IEC 63260:2020

IEC 63260 ®
Edition 1.0 2020-05
™
IEEE Std 1082
INTERNATIONAL
STANDARD
Guide for incorporating human reliability analysis into probabilistic risk
assessments for nuclear power generating stations and other nuclear facilities

IEC 60320:2020-05(en)  IEEE Std 1082-2017

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IEC 63260 ®
Edition 1.0 2020-05
IEEE Std 1082™
INTERNATIONAL
STANDARD
Guide for incorporating human reliability analysis into probabilistic risk

assessments for nuclear power generating stations and other nuclear facilities

INTERNATIONAL
ELECTROTECHNICAL
COMMISSION
ICS 27.120.20 ISBN 978-2-8322-7547-4

Contents
1. Overview . 11
1.1 Scope . 11
1.2 Purpose . 11
2. Definitions, acronyms, and abbreviations . 11
2.1 Definitions . 11
2.2 Acronyms and abbreviations . 13
3. Overview of an integrated HRA . 13
3.1 General . 13
3.2 Overall evaluation issues . 14
3.3 HRA process . 15
4. Details of the HRA process . 17
4.1 General . 17
4.2 Steps in the human reliability analysis (HRA) process . 17
5. Documentation . 26
5.1 Purpose . 26
5.2 Structure . 26
Annex A (informative) An example for documenting HRA data . 28
Annex B (informative) Bibliography . 32
Annex C (informative) IEEE list of participants . 34
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

IEEE Std 1082-2017
Guide for Incorporating Human Reliability Analysis into
Probabilistic Risk Assessments for Nuclear Power Generating Stations
and Other Nuclear Facilities
FOREWORD
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IEEE Standards documents are developed within IEEE Societies and Standards Coordinating Committees of the
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Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

– vi – IEC 63260:2020
IEEE Std 1082-2017
International Standard IEC 63260/IEEE Std 1082 has been processed through IEC technical
committee 45A: Instrumentation, control and electrical power systems of nuclear facilities,
under the IEC/IEEE Dual Logo Agreement.
The text of this standard is based on the following documents:
IEEE Std FDIS Report on voting
IEEE Std 1082-2017 45A/1285/FDIS 45A/1293/RVD
Full information on the voting for the approval of this standard can be found in the report on
voting indicated in the above table.
The IEC Technical Committee and IEEE Technical Committee have decided that the contents
of this publication will remain unchanged until the stability date indicated on the IEC web site
under "http://webstore.iec.ch" in the data related to the specific publication. At this date, the
publication will be
• reconfirmed,
• withdrawn,
• replaced by a revised edition, or
• amended.
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

– vii – IEEE Std 1082™-2017
IEEE Std 1082-2017
(Revision of IEEE Std 1082-1997)
IEEE Guide for Incorporating
Human Reliability Analysis into
Probabilistic Risk Assessments for
Nuclear Power Generating Stations
and Other Nuclear Facilities
Sponsor
Nuclear Power Engineering Committee
of the
IEEE Power and Energy Society
Approved 6 December 2017
IEEE-SA Standards Board
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

– viii –
IEEE Std 1082-2017
Abstract: A structured framework for the incorporation of human reliability analysis (HRA) into
probabilistic risk assessments (PRAs) is provided in this guide. To enhance the analysis of human/
system interactions in PRAs, to help ensure reproducible conclusions, and to standardize the
documentation of such assessments are the purposes of this guide. To do this, a specific HRA
framework is developed from standard practices. The HRA framework is neutral with respect to
specific HRA methods.
Keywords: HRA, human reliability analysis, IEEE 1082™, PRA, probabilistic risk assessment
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

