (see Significant) The term significant accident progression sequence is related to the term significant and is 

defined under “Significant.” 

Significant Accident Sequence 

 (see Significant) The term significant accident sequence is related to the term significant and is defined under 

“Significant.” 

Significant Basic Event 

(see Significant) The term significant basic event is related to the term significant and is defined under 

“Significant.” 

Significant Containment Challenge 

(see Significant) The term significant containment challenge is related to the term significant and is defined 

under “Significant.” 

Significant Contributor 

(see Significant) The term significant contributor is related to the term significant and is defined under 

“Significant.”

Significant Cutset 

(see Significant) The term significant cutset is related to the term significant and is defined under “Significant.” 

Skin Deposition

Exposure resulting In a Level 3 PRA, for the consequence calculation skin deposition is one of the assumed 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

from radioactive 

material deposited 

directly onto the 

surface of the body.

(see Exposure 

Pathways, 

Exposure, 

Exposure Time, 

Cloudshine, Water 

Immersion, 

Groundshine, 

Inhalation, 

Ingestion, Health 

Effects) 

pathways by which an individual can receive doses. The pathways of exposure include: (1) 

direct external exposure from radioactive material in a plume, principally due to gamma 

radiation (air immersion or cloudshine), (2) direct exposure from radioactive material in 

contaminated water given to an individual immersed in the water, (3) exposure from inhalation 

of radioactive materials in the cloud and resuspended material deposited on the ground, (4) 

exposure to radioactive material deposited on the ground (groundshine), (5) radioactive 

material deposited onto the body surfaces (skin deposition), and (6) ingestion from deposited 

radioactive materials that make their way into the food and water pathway. 

Small Early Release 

(see Radioactive 

Material 

Release) 

The term small early release is a type of radioactive material release and is defined in the 

discussion under “Radioactive Material Release.” 

Small Early Release Frequency 

(see Frequency) The term small early release frequency is a type of frequency used in PRA calculation and is 

defined in the discussion under “Frequency.” 

Small Early Release Frequency Analysis 

(see Radioactive 

Material Release 

Frequency 

Analysis) 

The term small early release frequency analysis is a type of radioactive material release 

frequency analysis and is defined under “Radioactive Material Release Frequency Analysis.” 

Small Late Release 

(see Radioactive 

Material Release)

The term small late release is a type of radioactive material release and is defined in the 

discussion under “Radioactive Material Release.” 

Small Late Release Frequency

 (see Frequency) The term small late release frequency is a type of frequency used in PRA calculation and is 

defined in the discussion under “Frequency.”

Small Late Release Frequency Analysis 

(see Radioactive 

Material Release 

Frequency 

Analysis) 

The term large late release frequency analysis is a type of radioactive material release 

frequency analysis and is defined under “Radioactive Material Release Frequency Analysis.” 

Source of Risk 

A substance that 

can pose danger 

or threat to public 

health. (see 

Hazard, Initiating 

Event)

In a PRA, sources of risk at nuclear power plants include, for example, the nuclear fuel 

contained within the reactor core and the spent fuel pool. These sources of risk could be 

affected by hazards which directly or indirectly cause initiating events and may further cause 

safety system failures or operator errors leading to core damage or radioactive material 

release. For instance, in a non-nuclear application, a leak in a pool may not cause a negative 

consequence other than having an empty pool. However, because the pool at a nuclear power 
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TERM AND 

DEFINITION DISCUSSION 

plant contains nuclear fuel, there could be a negative consequence if that pool drained and 

radioactive material (the source of risk) was released. 

The terms source of risk and hazard are sometimes incorrectly used as synonyms. A hazard is 

anything that has the potential to cause an undesired event. Intrinsically, a source of risk does 

not cause an event, but a hazard can cause an initiating event leading to core damage. For 

example, an earthquake (hazard) with particular frequency could cause a loss-of-coolant 

accident (initiating event) which may result in core damage of the nuclear fuel (source of risk). 

Source Term 

Types and 

amounts of 

radioactive or 

hazardous material 

released to the 

environment 

following an 

accident. (see 

Release Category, 

Mechanistic 

Source Term, 

Chemical Element 

Group, Release 

Fraction, Release 

Timing and 

Duration, Source 

Term Analysis)

In a Level 2 PRA, the source term is one of the end products of the analysis and involves the 

characterization of the release from containment to the environment. 

This characterization involves a description of the radionuclide release at a particular location, 

including the physical and chemical properties of released material, release magnitude, heat 

content (or energy) of the carrier fluid, location relative to local obstacles that would affect 

transport away from the release point, and the temporal variations in these parameters (e.g., 

time of release duration). 

The information used to define a source term can vary, depending on the objective and 

intended application of the PRA. For instance, if the Level 2 PRA results will be used in a Level 

3 consequence assessment, it may be necessary to provide more detailed source term 

information than if no Level 3 assessment will be performed. For a Level 3 assessment, the 

source term information needs to be sufficient to estimate offsite radiation doses and, in some 

cases, other radiological consequences such as land contamination. 

Source Term Analysis 

An analysis to 

determine the 

characteristics of 

the radioactive 

material released 

to the environment 

following an 

accident. (see 

Source Term)

In a Level 2 PRA, the source term analysis determines the release of radioactive material from 

the fuel or core debris and the transport of this material through the primary system and 

containment to the environment. (The scope of the PRA source term analysis usually does not 

include releases from the spent fuel pool.) 

NUREG-1489 (Ref. 54) states that there are three parts to a source term analysis: (1) the 

estimation of the release of radioactive material from the fuel and core debris, (2) the transport 

of this material through the primary system and the containment, and (3) the characterization of 

the release from containment to the environment. 

Split Fraction 

The likelihood that 

one specific 

outcome from a 

set of possible 

outcomes will be 

observed. (see 

Event Tree, 

Probability)

A split fraction is a unitless parameter (i.e., probability). This term typically is used with regard 

to the quantification of an event tree of a PRA model. It represents the fraction with which each 

possible outcome, or branch, of a particular top event in an event tree may be expected to 

occur. Split fractions are, in general, conditional on prior events. At any event tree branch 

point, the sum of all the split fractions representing the possible outcomes should be unity. 

The ASME/ANS PRA Standard (Ref. 2) defines the term split fraction as “a unitless quantity 

that represents the conditional (on preceding events) probability of choosing one direction 

rather than the other through a branch point of an event tree.”

State-of-Knowledge Correlation 

A type of 

dependency that 

arises when the 

same data is used 

to quantify the 

In a PRA, when the basic event mean values and uncertainty distributions are propagated 

without accounting for the state-of-knowledge correlation (SOKC), the calculated mean value of 

the relevant risk metric and the uncertainty about this mean value will be underestimated. 

When the same data is used to quantify the individual probabilities of two or more basic events, 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

individual 

probabilities of two 

or more basic 

events. (see 

Uncertainty)

the uncertainty associated with such basic event probabilities must be correlated to correctly 

propagate the parameter uncertainty through the risk calculation. The SOKC arises because, 

for identical or similar components, the state-of-knowledge about their failure parameters is the 

same. In other words, the data used to obtain mean values and uncertainties of the 

parameters in the basic event models of these components may come from a common source 

and, therefore, are not independent, but are correlated. 

The ASME/ANS PRA Standard (Ref. 2) defines the term SOKC as “the correlation that arises 

between sample values when performing uncertainty analysis for cut sets consisting of basic 

events using a sampling approach (such as the Monte Carlo method); when taken into account, 

this results, for each sample, in the same value being used for all basic event probabilities to 

which the same data applies.” 

State-of-Knowledge Uncertainty 

(see 

Uncertainty) 

The term state-of-knowledge uncertainty is related to epistemic uncertainty and defined under 

“Uncertainty.” 

Station Blackout 

The complete loss 

of alternating 

current electric 

power in a nuclear 

plant. (see 

Transient)

In a PRA, station blackout (SBO) accidents are analyzed because alternating current (AC) 

power is an important support system for numerous plant systems and components. A plant 

subjected to an SBO condition must achieve safe-shutdown by relying on mitigating systems 

and components that do not require AC power (e.g., steam-driven pumps and battery-powered 

valves and instrumentation). However, for operating plants, core cooling may not be 

indefinitely maintained without AC power. Important factors that influence the risk associated 

with SBO include the potential for recovery of AC power, battery depletion times, and the 

reliability of the mitigating systems and components that do not require AC power. 

10 CFR 50.2 (Ref. 13) defines the term station blackout as “the complete loss of alternating 

current (AC) electric power to the essential and nonessential switchgear buses in a nuclear 

power plant (i.e., loss of offsite electric power system concurrent with turbine trip and 

unavailability of the onsite emergency ac power system). SBO does not include the loss of 

available AC power to buses fed by station batteries through inverters or by alternate AC 

sources, nor does it assume a concurrent single failure or design basis accident.” 

The ASME/ANS PRA Standard (Ref. 2) defines the term SBO as “complete loss of AC electric 

power to the essential and nonessential switchgear buses in a nuclear power plant.” 

Steam Generator Tube Rupture 

A break or breach 

of a steam 

generator tube. 

(see 

Consequential 

(Induced) Steam 

Generator Tube 

Rupture) 

In a PRA for a pressurized-water reactor, steam generator tube ruptures (SGTRs) are modeled 

either as an initiating event or a subsequent failure as part of an accident sequence. If the 

SGTR occurs randomly while the plant is operating, it is an initiating event modeled in the PRA. 

However, if the SGTR occurs because of excessive conditions produced as a result of an 

accident, it is considered to be a consequential or induced SGTR. 

An SGTR allows reactor coolant to flow from the reactor vessel to the secondary side of the 

steam generator. As such, it can become a significant contributor to risk because an SGTR 

can serve as a possible mechanism for radioactive material transport to the environment 

because it can be a containment bypass mechanism. There is the potential that if a tube bursts 

or leaks while a plant is operating, radioactivity from the primary coolant system could escape 

directly to the atmosphere through the safety valves on the secondary side. 

Stochastic Uncertainty 

(see 

Uncertainty) 

The term stochastic uncertainty is related to aleatory uncertainty and defined under 

“Uncertainty.” 
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TERM AND 

DEFINITION DISCUSSION 

Structuralist 

An approach to 

defense-in-depth 

that relies on 

multiple strategies 

in the design and 

operation of a 

facility to 

compensate for 

both known and 

unknown 

uncertainties. (see 

Rationalist, 

Deterministic, 

Defense-in-Depth)

A PRA is not used in the structuralist approach to defense-in-depth, unlike the rationalist 

approach. Instead, the structuralist approach asserts that safety margins associated with 

defense-in-depth are embodied within the regulations and in the design of a facility built to 

comply with those regulations. 

The fundamental principle of the structuralist approach is that if a system is designed to 

withstand all the worst-case credible accidents, then it is by definition protected against any 

credible accident. It is a method that is solely based on deterministic analyses and principles to 

establish how precautions can be placed into a system, just in case an existing barrier or 

protective system fails. By comparison, a rationalist approach uses PRA methods to quantify 

and reduce system uncertainties, as opposed to relying on potentially overly conservative 

safety margins. 

Success Criteria 

The minimum 

combination of 

systems and 

components 

needed to carry 

out the safety 

functions given an 

initiating event.

In a PRA, success criteria are used at different places or levels in the analysis. At a high level, 

the success criteria define the safety functions that must be performed following an initiating 

event. Success criteria are then defined for each safety function, which are expressed in terms 

of requirements for the systems needed to support that function. Success criteria also are 

developed for the components within these systems. The success criteria specify how the 

systems and components must function, when they must begin to function, and how long they 

must function. Success criteria for PRA studies typically are developed through the use of 

deterministic analyses that represent the design and operation of the plant being evaluated. 

 

Success criteria may be defined in a number of ways, including the following: 

• In terms of the equipment required (e.g., one out of two service water pumps).

• In terms of equipment performance (e.g., at least 50 percent of the maximum system 

flow rate).

• In terms of the timing (e.g., system must be initiated within 30 minutes and operate 

for 24 hours).

The ASME/ANS PRA Standard (Ref. 2) defines the term success criteria as “criteria for 

establishing the minimum number or combinations of systems or components required to 

operate, or minimum levels of performance per component during a specific period of time, to 

ensure that the safety functions are satisfied.” 

Success Path 

A sequence of 

events (responding 

to an upset 

condition) that 

result in a 

successful state of 

a system, the 

reactor, or the 

containment. (see 

Event Tree, Safe 

Stable State)

In a PRA, the term success path often is used in the context of describing an event tree path 

that leads to a safe stable state of the reactor. Alternatively, a fault tree model can be 

transformed into its logical complement, a success tree that shows the specific ways (success 

paths) in which an undesired event (e.g., system failure) can be prevented from occurring. 

A successful state of a system occurs when the system is able to perform its intended function 

(e.g., provide injection water at a sufficient flow rate and pressure). A successful state of a 

reactor is achieved if adequate core cooling is maintained throughout the sequence of events 

following an upset condition. For the containment, a successful state is achieved if the 

containment pressure boundary remains intact throughout the sequence of events following an 

upset condition. 

The ASME/ANS PRA Standard (Ref. 2) defines a success path as “a set of systems and 

associated components that can be used to bring the plant to a stable hot or cold condition and 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

maintain this condition for at least 72 hrs.”

Supplementary Analysis 

Any evaluation that 

is performed to 

support another 

study or evaluation.

In a PRA context, the term supplementary analysis often is used to denote an evaluation made 

to facilitate the development or review of a PRA consistent with the ASME/ANS PRA Standard 

(Ref. 2). An example of a supplementary analysis would be an evaluation of plant-specific 

component failure data to support derivation of plant-specific component failure rates for use in 

a PRA. 

Sometimes the supplementary analysis is performed instead of following the specific 

requirements in the ASME/ANS PRA Standard. In this situation, the supplementary analysis is 

performed to meet the Standard’s intent, but it is outside the scope of the Standard. Therefore, 

performing a supplementary analysis does not meet all the Standard’s criteria. 

Support System 

A system that 

enables the 

operation of one or 

more systems. 

(see Front-Line 

System, Support 

System Initiating 

Event)

In a PRA, support system failures are evaluated to determine the effect of these failures on the 

operability of other plant systems and components. Often one support system, such as 

component cooling water, provides functionality to multiple systems or components, and 

therefore, needs to be considered in PRA modeling to assess what happens if that capability is 

lost to multiple systems. 

