Nuclear Safety
NEA/CSNI/R(2018)1
November 2018www.oecd-nea.org

Status of Practice for Level 3

Probabilistic Safety Assessments

Nuclear Energy Agency

NEA/CSNI/R(2018)1
Unclassified

English text only
19 December 2018

NUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

Status of Practice for Level 3 Probabilistic Safety Assessments

Please note that this document is available in PDF format only.

JT03441357

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2 │ NEA/CSNI/R(2018)1

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STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│3

Committee on the Safety of Nuclear Installations

The Committee on the Safety of Nuclear Installations (CSNI) is responsible for the Nuclear Energy
Agency (NEA) programmes and activities that support maintaining and advancing the scientific andtechnical knowledge base of the safety of nuclear installations.
The Committee constitutes a forum for the exchange of technical information and for collaboration
between organisations, which can contribute, from their respective backgrounds in research,development and engineering, to its activities. It has regard to the exchange of information between
member countries and safety R&D programmes of various sizes in order to keep all member countries
involved in and abreast of developments in technical safety matters.
The Committee reviews the state of knowledge on important topics of nuclear safety science andtechniques and of safety assessments, and ensures that operating experience is appropriately
accounted for in its activities. It initiates and conducts programmes identified by these reviews andassessments in order to confirm safety, overcome discrepancies, develop improvements and reachconsensus on technical issues of common interest. It promotes the co-ordination of work in different
member countries that serve to maintain and enhance competence in nuclear safety matters, including
the establishment of joint undertakings (e.g. joint research and data projects), and assists in the
feedback of the results to participating organisations. The Committee ensures that valuable endproducts of the technical reviews and analyses are provided to members in a timely manner, and madepublicly available when appropriate, to support broader nuclear safety.
The Committee focuses primarily on the safety aspects of existing power reactors, other nuclear
installations and new power reactors; it also considers the safety implications of scientific andtechnical developments of future reactor technologies and designs. Further, the scope for theCommittee includes human and organisational research activities and technical developments that
affect nuclear safety.

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Table of contents
LIST OF ABBREVIATIONS AND ACRONYMS ............................................................................. 5
EXECUTIVE SUMMARY ................................................................................................................... 8
ACKNOWLEDGEMENTS ................................................................................................................ 12
1. INTRODUCTION............................................................................................................................ 14
1.A. Background and motivation ...................................................................................................... 14
1.B. Objectives.................................................................................................................................. 14
1.C. Task group composition and co-ordination............................................................................... 15
1.D. Intended audience ..................................................................................................................... 15
1.E. Report structure ......................................................................................................................... 15
2. PROCESS AND METHODS .......................................................................................................... 17
2.A. Survey design ............................................................................................................................ 17
2.B. Survey administration ............................................................................................................... 18
2.C. Analysis of submitted survey responses ................................................................................... 18
2.D. Planned expanded task group workshop ................................................................................... 19
2.E. Final task report ........................................................................................................................ 19
3. SURVEY RESULTS AND INSIGHTS .......................................................................................... 20
3.A. Part 1: Respondent information ................................................................................................ 20
3.B. Part 2: Application of Level 3 PSA........................................................................................... 22
3.C. Part 3: Level 3 PSA modelling issues and technical challenges ............................................... 25
4. CONCLUSIONS AND RECOMMENDATIONS ......................................................................... 34
4.A. Typical practices ....................................................................................................................... 34
4.B. Common technical challenges or limitations ............................................................................ 37
4.C. Recommendation for future activities ....................................................................................... 38
5. REFERENCES ................................................................................................................................. 40
APPENDIX A: CSNI ACTIVITY PROPOSAL SHEET (CAPS) WGRISK(2015)3 ................... 41
APPENDIX B: FINAL SURVEY INSTRUMENT ......................................................................... 47
APPENDIX C: SUBMITTED SURVEY RESPONSES BY COUNTRY ...................................... 57
LIST OF TABLES
Table 1. Countries and organisation types represented by survey respondents ..................................... 21
Table 2. Uses of Level 3 PSA across all countries surveyed ................................................................. 23
Table 3. Probabilistic consequence analysis computer codes used for Level 3 PSA applications ........ 25

LIST OF FIGURES
Figure 1. Typical Level 3 PSA Technical elements and their relationships .......................................... 18
Figure 2. Distribution of organisation types represented........................................................................ 21

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│5

List of abbreviations and acronyms

ANVS

Authority for Nuclear Safety and Radiation Protection (Netherlands)

APR

Advanced power reactor

ARANO

Probabilistic consequence analysis computer code

ATD

Atmospheric transport and dispersion

BDBA

Beyond design-basis accident

CAPS

CSNI activity proposal sheet (NEA)

CCDF

Complementary cumulative distribution function

CDF

Core damage frequency OR cumulative distribution function

CFF

Containment failure frequency

CNRA

Committee on Nuclear Regulatory Activities

CNSC

Canadian Nuclear Safety Commission

COSYMA

Probabilistic consequence analysis code

CRPPH

Committee on Radiation Protection and Public Health (NEA)

CSNI

Committee on the Safety of Nuclear Installations (NEA)

DBA

Design-basis accident

DRA

Division of Risk Analysis (United States)

EG-COSTNA

Expert Group on Costs of Nuclear Accidents, Liability Issues
and their Impact on Electricity Costs

FN curve

Curve that illustrates frequency of N or more consequences (F) as afunction of total number of consequences (N) to express societal risk

HYSPLIT

HYbrid Single Particle Lagrangian Integrated Trajectory

IAEA

International Atomic Energy Agency

IRSN

Institut de radioprotection et de sûreté nucléaire (France)

JAEA

Japan Atomic Energy Agency

JNRA

Nuclear Regulation Authority (Japan)

KAERI

Korea Atomic Energy Research Institute

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT6 │ NEA/CSNI/R(2018)1
KI

Potassium iodide

KINS

Korea Institute of Nuclear Safety

KOSCA-OUTPUT Korea off-site consequence analysis software package – Output
post-processing module
LENA_P