– ix –
IEEE Std 1082-2017
,(((Introduction
This introduction is not part of IEEE Std 1082-2017, Guide for Incorporating Human Reliability Analysis into
Probabilistic Risk Assessments for Nuclear Power Generating Stations and Other Nuclear Facilities.
Any process that requires manual control to minimize public risk will require a high level of human reliability.
This reliability can be evaluated through the systematic application of a probabilistic risk assessment (PRA).
However, such an assessment requires a detailed understanding of human performance and human reliability
methods to form a reasonable reliability estimate.
The initial risk assessment made in the nuclear power plant industry, WASH-1400 [B17], recognized the
need for a discipline of human reliability analysis (HRA) to be systematically incorporated within the PRA
enterprise. But the methodology—both analyzing human failure events and identifying and incorporating
them appropriately in the PRA—was new, incomplete, and in several ways inadequate.
The limitations of the understanding of human reliability in the mid-1970s were vividly demonstrated by
the accident at Three Mile Island (TMI). Following TMI, the United States Nuclear Regulatory Commission
(NRC), in conjunction with The Institute of Electrical and Electronics Engineers (IEEE), immediately called
for a conference on the human factor issues raised by TMI. This conference has subsequently become a series.
Parallel to the initiation of the conference, Subcommittee 7, Human Factors and Control Facilities of the IEEE
Nuclear Power Engineering Committee began discussing the standardization of HRA technology. The PRA/
HRA interface of incorporating and performing an HRA in the context of a PRA was recognized as the most
mature of the efforts of HRA. A guide, the least mandating of the IEEE standards documents, was approved as
an IEEE standards project in 1984. The guide was revised in 1997.
This guide outlines the steps necessary to include human reliability in risk assessments. The intent of the guide
is not to discuss the details of specific HRA methods, but rather to affirm a method-neutral framework for
using a diverse range of HRA methods to support PRA. Since human error has been found to be an important
contributor to risk, this guide underscores the systematic integration of the HRA at the earliest stages and
throughout the PRA.
Since the 1997 revision of IEEE Std 1082™, there have been significant developments in HRA methods,
theories, and practices. A working group (WG) was convened in 2012 to reaffirm the guide. This WG
found numerous cases where the 1997 standard contained outdated references or failed to consider now-
commonplace aspects of HRA. The WG, however, confirmed the underlying practice of HRA espoused in
IEEE Std 1082-1997 is still contemporary and relevant to HRA practice. The WG has updated the guide, to
the extent necessary to reflect important advances in HRA. Thus, the framework for conducting HRA found in
IEEE Std 1082-1997 remains intact in this revision but has been augmented with references to contemporary
issues and practices.
IEEE Std 1082 remains a unique, concise guide for specifying the framework for conducting HRA as part of
PRA. Additional standard guidance documents are available beyond IEEE Std 1082. For example, the Electric
Power Research Institute (EPRI) released the Systematic Human Action Reliability Procedure (SHARP) and
revised SHARP1 approach [B4], which describes a detailed process of integrating quantitative HRA into
PRA, mirroring parts of IEEE Std 1082. The American Society of Mechanical Engineers (ASME) has created
the Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power
Plant Applications [B1], which outlines high level requirements for HRAs to be included in PRAs. The NRC
published Good Practices for Implementing Human Reliability Analysis [B13], which serves as a reference
for desirable, but not required aspects of HRA. These three guidelines and numerous recommended practices
found in specific HRA methods and texts, complement, but do not replace, IEEE Std 1082. For example,
SHARP1 [B4] elaborates on quantifying the HRA for inclusion in PRA but does not include the entire HRA
The numbers in brackets correspond to those of the bibliography in Annex B.
NUREG publications are available from the U.S. Nuclear Regulatory Commission (http://www .nrc .gov).
EPRI publications are available from the Electric Power Research Institute (http://epri .com).
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

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IEEE Std 1082-2017
process of IEEE Std 1082. The ASME PRA standard [B1] articulates quality requirements for HRA but
does not specify how the HRA should be conducted. NRC’s good practices [B13] parallel many aspects of
IEEE Std 1082 but does not provide an overall process flow for conducting HRA. IEEE Std 1082 remains
relevant as an overarching standard framework for conducting HRA.
IEEE Std 1082 is a method-neutral approach. It is beyond the scope of this guide to enumerate how the guidance
can be tied into different HRA methods. Recent reviews of HRA methods may be found in [B1], [B3], [B14],
[B15], and [B16]. HRA method development has been extensive, with new approaches that address cognition,
context, errors of commission, as well as approaches that span simplified HRA quantification, to dynamic
models of human performance. The framework for integrating HRA into PRA as outlined in this guide should
apply across HRA methods, although some adaptations may be necessary to meet the unique requirements
of specific methods. Such adaptations, especially when using simplified HRA methods, should not come as
efficiencies at the expense of performing an integrated and complete HRA process.
ASME publications are available from the American Society of Mechanical Engineers (http://www .asme .org/ ).
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