Examples of support systems include electrical power, cooling water, instrument air, and 

heating, ventilation, and air conditioning. Support systems (e.g., cooling water) can require 

other support systems for operation (e.g., electric power may be needed to operate the cooling 

water pumps). Front-line systems typically require one or more support systems. In some 

instances, a failed support system can lead to an undesired plant condition that requires 

successful mitigation by plant equipment and personnel to prevent core damage from 

occurring. In this situation, the support system failure would be characterized as a support 

system initiating event. 

The ASME/ANS PRA Standard (Ref. 2) defines the term support system as “a system that 

provides a support function (e.g., electric power, control power, or cooling) for one or more 

other systems.”

Support System Initiating Event 

A support system 

failure that perturbs 

the steady-state 

operation of the 

plant and could 

lead to an 

undesired plant 

condition. (see 

Initiating Event, 

Support System)

In a PRA, the failures of support systems are evaluated to determine if they could potentially 

cause an undesired plant condition (i.e., a manual trip or a reactor shutdown). At the same 

time, this failed support system also may have the potential to disable one or more systems 

that could be used to mitigate the undesired plant condition. 

An example of a support system initiating event would be the loss of the component cooling 

water (CCW) system at a pressurized-water reactor. The failure of this system would, in turn, 

lead to the consequential failure of a number of other important systems that depend on CCW, 

which might include the reactor coolant pumps (RCPs) and emergency core cooling system 

(ECCS) equipment. Loss of the RCPs would result in a plant trip, and loss of ECCS 

functionality would reduce the number of plant mitigating systems that could be used to 

maintain core cooling following the plant trip. 

Supporting Requirements 

Requirements that 

support the 

high-level 

requirements in 

defining the 

For a base PRA, NRC Regulatory Guide 1.200 (Ref. 91) defines a set of technical 

characteristics and associated attributes that make it technically acceptable. One approach to 

demonstrate a PRA is acceptable is to use a national consensus PRA standard, supplemented 

to account for the NRC staff’s regulatory positions. The ASME/ANS PRA Standard (Ref. 2) is 

one example of such a national consensus PRA standard. The ASME/ANS PRA Standard 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

minimum needed 

for a technically 

acceptable 

baseline PRA. 

(see High-Level 

Requirements, 

Capability 

Categories)

uses high-level requirements and supporting requirements. 

Regulatory Guide 1.200 (Ref.91) states, “Technical requirements may be defined at two 

different levels: (1) high-level requirements and (2) supporting requirements. High-level 

requirements are defined for each technical element and capture the objective of the technical 

element. These high-level requirements are defined in general terms, need to be met 

regardless of the level of analysis resolution and specificity (capability category), and 

accommodate different approaches. Supporting requirements are defined for each high-level 

requirement. These supporting requirements are those minimal requirements needed to satisfy 

the high-level requirement.” 

To use a PRA for a risk-informed application, it is recognized that not every PRA item will be, or 

needs to be, developed to the same level of detail, same degree of plant-specificity, or the 

same degree of realism. The ASME/ANS PRA Standard (Ref. 2) uses three capability 

categories to distinguish levels of detail, plant specificity, and realism. Furthermore, the 

supporting requirements are developed commensurate with each capability category. 

Therefore, while the high-level requirements are the same across all three capability 

categories, their supporting requirements reflect the differences in levels of detail, plant 

specificity, and realism across the three categories. 

Systems Analysis 

The evaluation of 

the reliability and 

availability of a 

system. (see 

Availability, 

Reliability) 

In a PRA, the term systems analysis can refer to a qualitative or quantitative evaluation of the 

failure modes of an individual system or group of systems (e.g., a fault tree analysis of 

a cooling water system or an electrical distribution system). 

Technical Acceptability, Technical Quality (PRA) 

Refers to a set of 

characteristics and 

related attributes 

that provide the 

minimum qualities 

a base PRA must 

satisfy to be used 

in risk-informed 

decisionmaking. 

(see Technical 

Adequacy)

For a PRA to be technically acceptable, it must satisfy a set of technical characteristics and 

associated attributes. Regulatory Guide (RG) 1.200 (Ref. 91) defines such a set of 

characteristics and accompanying attributes that need to be addressed in a technically 

acceptable base PRA (i.e., independent of the application for which the PRA is used). 

RG 1.200 guidance is for operating reactors and contains cautions for new advanced 

light-water reactors. 

Technical acceptability and technical quality mean the same thing and are used 

interchangeably. 

Technical Adequacy (PRA) 

Refers to the fact 

that the PRA has 

the scope and 

level of detail 

necessary to 

support the 

application for 

which it is being 

used and is also 

technically 

acceptable. (see 

Technical 

Acceptability) 

The scope of a PRA (i.e., risk characterization, level of detail, plant specificity and realism) 

needs to be commensurate with the scope of the specific risk-informed application that it is 

supporting. Some applications (e.g., extension of diesel generator allowed outage time) may 

only use a portion of the base PRA, whereas other applications (e.g., safety significance 

categorization of structures, systems, and components) may require the complete model. 

Regulatory Guide 1.200 (Ref. 91) provides guidance on an acceptable approach for 

demonstrating the technical adequacy of a PRA used to support a regulatory application. 

Central to this approach is the concept that the PRA needs to only have the scope and level of 

detail necessary to support the application for which it is being used, but it always needs to be 

technically acceptable. 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

Technical Elements 

(see PRA 

Technical 

Elements)

The term technical elements has the same meaning as PRA technical elements in the context 

of PRA and is defined under “PRA Technical Elements.” 

Technical Quality

(see Technical 

Acceptability) 

The term technical quality has the same meaning as technical acceptability and is defined the 

same as the term “Technical Acceptability.” 

Top Event (Event Tree Top Event) 

The events across 

the top of an event 

tree needed to 

mitigate an 

accident. (see 

Event Tree, Fault 

Tree) 

The NRC Web site Glossary (Ref. 36) defines top events as “the events across the top of the 

event tree, which graphically represent the systems needed to keep the plant in a safe state 

following an initiating event (i.e., a challenge to plant operation). A top event is the starting 

point of the fault tree, which identifies all of the pathways that lead to a system failure.” The 

fault tree starts with the top event, as defined by the event tree, and identifies what equipment 

and operator actions, if failed, would prevent successful operation of the system. 

The ASME/ANS PRA Standard (Ref. 2) includes two terms: event tree top event and top 

event. Event tree top event is defined as “the conditions (i.e., system behavior or operability, 

human actions, or phenomenological events) that are considered at each branch point in an 

event tree.” Top event is defined as the “undesired state of a system in the fault tree model 

(e.g., the failure of the system to accomplish its function) that is the starting point (at the top) of 

the fault tree.” 

An illustration of a top event is shown under the discussion for the term “Event Tree.” 

Total Effective Dose Equivalent 

(see Dose 

Equivalent)

The total effective dose equivalent is one measure of dose that can be used to calculate the 

effect of radiation received by an individual and is defined under “Dose Equivalent.” 

Transient, General Transient 

An event that 

could require a 

plant trip that might 

challenge safety 

systems but does 

not lead to a loss 

of significant 

quantities of 

reactor coolant. 

(see Initiating 

Event, Station 

Blackout)

In a PRA, two major categories of initiating events are evaluated; namely, transients and 

loss-of-coolant accidents. Transients can represent a variety of initiating events (e.g., manual 

reactor trip, loss of main feedwater, turbine trip, loss of offsite power, and loss of primary flow). 

Each of these initiating events subsequently leads to changes in reactor temperature or 

pressure that could demand functioning of safety systems. Transients are modeled in the PRA 

if they lead to a plant trip, thus challenging safety systems leading to positive or negative 

outcomes. The terms transient and general transient often are used interchangeably, which is 

appropriate and correct in a PRA context. 

NUREG/CR-6572 (Ref. 76) defines the term general transient as “events in which high 

pressure can be maintained in the primary system, active core cooling is required, and high 

pressure makeup may be needed.” 

The NRC Web site Glossary (Ref. 36) defines the term transient as “a change in the reactor 

coolant system temperature, pressure, or both, attributed to a change in the reactor’s power 

output. Transients can be caused by (1) adding or removing neutron poisons, (2) increasing or 

decreasing electrical load on the turbine generator, or (3) accident conditions.” 

Truncation Limit 

The minimum In a PRA, a truncation limit is a numerical criterion that defines the boundaries, in terms of 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

value of 

contributors 

retained in the 

PRA quantification 

process. (see 

Accident 

Sequence, Cutset)

frequencies or probabilities, of what is retained and what is screened out. The truncation limit 

determines what accident sequences or cutsets are retained for or excluded from further 

analysis. 

Since truncation limit affects PRA quantification, Regulatory Guide 1.200 (Ref. 91) notes that 

truncation values should be set relative to the total plant core damage frequency (CDF) such 

that the CDF is stable with respect to further reduction in the truncation value. 

The ASME/ANS PRA Standard (Ref. 2) defines truncation limit as “the numerical cutoff value of 

probability or frequency below which results are not retained in the quantitative PRA model or 

used in subsequent calculations (such limits can apply to accident sequences-cutsets, system 

level cutsets, and sequence-cutset database retention).” 

Unavailability 

(see Availability) The term unavailability is the opposite of availability and is defined under “availability.” 

Uncertainty (Aleatory, Random, Stochastic, Epistemic, State-of-Knowledge, Model, 

Source of Model, Key Source of Model, Parameter, Completeness)

Variability in an 

estimate because 

of the randomness 

of the data or the 

lack of knowledge. 

When used in the context of a PRA, the term uncertainty is associated with the lack of 

information or knowledge, or the random behavior of a system or model that is taken into 

account in the PRA in different ways. 

In defining uncertainty, there are two types: aleatory and epistemic. Aleatory uncertainty is 

based on the randomness of the nature of the events or phenomena and cannot be reduced by 

increasing the analyst’s knowledge of the systems being modeled. Therefore, it is also known 

as random uncertainty or stochastic uncertainty. Epistemic uncertainty is the uncertainty 

related to the lack of knowledge or confidence about the system or model and is also known as 

state-of-knowledge uncertainty. 

The PRA model itself reflects aleatory uncertainty. The PRA model contains epistemic 

uncertainty that includes model uncertainty, parameter uncertainty, or completeness 

uncertainty. 

In the ASME/ANS PRA Standard (Ref. 2), uncertainty is defined as “a representation of the 

confidence in the state-of-knowledge about the parameter values and models used in 

constructing the PRA.” 

In the ASME/ANS PRA Standard (Ref. 2), aleatory uncertainty is defined as “the uncertainty 

inherent in a nondeterministic (stochastic, random) phenomenon. Aleatory uncertainty is 

reflected by modeling the phenomenon in terms of a probabilistic model. In principle, aleatory 

uncertainty cannot be reduced by the accumulation of more data or additional information. 

(Aleatory uncertainty is sometimes called ‘randomness.’)” 

In the ASME/ANS PRA Standard (Ref. 2), epistemic uncertainty is defined as “the uncertainty 

attributable to incomplete knowledge about a phenomenon that affects our ability to model it. 

Epistemic uncertainty is reflected in ranges of values for parameters, a range of viable models, 

the level of model detail, multiple expert interpretations, and statistical confidence. In principle, 

epistemic uncertainty can be reduced by the accumulation of additional information. (Epistemic 

uncertainty is sometimes also called ‘modeling uncertainty.’)” 

Model uncertainty is discussed in NUREG-1855 (Ref. 60) as follows: 

“Model uncertainty is related to an issue for which no consensus approach or model exists and 

where the choice of approach or model is known to have an effect on the PRA model (e.g., 

introduction of a new basic event, changes to basic event probabilities, change in success 

criterion, and introduction of a new initiating event). A model uncertainty results from a lack of 

knowledge of how structures, systems and components (SSCs) behave under the conditions 

arising during the development of an accident. A model uncertainty can arise for the following 
4. GLOSSARY

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TERM AND 

DEFINITION DISCUSSION 

reasons: 

• The phenomenon being modeled is itself not completely understood (e.g., behavior of 

gravity-driven passive systems in new reactors, or crack growth resulting from 

previously unknown mechanisms). For some phenomena, some data or other 

information may exist, but it needs to be interpreted to infer behavior under conditions 

different from those in which the data were collected (e.g., RCP seal LOCA 

information).

• The nature of the failure modes is not completely understood or is unknown 

(e.g., digital instrumentation and controls).”

In the ASME/ANS PRA Standard (Ref. 2), source of model uncertainty is defined as: “a source 

that is related to an issue in which there is no consensus approach or model and where the 

choice of approach or model is known to have an effect on the PRA model (e.g., introduction of 

a new basic event, changes to basic event probabilities, change in success criterion, 

introduction of a new initiating event). A source of model uncertainty is labeled “key” when it 

could impact the PRA results that are being used in a decision, and consequently, may 

influence the decision being made. Therefore, a key source of model uncertainty is identified in 

the context of an application. This impact would need to be significant enough that it changes 

the degree to which the risk acceptance criteria are met, and therefore, could potentially 

influence the decision.” 

NUREG-1855 (Ref. 62) has additional discussion on key sources of model uncertainty. The 

terms key model uncertainty and key sources of model uncertainty have the same meaning. 

Parameter uncertainty is the uncertainty in the values of the parameters of a model 

represented by a probabilistic distribution. Examples of parameters that could be uncertain 

include initiating event frequencies, component failure rates and probabilities, and human error 

probabilities that are used in the quantification of the accident sequence frequencies. 

Completeness uncertainty is caused by the limitations in the scope of the model, such as 

whether all applicable physical phenomena have been adequately represented, and all 

accident scenarios that could significantly affect the determination of risk have been identified. 

Completeness uncertainty also can be thought of as a type of model uncertainty. However, 

completeness uncertainty is separated from model uncertainty because it represents a type of 

uncertainty that cannot be quantified. It also represents those aspects of the system that are, 

either knowingly or unknowingly, not addressed in the model. (Ref. 62) 

Uncertainty Analysis 

A process for 

determining the 

level of imprecision 

in the results of the 

PRA and its 

parameters. 

In a PRA, the ways in which the uncertainty in the results is presented includes the following: 

• A continuous probability distribution on numerical results.

• A discrete probability distribution representing the impact of different models or 

assumptions.

• Sensitivity studies that provide a discrete set of results that represent the results of 

making different assumptions or using different models, or that represent the impact 

of varying key parameters in the model that have significant uncertainty, without 

providing weights or probabilities to the members of the set.

• Bounds or ranges of results that represent the results of the extreme assumptions.