Probabilistic version of LENA consequence analysis computer code

LERF

Large early release frequency

LNT

Linear no threshold

LRF

Large release frequency

MAAP

Modular accident analysis program

MACCS

MELCOR accident consequence code system

MELCOR

Severe accident analysis computer code

MTA EK

Hungarian Academy of Sciences, Centre for Energy Research

NDC

Nuclear Development Committee (NEA)

NEA

Nuclear Energy Agency

NPP

Nuclear power plant

NRC

Nuclear Regulatory Commission (United States)

NRG

Nuclear Research and Consultancy Group (Netherlands)

NSSC

Nuclear Safety and Security Commission (Korea)

NUBIKI

Nuclear Safety Research Institute (Hungary)

NUDOS2

NUclide DOSes version 2 codeOECD

Organisation for Economic Co-operation and Development

OSCAAR

Off-site consequence analysis code for atmospheric release in reactor
accident

PDF

Probability density function

PSA

Probabilistic safety assessment

QHO

Quantitative health objective

RES

Office of Nuclear Regulatory Research (United States)

SILAM

System for Integrated modeLling of Atmospheric coMposition

SINAC

Simulator software for interactive
consequences of nuclear accidents

SSM

Swedish Radiation Safety Authority

STUK

Radiation and Nuclear Safety Authority (Finland)

TSO

Technical support organisation

modellingofenvironmental

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1

VALMA

Atmospheric dispersion and dose assessment code

VTT

VTT Technical Research Centre of Finland (Finland)

WGAMA

Working Group on Analysis and Management of Accidents (NEA)

WGPCWorking Group on Public Communication (NEA)WGRISK

Working Group on Risk Assessment (NEA)

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Executive summary
Introduction
A Level 3 probabilistic safety assessment (PSA) is an assessment of the off-site public risks
attributable to a spectrum of possible accident scenarios involving a nuclear installation. In the
traditional PSA framework for commercial nuclear power plants, a Level 3 PSA includes threeprogressive levels of analysis:
1) Level 1 PSA:
2) Level 2 PSA:
3) Level 3 PSA:

Fuel damage accident or plant damage state frequency analysis.
Level 1 analysis plus accident progression, containment performance, andradiological release frequency analysis.
Level 2 analysis plus off-site radiological consequence analysis.

An increasing number of organisations in many countries are pursuing the development andapplication of Level 3 PSAs for nuclear installations. However, there are several modelling issues andother technical challenges that would benefit from increased information exchange and sharing of
methods and practices in this area. For this reason, the Nuclear Energy Agency (NEA) Committee on
the Safety of Nuclear Installations (CSNI) approved a proposed Working Group on Risk Assessment(WGRISK) Level 3 PSA activity in June 2015. The objectives of this activity were to: 1) survey
member and partner countries to determine current methodological practices in Level 3 PSA; 2)
identify common challenges and typical practices; and 3) summarise the results.

Process and methods
A survey instrument was developed to elicit and characterise international practices with respect toselected modelling issues and other technical challenges. The final survey instrument – which wasadministered by e-mail – consisted of three parts:
1) Part 1: Respondent information. Part 1 was designed to obtain respondent information thatwould allow the task group to determine whether responses differ across different countries ortypes of organisations.
2) Part 2: Application of Level 3 PSA. Part 2 was designed to obtain information about whether
and how respondents or organisations use or intend to use Level 3 PSA. Since the purpose orintended use of any PSA can influence multiple aspects of a PSA model, it was judged that
responses to Part 2 would provide important contextual information for understanding andevaluating responses to Part 3.
3) Part 3: Level 3 PSA Modelling issues and technical challenges. Part 3 was designed to elicit
information about whether and how respondents or organisations are addressing specificmodelling issues and other technical challenges that practitioners encounter in performing offsite radiological consequence analyses in support of Level 3 PSA applications for nuclear
installations.

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Survey responses
Representatives from ten WGRISK member countries submitted their responses to the survey.
Representatives from diverse organisation types within each country often collaborated to submit a
joint response to the survey, thereby resulting in multiple organisation types being represented in asingle survey response. Fifty per cent of survey respondents represented regulatory authorities, withthe other 50% representing a mix of academic/research, industry, and technical support organisations(TSOs).
Two out of the ten of surveyed countries indicated that there is a requirement to perform
Level 3 PSAs for nuclear installations in their country, while 7 out of the 10 surveyed countries
indicated that they are currently performing, planning to perform, or considering performing
Level 3 PSAs for nuclear installations: 1) Finland (performing); 2) Hungary (planning to perform);
3) Japan (performing); 4) Korea (required and performing); 5) Netherlands (required and performing);
6) Sweden (considering performing); and 7) United States (performing).
Analysis of submitted survey responses produced many insights and resulted in the identification of
typical practices and common technical challenges or limitations across several aspects of Level
3 PSA. These typical practices and common technical challenges or limitations are summarised in thesections that follow.

Typical practices
Typical practices identified through analysis of submitted survey responses can be grouped into four
broad categories: 1) Level 3 PSA applications; 2) Level 3 PSA scope considerations; 3) modelling
practices; and 4) presentation of risk results and uncertainties in risk results.

Level 3 PSA applications• Among countries that are currently performing, planning to perform, or considering performing
Level 3 PSAs for nuclear installations, five broad types of Level 3 PSA applications wereidentified: 1) applied research; 2) comparison with risk acceptance criteria, safety goals, orother quantitative objectives; 3) evaluation of protective action effectiveness to informemergency preparedness and response guidelines; 4) information for the development of severe
accident management plans; and 5) environmental assessments.
• Although survey respondents generally acknowledged the potential benefit of using resultsfrom Level 3 PSAs to inform nuclear installation siting decisions, most indicated that Level 3
PSAs are not being used when considering such applications. Results from deterministic
design-basis accident (DBA) dose consequence assessments are typically used to demonstratethat doses following postulated fission product releases are within prescribed limits.
• Results from deterministic analyses provide the primary analytical basis for establishing
emergency planning or protective action zones. If Level 3 PSA insights are used in developing
requirements for the sizes and boundaries of emergency planning or protective action zones,
such risk insights are typically considered as a complement to results from deterministic
analyses as part of a risk-informed approach.