– 11 –
IEEE Std 1082-2017
IEEE Guide for Incorporating
Human Reliability Analysis into
Probabilistic Risk Assessments for
Nuclear Power Generating Stations
and Other Nuclear Facilities
1. Overview
1.1 Scope
This guide provides a structured framework for the incorporation of human reliability analysis (HRA) into
probabilistic risk assessments (PRAs).
1.2 Purpose
The purpose of this guide is to enhance the analysis of human-system interactions in PRAs, to help ensure
reproducible conclusions, and to standardize the documentation of such assessments. To do this, a specific
HRA framework is developed from standard practices to serve as a benchmark to assess alternative ways of
incorporating HRA into PRA.
2. Definitions, acronyms, and abbreviations
2.1 Definitions
For the purposes of this document, the following terms and definitions apply. The IEEE Standards Dictionary
Online should be consulted for terms not defined in this clause.
NOTE—Several terms used in this guide and in the field of HRA are important, yet are ambiguous in common usage or not
used frequently enough to be well known. They are defined in this clause for the use in understanding and following this
guide.
basic event: An element of the probabilistic risk assessment model for which no further decomposition is
performed because it is at the limit of resolution consistent with available data.
IEEE Standards Dictionary Online subscription is available at: http:// dictionary.ieee .org.
Notes in text, tables, and figures of a standard are given for information only and do not contain requirements needed to implement this
standard.
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

– 12 –
IEEE Std 1082-2017
consequences: The result(s) of (i.e., events that follow and depend upon) a specified event.
cutset: A group of events that, if all occur, would cause occurrence of the top event (the outcome of interest
such as that investigated by means of a fault tree).
dependence: The relationship between two or more human failure events, which may result in an adjustment
to the model or the human error probability.
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 help ensure public health and safety.
dominant sequence: A sequence of events that constitutes a dominant contributor to overall risk.
event: (A) Any change in conditions or performance of interest. (B) An occurrence at a specific point in time.
event tree: A graphical representation of the logical progression of the possible scenarios through a multiple
series of events that may or may not occur.
fault tree: A graphical representation of an analytical technique whereby an undesired state of a system is
specified and the patterns leading to that state can be evaluated to determine how the undesirable system
failure can occur.
human action: The observable result (often a bodily movement) of a person’s intention.
human error: Failure of human task performance to meet specified criteria of accuracy, completeness,
correctness, appropriateness, or timeliness.
human error probability (HEP): The quantitative estimation of the likelihood of a human error.
human failure event (HFE): A basic event that pertains to a human error.
human interaction: A human action or set of actions that affects equipment, response of systems, or other
human actions.
human reliability analysis (HRA): Any number of formal approaches and methods used to identify sources
of human error and quantify their accompanying human error probabilities.
initiating event: An event either internal or external to the plant that 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.
operating crew: Plant personnel working on shift to operate the plant. They include control room personnel
and those support personnel who directly support the control room personnel in operating the plant.
performance shaping factor (PSF): A factor that influences human reliability through its effects on
performance. These include factors such as environmental conditions, human-system interface design,
procedures, training, and supervision.
probabilistic risk assessment (PRA): 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
This definition differs from the one(s) found in previous IEEE guidance. The current definition has been tailored to match the specific
use in human reliability analysis.
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

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IEEE Std 1082-2017
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)].
recovery: A set of interactions intended to restore failed equipment or to find alternatives to achieve its
function.
risk: Probability and consequences of an event, as expressed by the answer to the following three questions:
(1) What can go wrong?, (2) How likely is it?, and (3) What are the consequences if it occurs?
screening: A type of analysis aimed at eliminating from further consideration factors that are less significant
for protection or safety in order to concentrate on the more significant factors.
screening value: A rough but conservative point estimate of the probability of a specific human failure event.
uncertainty interval: The confidence in the human error probability estimate as expressed in a confidence
bound around the single-point estimate.
walkthrough: A systematic process by which the actions required of operators are checked against the real
plant or against a model, mock-up, or simulation of the real plant. A walkthrough is typically used to identify
performance shaping factors.
2.2 Acronyms and abbreviations
DOE U.S. Department of Energy
I&C instrumentation and control
INPO Institute of Nuclear Power Operations
LOCA loss of coolant accident
NRC U.S. Nuclear Regulatory Commission
3. Overview of an integrated HRA
3.1 General
3.1.1 Importance of human reliability
In assessing the risk associated with a nuclear power plant, the analyst should consider not only the reliability
of plant hardware systems but also the reliability of people’s interactions with other plant or support personnel
and with the plant’s equipment and systems. The scope of interactions with plant equipment and systems
should include those in the control room and at local control stations and with both manually controlled and
automated systems.
3.1.2 Importance of integrated HRA and PRA
An HRA should be an integral part of a PRA. In PRAs, the quality of the analysis (e.g., quantification of human
error) is dependent upon the analyst’s ability to identify scenarios and the expected human actions. This guide
provides a specific approach that, if applied, will standardize the integration of HRA into the PRA process. The
breakdown and order of the steps presented are not so important; all of the steps and their activities, however,
should be found within any HRA. This approach is well established for design-basis PRAs. The approach
applies to beyond design-basis analyses such as those used for severe accidents. However, as the uncertainty
and variability of the plant state and accident scenario evolution increase, so too does the complexity of
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