• An identification of limitations in the scope of the model (e.g., incompleteness) and 

how they might influence the applicability of the PRA.

The ASME/ANS PRA Standard (Ref. 2) defines uncertainty analysis as “the process of 

identifying and characterizing the sources of uncertainty in the analysis, and evaluating their 
4. GLOSSARY

 4-101 

TERM AND 

DEFINITION DISCUSSION 

impact on the PRA results and developing a quantitative measure to the extent practical.”

Uncertainty Distribution 

(see Probability 

Distribution) 

The term uncertainty distribution is related to the term probability distribution and is defined 

under “Probability Distribution.” 

Uncertainty Interval, Uncertainty Range

A range that 

bounds the 

uncertainty 

value(s) of a 

parameter or 

analysis result by 

establishing upper 

and lower limits. 

(see Confidence 

Interval, Probability 

Distribution) 

In a PRA, uncertainty intervals can provide the range of the frequency or probability of the 

various inputs (e.g., initiating event frequencies, component failure probabilities, human error 

probabilities), as well as outputs of the analysis (e.g., core damage frequency, conditional 

containment failure probability). However, in most cases, a probability distribution of the 

uncertainty around a mean value is preferred. 

NUREG 1855 (Ref. 62) defines uncertainty interval as “a characterization of the uncertainty. 

This characterization could, in the simplest approach, take the form of an interval (i.e., a range 

of values within which the value lies). However, it is more usual to characterize the uncertainty 

in terms of a probability distribution on the value of the quantity of concern, whether it is a 

parameter, accident sequence frequency, or a core damage frequency.” 

The NRC Web site Glossary (Ref. 34) defines uncertainty range as “an interval within which 

a numerical result is expected to lie within a specified level of confidence. The interval often 

used is the 5–95 percentile of the distribution reporting the uncertainty.” 

The definition provided was based on definitions in the NRC Web site Glossary (Ref. 36) and in 

NUREG-1855 (Ref. 62).

Uncertainty Range 

(see Uncertainty 

Interval) 

The term uncertainty range has the same meaning as uncertainty interval and is defined under 

“Uncertainty Interval.” 

Unreliability 

(see Reliability) The term unreliability is the opposite of reliability and is defined under “Reliability.” 

Up-to-Date 

(see PRA 

Configuration 

Control, As-Built 

As-Operated) 

The term up-to-date is related to PRA configuration control and is defined under “PRA 

Configuration Control” or “As-Built As-Operated.” 

Vulnerability 

Weakness in the 

design or 

operation of a 

system, 

component, or 

structure that could 

disable its function.

Results from a PRA of a nuclear power plant (NPP) model can be used to identify plant 

vulnerabilities (e.g., vulnerabilities related to system design or plant operations). The term 

vulnerability was used in Generic Letter (GL) 88-20, “Individual Plant Examination For Severe 

Accident Vulnerabilities” (Ref. 40). As part of GL 88-20, each licensee was asked to perform a 

systematic examination of its NPP to identify any plant-specific vulnerabilities to severe 

accidents. The NRC, however, did not define vulnerability; it was the licensee’s responsibility 

to define vulnerability. The method all licensees used to identify vulnerabilities was a PRA. 

For some licensees, vulnerabilities were based on the contribution of accident sequence types 

or individual failure events (e.g., fault tree basic events) to overall plant core damage frequency 

(CDF) or a percent contribution to CDF (e.g., a functional accident sequence with a CDF that 

exceeds 1E-04/yr, or one that contributes more than 50% to the total plant CDF). 
4. GLOSSARY

 4-102 

TERM AND 

DEFINITION DISCUSSION 

Water Immersion

Direct exposure 

from radioactive 

material in 

contaminated 

water given to an 

individual 

immersed in the 

water. (see 

Exposure 

Pathways, 

Cloudshine, 

Groundshine, 

Inhalation, 

Ingestion, Skin 

Deposition)

In a Level 3 PRA, for the consequence calculation, water immersion, is one of the assumed 

pathways by which an individual can receive doses. The pathways of exposure include: (1) 

direct external exposure from radioactive material in a plume, principally due to gamma 

radiation (air immersion or cloudshine), (2) direct exposure from radioactive material in 

contaminated water given to an individual immersed in the water, (3) exposure from inhalation 

of radioactive materials in the cloud and resuspended material deposited on the ground, (4) 

exposure to radioactive material deposited on the ground (groundshine), (5) radioactive 

material deposited onto the body surfaces (skin deposition), and (6) ingestion from deposited 

radioactive materials that make their way into the food and water pathway. 
5. REFERENCES 

5-1 

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(11) NFPA 805, “Performance-Based Standard for Fire Protection for Light Water Reactor 

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5. REFERENCES 

5-2 

Code of Federal Regulations

(12) Title 10 of the Code of Federal Regulations (10 CFR) 20.1003, 2010. 

(13) 10 CFR 50.2, “Definitions,” 2010. 

(14) 10 CFR 50.34, “Contents of Applications; Technical information,” 2010. 

(15) 10 CFR 50.36, “Technical Specifications,” 2010. 

(16) 10 CFR 50.44, “Combustible Gas Control for Nuclear Power Reactors,” 2010. 

(17) 10 CFR 50.48, “Fire Protection,” 2010. 

(18) 10 CFR 50.49, “Environmental qualification of electric equipment Important to Safety for 

Nuclear Power Plants,” 2010. 

(19) 10 CFR 50.62, “Requirements for Reduction of Risk from Anticipated Transients without 

Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants,” 2010. 

(20) 10 CFR 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at 

Nuclear Power Plants,” 2010. 

(21) 10 CFR 50.69, “Risk-informed Categorization and Treatment of Structures, Systems and 

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(22) Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR Part 50, 

“Domestic Licensing of Production and Utilization Facilities,” 2010. 

 

(23) Appendix E, “Emergency Planning and Preparedness for Production and Utilization 

Facilities,” to 10 CFR Part 50, 2010. 

(24) Appendix R, “Fire Protection Program for Nuclear Power Facilities Operating Prior to 

January 1, 1979,” to 10 CFR Part 50, 2010. 

(25) Appendix S, “Earthquake Engineering Criteria for Nuclear Power Plants,” to 10 CFR 

Part 50, 2010. 
5. REFERENCES 

5-3 

(26) 10 CFR 100.11, “Determination of Exclusion Area, Low Population Zone, and Population 

Center Distance,” 2010. 

(27) 10 CFR 54.3, “Definitions,” 2010. 

Federal Guidance Report

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Environmental Exposure to Radionuclides; Updates and Supplements,” 

U.S. Environmental Protection Agency, 2006. 

Federal Register

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Under the National Environment Policy Act of 1969,” Federal Register (45 FR 40101). 

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Designs and Existing Plants,” Federal Register (50 FR 32138). 

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Plants, Statement of Policy,” Federal Register (51 FR 24643). 

(33) U.S. Nuclear Regulatory Commission, “Use of Probabilistic Risk Assessment Methods 

in Nuclear Activities: Final Policy Statement,” Federal Register (60 FR 42622). 

NRC Web Sites and Documents

(34) NRC Web: “Fact Sheet on Nuclear Reactor Risk” 

http://www.nrc.gov/reading-rm/doc-collections/fact-sheets/reactor-risk.html 

(35) NRC Web: “Fact Sheet on Probabilistic Risk Assessment” 

http://www.nrc.gov/reading-rm/doc-collections/fact-sheets/probabilistic-risk-asses.html 

(36) NRC Web: “Glossary” 

http://www.nrc.gov/reading-rm/basic-ref/glossary.html 

(37) NRC Web: “Risk Assessment in Regulation” 

http://www.nrc.gov/about-nrc/regulatory/risk-informed.html 
5. REFERENCES 

5-4 

(38) NRC Web: “Risk and Performance Concepts in the NRC’s Approach to Regulation” 

http://www.nrc.gov/about-nrc/regulatory/risk-informed/concept.html

(39) ACRS Letter Report, “The Role of Defense in Depth in a Risk-Informed Regulatory 

System,” U.S. Nuclear Regulatory Commission, May 19, 1999. 

(40) Generic Letter No. 88-20, “Individual Plant Examination For Severe Accident 

Vulnerabilities–10 CFR 50.54(f),” U.S. Nuclear Regulatory Commission, 

November 23, 1988. 

(41) IMC 0308, “Technical Basis for Maintenance Risk Assessment and Risk Management 

SDP, Attachment 3, Appendix K,” U.S. Nuclear Regulatory Commission, Washington 

DC, May 19, 2005. 

(42) IMC 0609, “Shutdown Operations Significance Determination Process, Appendix G,” 

U.S. Nuclear Regulatory Commission, Washington DC, February 28, 2005. 

(43) Management Directives 8.3, “NRC Incident Investigation Program,” U.S. Nuclear 

Regulatory Commission, Washington, DC, March 27, 2001. 

(44) NUREG 75/014 (WASH-1400), “Reactor Safety Study: An Assessment of Accident 

Risks in U.S. Commercial Nuclear Power Plants,” U.S. Nuclear Regulatory Commission, 

Washington, DC, October 1975. 

(45) NUREG-0713, “Occupational Radiation Exposure at Commercial Nuclear Power 

Reactors and Other Facilities,” Volume 28, U.S. Nuclear Regulatory Commission, 

Washington, DC, December, 2007. 

(46) NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for 

Nuclear Power Plants: LWR Edition—Engineered Safety Features,” Chapter 6, 

U.S. Nuclear Regulatory Commission, Washington, DC, March 2007. 

(47) NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for 

Nuclear Power Plants: LWR Edition—Transient and Accident Analysis,” Chapter 15, 

U.S. Nuclear Regulatory Commission, Washington, DC, September 2009. 

(48) NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for 

Nuclear Power Plants: LWR Edition—Severe Accidents,” Chapter 19, Sections 19.0 and 

19.1, U.S. Nuclear Regulatory Commission, Washington, DC, September 2009. 
5. REFERENCES 

5-5 

(49) NUREG-0492, “Fault Tree Handbook,” U.S. Nuclear Regulatory Commission, 

Washington, DC, January 1981. 

(50) NUREG-0700, “Human-System Interface Design Review Guidelines,” Revision 2, 

U.S. Nuclear Regulatory Commission, Washington, DC, May 2002. 

(51) NUREG-1150, “Severe Accident Risks: An Assessment for Five U.S. Nuclear Power 

Plants,” U.S. Nuclear Regulatory Commission, Washington, DC, December 1990. 

(52) NUREG-1335, “Individual Plant Examination: Submittal Guidance,” U.S. Nuclear 

Regulatory Commission, Washington, DC, August 1989. 

(53) NUREG-1407, “Procedural and Submittal Guidance for the Individual Plant Examination 

of External Events (IPEEE) for Severe Accident Vulnerabilities,” U.S. Nuclear Regulatory 

Commission, Washington, DC, June 1991. 

(54) NUREG-1489, “A Review of NRC Staff Uses of Probabilistic Risk Assessment,” 

U.S. Nuclear Regulatory Commission, Washington, DC, March 1994. 

(55) NUREG-1542, “Performance and Accountability Report—NRC Summary of 

Performance And Financial Information Fiscal Year 2009,” Volume 15, Supplement 1, 

U.S. Nuclear Regulatory Commission, Washington, DC, February 2010. 

(56) NUREG-1560, “Individual Plant Examination Program: Perspectives on Reactor Safety 

and Plant Performance,” U.S. Nuclear Regulatory Commission, Washington, DC, 

December 1997. 

(57) NUREG-1563, “Branch Technical Position on the Use of Expert Elicitation in the 

High-Level Radioactive Waste Program,” U.S. Nuclear Regulatory Commission, 

Washington, DC, November 1996. 

(58) NUREG-1649, “Reactor Oversight Process,” U.S. Nuclear Regulatory Commission, 

Revision 4, Washington, DC, December 2006. 

(59) NUREG-1742, “Perspectives Gained from the Individual Plant Examination of External 

Events,” U.S. Nuclear Regulatory Commission, Washington, DC, April 2002. 

(60) NUREG-1805, “Fire Dynamics Tools (FDTs) Quantitative Fire Hazard Analysis Methods 

for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program,” 

U.S. Nuclear Regulatory Commission, Washington, DC, December 2004. 
5. REFERENCES 

5-6 

(61) NUREG-1816, “Independent Verification of the Mitigating Systems Performance Index 

(MSPI) Results for the Pilot Plants,” U.S. Nuclear Regulatory Commission, 

Washington, DC, February 2005. 

(62) NUREG-1855, “Guidance on the Treatment of Uncertainties Associated with PRAs in 

Risk-Informed Decisionmaking,” U.S. Nuclear Regulatory Commission, Washington, DC, 

March 2009. 

(63) NUREG-1860, “Feasibility Study for a Risk-Informed and Performance-Based 

Regulatory Structure for Future Plant Licensing,” U.S. Nuclear Regulatory Commission, 

Washington, DC, December 2007. 

(64) NUREG-1925, “Research Activities,” U.S. Nuclear Regulatory Commission, Washington, 

DC, Revision 1, December, 2010. 

(65) NUREG-1934, “Nuclear Power Plant Fire Modeling Application Guide (NPP Fire MAG)”, 

U.S Nuclear Regulatory Commission, Washington, DC, August 2011. 

(66) NUREG/BR-0318, “Effective Risk Communication: The Nuclear Regulatory 

Commission’s Guidelines for Internal Risk Communication,” U.S. Nuclear Regulatory 

Commission, Washington, DC, December 2004. 

(67) NUREG-2150, “A Proposed Risk Management Regulatory Framework,” U.S. Nuclear 

Regulatory Commission, Washington, DC, April 2012. 

(68) NUREG/CR-1278 (SAND80-0200), “Handbook of Human Reliability Analysis with 

Emphasis on Nuclear Power Plant Applications,” U.S. Nuclear Regulatory Commission, 

Washington, DC, August 1983. 

(69) NUREG/CR-2300, “PRA Procedures Guide: A Guide to the Performance of Probabilistic 

Risk Assessments for Nuclear Power Plants,” U.S. Nuclear Regulatory Commission, 

Washington, DC, January 1983. 

(70) NUREG/CR-3673 (SAND84-0178), “Economic Risks of Nuclear Power Reactor 

Accidents,” U.S. Nuclear Regulatory Commission, Washington, DC, May, 1984. 