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Level 3 PSA scope considerations• Since the objective of Level 3 PSA is to assess the risks to the off-site public that are
attributable to a spectrum of possible accident scenarios involving a nuclear installation,
countries that perform Level 3 PSA typically exclude the on-site worker population.
• Radiological releases to aqueous pathways are typically screened out and excluded fromLevel 3 PSAs for nuclear installations that have the potential for airborne radiological releases,
since airborne releases are expected to dominate radiological health risks.
• Among most countries that perform Level 3 PSAs, economic models are generally used toestimate the costs attributed to the implementation of modelled protective actions to reduce theradiological dose to the off-site public from the accidental release of radiological materials.

Modelling practices• Level 2 PSA analysts typically work closely with Level 3 PSA analysts to: 1) defineradiological release categories; 2) select a representative accident sequence for eachradiological release category to estimate source term characteristics; and 3) perform sensitivity
or uncertainty analyses on release categories or release fractions to evaluate the impact on risk
results.
• The size of the modelled region and spatial intervals around a site are typically application- andsite-specific. In general, a larger number of spatial intervals with finer resolution grid elements
are used for the region close to the site, with the number of spatial intervals and grid resolution
decreasing as the distance from the site increases.
• Most countries that perform Level 3 PSA: 1) use site-specific data that includes hourly
observations for an entire year; 2) implement code-specific weather binning and samplingstrategies to account for the impact of variability in weather conditions over time; and 3) use
site-specific information sources to model various other site characteristics.
• A range of protective actions is typically modelled for the early (emergency), intermediate, andlate (recovery) phases of accident response. Most countries that model protective actions
typically do not: 1) model population groups with different protective action behaviours; or2) use models to account for the probabilities of success or failure of protective actions.
• In general, a range of early radiological health effects are estimated using deterministic models.
Fatal and non-fatal cancers involving multiple organs generally represent the latent radiologicalhealth effects that are estimated using stochastic models. Respondents from all countries that
perform Level 3 PSA indicated they use a default dose-response model based on the linear nothreshold (LNT) hypothesis to estimate the excess cancer risk attributable to radiological doses
caused by accidental releases from nuclear installations.

Presentation of risk results and uncertainties in risk results• In general, risk results and uncertainties in risk results are presented using a variety of formats
and graphical displays.

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Common technical challenges or limitations
Analysis of submitted survey responses also resulted in identification of common technical challenges
or limitations:• Among countries that are not currently performing Level 3 PSAs for nuclear installations, four
potential barriers to performing Level 3 PSA were identified: 1) absence of a technical or legal
framework to perform Level 3 PSA; 2) large uncertainties in Level 3 PSA results, especiallywhen combined with uncertainties propagated from Level 1 and Level 2 analyses; 3) limited
expected benefit in terms of potential back fits or safety improvements; and 4) additional
resources required to perform detailed off-site radiological consequence analyses.
• Most computer codes for performing probabilistic analyses of off-site radiologicalconsequences: 1) use relatively simple atmospheric transport and dispersion (ATD) models;and 2) either do not support terrain modelling or rely on the use of simple terrain models.
• None of the available probabilistic consequence analysis computer codes can directly perform
risk characterisation by mathematically combining the radiological release category frequencies
with the corresponding conditional consequences.
• Most countries that perform Level 3 PSA do not have a communication plan or guidelines forrisk communication. A notable exception is the US Nuclear Regulatory Commission (NRC)that has developed guidance documents aiming to enhance risk communication with internaland external stakeholders.

Recommendation for future activities
A key finding from analysis of submitted survey responses is that many typical practices or common
technical challenges identified as part this WGRISK activity are driven by the limitations orcapabilities of available probabilistic consequence analysis codes. This finding underscores thepotential need for future research and leads to the following recommendation stemming from thisWGRISK activity.

Recommendation: Consider performing a study to benchmark available probabilistic
consequence analysis computer codes used for Level 3 PSA applications.
The NEA previously performed similar studies to benchmark probabilistic consequence analysis
codes. Yet the most recent of these studies was completed more than 20 years ago. Since probabilistic
consequence analysis codes have evolved considerably during the past two decades, it seems
reasonable to perform a follow-on benchmarking study to compare more recent versions of thecomputer codes. However, many of these probabilistic consequence analysis codes are currently beingenhanced to address various modelling issues and technical challenges, some of which were brought
into focus by the 2011 accident at the Fukushima Daiichi nuclear power plant in Japan. Based onthese ongoing code enhancements, a logical path forward may to be to delay any future benchmarking
study (if one is to be undertaken) until after these enhancements have been implemented, withverification and validation testing completed, as appropriate.

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Acknowledgements
This project was carried out by the Nuclear Energy Agency (NEA) Committee on the Safety of
Nuclear Installations (CSNI) Working Group on Risk Assessment (WGRISK). Technical staff in the
US Nuclear Regulatory Commission (USNRC)/Office of Nuclear Regulatory Research(RES)/Division of Risk Analysis (DRA) led this project, with support from technical staff assigned toother USNRC offices, as well as from individuals from other member countries who volunteered toserve on the core task group and expanded task group for this WGRISK activity.
As the lead organisation for this project, the USNRC established an internal working group that:
1) identified and prioritised specific modelling issues and other technical challenges in performing
Level 3 PSA off-site radiological consequence analyses that could benefit from information exchange
and sharing of methods and practices across the international PSA community; 2) developed an initial
draft of the survey instrument that would be used to elicit and characterise international practices withrespect to selected modelling issues and other technical challenges; and 3) developed an example
response that would be distributed with the final survey instrument to provide survey respondents
with guidance by way of an example on the scope and level of detail that would be expected for
responses to each of the survey items. The following USNRC staff members served on or supported
this working group: Keith Compton, Kevin Coyne, Elijah Dickson, Daniel Hudson, Tony Nakanishi,
Donald Palmrose, Jason Schaperow, and Nathan Siu.
The following individuals represented their respective organisations and countries as members of thecore task group that was responsible for project planning, survey design, survey administration,
analysis of submitted survey responses, and development of this final report:
Name