ng Statio
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IEEE Std 1082-2017
performing the analysis. The steps outlined in this guide should be considered at a minimum; additional steps
may be appropriate for certain cases, such as severe accident analyses.
3.2 Overall evaluation issues
The focus of this guide is restricted to the incorporation of the HRA integrally into a PRA. This includes the
following issues:
a) The compatibility of an HRA with the PRA of which it is a part;
b) The relationship between the way in which an HRA is performed, its philosophy, and the results or
insights that may be obtained;
c) Matching the best suited HRA method to the analysis requirements; and
d) The limits of an HRA or its results.
3.2.1 PRA compatibility
The HRA process proposed is suitable to all levels of a PRA given defined human failure events. If these are not
defined, then this guidance cannot be applied successfully. The risk focus of a PRA requires the quantitative
results of an HRA to be probabilistic in nature. Applications of PRAs to risk management efforts require
that the HRA documents in sufficient detail the analyst’s human factors considerations for the human failure
events. The PRA can have a diverse range of applications, the objectives of which may not be completely
identified prior to the assessment. The HRA process should be flexible enough to anticipate some of the likely
applications of the results of the HRA. For example, this may include design changes, procedure changes,
training development, safety evaluations, or technical specification modification.
3.2.2 Qualitative HRA
While the approach identified in this guide supports HRA quantification as part of the PRA, it should be noted
that there is an increasing emphasis on the importance of qualitative HRA [B15], i.e., HRA that does not
produce a human error probability, but rather insights into the human’s role and contribution to overall system
performance. The HRA approach in this guide supports both qualitative and quantitative aspects of HRA.
For PRA, the quantitative approach should be adopted. For non-PRA applications of HRA, steps relevant to
quantification should be omitted as appropriate.
3.2.3 The relationship of approach to results
Assumptions made by human reliability analysts about the relative importance of various human activities
will influence the breadth and detail of models developed for the HRA. The data and chosen method of
quantifying human interactions will influence the specific estimates or ranges of uncertainty obtained,
although there is generally good agreement between HRA methods. If results point to the need to improve
the reliability of selected systems and accompanying human interactions, these improvements should either
be readily identifiable from the documented HRA or should be the subject of further or different analytical
methods that will allow improvements to be identified as described in method-specific documents. In addition
to method-specific guidance, general guidance on selection of appropriate HRA approaches can be found in
cross-method overview documents such as [B2], [B3], and [B14]. The HRA analyst should be mindful of this
when considering the specific approach to be taken.
3.2.4 Matching the method to the application
Various HRA-related methods are available and being developed (e.g., cognitive approaches to human error
or approaches that address errors of commission). HRAs should be flexible enough to accommodate new
findings and model developments, while structured enough to be repeatable and traceable. HRA methods
were developed for different purposes, and they feature different strengths [B14]. One emerging practice is
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IEEE Std 1082-2017
that multiple HRA methods may be used within a single analysis to reflect the strengths of those approaches
and the needs of the analysis [B7]. This guide remains method-neutral but encourages analysts to be flexible in
considering the best HRA methods for particular applications.
3.2.5 Limits
To be effectively incorporated into a PRA, an HRA should provide a realistic-as-possible interpretation of the
role of plant personnel in accident prevention and mitigation. Accordingly, the results of an HRA should be
documented in a format such that the basic assumptions, models, and data sources are clearly documented and
the limitations of the analysis are understandable to the user (e.g., PRA analyst).
Figure 1—General HRA process
3.3 HRA process
3.3.1 General
A general HRA process is depicted in Figure 1. It parallels the typical PRA process (see [B1] and [B6]), but
is not organizationally related to it. The chosen HRA process is an adaptation of a process that has been used
in many PRAs (see [B9]). The general structure is described in 3.3.3, the specific steps are described in 3.3.4
and detailed in Clause 4, and the outputs at each step are described in 3.3.5. The timing and level of detail of
the HRA should synchronize with the PRA. Therefore, the HRA structure discussed in this guide should be
applied with modification as needed. In its general form, Figure 1 can also show the initial portions of the
PRA process, so that both efforts are truly integrated to form an integrated analysis.
For example, HRA for external and area initiating events may deviate from the prescribed HRA process. In such a case, typically the
analyst takes the initiating event HRA as the basis, checks the feasibility of actions for external and area events, and then modifies the
analyses as necessary.
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