(71) NUREG/CR-3385 (BMI-2103), “Measures of Risk Importance and Their Applications,” 

U.S. Nuclear Regulatory Commission, Washington, DC, July 1983. 
5. REFERENCES 

5-7 

(72) NUREG/CR-4772, “Accident Sequence Evaluation Program Human Reliability Analysis 

Procedure,” U.S. Nuclear Regulatory Commission, Washington, DC, February 1987. 

(73) NUREG/CR-5485 (INEEL/EXT-97-01327), “Guidelines on Modeling Common-Cause 

Failures in Probabilistic Risk Assessment,” U.S. Nuclear Regulatory Commission, 

Washington, DC, November 1998. 

(74) NUREG/CR-5695, “A Process for Risk-Focused Maintenance,” U.S. Nuclear Regulatory 

Commission, Washington, DC, March 1991. 

(75) NUREG/CR-6268 (INEEL/EXT-97-00696), “Common-Cause Failure Database and 

Analysis System: Software Reference Manual,” Vol. 4, U.S. Nuclear Regulatory 

Commission, Washington, DC, June 1998. 

(76) NUREG/CR-6572 (BNL-NUREG-52534-R1), “Kalinin VVER-1000 Nuclear Power Station 

Unit 1 PRA: Procedure Guides for a Probabilistic Risk Assessment,” U.S. Nuclear 

Regulatory Commission, Washington, DC, December 2005. 

(77) NUREG/CR-6595, “An Approach for Estimating the Frequencies of Various Containment 

Failure Modes and Bypass Events,” U.S. Nuclear Regulatory Commission, 

Washington, DC, October 2004. 

(78) NUREG/CR-6823 (SAND2003-3348P), “Handbook of Parameter Estimation for 

Probabilistic Risk Assessment,” U.S. Nuclear Regulatory Commission, Washington, DC, 

September 2003. 

(79) NUREG/CR-6850 (EPRI TR-1011989), “Fire PRA Methodology for Nuclear Power 

Facilities,” U.S. Nuclear Regulatory Commission, Washington, DC, September 2005. 

(80) NUREG/CR-6890, “Reevaluation of Station Blackout Risk at Nuclear Power Plants,” 

U.S. Nuclear Regulatory Commission, Washington, DC, December 2005. 

(81) NUREG/CR-6901, “Current State of Reliability Modeling Methodologies for Digital 

Systems and Their Acceptance Criteria for Nuclear Power Plant Assessments,” 

U.S. Nuclear Regulatory Commission, Washington, DC, February 2006. 

(82) NUREG/CR-6928, “Industry-Average Performance for Components and Initiating Events 

at U.S. Commercial Nuclear Power Plants,” U.S Nuclear Regulatory Commission, 

Washington, DC, January 2007. 
5. REFERENCES 

5-8 

(83) NUREG/CR-6952, Vol.2, “Systems Analysis Programs for Hands-on Integrated 

Reliability Evaluations (SAPHIRE): Technical Reference,” U.S. Nuclear Regulatory 

Commission, Washington, DC, October 2007. 

(84) Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in 

Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 1, 

U.S. Nuclear Regulatory Commission, Washington, DC, November 2002. 

(85) Regulatory Guide 1.175, “An Approach for Plant-Specific, Risk-Informed 

Decisionmaking: Inservice Testing,” U.S. Nuclear Regulatory Commission, 

August 1998. 

(86) Regulatory Guide 1.177, “An Approach for Plant-Specific, Risk-Informed 

Decisionmaking: Technical Specifications,” U.S. Nuclear Regulatory Commission, 

Washington, DC, August 1998. 

(87) Regulatory Guide 1.178, “An Approach for Plant-Specific, Risk-Informed 

Decisionmaking for Inservice Inspection of Piping,” U.S. Nuclear Regulatory 

Commission, Washington, DC, September 2003. 

(88) Regulatory Guide 1.182, “Assessing and Managing Risk Before Maintenance Activities 

at Nuclear Power Plants,” U.S. Nuclear Regulatory Commission, Washington, DC, 

May 2000. 

(89) Regulatory Guide 1.187, “Guidance for Implementation of 10 CFR 50.59, Changes, 

Tests, and Experiments,” U.S. Nuclear Regulatory Commission, Washington, DC, 

November 2000. 

(90) Regulatory Guide 1.189, “Fire Protection for Nuclear Power Plants”, U.S Nuclear 

Regulatory Commission, Washington, DC, April 2009. 

(91) Regulatory Guide 1.200, “An Approach for Determining the Technical Adequacy of 

Probabilistic Risk Assessment Results for Risk-Informed Activities,” Revision 2, 

U.S. Nuclear Regulatory Commission, Washington, DC, March 2009. 

(92) Regulatory Guide 1.201, “Guidelines for Categorizing Structures, Systems, and 

Components in Nuclear Power Plants According to Their Safety Significance,” 

Revision 1, U.S. Nuclear Regulatory Commission, Washington, DC, May 2006. 
5. REFERENCES 

5-9 

(93) Regulatory Guide 1.205, “Risk-Informed, Performance-Based Fire Protection for 

Existing Light-Water Nuclear Power Plants,” Revision 1, U.S. Nuclear Regulatory 

Commission, Washington, DC, December 2009. 

(94) Regulatory Guide 1.206, “Combined License Applications for Nuclear Power Plants,” 

U.S. Nuclear Regulatory Commission, Washington, DC, June 2007. 

(95) Regulatory Guide 1.208, “A Performance-Based Approach to Define the Site-Specific 

Earthquake Ground Motion,” U.S. Nuclear Regulatory Commission, Washington, DC, 

March 2007. 

(96) SECY-98-144, “White Paper on Risk-Informed and Performance-Based Regulation,” 

U.S. Nuclear Regulatory Commission, Washington, DC, June 22, 1998. 

(97) SECY-99-100, “Framework for Risk-Informed Regulation in the Office of Nuclear 

Material Safety and Safeguards,” U.S. Nuclear Regulatory Commission, Washington, 

DC, March 31, 1999. 

(98) SECY-00-0162, “Addressing PRA Quality in Risk-Informed Activities,” U.S. Nuclear 

Regulatory Commission, Washington, DC, July 28, 2000.

(99) SECY-06-0217, “Improvement to and Update of the Risk-Informed Regulation 

Implementation Plan,” U.S. Nuclear Regulatory Commission, Washington, DC, 

October 25, 2006. 

(100) SECY-13-0029, “History of the Use and Consideration of the Large Release Frequency 

Metric”, U.S Nuclear Regulatory Commission, Washington, DC, March 22, 2013.

(101) Staff Requirements Memorandum M060503B, “Briefing on Status of Risk-Informed 

and Performance-Based Reactor Regulation,” U.S. Nuclear Regulatory Commission, 

Washington, DC, June 1, 2006. 

APPENDIX A 

A-1 

INTERNAL FIRE GLOSSARY 

Table A-1 provides internal fire terms and their definitions with the associated discussion. The 

terms are listed alphabetically. 

Table A-1 Internal Fire Terms and Definition 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Active Fire Barriers 

A fire barrier that must 

be physically 

repositioned from its 

normal configuration to 

an alternate 

configuration in order 

to provide its protective 

function. 

In a fire PRA, fire barriers impede the spread of fires and limit potential damage to safety 

equipment, thus reducing probabilities of fire spread to additional components and the 

probability of accident sequences. Ventilation system fire dampers, normally open fire doors, 

and water curtains are examples of passive fire barriers. 

The definition provided was based on the definition in NUREG-1805 (Ref. 60).

Algebraic Fire Models 

A type of fire model 

that provides a method 

for calculating simple 

fire phenomena based 

on a closed-form 

algebraic formulation.

In a fire PRA, fire models predict fire damage of components, and thus contribute to the 

failure of those components, given failure of suppression. 

Algebraic models may be standalone equations found in the literature or may be contained 

within spreadsheets, such as the NRC’s fire dynamics tools (FDTs). These equations are 

typically closed-form algebraic expressions, many of which were developed as correlations 

from empirical data. In some cases, they may take the form of a first-order ordinary 

differential equation and can provide an estimate of fire variables, such as hot gas layer 

(HGL) temperature, heat flux from flames or the HGL, smoke production rate, depth of the 

hot gas layer, and the actuation time for detectors. 

Algebraic models are helpful because they require minimal computational time and a limited 

number of input variables. Other than for very simple situations, algebraic models are useful 

primarily as screening tools. 

The definition provided was based on the definition in NUREG-1934 (Ref. 65).

Authority Having Jurisdiction 

The organization, 

office, or individual 

responsible for 

approving equipment, 

materials, an 

installation, or a 

procedure. 

The NRC is the authority having jurisdiction for NFPA 805 as it is applied under 10 CFR 

50.48 (Ref.17). 

The definition provided was based on the definition in the NFPA 805 Standard (Ref. 11). 

Cable and Raceway (Database) System

Cross-reference of 

power, control, or 

instrument cables 

associated with certain 

components or systems 

and their location 

throughout the plant, as 

it relates to specific 

cable raceways, tracks, 

or conduits where they 

may be situated. 

The Cable and Raceway System (CRS) generally correlates cables to raceways, raceways 

to locations within the plant, and tracks basic cable and raceway attributes. Newer CRSs 

typically contain sophisticated database sort and query features. 

The information in the CRS may be used to determine how a fire in a certain location may 

affect the cables nearby and thus determine which components and systems may be 

affected. The location of cables is then used for the development of fire scenarios that are 

quantified in the fire PRA. This is then used in a PRA as input in constructing and calculating 

accident sequences. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79).
APPENDIX A 

A-2 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Cable Failure Mode 

The behavior of an 

electrical cable upon 

fire-induced failure. 

(see Intercable 

Shorting, Intracable 

Shorting)

In a fire PRA, component failure modes can be attributed to cable failure modes resulting 

from fire. The ASME/ANS PRA Standard (Ref. 2) indicates that “failure modes for electrical 

cables include intractable shorting, intercable shorting, open circuit (loss of conductor 

continuity), and/or shorts between a conductor and an external ground.” 

Ceiling Jet 

The relatively rapid gas 

flow in a shallow layer 

beneath the ceiling 

surface that is driven by 

the buoyancy of hot 

combustion products. 

Typically, a fire plume will form above a burning object. The fire plume will rise until 

obstructed by a horizontal surface, such as a ceiling. Upon hitting the ceiling, the hot gases 

in the fire plume will turn and flow along the ceiling in the form of a ceiling jet. When the 

ceiling jet gases are blocked by vertical surfaces, such as walls, they will accumulate into a 

hot gas layer or smoke layer. As more hot gas accumulates in the layer, the interface 

between the hot gas layer and cooler layer below will continue to drop toward the floor of the 

enclosure. As stated in NUREG/CR-6850 (Ref. 79), “ceiling jets form when a fire plume 

impinges under a ceiling and hot gases spread away.” 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Circuit Failure Analysis 

The evaluation of 

electrical circuits to 

determine both the 

potential failure modes 

and their impact on the 

systems and equipment 

supported by the 

circuit. 

Circuit failure analysis can include the assignment of probabilities to the likelihood of the 

cable failure modes of concern. Circuit failure analysis would include consideration of the 

impact of cable failures on circuit function. The equipment failures associated with those 

circuit failure modes would be input to the PRA and contribute to accident sequence 

quantification. 

Circuit Failure Mode 

The manner in which 

conductor failures from 

an electrical cable are 

manifested in the 

circuit. (see Cable 

Failure Mode)

In a fire PRA, equipment failures associated with circuit failure modes are analyzed and 

contribute to accident sequence quantification. Examples of circuit failure modes include 

loss of motive power, loss of control, loss of or false indication, open circuit conditions, and 

spurious operation. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 
APPENDIX A 

A-3 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Code of Record 

The edition of the code 

or standard in effect at 

the time the fire 

protection systems or 

feature was designed 

or specifically 

committed to the 

authority having 

jurisdiction. (see 

Authority Having 

Jurisdiction)

If the 1996 edition of NFPA 13 was in effect at the time a sprinkler system was designed, the 

code of record would be NFPA 13, Standard for the Installation of Sprinkler Systems – 1996 

edition. 

The definition provided was based on the definition in the NFPA 805 Standard (Ref.11). 

Compensatory Actions 

Actions taken to 

counteract or reduce an 

impairment to a 

required fire protection 

system, feature, or 

component. 

In the NFPA 805 Standard (Ref.11), compensatory actions are described as “actions taken if 

an impairment to a required system, feature, or component prevents that system, feature, or 

component from performing its intended function. These actions are a temporary alternative 

means of providing reasonable assurance that the necessary function will be compensated 

for during the impairment, or an act to mitigate the consequence of a fire. Compensatory 

measures include, but are not limited to, actions such as fire watches, administrative 

controls, temporary systems, and features of components.” 

The term compensatory measures may be used in place of compensatory actions (e.g., fire 

watch compensatory actions may improve detection in the affected vicinity). 

The definition provided was based on the definition in the NFPA 805 Standard (Ref. 11). 

Concurrent Hot Shorts 

The occurrence of two 

or more hot shorts such 

that the shorts overlap 

in time. (see 

Conductor-toConductor Short) 

In a fire PRA, concurrent hot shorts are important because they can cause multiple 

equipment failures, complicate operator response, and increase human error probabilities in 

a fire PRA. These challenges may be more difficult to overcome than would be the case 

given only a single spurious operation at a time. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Conductor-to-Conductor Short 

An abnormal 

connection (including 

an arc) of relatively low 

impedance between 

two conductors. 

In a fire PRA, conductor-to-conductor shorts may be caused by fire and in turn may cause 

failure of equipment, thus contributing to accident sequences. 

As described in NUREG/CR-6850 (Ref. 76), a conductor-to-conductor short can occur in the 

following manner: “a conductor-to-conductor short between an energized conductor of a 

grounded circuit and a grounded conductor results in a ground fault. A conductor-toconductor short between an energized conductor and a non-grounded conductor results in a 

hot short. A conductor-to-conductor short between an energized conductor of an 

ungrounded circuit and a neutral conductor has the same functional impact as a ground 

fault.” 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 
APPENDIX A 

A-4 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Damage Criteria 

Those characteristics 

of the fire-induced 

environment that are 

specified as indicating 

failure of a damage 

target or set of damage 

targets. (see Damage 

Target, Damage 

Threshold)

In a fire PRA, cables and their associated components are failed in the PRA model upon 

damage. Damage criteria commonly refer to certain temperatures or heat fluxes at target 

locations that when exceeded indicate failure of the targets. The damage target may be a 

cable, set of cables, or a component in a location near the fire. The damage criteria also 

may be based on any other environmental effect of the fire (e.g., smoke density). 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Damage Target 

Any cable, equipment, 

or structural element in 

the fire PRA whose 

function can be 

adversely affected by 

the modeled fire. 