Organisation

Country

Usha Menon
Chantal Morin

Canadian Nuclear Safety Commission (CNSC)

CanadaTams Pázmándi

Hungarian Academy of Sciences, Centre for Energy Research(MTA EK)

Hungary

Yu Maruyama

Japan Atomic Energy Agency (JAEA)

Japan

Kwang-Il Ahn
Seok-Jung Han

Korea Atomic Energy Research Institute (KAERI)

Korea

J.L. (Hans) Brinkman
J.B. (Jacques) Grupa

Nuclear Research and Consultancy Group (NRG)

Netherlands

Daniel Hudson(Project lead)

US Nuclear Regulatory Commission (USNRC)

United States

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The following individuals represented their respective organisations or countries as members of the
expanded task group and added valuable contributions to this project by reviewing and providing
constructive feedback on the draft survey instrument or draft final report:
Name

Organisation

Country

Shahen Poghosyan

International Atomic Energy Agency (IAEA)

Aurélie Lorin

NEA/Committee on Nuclear Regulatory Activities(CNRA)/Working Group on Public Communication(WGPC)

Neil Higgins

NEA/Nuclear Development Committee (NDC)/Expert
Group on Costs of Nuclear Accidents, Liability Issues
and their Impact on Electricity Costs (EG-COSTNA)

Ilkka Karanta
Markus Porthin

VTT Technical Research Centre of Finland

FinlandAttila Bareith

NUBIKI Nuclear Safety Research Institute

Hungary

Per Hellström

Swedish Radiation Safety Authority (SSM)

SwedenN/A

The members of the core task group wish to acknowledge and extend their sincere gratitude to thecountries, organisations, and individuals who supported this project by providing high-quality
responses to the survey and by responding to subsequent correspondence to obtain additional
information, as needed. These countries, organisations, and individuals are identified in each of thesubmitted survey responses provided in Appendix C.
The group also wishes to acknowledge the leadership and vision for this project that Dr Kevin Coyne
provided throughout the conception, development, and implementation phases of this project. Lastly,
this task would not have been possible without the significant contributions, leadership, and coordination provided by the task project leader, Dr Daniel Hudson. The valuable insights about current
methodological practices in Level 3 PSA that were developed from this project would not have beenrealised without Dr Hudson’s substantial efforts.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT14 │ NEA/CSNI/R(2018)1

1. INTRODUCTION

1.A.

Background and motivation

The mission of the Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations(CSNI) Working Group on Risk Assessment (WGRISK) is to advance understanding of probabilistic
safety assessment (PSA) and to enhance its utilisation for: 1) improving the safety of nuclear
installations; 2) improving the design and operation of nuclear installations; and 3) increasing
regulatory effectiveness through risk-informed approaches. In support of this mission, WGRISK
routinely collects and shares information among member countries regarding PSA methods andpractices used to estimate nuclear accident risks and to identify significant risk contributors in supportof nuclear installation applications.
A Level 3 PSA is an assessment of the off-site public risks attributable to a spectrum of possible
accident scenarios involving a nuclear installation. In the traditional PSA framework for commercial
nuclear power plants, a Level 3 PSA includes three progressive levels of analysis: 1) core damage
accident or plant damage state frequency analysis; 2) accident progression, containment performance,and radiological release frequency analysis; and 3) off-site radiological consequence analysis. A Level
1 PSA includes only the first level of analysis and estimates the frequency of core damage accidents
or plant damage states and identifies significant contributors to these frequencies. A Level 2 PSA
includes the first and second levels of analysis and - in addition to Level 1 PSA risk metrics estimates the frequencies of defined radiological release categories and identifies significant
contributors to these frequencies. Finally, a Level 3 PSA includes all three levels of analysis and - inaddition to Level 2 PSA metrics - estimates the frequencies of off-site public health, environmental,
and economic consequences attributable to the set of radiological release categories andcorresponding source terms defined in the Level 2 analysis. Compared to Level 1 and Level 2 PSAs, a
Level 3 PSA thus represents a more complete characterisation of the off-site public risks attributable
to a spectrum of possible accident scenarios involving a nuclear installation and provides an important
input to cost-benefit analyses used to evaluate proposed risk management options. Although the term
Level 3 PSA has been used in some cases to refer only to the off-site radiological consequence
analysis element, the focus of this activity is on off-site radiological consequence analyses that are
performed as part of an integrated PSA that includes all three analysis levels.
An increasing number of organisations in many countries are pursuing the development andapplication of Level 3 PSAs for nuclear installations [1,2]. However, the experience of several
WGRISK member countries indicates there are several modelling issues and other technicalchallenges that would benefit from increased information exchange and sharing of methods andpractices in this area. Such co-operation could: 1) help harmonise international methods and practices;and 2) assist organisations in developing new Level 3 PSAs or in improving upon existing Level 3
PSAs.

1.B.

Objectives

Given the increasing interest in Level 3 PSA and the potential benefits of information exchange andsharing of methods and practices in this area, the CSNI approved a proposed WGRISK Level 3 PSA
STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 15

activity in June 2015. The approved CSNI Activity Proposal Sheet (CAPS) for this activity isprovided in Appendix A.
The objectives of this activity were to: 1) survey member and observer countries to determine current
methodological practices in Level 3 PSA; 2) identify common challenges and typical practices; and3) summarise the results of this activity in a final task report. This report represents the final productof this CSNI-approved WGRISK Level 3 PSA activity.

1.C.

Task group composition and co-ordination

Representatives from six WGRISK member countries comprised the core task group for this activity,with the United States serving as the lead country. Countries and organisations that contributed to thecore task group included:
1.
2.
3.
4.
5.
6.