s
ng Statio IEC 63260:2020
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IEEE Std 1082-2017
3.3.2 Existing versus new HRA
The process outlined in Figure 1 represents an approach that may be used to create a new HRA. In creating
a new HRA, the HRA should be integrated into the PRA process and not viewed as an independent analysis
from the PRA. In practice, it is common to start with an existing HRA for a similar system or plant and then
reuse and modify it with details specific to the new analysis. The proper documentation of the initial HRA
across all steps helps to ensure that reuse is viable. When reusing and modifying an existing HRA, analysts
should carefully document any changes in assumptions and their consequences in the new analysis. Steps in
Figure 1 may be omitted when feasible as part of the reuse process. The applicability of specific steps will vary
according to the types of modifications made to the existing HRA.
3.3.3 HRA structure
The HRA should be incorporated within the PRA in a stepwise manner. The HRA should begin at a level that
is broad enough to help ensure completeness without including unnecessary details. As the analysis moves
forward, the focus on risk-significant aspects of the plant that affect the core damage frequency becomes
possible, and breadth can be traded for depth in the analysis. This narrowing of the analysis requires evaluation
and judgment. This focusing of the analysis is referred to as “screening” by PRA practitioners.
3.3.4 Summary of HRA steps
The HRA process depicted in Figure 1 is a process of eight tasks and two major decisions. These steps are
detailed in Clause 4. These steps form a basis for auditable documentation of the HRA. If the team chooses to
deviate from this process, then the deviations from this framework can also be documented for auditability as
follows:
a) The HRA process begins with the selection and training of the joint HRA-PRA team.
b) The team should familiarize itself with the plant and its systems, functions, and procedures.
c) The first models of the plant should be developed jointly by the team. This model identifies the major
human interaction events in functional terms.
d) The results form a list of interactions that are screened by restricting the analysis to candidate- dominant
sequences. The HRA supports screening by providing rough but conservative point estimates of the
probabilities of all the human interaction events called “screening values” into the models at this stage.
e) Human actions whose failures are found to be significant contributors to risk are then characterized in
detail as human failure events to support quantification and application of the PRA results.
f) Each human failure event is then quantified. The result of the quantification step often changes the
event description or adds a more detailed representation of the event; the result should be incorporated
into the update of the model.
g) Dependence between human failure events is considered. If recovery events are identified, then the
models of these events, including the models of the human interactions, are incorporated into the
updated models.
h) The final review step ends the activity and, if the HRA-PRA team includes people from plant
operations, should only be confirmatory.
Finally, the documentation of the HRA, like the rest of the PRA, is a crucial element that is ongoing throughout
the HRA. Documentation is not shown on the flow chart in Figure 1 but is detailed separately in Clause 5.
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IEEE Std 1082-2017
3.3.5 Outputs
The HRA process is structured so as to be able to yield useful results at each step of its production. Table 1
lists the kinds of information produced at each step. Generally, all steps of the HRA process will be performed,
since they together represent the steps needed to support a PRA.
Table 1—Products of the HRA process steps
Step Product
1. Select and train HRA team Integrated team with requisite PRA and HRA skills.
2. Familiarize team with plant Initial identification of human functions and activities. Potential human
errors can be spotted although without much risk context yet.
3. Build initial plant model All major systems modeled.
System interactions identified.
Defense barriers against off-normal events described.
Most key human interactions identified.
4. Screen human interactions Key human interactions identified, screening values
chosen, and initial quantification performed.
5. Characterize human interactions Failure modes, mechanisms, causes, effects, and
influences of the key human interactions are determined.
Initial estimates of time required to take action.
6. Quantify human interactions Importance ranking, likelihood, and uncertainties
of key human interactions.
7. Update plant model with Model with recovery actions and dependencies included.
dependence and recovery
8. Review results Confidence that results make sense and can be used
by plant staff in risk reduction efforts.
4. Details of the HRA process
4.1 General
The HRA process, as summarized in Clause 3, is detailed below and includes each step and a statement of
purpose, a description of the step, and a statement of output. A strong technical interface between the human
reliability analysts and the rest of the PRA team performing the equipment reliability modeling is needed. This
team should be a joint team whenever possible. The HRA-PRA team may be a subset of the overall PRA team,
including members specialized in human factors and human reliability.
4.2 Steps in the human reliability analysis (HRA) process
4.2.1 Step 1: Select and train HRA team
4.2.1.1 General
The HRA process requires contributions from a wide range of skills in engineering, plant performance, human
factors, and mathematical or statistical analysis. It is important that the team that performs the HRA possesses
these skills. The number of people on the team and their formal backgrounds are relatively unimportant, so
long as the aggregate represents the skills listed in 4.2.1.3. A senior person should be appointed from within
the organization to lead the HRA-PRA team and in particular, to lead the integration of the team skills and
champion their interaction with the broader PRA team. The required training should include methods for
cooperation and communication across disciplines.
Published by IEC under license from IEEE. © 2017 IEEE. All rights reserved.