In a fire PRA, cables and their associated components are failed in the PRA model upon 

damage. 

The ASME/ANS PRA Standard (Ref. 2) defines the term damage target as “a cable or 

equipment item that belongs to the Fire PRA cable or equipment list and that is included in 

event trees and fault trees for fire risk estimation. Damage targets also may include 

structural elements (e.g., structural steel) in the case of certain high-hazard fire sources, 

such as very large oil spills.” 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Damage Threshold 

The values 

corresponding to the 

damage criteria that will 

be taken as indicative 

of the onset of fireinduced failure of a 

damage target or set of 

damage targets. (see 

Damage Criteria)

An example of a damage threshold would be the temperature at a cable location that when 

exceeded would indicate failure of the cable. 

The definition provided was based on the definition in the ASME/ANS PRA Standard 

(Ref. 2). 

Electrical Cable 

A construct consisting 

of one or more 

insulated conductors 

designed to carry 

signals or power 

between points in a 

circuit.

In a fire PRA, fire damage to a cable may result in disablement or spurious operation of 

safety-related equipment (affecting probability of failure of safety systems) and/or generation 

of an initiating event. Cables are used to connect points in a common electrical circuit and 

may be used to transmit power, control signals, indications, or instrument signals. Cables are 

important to risk because they connect equipment necessary for safe operation of the plant 

to sources of power and control over relatively long distances in the plant. This increases the 

possibility that an undesired event (e.g., a fire) at an intervening location will affect the cable 

and disrupt the continued operation of equipment. 

Electrical Raceway Fire Barrier System 

Non-load-bearing 

partition type envelope 

system installed around 

electrical components 

and cabling that are 

rated by test 

laboratories in hours of 

fire resistance and 

In a fire PRA, electrical raceway fire barrier systems (ERFBSs) are modeled because they 

provide protection for electrical cables and delay or prevent damage from fires. A fire rated 

ERFBS provides additional time before damage for those protected cables in a fire PRA. 

The definition provided was based on the definition in Regulatory Guide 1.189 (Ref. 90). 
APPENDIX A 

A-5 

TERM AND 

DEFINITION (S) 

DISCUSSION 

used to maintain safeshutdown functions free 

of fire damage. (see 

Wrap)

External Hot Short 

A hot short in which the 

source conductor and 

target conductor are 

from separate cables. 

(see Hot Short, 

Intercable Short Circuit)

The term external hot short can be used interchangeably and correctly with intercable short 

circuit, which is also referred to as intercable conductor-to-conductor short circuit. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Field Models 

A type of fire model that 

provides a method for 

calculating fluid flow 

through a volume using 

numerical solutions of 

the governing 

equations for 

conservation of total 

mass, chemical 

species, momentum, 

and energy. 

In a fire PRA, the results from a field model can be used as input in determining the 

probability of damage from a particular fire to targets nearby and to associated safety-related 

equipment. 

Field models are computational fluid dynamics models that can be used to predict fireinduced environmental conditions (e.g., temperature at different times). The equations used 

in field models are approximated using finite differences over discrete control volumes, and 

the solution is obtained using the discretized equations. The calculations are performed over 

a period of time to obtain a transient (time-dependent) solution, or iterated over many times 

to provide a steady-state (time-independent) solution. The model typically is comprised of a 

large number of control volumes from thousands to millions. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Fire Analysis Tool 

A method used to 

estimate or calculate 

one or more physical 

fire effects. (see Field 

Model, Zone Model, 

Algebraic Fire Model)

Fire analysis tools include, but are not limited to, computerized compartment fire models, 

such as zone or field models, closed-form algebraic fire models, empirical correlations such 

as those provided in a handbook, and lookup tables that relate input parameters to a 

predicted output. The fire analysis tool used is based on the objectives of the specific 

analysis and a predefined set of input parameter values as defined by the fire scenario being 

analyzed. 

Examples of calculated physical fire effects are temperature, heat flux, time to failure of a 

damage target, rate of flame spread over a fuel package, heat release rate for a burning 

material, and smoke density. 

The ASME/ANS PRA Standard (Ref. 2) defines the term fire analysis tool as “any method 

used to estimate or calculate one or more physical fire effects (e.g., temperature, heat flux, 

time to failure of a damage target, rate of flame spread over a fuel package, heat release rate 

for a burning material, smoke density, etc.) based on a predefined set of input parameter 

values as defined by the fire scenario being analyzed. Fire analysis tools include, but are not 

limited to, computerized compartment fire models, closed-form analytical formulations, 

empirical correlations such as those provided in a handbook, and lookup tables that relate 

input parameters to a predicted output.” 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Fire Area 

An area enclosed by 

rated fire barriers 

capable of preventing 

In a fire PRA, the spread of fire and fire effects is limited (reduced probability of propagation) 

across fire areas. A multicompartment fire analysis is done across fire areas to evaluate the 

risk significance of these fire scenarios. 
APPENDIX A 

A-6 

TERM AND 

DEFINITION (S) 

DISCUSSION 

or inhibiting spread of 

fires to and from the 

outside. (see Fire 

Barrier)

A fire area must be made up of rated fire barriers with openings in the barriers provided with 

fire doors, fire dampers, and fire penetration seal assemblies with a fire resistance rating at 

least equivalent to the barrier in which it exists (e.g., this term is defined in the analysis in 

Appendix R to 10 CFR Part 50 (Ref. 24)). Fire areas tend to confine most fires within the 

area. In a PRA, the fire area concept may simplify analysis, as each fire area generally may 

be treated independently from others. Fires may spread from one area to the next should a 

portion of the barrier be defeated (e.g., fire door left open). 

Regulatory Guide 1.189 (Ref. 90) defines the term fire area as “the portion of a building or 

plant that is separated from other areas by rated fire barriers adequate for the fire hazard.” 

Fire Barrier 

A component intended 

to impede spreading of 

a fire and its effects. 

(see Passive Fire 

Barrier, Active Fire 

Barrier)

In a fire PRA, fire barriers are modeled to prevent or reduce the spread of fires between fire 

areas. Therefore, fire barriers reduce the probability of damage to safety-related equipment 

in adjacent areas, and thus reduce the frequency of undesired end states. Fire barriers can 

be active, indicating the barrier requires some physical repositioning to function, or passive, 

indicating the barrier provides protection in its normal orientation. 

Certification of a fire barrier’s fire resistance endurance rating typically is based on 

standardized tests, such as the American Society of Testing and Materials (ASTM) Standard 

E-119. Examples of solid construction made of fire-resistant material could be a wall or door.

NUREG/CR-6850 (Ref. 79) defines the term fire barrier as “components of construction 

(walls, floors, and their supports), including beams, joists, columns, penetration seals or 

closures, fire doors, and fire dampers that are rated by approving laboratories in hours of 

resistance to fire, that are used to prevent the spread of fire and restrict spread of heat and 

smoke.” 

Fire Compartment 

A subdivision of a 

building or plant that is 

a well-defined enclosed 

room, not necessarily 

bounded by rated fire 

barriers, which 

essentially confines the 

fire. 

In a fire PRA, fire compartments are modeled because they reduce the probability of fire 

spread across boundaries. Boundaries of a fire compartment may have open equipment 

hatches, stairways, doorways, or unsealed penetrations. 

As discussed in the ASME/ANS PRA Standard (Ref. 2), “a fire compartment generally falls 

within a fire area and is bounded by noncombustible barriers where heat and products of 

combustion from a fire within the enclosure will be substantially confined.” 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Fire Control

The stage of firefighting 

in which a fire incident 

is controlled and not 

allowed to escalate in 

magnitude. 

In current fire PRA practice, the concept of fire control generally is not used because there is 

large uncertainty associated with declaring when a fire has been brought under control as 

opposed to having been fully extinguished. Also, fire control is not modeled in fire models. 

Fire control can be achieved by water-based fixed systems or through the application of 

other fire suppression means (e.g., hose streams, portable extinguishers). Furthermore, 

gaseous fixed systems can prevent fire damage from extending beyond the locations 

damaged when the system is actuated. The concept of fire control may also include 

managed fire burnout whereby a fire is allowed to continue burning until the fuel source is 

exhausted (e.g., in the case of a leak of flammable compressed gases such as hydrogen). 

The definition provided was based on the definition in NUREG-1805 (Ref. 60). 
APPENDIX A 

A-7 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Fire Event 

A particular case where 

a fire has occurred in a 

nuclear power plant. 

Fire events are characterized in the fire events database. A fire event is described by its 

initiation, the progression of the fire, detection and suppression, and the impact on plant 

systems. 

Fire Events Database 

A collection of fire 

events that indicates 

characteristics of the 

fire and response by 

fire protection systems 

and plant personnel as 

well as the impact of 

the fire on plant 

equipment and 

operations. 

In a fire PRA, the fire events database is used to provide raw data to calculate fire ignition 

frequencies and manual suppression reliability for different types of fires. 

Fire Extinguishment 

The stage of a fire 

when combustible 

materials are no longer 

burning. 

In a fire PRA, fire extinguishment concludes the duration of a fire and implies that all burning 

materials have been fully suppressed. Fire damage generally is modeled in fire PRA until 

fire extinguishment. 

Fire Hazard Analysis 

An analysis to evaluate 

potential fire sources 

and combustibles, and 

appropriate fire 

protection systems, and 

features used to 

mitigate the effects. 

Fire hazards analyses are generally of a qualitative or semi-quantitative nature as compared 

to a quantitative PRA. 

Regulatory Guide 1.189 (Ref. 90) defines fire hazard analysis as “an analysis used to 

evaluate the capability of a nuclear power plant to perform safe-shutdown functions and 

minimize radioactive releases to the environment in the event of a fire. The analysis includes 

the following features: identification of fixed and transient fire hazards; identification and 

evaluation of fire prevention and protection measures relative to the identified hazards; 

evaluation of the impact of fire in any plant area on the ability to safely shut down the reactor 

and maintain shutdown conditions, as well as to minimize and control the release of 

radioactive material.” 

The definition provided was based on the definition in the NFPA 805 Standard (Ref. 11). 

Fire Human Reliability Analysis 

A structured approach 

used to identify 

potential human error 

events that may occur 

in a sequence of events 

following a fire and to 

systematically estimate 

the probability of those 

errors using data, 

models, or expert 

judgment as applied to 

a fire. 

Fire human reliability analysis is used to quantify the potential impact of fire-generated 

environmental effects and stressors on human performance and the likelihood that errors 

might occur during execution of fire response procedures for specific areas of the plant, 

including control room evacuation. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 
APPENDIX A 

A-8 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Fire Ignition Frequency 

Frequency of fire 

occurrence generally 

expressed as fire 

ignitions per reactoryear. 

In a fire PRA, fire ignition frequency is normally calculated based on fires events that have 

the potential to cause damage to targets outside the ignition source. Fire ignition frequency 

is the factor that, in quantification, introduces the frequency element into the fire-induced 

core damage frequency. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2).

Fire-Induced Initiating Event 

The initiating event 

assigned to occur in the 

fire PRA plant response 

model for a given fire 

scenario. (see Fire 

Plant Response Model)

 

The term initiating event is defined in the exact same context as is used in internal events 

PRA. That is, the initiating event is not the fire, it is induced by the fire. For example, a fire 

affects a pilot operated relief valve (PORV) control cable, causing spurious operation of a 

PORV, and thus an initiating event. 

Fire-induced initiating events trigger sequences of events that challenge plant control and 

safety systems whose failure potentially could lead to core damage or large early release. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Fire Model 

A mathematical 

prediction of fire 

growth, environmental 

conditions, and 

potential effects on 

structures, systems, or 

components based on 

the conservation 

equations or empirical 

data. 

The American Society of Testing and Materials (ASTM) Standard E176-10a, “Standard 

Terminology of Fire Standards” (Ref. 3), defines fire model as “a physical representation or 

set of mathematical equations that approximately simulate the dynamics of burning and 

associated processes.” 

The definition provided was based on the definition in the NFPA 805 Standard (Ref. 11). 

Fire Plant Response Model 

A representation of a 

combination of 

equipment, cable, 

circuit, and system 

function, and operator 

failures or successes, 

of an accident that 

when combined with a 

fire-induced initiating 

event can lead to 

undesired 

consequences, with a 

specified end state 

(e.g., core damage or 

large early release). 

In a fire PRA, the fire plant response model contains the event trees and fault trees that will 

be used to analyze fire-induced initiating events. Given a fire scenario leading to fire-induced 

failure of a fire damage target set, a plant damage state (fire-induced damage to plant 

systems and components including equipment failure modes) is defined and incorporated 

into the fire plant response model. The event tree/fault tree models are then manipulated to 

depict the logical relationships among equipment failures (both random and fire-induced) and 

human failure events. As in internal events, the fire plant response model estimates the 

conditional core damage probability given loss of a fire damage target set. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Fire Plume 

Buoyant stream of hot 

gases rising above a 

localized area 

undergoing combustion 

Typically, a fire plume will form above a burning object. The fire plume will rise until 

obstructed by a horizontal surface, such as a ceiling. Upon hitting the ceiling, the hot gases 

in the fire plume will turn and flow along the ceiling in the form of a ceiling jet. When the 

ceiling jet gases are blocked by vertical surfaces, such as walls, they will accumulate into a 
APPENDIX A 

A-9 

TERM AND 

DEFINITION (S) 

DISCUSSION 

into surrounding space 

of essentially 

uncontaminated air. 

hot gas layer or smoke layer. As more hot gas accumulates in the layer, the interface 

between the hot gas layer and cooler layer below will continue to drop toward the floor of the 

enclosure. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79).

Fire PRA, Fire Probabilistic Safety Assessment 

An approach to 

quantitatively evaluate 

the risk from hazards 

associated with a fire. 

(see Main Glossary: 

PRA)

This quantitative approach consists of fire ignition frequencies, the associated initiating event 

produced by the ignition, the probability of fire damage from those ignition sources, and the 

resulting impact on the plant. 

The term probabilistic safety assessment is another term that can be used interchangeably 

and correctly with PRA. Typically, the term probabilistic safety assessment is used 

internationally. 

Fire Prevention 

Measures directed 

toward reducing the 

likelihood of fire. 