Canada: Canadian Nuclear Safety Commission (CNSC)
Hungary: Hungarian Academy of Sciences, Centre for Energy Research (MTA EK)Japan: Japan Atomic Energy Agency (JAEA)
Korea: Korea Atomic Energy Research Institute (KAERI)
Netherlands: Nuclear Research and Consultancy Group (NRG)
United States: US Nuclear Regulatory Commission (USNRC)

The expanded task group included core task group members and other interested WGRISK
participants who are familiar with Level 3 PSA, and additional NEA organisations. In particular, since
Level 3 PSA involves multiple disciplines and areas of interest to other groups, representatives fromfive groups within and beyond the NEA were also engaged in this activity to ensure effective coordination between related activities. These groups included:
1.
2.
3.
4.
5.

1.D.

NEA/CSNI/Working Group on Analysis and Management of Accidents (WGAMA)
NEA/Committee on Nuclear Regulatory Activities (CNRA)/Working Group on Public
Communication (WGPC)
NEA/Committee on Radiation Protection and Public Health (CRPPH)
NEA/Nuclear Development Committee (NDC)/Expert Group on Costs of Nuclear
Accidents, Liability Issues and their Impact on Electricity Costs (EG-COSTNA)
International Atomic Energy Agency (IAEA)

Intended audience

The intended audience for this report includes a broad spectrum of individuals and entities across theinternational community with an interest in the assessment of off-site public risks attributable to aspectrum of possible accident scenarios involving a nuclear installation

1.E.

Report structure

The main report is organised into five chapters and is supported by three appendices. This chapter
introduces: 1) the background and motivation for this CSNI-approved activity; 2) the objectives ofthis activity; 3) the composition of task groups assigned to this activity, along with identification ofother groups within and beyond NEA that were engaged in this activity to ensure effective coordination; 4) the intended audience for this report; and 5) the structure of this report.
Chapter 2 summarises the process and methods used to complete this activity and includes fivesections that respectively address: 1) design of a survey instrument that was used to elicit andcharacterise international practices with respect to selected modelling issues and other technical

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challenges; 2) survey administration; 3) analysis of submitted survey responses to develop results andinsights; 4) a planned expanded task group workshop; and 5) development of the final task report.
Chapter 3 summarises results and insights obtained from analysis of submitted responses to the surveythat was administered by e-mail to elicit and characterise international practices with respect toselected modelling issues and other technical challenges. Results and insights are grouped by survey
part and associated Level 3 PSA technical element.
Chapter 4 identifies and summarises key conclusions from this activity, including typical practices
and common technical challenges or limitations that were identified through analysis of submitted
survey responses. The chapter also provides recommendations for future activities that follow fromthese conclusions.
Chapter 5 identifies the list of reference documents cited throughout the main report.
Appendix A provides the CSNI-approved CAPS for this WGRISK activity.
Appendix B provides the final survey instrument that was distributed by e-mail to WGRISK member
and observer countries to elicit and characterise international practices with respect to selectedmodelling issues and other technical challenges.
Appendix C provides the ten survey responses that were submitted by representatives of ten countries
in support of this activity. Other than changing the formatting to be consistent with the rest of thereport and correcting any obvious typographical errors, the detailed responses that survey respondents
submitted have been preserved in their entirety.

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2. PROCESS AND METHODSThis chapter summarises the process and methods used to complete this activity and includes fivesections that respectively address: 1) design of a survey instrument that was used to elicit andcharacterise international practices with respect to selected modelling issues and other technicalchallenges; 2) survey administration; 3) analysis of submitted survey responses to develop results andinsights; 4) a planned expanded task group workshop; and 5) development of the final task report.

2.A.

Survey design

As the lead organisation for this activity, the US Nuclear Regulatory Commission (NRC) established
an internal working group comprised of technical staff with diverse expertise pertaining todevelopment and application of the off-site radiological consequence analysis element of a Level 3
PSA. The purposes of this working group were to: 1) identify and prioritise specific modelling issues
and other technical challenges in performing Level 3 PSA off-site radiological consequence analyses
that could benefit from information exchange and sharing of methods and practices across theinternational PSA community; 2) develop an initial draft of the survey instrument that would be used
to elicit and characterise international practices with respect to selected modelling issues and othertechnical challenges; and 3) develop an example response that would be distributed with the final
survey instrument to provide survey respondents with guidance by way of an example on the scopeand level of detail that would be expected for responses to each of the survey items.
The draft survey instrument that the USNRC working group developed was initially distributed bye-mail to members of both the core task group and expanded task group for review in September
2015. Two specific questions were posed to task group members for consideration as they performed
their review of the draft survey instrument:
1.
2.

Does the survey fail to address any important Level 3 PSA modelling issues or technicalchallenges that you believe should be included?
Are there any questions (or groups of questions) that you believe can be removed from thesurvey to reduce the burden on respondents and to maximise our response rate?

From October 2015 to March 2016, the draft survey instrument was then iteratively revised to address
the feedback received from members of both the core task group and expanded task group after theirreview. In April 2016, after completion of these revisions and before distribution to WGRISK
member and observer countries, the survey instrument and example response that the USNRC
working group had developed were distributed by e-mail for review to representatives of thefollowing five groups within and beyond the NEA that were identified in Chapter 1:
1) NEA/CSNI/WGAMA; 2) NEA/CNRA/WGPC; 3) NEA/CRPPH; 4) NEA/NDC/EG-COSTNA; and
5) IAEA. No additional feedback was received after this review.
The final survey instrument, which is provided in Appendix B, included three parts. Part 1(Respondent information) was designed to obtain respondent information that would allow the task
group to determine whether responses differ across different countries or types of organisations. Part 2(Application of Level 3 PSA) was designed to obtain information about whether and how respondents
or organisations use or intend to use Level 3 PSA. Since the purpose or intended use of any PSA can

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influence multiple aspects of a PSA model (e.g. scope, level of detail, assumptions, or other modelling
choices), it was judged that responses to Part 2 would provide important contextual information forunderstanding and evaluating responses to Part 3. Part 3 (Level 3 PSA Modelling Issues andTechnical Challenges) was designed to elicit information about whether and how respondents ororganisations are addressing specific modelling issues and other technical challenges that practitioners
encounter in performing off-site radiological consequence analyses in support of Level 3 PSA
applications for nuclear installations. To facilitate identification and organisation of this information,
survey items in Part 3 were grouped by the associated major technical elements or tasks that typicallycomprise such off-site radiological consequence analyses. These technical elements and therelationships between them are illustrated below in Figure 1.
Figure 1. Typical Level 3 PSA Technical Elements and Their Relationships

Note: This figure illustrates technical elements that typically comprise off-site radiological consequence
analyses performed as part of Level 3 PSAs and the relationships between elements. Survey items
pertaining to Level 3 PSA modelling issues and technical elements were grouped by associated technicalelement to facilitate identification and organisation of this information.