ng Statio
– 18 –
IEEE Std 1082-2017
4.2.1.2 Purpose
The purpose of this step is to assemble a highly qualified team that works well together and integrates several
important skills and disciplines. The quality of the HRA portion of a PRA depends, in large part, on ensuring
that the individuals selected as team members possess the necessary technical skills and ability to work and
communicate effectively as a team. This teamwork, early in the PRA, will facilitate the development of the
HRA and the accuracy of the PRA.
4.2.1.3 Description
The expertise required for the HRA-PRA team includes items a) through g) below. These experts interface
with other experts on the PRA team and at the plant as needed. This list does not imply that individuals in each
of the specific areas are required. Individuals may have multiple areas of expertise, e.g., a PRA analyst with
operations background who is trained in HRA methods. To help ensure proper consideration of human aspects
of risk, HRA expertise should be part of the PRA team throughout the analysis. The team should have the
following cumulative experience:
a) Plant operations personnel provide experienced insight into the way(s) in which people conduct
their jobs/tasks in light of training, procedures, and operating experience. As such, they provide
added reality to the analysis of human-system interaction and can identify information about task
requirements that may not be apparent from system operating descriptions and procedures.
b) Human factors engineering personnel provide information about the expected effects of task and
workplace characteristics on human performance, given human capabilities and limitations. Human
factors engineers augment the team by identifying attributes of tasks and task environments that should
be taken into account when evaluating performance shaping factors that contribute to the reliability or
risk of an operation.
c) Human reliability analysts develop or work with the developed detailed qualitative models of the
human-system analysis to provide both qualitative and quantitative estimations of human reliability.
The human reliability analysts work with a variety of quantitative techniques to address performance
shaping factors, time, and training constraints, and their impact on the reliability of human
performance.
d) Nuclear power plant system engineering personnel bring a broad range of knowledge of how the
systems examined are designed, operated, and maintained. These personnel are knowledgeable
of general operating practices and the details of equipment capacities, operating envelopes, and
equipment performance, and they support the analysis of human-system interaction.
e) System safety analysts provide evaluations of system performance and human performance
requirements for situations in which system conditions transition from normal to abnormal and
beyond technical specification operating envelopes. The system safety analysis may include thermal
hydraulic, fuel, and fission product behavior, and other phenomenological assessment methods to
identify realistic functional requirements and success criteria for the human.
f) Probabilistic risk assessment personnel, who develop equipment reliability models and data,
provide the detailed information on failure rates (or probabilities) for hardware caused by internally
and externally initiated events. Output of the PRA can consist of fault trees, cut sets, event trees,
Bayesian networks, dynamic event trees, system dynamics models, or Petri nets, to which the human
reliability analyst prov
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