Fire prevention is not generally modeled in fire PRA, although it is reflected in fire ignition 

frequency. Lower fire frequencies could be due, at least in part, to an effective fire 

prevention program. 

The definition provided was based on the definition in the NFPA 805 Standard (Ref. 11). 

Fire Probabilistic Safety Assessment 

 (see Fire PRA) The term fire probabilistic safety assessment has the same meaning as fire PRA and is 

defined under “Fire PRA.” 

Fire Protection Defense-In-Depth 

The principle of 

providing multiple and 

diverse fire protection 

systems and features. 

Fire protection defense-in-depth is modeled explicitly in fire PRA. In particular, fire PRA will 

credit defense-in-depth fire protection measures and will predict the likelihood that those 

measures fail to prevent fire-induced damage to plant equipment and cables. 

The fire protection defense-in-depth objectives, as indicated in Appendix R to 10 CFR Part 

50, (Ref. 24), are “(1) to prevent fires from starting; (2) to detect rapidly, control, and 

extinguish promptly those fires that do occur; and (3) to provide protection for structures, 

systems, and components important to safety so that a fire that is not promptly extinguished 

by the fire suppression activities will not prevent the safe shutdown of the plant.” Multiple 

and diverse fire protection systems and features attain these objectives. 

Fire Protection Design Elements 

Any aspect of the fire 

protection program 

supported by specific 

Fire protection design elements can include active fire protection systems such as sprinkler 

or smoke detector systems, passive systems such as electrical raceway fire barriers, and 

programmatic elements. 
APPENDIX A 

A-10 

TERM AND 

DEFINITION (S) 

DISCUSSION 

design requirements 

and/or analyses. 

The definition provided was based on the definition in the ASME/ANS PRA Standard 

(Ref. 2). 

Fire Protection Feature 

Administrative controls, 

emergency lighting, fire 

barriers, fire detection 

and suppression 

systems, fire brigade 

personnel, and other 

features provided for 

fire protection 

purposes. 

In a fire PRA, fire protection features would be credited in accident sequences in which a fire 

endangers stable operation of the plant. Fire protection features are important to risk 

because they reduce damage due to fire and thus the frequency of accidents with undesired 

consequences because of fires. 

The definition provided was based on the definition in Regulatory Guide 1.189 (Ref. 90). 

Fire Protection Program 

The integrated effort 

involving equipment, 

procedures, and 

personnel used in 

carrying out all 

activities of fire 

protection. 

The ASME/ANS PRA Standard (Ref. 2) states that the fire protection program includes 

“system and facility design, fire prevention, fire detection, annunciation, confinement, 

suppression, administrative controls, fire brigade organization, inspection and maintenance, 

training, quality assurance, and testing.” 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Fire Protection Program Element 

Any specific aspect or 

provision included as a 

part of the fire 

protection program. 

As described in the ASME/ANS Standard (Ref. 2), fire protection program elements include 

“system and facility design, fire prevention, fire detection, annunciation, confinement, 

suppression, administrative controls, fire brigade organization, inspection and maintenance, 

training, quality assurance, and testing.” 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Fire Protection System 

Fire detection, 

notification, and fire 

suppression systems 

designed, installed, and 

maintained in 

accordance with the 

applicable National Fire 

Protection Association 

codes and standards. 

Fire protection systems are systems installed to provide detection, warning, or suppression 

of fires. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Fire Response Procedure 

A procedure 

established for 

operators to respond to 

a fire. 

An example of a fire response procedure is to evacuate the control room when certain 

environmental conditions are reached due to a control room fire. 

Specific facilities may have alternate names for the fire response procedures such as fire 

emergency procedures, pre-fire plans, or emergency response procedures. The fire 

response procedures also may be embedded within a more general set of emergency 

operating procedures designed to deal with a range of potential off-normal plant operating 

states, including fires. 
APPENDIX A 

A-11 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Fire Risk Analysis 

 (see Fire PRA) The term fire risk analysis has the same meaning as fire PRA and is defined under “Fire 

PRA.”

Fire Safe-Shutdown Analysis 

The deterministic 

process or method 

conducted to identify 

and evaluate the 

capability of structures, 

systems, and 

components necessary 

to accomplish and 

maintain safe shutdown 

conditions in the event 

of a fire. 

Fire safe shutdown analysis is conducted based on a fire scenario in fire PRA and affects the 

plant response mode. 

For fire events, safe shutdown are those plant conditions specified in the plant technical 

specifications as hot standby, hot shutdown, or cold shutdown. 

The definition provided was based on the definition in Regulatory Guide 1.189 (Ref. 90). 

Fire Scenario 

A set of elements that 

describe a fire event.

A fire scenario includes a description of the fire and any factors affecting it from ignition to 

suppression. As a result, the fire scenario describes the progression of the fire from ignition 

to damage in the fire PRA. 

The ASME/ANS Standard (Ref. 2) states that the elements of a fire scenario include “a 

physical analysis unit, a source fire location and characteristics, detection and suppression 

features to be considered, damage targets, and intervening combustibles.” 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Fire Suppression 

The process of 

controlling and 

ultimately extinguishing 

fires. 

In fire PRA, fire suppression is a process, but successful completion of that process implies 

fire extinguishment, which represents the termination of the fire itself. An accident sequence 

caused by the fire may continue beyond extinguishment of the fire. Traditional fire protection 

definitions refer to fire suppression as controlling and extinguishing fires, which is consistent 

with the term as applied in fire PRA. 

Fire suppression can be either manual or automatic. Manual fire suppression is the use of 

hoses, portable extinguishers, or manually actuated fixed suppression systems by plant 

personnel. Automatic fire suppression is the use of automatic fixed systems, such as 

sprinkler, Halon, and CO2 systems. 

Manual fire suppression is modeled as a time-dependent activity in fire PRA, occurring at 

potentially different times in the scenario, in which automatic fixed suppression is modeled as 

occurring early in the scenario and often can be treated as time-independent. 

Fire Suppression System 

Typically, permanently 

installed fire protection 

systems provided for 

the express purpose of 

suppressing fires. 

In a fire PRA, the effectiveness of the fire suppression system is an important consideration, 

in addition to the system availability and reliability. The ASME/ANS Standard (Ref. 2) states 

that a fire suppression system “may be either automatically or manually actuated. However, 

once activated, the system should perform its design function with little or no manual 

intervention.” 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 
APPENDIX A 

A-12 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Fire Wrap 

A localized protective 

covering designed to 

protect cables, cable 

raceways, or other 

equipment from 

fire-induced damage. 

Fire wrap, used to protect against thermal damage, is the common term usually used to 

denote a type of electronic raceway fire barrier system. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Fire Zone 

1. Subdivisions of a fire 

area not necessarily 

bounded by fire rated 

assemblies. 

2. Subdivisions of a fire 

detection or 

suppression systems, 

which provide alarm 

indications at the 

central alarm panel. 

The term fire zone is not widely used in current fire PRA practice but, when used, can have 

different meanings. A fire zone may be a loosely defined spatial area such as a partially 

enclosed space within a larger fire compartment or fire area (per definition (1)). The term 

also may be used in the more traditional context of a zone of coverage for fixed fire 

protection features such as fire detection and fire suppression (per definition (2)). The term 

fire zone may also be encountered in older fire PRAs in which terminology was as yet 

unsettled. That is, some older fire PRAs may use the term fire zone in the same context that 

the ASME/ANS Standard uses the term physical analysis unit. 

The definition provided was based on the definition in the NFPA 805 Standard (Ref. 11). 

Fire-Resistance Rating 

The time that 

materials or 

assemblies have 

withstood a fire 

exposure as 

established in 

accordance with an 

approved test 

procedure appropriate 

for the structure, 

building material, or 

component under 

consideration. 

In a fire PRA, the greater the fire-resistance rating, the longer time to damage is modeled. 

American Society of Testing and Materials (ASTM) Standard E-119 is the test standard for 

determining fire resistance. The fire-resistance rating is provided in units of minutes or 

hours. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Flame Spread Rating 

A relative 

measurement of the 

surface burning 

characteristics of 

building materials. 

The flame spread rating is tested in accordance with NFPA 255, “Standard Method of Test 

Surface Burning Characteristics of Building Materials” (Ref. 10). 

The definition provided was based on the definition in the NFPA 805 Standard (Ref. 11). 

Free of Fire Damage 

The structure, system, 

or component under 

consideration remains 

capable of performing 

its intended function 

during and after the 

postulated fire. 

A component free of fire damage in the fire PRA model is given full credit to performing its 

function. 

The definition provided was based on the definition in the NFPA 805 Standard (Ref. 11). 
APPENDIX A 

A-13 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Ground Fault 

A type of short circuit 

involving an abnormal 

connection between a 

conductor and a 

grounded conducting 

medium. 

NUREG/CR-6850 (Ref. 79) describes a ground fault as being characterized by “an abnormal 

current surge (fault current) attributable to the lack of any significant circuit burden (i.e., load). 

A ground fault should trigger over-current protective action for a properly designed circuit.” 

As used in the definition, the grounded conducting medium refers to any conduction path 

associated with the reference ground of the circuit. This might include structural elements 

(e.g., tray, conduit, enclosures, metal beams) or intentionally grounded conductors of the 

circuit (neutral conductor). 

The term ground fault is used interchangeably and correctly with the term short-to-ground. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Heat Release Rate 

The amount of heat 

generated by a burning 

object per unit time. 

The heat release rate (HRR) is the key driver in determining the extent of damage in a fire 

scenario and is usually expressed in units of kW. An example of an HRR can be found in an 

HRR profile. An HRR profile refers to the behavior of the HRR as a function of time (an HRR 

versus time plot). For example, a fire with a constant HRR has an intensity that does not 

change. 

The American Society of Testing and Materials (ASTM) Standard E176-10a, “Standard 

Terminology of Fire Standards” (Ref. 3), defines heat release rate as “the thermal energy 

released per unit time by an item during combustion under specified conditions.” The 

following figure represents an HRR curve. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref.79). 

High-Energy Arcing Fault 

A high-current, 

electrical fault that 

produces an energetic 

discharge of electrical 

and thermal energy and 

may be followed by a 

fire. 

High-energy arcing faults are unique in fire PRA since damage is assumed to occur 

instantaneously to targets, regardless of the potential presence of a fixed suppression 

system. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

High-Hazard Fire Source 

A fire source that can 

lead to fires of a 

particularly severe and 

challenging nature. 

In a fire PRA, high-hazard fire sources may cause extensive damage, potentially including 

the failure of structural elements such as steel, which is mapped into failures of equipment. 

Examples of high-hazard fire sources include catastrophic failure of an oil-filled transformer, 

an unconfined release of flammable or combustible liquid, leaks from a pressurized system 
APPENDIX A 

A-14 

TERM AND 

DEFINITION (S) 

DISCUSSION 

containing flammable or combustible liquids, and significant releases or leakage of hydrogen 

or other flammable gases (Ref. 2). 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

High-Low Pressure Interface 

Interface between the 

reactor coolant system 

and lower-pressure 

systems. 

In a fire PRA, regulations stipulate that at least one isolation valve at the interface of highand low-pressure systems must remain closed despite any damage that may be caused by 

fire. 

Hot Gas Layer 

The volume under the 

ceiling of a fire 

enclosure where smoke 

accumulates and high 

gas temperatures are 

observed. 

Typically, a fire plume will form above a burning object. The fire plume will rise until 

obstructed by a horizontal surface, such as a ceiling. Upon hitting the ceiling, the hot gases 

in the fire plume will turn and flow along the ceiling in the form of a ceiling jet. When the 

ceiling jet gases are blocked by vertical surfaces, such as walls, they will accumulate into a 

hot gas layer or smoke layer. As more hot gas accumulates in the layer, the interface 

between the hot gas layer and cooler layer below will continue to drop toward the floor of the 

enclosure. Hot gas layer is the upper zone in a two-zone fire model formulation. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Hot Short 

The condition in which 

individual conductors of 

the same or different 

cables come in contact 

with each other. At 

least one of the 

conductors involved in 

the shorting is 

energized, resulting in 

an impressed voltage 

or current on the circuit 

being analyzed. 

In a fire PRA, a hot short can cause a spurious operation, which is one possible failure mode 

considered in the accident sequence model. Hot shorts also can cause misleading 

instrumentation and indication signals. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Ignition Source 

A piece of equipment or 

activity that causes a 

fire. 

Ignition source is the first link to an accident sequence caused by fire. A fire started by an 

ignition source may damage equipment, causing an initiating event, and possibly damaging 

safety systems required for response. 
APPENDIX A 

A-15 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Fixed ignition sources are permanently installed, and transient ignition sources are 

temporarily located. Examples of transient ignition sources are a welder or grinder being 

used for hot work. Examples of fixed ignition sources are switchgear cabinets, transformers, 

pumps, and cables. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 2).

Intercable Short Circuit 

Electrical contact 

between individual 

conductors in two or 

more separate cables 

due to damaged 

insulation and cable 

wrapping. (see 

Intracable Short Circuit)

As analyzed in a PRA, an intercable short circuit may lead to any one of several possible 

conductor fault modes including hot shorts and ground faults. Such faults may disable 

safety-related systems, cause the spurious operation of plant components, and may lead to 

or contribute to an accident sequence. An intercable short circuit may be caused by fireinduced damage to grouped electrical cables. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Internal Fire 

A hazard group in 

which a fire occurs from 

within the plant that is 

evaluated in fire PRA. 

For fire PRA, the phrase within the plant as used in this definition is any location that lies 

within the global analysis boundary as defined by the plant partitioning technical element 

under Part 4 of the ASME/ANS Standard (Ref. 2). Examples of internal fires are fires that 

occur in the confines of the plant, including any buildings associated with plant operations, 

the switchyard, transformer yard, and service water supply. Forest fires are classified as 

external fires. 

Internal Hot Short 

A hot short in which 

both the source 

conductor and target 

conductor are in the 

same multi-conductor 

cable. (see Hot Short, 

Intracable Short Circuit) 

Internal hot shorts have greater probabilities of occurrence than external hot shorts. The 

term internal hot short can be used interchangeably and correctly with intracable short circuit, 

which is also referred to as intracable conductor-to-conductor short circuit. 

 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Intervening Combustibles 

Materials that may burn 

but are not ignition 

sources. 