2.B.

Survey administration

The WGRISK Secretariat distributed the final survey instrument and example response to WGRISK
member and observer countries by e-mail in May 2016. Completed survey responses were requested
to be submitted directly to the USNRC point of contact for the activity by July 2016. However, basedon requests from individual survey respondents who had expressed an intent to complete the survey,
but who needed additional time to do so, the deadline for accepting completed survey responses was
extended to December 2016.

2.C.

Analysis of submitted survey responses

From January 2017 to March 2017, the core task group analysed submitted survey responses tosummarise results and to develop key insights. The first step was to develop a preliminary summaryof survey responses to: 1) determine whether and to what extent a planned expanded task group
workshop described in the next section would be needed; and 2) identify and prioritise the key
technical challenges or modelling issues that would be addressed in the planned expanded task group
workshop, if the group determined a workshop was needed or would add significant value.
Once it was determined whether an expanded task group workshop would be held, the core task groupperformed more detailed analysis of submitted survey responses to summarise results and to develop
key insights derived from items in each part of the survey.
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2.D.

│ 19

Planned expanded task group workshop

As described in the CAPS provided in Appendix A, the original plan was to use the survey to generate
results and insights for use in a subsequent expanded task group workshop that would provide a forum
for more detailed discussions and information exchange on key Level 3 PSA topics. However, it waslater determined that the time and resources needed to conduct an expanded task group workshop
would not be justified for two main reasons: 1) the collective information provided in the submitted
survey responses was generally judged to be of sufficient scope and level of detail to satisfy theobjectives of this activity; and 2) where additional information was needed, this information could becollected by direct e-mail or telephone correspondence with individual survey respondents. The task
group thus agreed that the additional benefit that could be obtained from an expanded task group
workshop did not warrant the use of additional time and resources.

2.E.

Final task report

Based on a review and analysis of the submitted survey responses, the core task group developed thisfinal task report to summarise the results. This final task report represents the final deliverable for thisWGRISK activity.

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3. SURVEY RESULTS AND INSIGHTSThis chapter summarises results and insights obtained from analysis of submitted responses to thesurvey that was administered by e-mail to elicit and characterise international practices with respect toselected modelling issues and other technical challenges. Results and insights are grouped by survey
part and – for the Level 3 PSA modelling issues and technical challenges addressed in Part 3 – by the
associated Level 3 PSA technical element.

3.A.

Part 1: Respondent Information

This section summarises results and insights from Part 1 of the survey, which was designed to obtain
respondent information that would allow the task group to determine whether responses differ acrossdifferent countries or types of organisations.

3.A.1. Countries represented
Representatives from ten WGRISK member countries submitted a total of ten responses to the survey:
one survey response per country. Table 1 summarises the countries represented, along with theorganisation types that were represented by survey respondents.
All ten survey responses that were submitted in support of this activity, including the example
response developed by the USNRC, are provided in Appendix C to this report.

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Table 1. Countries and Organisation Types Represented by Survey Respondents
Country
Belgium

Organisation Types Represented by Survey Respondents
Technical Support Organisation (Bel V)
Industry (Tractebel)

CanadaRegulatory Authority (CNSC)

Finland

Academic/Research (VTT)Regulatory Authority (STUK)

Hungary

Academic/Research (MTA EK)Technical Support Organisation (NUBIKI)
Regulatory Authority (HAEA)

Japan

Academic/Research (JAEA)

Korea

Academic/Research (KAERI)
Regulatory Authority (KINS)

Netherlands

Technical Support Organisation (NRG)
Regulatory Authority (ANVS)

SwedenRegulatory Authority (SSM)

Switzerland

Regulatory Authority (ENSI)

United States

Regulatory Authority (USNRC)

3.A.2. Organisation types represented
Table I shows that representatives from multiple organisation types often collaborated to submit a
joint response to the survey, thereby resulting in multiple organisation types being represented in asingle survey response. Figure 2 further illustrates that survey respondents represented a diverse set oforganisation types, with 50% representing regulatory authorities and the other 50% representing a mixof academic/research, industry, and technical support organisations (TSOs).
Figure 2. Distribution of Organisation Types Represented

Note: Survey respondents represented a diverse set of organisation types, with 50% representing regulatory
authorities and the other 50% representing a mix of academic/research, industry, and TSOs.

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3.B. Part 2: Application of Level 3 PSA
This section summarises results and insights from Part 2 of the survey, which was designed to obtain
information about whether and how respondents or organisations use or intend to use Level 3 PSA.
Since the purpose or intended use of any PSA can influence multiple aspects of a PSA model(e.g. scope, level of detail, assumptions, or other modelling choices), it was judged that responses toPart 2 would provide important contextual information for understanding and evaluating responses toPart 3 of the survey (Level 3 PSA Modelling Issues and Technical Challenges).

3.B.1. Level 3 PSA Requirements
20% (2 out of 10) of countries surveyed indicated that there is a requirement to perform Level 3 PSAs
for nuclear installations in their country. The two countries that require the performance of Level 3
PSAs for nuclear installations are: 1) Korea and (2) Netherlands.
Korea requires applicants for an operating licence for new nuclear power plants to perform a Level 3
PSA that demonstrates two safety goal quantitative health objectives (QHOs) are satisfied:
1.
2.

Early fatality risk from nuclear accidents shall not exceed 0.1% of the total early fatality
risk arising from other accidents (4E-07 per year).
Cancer fatality risk from nuclear accidents shall not exceed 0.1% of the total cancer
fatality risk (1E-06 per year).

Emergency response protective actions (e.g. evacuation, sheltering, potassium iodide) are not credited
in determining whether these risk acceptance criteria are satisfied. However, the regulatory authority,
Nuclear Safety and Security Commission (NSSC), recently required confirmation of effectiveness of
emergency response protective actions in the licensing process for Shin-Hanul Units 1 and 2(Advanced Power Reactor (APR)-1400 series). The licensee is thus assessing the effectiveness of
relevant protective actions in sensitivity analyses.
Netherlands requires a Level 3 PSA to show that a nuclear installation containing fissile materials
and/or ores complies with certain risk acceptance criteria when applying for a licence (or a change ofthe license) to establish, construct, commission, operate, or decommission the installation. These risk
acceptance criteria include:
1.

The probability that an individual who resides permanently and unprotected outside thefacility grounds dies from a beyond design-basis accident (BDBA) is less than 1E-06 per
year. To calculate this individual risk, the characteristics of this individual (e.g. age,location, habits) are chosen to result in the worst-case scenario that would still bereasonable, but does not have to be a real person. In practice, this is a 1-year-old infant
residing next to the edge of the nuclear installation. In determining the total dose resulting
from a BDBA, long-term (stochastic) effects for at least 50 years are considered, including
ongoing exposures over that time-period due to contamination of the environment.

2.

The probability that a BDBA leads to at least 10 direct fatalities within a few weeks, due todeterministic effects, is less than 1E-05 per year, and the probability of n times more
fatalities is n2 times smaller. For example, the probability of at least 100 fatalities is lessthan 1E-07 per year and the probability of at least 1 000 fatalities is less than 1E-09 per
year. To calculate this group (societal) risk, the actual population distribution around thenuclear installation is to be considered.

Like Korea, emergency response protective actions are also not credited in determining whether theserisk acceptance criteria are satisfied in the Netherlands.

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3.B.2. Use of Level 3 PSA
70% (7 out of 10) countries surveyed indicated they are currently performing, planning to perform, orconsidering performing Level 3 PSAs for nuclear installations. These seven countries include:
1) Finland (performing); 2) Hungary (planning to perform); 3) Japan (performing); 4) Korea (required
and performing); 5) Netherlands (required and performing); 6) Sweden (considering performing); and
7) United States (performing).
Table 2 summarises uses of Level 3 PSA across all countries surveyed, including any specificapplications that were identified by survey respondents. Specific risk metrics for each type of
application and any corresponding criteria used to evaluate results are identified in the individual
responses provided in Appendix C. In general, risk metrics used are driven by the objectives of theLevel 3 PSA. Commonly used risk metrics include: 1) population dose risk; 2) individual risk ofradiological health effects (e.g. individual early fatality risk, individual latent cancer fatality risk); and3) group (societal) risk of radiological health effects (e.g. risk of exceeding a specified number offatalities).
Table 2. Uses of Level 3 PSA across all countries surveyedCountry

Use of Level 3 PSA

Applications

Belgium
CanadaFinland
Hungary
Japan

No
No
Yes
Planning
Yes

Not applicable.
Not applicable.
Applied research.
Applied research.

Korea

Yes
Required

Netherlands

Yes
RequiredConsideringNo
Yes

Comparison with risk acceptance criteria, safety goals,
or other quantitative objectives.
Inform development of severe accident management
plans.
Comparison with risk acceptance criteria, safety goals,
or other quantitative objectives.

SwedenSwitzerland
United States

Comparison with risk acceptance criteria, safety goals,
or other quantitative objectives.
Evaluation of protective action effectiveness to informemergency preparedness and response guidelines.

Applied research.
Not applicable.
Comparison with risk acceptance criteria, safety goals,
or other quantitative objectives as part of regulatory or
backfit analyses.
Environmental assessments.
Applied research.

Among countries that are not currently performing, planning to perform, or considering performing
Level 3 PSAs for nuclear installations (Belgium, Canada, Switzerland), four potential barriers to
performing Level 3 PSA were identified: 1) absence of a technical or legal framework to perform
Level 3 PSA; 2) large uncertainties in Level 3 PSA results, especially when combined withuncertainties propagated from Level 1 and Level 2 analyses; 3) limited expected benefit in terms ofpotential back fits or safety improvements; and 4) additional resources required to perform detailed
off-site radiological consequence analyses. These countries further identified a range of alternative

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methods to estimating off-site public risks attributable to accidental releases of radiological materials
from nuclear installations. These alternative methods include use of: 1) surrogate metrics based on
results from Level 1 PSAs (e.g. core damage frequency (CDF)) and Level 2 PSAs (e.g. large early
release frequency (LERF), large release frequency (LRF)); 2) deterministic design-basis accident(DBA) dose consequence assessments with conservative assumptions to demonstrate compliance withprescribed dose limits; and 3) dose consequence assessments to demonstrate compliance with a rangeof specified dose limits that vary with the expected frequency of events.

3.B.3. Siting of nuclear installations
Although survey respondents generally acknowledged the potential benefit of using results fromLevel 3 PSAs to inform decisions with respect to siting of nuclear installations, most indicated that
Level 3 PSAs are not being used to support such applications. Instead, respondents stated that resultsfrom deterministic DBA dose consequence assessments are used to demonstrate that doses toindividuals located at any point along the boundaries of defined areas around a nuclear installation for
specified periods of time following postulated fission product releases are within prescribed limits.

3.B.4. Establishment of emergency planning or protective action zones
Practices appear to vary with respect to consideration of Level 3 PSA insights in establishing the sizes
and boundaries of emergency planning or protective action zones, with some countries actively reevaluating the technical basis for their existing zones. Some countries relied solely on deterministic
DBA analyses and various dose-distance criteria in establishing existing zones. Others supplemented
such analyses with risk insights from previous Level 3 PSAs (e.g. WASH-1400) or considered theeffect of weather variability and assessed the probability of exceeding specified dose thresholds for
implementing various protective actions.
A typical practice with respect to establishing emergency planning or protective action zones is that
results from deterministic analyses provide the primary analytical basis for established zones. If
Level 3 PSA insights are used in developing requirements for the sizes and boundaries of emergency
planning or protective action zones, such risk insights are typically considered as a complement toresults from deterministic analyses as part of a risk-informed approach.