The fire scenario becomes more extensive in the presence of intervening combustibles. This 

is because intervening combustibles, located between the ignition source and target, 

contribute to fire propagation along this path. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Intracable Short Circuit 

Electrical contact 

between individual 

conductors in a cable 

due to damaged 

insulation between the 

conductors. (see 

Intercable Short Circuit)

As analyzed in a PRA, intractable short circuits may lead to any of the defined cable and 

circuit failure modes, including hot shorts and ground faults. Such faults may cause the 

spurious operation of plant components, disable safety-related systems, and lead to or 

contribute to an accident sequence. Intracable short circuits may occur because of a fire 

damaging insulation between the conductors of any multi-conductor cable, or they may occur 

because of insulation faults. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 
APPENDIX A 

A-16 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Limiting Fire Scenario 

Fire scenario(s) in 

which one or more of 

the inputs to the fire 

modeling calculation 

are varied to the point 

that particular 

equipment is failed. 

The intent of the limiting fire scenario is to determine that there is a reasonable margin 

between the expected fire scenario conditions and the point of this failure. Examples of fire 

modeling inputs that could be varied include heat, release rate, initiation location, or 

ventilation rate. 

The definition provided was based on the definition in the NFPA 805 Standard (Ref. 11). 

Maximum Expected Fire Scenario 

Scenarios that 

represent the most 

challenging fire that 

could be reasonably 

anticipated for the 

occupancy type and 

conditions in the space. 

Maximum expected fire scenario is a term for an analysis in the fire modeling track of NFPA 

805 and is not specifically related to fire PRA. Maximum expected fire scenarios can be 

based on industry experience using plant-specific conditions and fire experience (Ref. 11). 

The definition provided was based on the definition in the NFPA 805 Standard (Ref. 11). 

Multiple Spurious Operations 

Concurrent spurious 

operations of two or 

more equipment items. 

(see Concurrent Hot 

Shorts) 

Multiple spurious operations may cause multiple equipment failures and complicate operator 

actions in a fire accident sequence in comparison to single spurious operations. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Natural Ventilation 

The condition in which 

gas flows into or out of 

the room because of 

density differences 

between the fluids. 

Ventilation (supplying fresh air) may cause the fire to burn more intensely, while at the same 

time potentially removing part of the hot gas layer. Therefore, ventilation may affect the 

probability of damage to equipment, given a fire in a certain location. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Open Circuit 

A loss of electrical 

continuity in an 

electrical circuit, either 

intentional or 

unintentional. 

In a fire PRA, open circuits will cause the associated electrical equipment to be inoperable. 

This may increase the probability of system failures and probabilities of relevant accident 

sequences. Open circuits could result from a loss of conductor continuity or from the 

triggering of circuit protection devices such as a blown fuse or open circuit breaker, or 

because of a loss of physical continuity in one or more cable conductors (Ref. 79). 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Passive Fire Barriers 

A fire barrier that 

provides its protective 

function while in its 

normal orientation, 

without any need to be 

repositioned. 

In a fire PRA, fire barriers impede the spread of fires and limit potential damage to safety 

equipment, thereby reducing probabilities of fire spread to additional components and the 

probability of accident sequences. Walls and normally closed fire doors are examples of 

passive fire barriers. 

The definition provided was based on the definition in NUREG-1805 (Ref. 60). 

Physical Analysis Unit

A spatial subdivision of 

the plant on which the 

fire PRA is based. 

In a fire PRA, the physical analysis units are the fundamental spatial element considered as 

being affected by fires. While the fire PRA will include consideration of fires affecting more 

than one physical analysis unit at a time (the multicompartment analysis), most fire scenarios 
APPENDIX A 

A-17 

TERM AND 

DEFINITION (S) 

DISCUSSION 

are assumed to remain confined to one physical analysis unit. Physical analysis units 

usually are based on fire areas or fire compartments, but they also may be based on factors 

such as spatial separation (as opposed to physical barriers), nonrated partitioning elements, 

and active fire barrier systems (e.g., a water curtain). Since a physical analysis unit 

substantially contains the effects of a fire, it generally reduces the probability of additional 

component damage. 

This term was coined in relation to the fire portion of the ASME/ANS PRA Standard to refer 

generally to fire compartments, fire zones, and fire areas. 

The definition provided was based on the definition in the ASME/ANS PRA Standard 

(Ref. 2).

Probability of Nonsuppression 

Probability of failing to 

suppress a fire before 

target damage occurs. 

In a fire PRA, probability of nonsuppression is used to calculate the probability of target 

damage (and, consequently, probability of component or system failure), given a fire of a 

certain intensity in a certain location. Probability of nonsuppression depends on the 

characteristics of the fire, fire suppression method, and the time available until target 

damage. 

The definition provided was based on the definition in the ASME/ANS PRA Standard 

(Ref. 2). 

Qualified Cable 

A cable that has been 

tested and certified as 

meeting all aspects of 

IEEE-383 standard 

including both the 

equipment qualification 

and flame spread 

elements. 

The IEEE-383 standard primarily deals with the equipment qualification issues of cable aging 

and severe accident environmental exposures. The standard also includes a vertical flame 

spread test. In practice, cables that have been only tested against the flame spread portion 

of the standard, but have not been subjected to the equipment qualification elements, may 

be referred to as low flame spread cables, but they would not be considered fully qualified. 

A cable that does not meet this criterion is referred to as unqualified or nonqualified. 

Raceway 

An enclosed channel of 

metallic or nonmetallic 

materials designed 

expressly for holding 

wires, cables, or bus 

bars, with additional 

functions as permitted 

by code. 

In a fire PRA, generally all cables in a raceway are affected equally by the modeled fire. 

Open cable trays (e.g., ladder style trays) also are referred to as raceways. 

The ASME/ANS PRA Standard (Ref. 2) states that raceways include, but are not limited to, 

“rigid metal conduit, rigid nonmetallic conduit, intermediate metal conduit, liquid-tight flexible 

conduit, flexible metallic tubing, flexible metal conduit, electrical nonmetallic tubing, electrical 

metallic tubing, underfloor raceways, cellular concrete floor raceways, cellular metal floor 

raceways, surface raceways, wireways, and busways.” 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Short Circuit 

An abnormal 

connection (including 

an arc) of relatively low 

impedance between 

two conductors or 

points of different 

potential. 

With regard to control circuit failures, short circuits could involve a ground fault or hot short. 

Either may cause disablement or undesired operation of safety-related equipment and 

contribute to initiation or propagation of an accident sequence. Short circuits also can cause 

the failure or maloperation of the indication elements of a control circuit, instrument circuits, 

and power circuits. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 
APPENDIX A 

A-18 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Short-to-Ground 

A type of short circuit 

involving an abnormal 

connection between a 

conductor and a 

grounded conducting 

medium. 

NUREG/CR-6850 (Ref. 79) describes a ground fault as being characterized by “an abnormal 

current surge (fault current) attributable to the lack of any significant circuit burden (i.e., load). 

A ground fault should trigger over-current protective action for a properly designed circuit.” 

As used in the definition, the grounded conducting medium refers to any conduction path 

associated with the reference ground of the circuit. This might include structural elements 

(e.g., tray, conduit, enclosures, metal beams) or intentionally grounded conductors of the 

circuit (neutral conductor). The term short-to-ground is used interchangeably and correctly 

with the term ground fault. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Smoke Layer 

The volume under the 

ceiling of a fire 

enclosure where smoke 

accumulates and high 

gas temperatures are 

observed. (see Upper 

Layer, Hot Gas Layer)

Typically, a fire plume will form above a burning object. The fire plume will rise until 

obstructed by a horizontal surface, such as a ceiling. Upon hitting the ceiling, the hot gases 

in the fire plume will turn and flow along the ceiling in the form of a ceiling jet. When the 

ceiling jet gases are blocked by vertical surfaces, such as walls, they will accumulate into a 

hot gas layer or smoke layer. As more hot gas accumulates in the layer, the interface 

between the hot gas layer and cooler layer below will continue to drop toward the floor of the 

enclosure. The smoke layer is the upper zone in a two-zone model formulation. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79). 

Spurious Operation 

The undesired 

operation of equipment 

resulting from a fire that 

could affect the 

capability to achieve 

and maintain safe 

shutdown. 

Spurious operation results from a hot short and may result in undesired change of state or 

disablement of safety-related equipment, thereby resulting in initiation of an accident 

sequence or damage to a component within the accident sequence. In some cases, ground 

faults or open circuits also may cause spurious operation, depending on the specific circuit 

design. 

The definition provided was based on the definition in the ASME/ANS PRA Standard (Ref. 

2). 

Transient Combustible 

Combustible materials 

placed in a temporary 

location. 

In a fire PRA, a transient combustible is one of many potential ignition sources. As 

discussed in NUREG/CR-6850 (Ref. 79), transient combustibles “are usually associated with 

(but not limited to) maintenance or modifications involving combustible and flammable 

liquids, wood and plastic products, waste, scrap, rags, or other combustibles resulting from 

the work activity.” 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79).
APPENDIX A 

A-19 

TERM AND 

DEFINITION (S) 

DISCUSSION 

Upper Layer 

The volume under the 

ceiling of a fire 

enclosure where smoke 

accumulates and high 

gas temperatures are 

observed. (see Smoke 

Layer, Hot Gas Layer)

Typically, a fire plume will form above a burning object. The fire plume will rise until 

obstructed by a horizontal surface, such as a ceiling. Upon hitting the ceiling, the hot gases 

in the fire plume will turn and flow along the ceiling in the form of a ceiling jet. When the 

ceiling jet gases are blocked by vertical surfaces, such as walls, they will accumulate into a 

hot gas layer or smoke layer. As more hot gas accumulates in the layer, the interface 

between the hot gas layer and cooler layer below will continue to drop toward the floor of the 

enclosure. The smoke layer is the upper zone in a two-zone model formulation. 

The definition provided was based on the definition in NUREG/CR-6850 (Ref. 79).

Ventilation Rate 

Amount of air injected 

or extracted by a 

mechanical ventilation 

system into or from a 

location, respectively. 

The ventilation rate is usually measured in cubic meters per second (m3/sec). 

Zone Model 

A type of fire model that 

provides a method for 

calculating fire 

environment conditions 

in control volumes, or 

zones, within a space 

by applying 

conservation equations 

and the ideal gas law. 

The fundamental idea behind a zone model is that each zone is well-mixed and that all fire 

environment variables (e.g., temperature, smoke concentration), therefore, are uniform 

throughout the zone. The variables in each zone change as a function of time and rely on 

the initial conditions that the user specifies. It is assumed that there is a well-defined 

boundary separating the two zones, though this boundary may move up or down throughout 

the simulation. 

Zone models can easily analyze conditions resulting from fires involving single compartments 

or compartments with adjacent spaces, and they are often used to compute the hot gas layer 

temperature, hot gas layer composition, and target heat fluxes. Zone models also are 

capable of modeling some effects of natural and mechanical ventilation in both horizontal 

and vertical directions. Smoke production, fire plume dynamics, ceiling jet characteristics, 

heat transfer, and ventilation flows are all algebraic models embedded within zone models. 

The definition provided was based on the definition in NUREG-1934 (Ref. 65). 

Zone of Influence 

That vicinity of the fire 

in which fire damage or 

fire spread to 

secondary 

combustibles is 

possible. 

Fire damage or spread may require some time to occur. The zone of influence is associated 

with the potential for fire damage or fire spread, regardless of the time available. Zone of 

influence generally does not encompass hot gas layer effects; instead, it focuses on direct 

radiant heating, plume, and ceiling jet effects. 

Typically a component is not damaged initially in the fire scenario if it is outside the zone of 

influence for an ignition source.

APPENDIX B 

B-1 

PRA TECHNICAL ELEMENTS 

Table B-1 provides the technical elements as defined in the ASME PRA Standard for Level 1, 

Level 2 and Level 3 PRA with the associated discussion. The technical elements are listed 

alphabetically by level of the PRA and hazard groups. 

 

Table B-1 Technical Elements and Discussion 

TECHNICAL 

ELEMENT DISCUSSION 

Level 1 Internal Events 

Accident 

Sequence 

Analysis 

The term accident sequence analysis is a technical element in the ASME/ANS PRA 

Standard whose objectives are to ensure that the response of the plant’s systems and 

operators to an initiating event is reflected in the assessment of CDF and LERF. 

Data Analysis The term data analysis is a Level 1 technical element in the ASME/ANS PRA Standard 

(Ref. 2) whose objectives are to provide estimates of the parameters used to determine the 

probabilities of the basic events representing equipment failures and unavailabilities 

modeled in the PRA. 

Human Reliability 

Analysis 

The term human reliability analysis is a Level 1 technical element in the ASME/ANS PRA 

Standard whose objective is to ensure that the impacts of plant personnel actions are 

reflected in the risk assessment. 

Initiating Event 

Analysis 

The term initiating event analysis is a technical element in the ASME/ANS PRA Standard 

(Ref.2) whose objective is to identify and quantify events that could lead to core damage. 

Large Early 

Release 

Frequency 

Analysis 

The term large early release frequency (LERF) analysis is a technical element of Part 2 of 

the ASME/ANS “Combined Standard: Requirements for Internal Events At-Power PRA.” 

The objectives of the LERF analysis element are to identify and quantify the contributors to 

large early releases based on the plant-specific core damage scenarios. 

Quantification The term quantification is a technical element in the ASME/ANS Level 1 PRA Standard 

(Ref. 2) whose objective is to provide an estimate of core damage frequency (and support 

the quantification of large early release frequency) based on the plant-specific core damage 

scenarios. 

Success Criteria The term accident success criteria is a technical element in the ASME/ANS PRA Standard 

whose objectives are to define the plant-specific measures of success and failure that 

support the other technical elements of the PRA.

Systems Analysis The term systems analysis is also a technical element in the ASME/ANS PRA Standard 

(Ref.2) whose objectives are to identify and quantify the causes of failure for each plant 

system represented in the initiating event analysis and accident sequence analysis. 

Level 1 Internal Flood At-Power 

Internal Flood 

Accident 

Sequences and 

Quantification 

The term internal flood accident sequences and quantification is a technical element in the 

ASME/ANS Level 1 PRA Standard (Ref. 2) whose objective is to quantify the core damage 

frequency and large early release frequency for the internal flood plant response 

sequences. 

Internal Flood 

Plant Partitioning 

The term internal flood plant partitioning is a technical element in the ASME/ANS Level 1 

PRA Standard whose objectives are to identify plant areas where internal floods could lead 

to core damage in such a way that plant-specific physical layouts and separations are 

accounted for. 
APPENDIX B 

B-2 

Internal Flood 

Scenarios 

The term internal flood scenarios is a technical element in the ASME/ANS Level 1 PRA 

Standard (Ref. 2) whose objective is to develop a set of internal flood scenarios relating 

flood source, propagation path(s), and affected equipment. 

Internal Flood 

Source 

Identification and 

Characterization 

The term internal flood source identification and characterization is a technical element in 

the ASME/ANS Level 1 PRA Standard (Ref. 2) whose objective is to identify the various 

sources of floods and equipment spray within the plant, along with the mechanisms 

resulting in flood or spray from the sources, and a characterization of the flood/spray 

sources is made. 

Internal FloodInduced Initiating 

Events 

The term internal flood-induced initiating events is a technical element in the ASME/ANS 

Level 1 PRA Standard (Ref. 2) whose objective is to determine the expected plant response 

to the selected set of flood scenarios, and an accident sequence from the internal event at 

power PRA that is reasonably representative of this response is selected for each scenario. 

Internal Fire At-Power

Circuit Failure 

Analysis 

The term circuit failure analysis is a technical element in the ASME/ANS Level 1 PRA 

Standard (Ref. 2) whose objectives are to treat fire-induced cable failures and their impact 

on the plant equipment, systems, and functions, and estimate the relative likelihood of 

various circuit failure modes. 

Fire Ignition 

Frequency 

The term fire ignition frequency is a technical element in the ASME/ANS Level 1 Internal 

Fire PRA Standard (Ref. 2) whose objective is to estimate the frequency of fires (expressed 

as fire ignitions per reactor-year). 

Fire PRA Cable 

Selection 

The term fire probabilistic risk assessment cable selection is a technical element in the 

ASME/ANS Level 1 Internal Fire PRA Standard (Ref. 2) whose objectives are to identify and 

locate cables required to support the operation of fire PRA equipment selected and cables 

whose failure could adversely affect credited systems and functions. 

Fire PRA 

Equipment 

Selection 

The term fire probabilistic risk assessment equipment selection is a technical element in the 

ASME/ANS Level 1 Internal Fire PRA Standard (Ref. 2) whose objective is to identify the 

set of plant equipment that will be included in the fire PRA. 

Fire PRA Plant 

Response Model 

The term fire probabilistic risk assessment plant response model is a technical element for 

internal fires in the ASME/ANS PRA Standard (Ref. 2) whose objective is to identify the 

initiating events that can be caused by a fire event and develop a related accident sequence 

model; and to depict the logical relationships among equipment failures (both random and 

fire-induced) and human failure events for core damage frequency and large early release 

frequency assessment when combined with the initiating event frequencies. 

Fire Risk 

Quantification 

The term fire risk quantification is a technical element in the ASME/ANS Level 1 Internal 

Fire PRA Standard (Ref. 2) whose objective is to quantify and present fire risk results. 

Fire Scenario 

Selection and 

Analysis 

The term fire scenario selection and analysis is a technical element in the ASME/ANS 

Level 1 Internal Fire PRA Standard (Ref. 2) whose objectives are to select a set of fire 

scenarios for each unscreened physical analysis unit upon which fire risk estimates will be 

based, characterize the selected fire scenarios, determine the likelihood and extent of 

risk-relevant fire damage for each select fire scenario, and examine multicompartment fire 

scenarios. 

Plant Boundary 

Definition and 

Partitioning 

The term plant boundary definition and partitioning is a technical element in the ASME/ANS 

PRA Standard (Ref. 2) for internal fire whose objective is to define the physical boundaries 

of the analysis and divide the various volumes within that boundary into physical analysis 

units. 

Post-Fire Human 

Reliability 

Analysis 

The term post-fire human reliability analysis is a technical element in the ASME/ANS PRA 

Standard (Ref. 2) whose objective is to consider the operator actions as needed for safe 

shutdown, including those called out in the relevant plant fire response procedures. 

Qualitative 

Screening 

The term fire probabilistic risk assessment cable selection is a technical element in the 

ASME/ANS Level 1 Internal PRA Standard whose objective is to identify physical analysis 
APPENDIX B 

B-3 

units whose potential fire risk contribution can be judged negligible without quantitative 

analysis

Quantitative 

Screening 

The term fire ignition frequency is a technical element in the ASME/ANS Level 1 Internal 

Fire PRA Standard (Ref. 2) whose objective is to screen physical analysis units from further 

consideration based on preliminary estimates of fire risk contribution and using established 

quantitative screening criteria.

Seismic/Fire 

Interactions 

The term seismic/fire interactions is a technical element in the ASME/ANS Level 1 PRA 

Standard (Ref. 2) whose objective is to provide a qualitative review of potential interactions 

between an earthquake and fire that might contribute to plant risk. 

Uncertainty and 

Sensitivity 

Analyses 

The term uncertainty and sensitivity analysis is a technical element in the ASME/ANS 

Level 1 Internal Fire PRA Standard (Ref. 2) whose objectives are the identification and 

treatment of uncertainties throughout the Fire PRA process. 

Seismic Events

Probabilistic 

Seismic Hazard 

Analysis 

The term probabilistic seismic hazard analysis is a technical element for seismic PRA in the 

ASME/ANS PRA Standard (Ref. 2) whose objective is to estimate the probability or 

frequency of exceeding different levels of vibratory ground motion.

Seismic Fragility 

Analysis 

The term seismic fragility analysis is a technical element for seismic PRA in the ASME/ANS 

PRA Standard (Ref. 2) whose objective is to determine the plant-specific failure probabilities 

of structures, systems, and components as a function of the seismic event intensity level, 

usually given in peak ground acceleration. 

Seismic Plant 

Response 

Analysis

The term seismic plant response analysis is a technical element in seismic PRA in the 

ASME/ANS PRA Standard (Ref. 2) whose objective is to develop a plant response model 

that addresses the initiating events and other failures resulting from the effects of the 

seismic hazard that can lead to core damage or large early release. The model usually is 

based on the internal events, at-power PRA model to incorporate those aspects that are 

different, because of the seismic hazard’s effects, from the corresponding aspects of the 

at-power, internal events model. 

High Winds

High Wind 

Fragility Analysis 

The term high wind fragility analysis is a technical element for high wind hazards in the 

ASME/ANS PRA Standard (Ref. 2) whose objective is to identify those structures, systems, 

and components susceptible to the effects of high winds and to determine their 

plant-specific failure probabilities as a function of the wind intensity.

High Wind Plant 

Response 

Analysis

The term high wind plant response analysis is a technical element for high winds PRA in the 

ASME/ANS PRA Standard (Ref. 2). The objective is: (1) to modify the internal events of 

the at-power PRA model to include the effects of high wind events in terms of the initiating 

events and failures induced, and (2) to exercise the resulting model to obtain quantitative 

results in terms of core damage frequency and large early release frequency. 

High Winds 

Hazard Analysis 

The term high winds hazard analysis is a technical element for high wind hazards in the 

ASME/ANS PRA Standard (Ref. 2) whose objective is to assess the frequency of 

occurrence of high wind as a function of intensity on a site-specific basis.

External Floods

External Flood 

Fragility Analysis 

The term external flood fragility analysis is a technical element for external floods in the 

ASME/ANS PRA Standard (Ref. 2) whose objective is to identify those structures, systems, 

and components susceptible to the effects of external floods and to determine their 

plant-specific failure probabilities as a function of the severity of the external flood. 

External Flood 

Hazard Analysis 

The term external flood hazard analysis is a technical element for external floods in the 

ASME/ANS PRA Standard (Ref. 2) whose objective is to assess the frequency of 

occurrence of external floods as a function of severity on a site-specific basis.
APPENDIX B 

B-4 

External Flood 

Plant Response 

Model and 

Quantification 

The term external flood plant response model and quantification is a technical element for 

external floods in the ASME/ANS PRA Standard (Ref. 2) whose objectives are to:

• develop an external flood plant response model by modifying the internal events 

at-power PRA model to include the effects of the external flood in terms of 

initiating events and failures caused;

• quantify this model to provide the conditional core damage probability (CCDP) 

and conditional large early release probability (CLERP) for each defined external 

flood plant damage state;

• evaluate the unconditional CDF and LERF by integrating the CCDP/CLERP with 

the frequencies of the plant damage states obtained by combining the external 

flood hazard analysis and external flood fragility analysis. 

Other External Hazards

External Hazard 

Analysis 

The term external hazard analysis is also a technical element for other external hazards in 

the ASME/ANS PRA Standard (Ref. 2) whose objective is to assess the frequency of 

occurrence of the external hazard as a function of intensity on a site-specific basis. 

External Hazard 

Fragility 

Evaluation/ 

Analysis 

The term external hazard fragility evaluation is also a technical element for other external 

hazards in the ASME/ANS PRA Standard (Ref. 2) whose objective is to identify those 

structures, systems, and components susceptible to the effects of the other external hazard 

and to determine their plant-specific failure probabilities as a function of the intensity of the 

hazard. 

External Hazard 

Plant Response 

Model/Analysis

The term external hazard plant response model is a technical element for other external 

hazards in the ASME/ANS PRA Standard (Ref. 2) whose objective is to develop a plant 

response model that addresses the initiating events and other failures resulting from the 

effects of the external hazard that can lead to core damage or large early release. The 

model is based on the internal events, at-power PRA model to incorporate those aspects 

that are different, because of the external hazard’s effects, from the corresponding aspects 

of the at-power, internal events model. 

Level 2

Containment 

Capacity Analysis 

The term containment capacity analysis is a technical element of a Level 2 PRA whose 

objective is to select an analysis method and calculate the ability of the containment to 

withstand challenges. 

Interface Between 

a Level 2 and 

Level 3 PRA 

The term interface between Level 2 and Level 3 PRA is a technical element of a Level 2 

PRA whose objectives are to provide clear traceability of the release category quantification 

back to the Level 2 analysis, to assure that initiating event information that could affect the 

Level 3 analysis is communicated, and to assure that all information required for the Level 3 

analysis is provided in suitable form. 

Level 1-2 Interface The term level 1-2 interface is a technical element of a Level 2 PRA whose objective is to 

consolidate or group accident sequences (or individual cutsets) from the Level 1 PRA in a 

way that reduces the number of unique scenarios for evaluation, but preserves initial and 

boundary conditions to the analysis of plant response (i.e., plant damage states or 

equivalent).

Probabilistic 

Treatment of Event 

Progression and 

Source Terms 

The term probabilistic treatment of event progression and source terms is a technical 

element of a Level 2 PRA whose objective is to establish a framework to support the 

systematic quantification of the potential severe accident sequences evolving from each 

Level 2 core damage sequence in sufficient detail. 

Radiological 

Source Term 

Analysis 

The term radiological source term analysis is a technical element in the draft Level 2 PRA 

whose objective is to develop a quantitative basis for associating a unique radiological 

source term to the environment for each accident progression sequence and release 

category. The metrics used to define a source term can vary, depending on the objective 

and intended application of the PRA. 
APPENDIX B 

B-5 

Severe Accident 

Progression 

Analysis 

The term severe accident progression analysis is a technical element of a Level 2 PRA 

whose objective is to generate a technical basis, rooted in realistic deterministic analysis for 

describing the chronology of postulated accident involving significant damage to reactor 

fuel, quantitatively characterizing thermal and mechanical challenges to engineered barriers 

to fission product release to the environment, and generating quantitative estimates of 

radioactive material release to the environment for accident sequences identified as 

contributors to the frequency of release. 

Level 3 PRA

Atmospheric 

Transport and 

Diffusion 

The term atmospheric transport and diffusion (ATD) is a technical element of a Level 3 PRA 

that refers to the process by which material that has been released from containment, 

moves through and spreads upon release to the atmosphere. The objective of ATD is to 

model the transport of radioactive material as it travels for many hours in the atmosphere 

under the meteorological conditions prevailing at and beyond the site that can change in 

both space and time. ATD models range from simple straight-line, steady-state Gaussian 

dispersion models that calculate ground-level instantaneous and time-integrated airborne 

concentrations in the plume, to more sophisticated models that allow terrain-dependent 

effects and temporal variations in wind speed and atmospheric stability. 

Probabilistic consequence modeling codes typically include sampling of meteorological data 

from a site-specific annual data base of hourly weather data to determine appropriately 

weighted scenarios of plume transport under different weather conditions to provide 

probabilistic results, model ATD for accident- and site-specific input parameters, 

accommodate temporal and spatial changes in meteorological conditions, calculate wet and 

dry deposition of particulate and halogen radionuclides, and document algorithms, 

assumptions, limitations, and uncertainties. 

Dosimetry The term dosimetry is a technical element in a Level 3 PRA whose objectives are to 

determine dose by including all applicable dose pathways such as cloudshine, groundshine, 

skin deposition, inhalation and ingestion; apply the effect of mitigation actions such as 

shielding; apply recognized dose conversion factors; and document assumptions, limitation 

and uncertainties associated with dosimetry. 

Economic 

Factors 

The term economic factor is a technical element in a Level 3 PRA whose objective is to 

determine the economic impacts of the release on the surrounding land and the population. 

Meteorological 

Data 

The term meteorological data is a technical element of a Level 3 PRA whose objective is to 

provide valid and representative meteorological data that are input into the atmospheric 

transport and dispersion codes, which provide the basis for consequences analysis 

calculations. 

Protective Action 

Parameters and 

Other Site Data 

The term protective action parameters and other site data is a technical element in a 

Level 3 PRA whose objectives are to model appropriate emergency response actions and 

protective actions; use appropriate site, local, and regional data; and document site-specific 

data, emergency response planning modeling, assumptions, limitations, and uncertainties. 

Quantification and 

Reporting 

The term quantification and reporting is a technical element of a Level 3 PRA whose 

objectives are to ensure that the Level 3 model executes properly, proves appropriate 

results, and is documented in a manner that facilitates risk assessments, PRA applications, 

upgrades and peer reviews. 

Risk Integration The term risk integration is a technical element of a Level 3 PRA whose objective is to 

combine the Level 3 analyses with the results from the Level 1-2 analyses to obtain a 

characterization of the overall risk, including uncertainty. 

Transition from 

the Radionuclide 

(Radioactive 

Material) Release 

to Level 3 

The term transition from radioactive material release to Level 3 is a technical element of a 

Level 3 PRA whose objectives are to provide clear traceability of the release category 

quantification back to the radioactive material release analysis, to ensure that initiating 

event information that could affect the Level 3 analysis is communicated, and to ensure that 

all information required for the Level 3 analysis is provided in suitable form.