3.B.5. Development of safety goals or risk acceptance criteria
A previous WGRISK project addressed the development and application of probabilistic risk criteria
and safety goals in many countries [3]. Among countries with defined risk acceptance criteria, safety
goals, or other quantitative objectives that participated in the current project (Japan, Korea,
Netherlands, United States), survey respondents generally indicated that results from Level 3 PSAs
are used to evaluate compliance with or attainment of established criteria related to radiological healtheffects, rather than to inform development of the established criteria. In particular, for those countries
that identified criteria related to individual and/or group (societal) risk of radiological health effects,
two respondents (Korea and United States) indicated that these criteria reflect a value judgement
about the level of incremental (additional) risk to the public from nuclear installation operations
relative to background risk that was judged to be acceptable; in both countries, a level of 0.1% of the
background risk was selected to establish QHOs for individual early fatality risk and individual latent
cancer fatality risk.
Although results from Level 3 PSAs were not explicitly used to develop these QHOs, respondents
from three out of these four countries (Japan, Korea, United States) indicated that Level 3 PSA results
are used in defining and establishing criteria with respect to subsidiary or surrogate risk metrics thatare calculated as part of Level 1 PSAs (e.g. CDF) or Level 2 PSAs (e.g. LERF, LRF, containment
failure frequency (CFF)) to support risk-informed decision making.
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3.B.6.

│ 25

Other applications of off-site radiological consequence analyses

Survey respondents identified a wide range of other uses of off-site radiological consequence analyses
that are performed independently of Level 3 PSA. Notable examples of these other uses include:
1) deterministic DBA dose consequence assessments used to demonstrate that doses to individuals
located at any point along the boundaries of defined areas around a nuclear installation for specified
periods of time following postulated fission product releases are within prescribed limits as part ofsiting or emergency planning applications; (2) decision support for real-time response to nuclear
incidents or accidents, including dose projections to inform or evaluate protective action
recommendations; and (3) environmental assessments in support of licensing applications.

3.C. Part 3: Level 3 PSA modelling issues and technical challenges
3.C.1.

General

3.C.1.1. Computer codes for off-site radiological consequence analysis
Several computer codes for performing off-site radiological consequence analyses were identified insurvey responses. Some of these identified codes are used exclusively for applications not related to
Level 3 PSA that were described in Section 3.B.6. Table 3 lists the probabilistic consequence analysis
codes that respondents from each country identified for use as part of Level 3 PSA. A key finding
from the analysis of survey responses is that many typical practices identified as part this WGRISK
activity are driven by the limitations or capabilities of these probabilistic consequence analysis codes.
This finding is described in more detail in Chapter 4.
Table 3. Probabilistic consequence analysis computer codes usedfor Level 3 PSA applications
Country

Level 3 PSA Probabilistic Consequence Analysis Computer Code(s)Belgium

Not applicable.

CanadaNot applicable.

FinlandARANOMELCOR Accident Consequence Code System (MACCS)

Hungary

COSYMAJapan

Off-Site Consequence Analysis code for Atmospheric Release in reactor
accident (OSCAAR)

Korea

MELCOR Accident Consequence Code System (MACCS)

Netherlands

COSYMA
NUDOS2
MELCOR Accident Consequence Code System (MACCS)

SwedenLENA_P

Switzerland

Not applicable.

United States

MELCOR Accident Consequence Code System (MACCS)

3.C.1.2. Computer codes for risk characterisation
A notable finding from analysis of the survey responses is that none of these probabilistic
consequence analysis codes can directly perform risk characterisation. Risk characterisation is the

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process of developing and evaluating risk triplets comprised of accident scenarios, frequencies, andconditional consequences to produce: 1) qualitative insights about accident scenarios and significant
risk contributors; and 2) quantitative estimates for risk metrics of interest. These quantitative
estimates are typically calculated by combining estimates of radiological release category frequencies
from Level 2 PSA analyses with conditional consequence metric estimates from Level 3 PSA
analyses for each radiological release category.
Among countries that perform Level 3 PSA, three countries (Korea, Netherlands, and United States)
indicated that risk characterisation is performed by post processing of results using an application
external to the probabilistic consequence analysis code (e.g. spread sheet software or graphical userinterface software). For example, Korea (KAERI) utilises KOSCA-OUTPUT (Korea off-site
consequence analysis software package – Output post-processing module) for risk characterisation.
KOSCA-OUTPUT is a graphical user interface module that is used to post-process the consequence
results from MACCS and release frequency data from the Level 2 PSA to calculate selected risk
metrics.

3.C.1.3. Scope of Level 3 PSAs
Among countries that perform Level 3 PSA, survey respondents identified a range of potential
accident sequence initiating event hazards that would be considered as part of the Level 1 analysis inestablishing the scope of a Level 3 PSA. Some individual responses in Appendix C identify specificinitiating event hazards or hazard groups that are typically considered. Although some countries do
not have experience with considering external events in the context of Level 3 PSA, most countries
indicated that both internal and external initiating event hazards should be considered.
Practices vary with respect to the treatment of potential correlation between the causes of accident
sequence initiating events and off-site phenomenological and consequence modelling. Finland
indicated it accounts for this potential correlation and cited an example of the correlation between thetime of year in which an accident sequence is initiated and use of appropriate meteorological data forthat time of year. Netherlands indicated it only accounts for this potential correlation if failure toaccount for the correlation would result in a non-conservative estimate of off-site consequences.
Korea and United States described a partial treatment of this correlation by developing uniqueconsequence models for seismic events that impact protective action parameters (e.g. delay time toevacuation, evacuation speed, shielding parameters). Other countries stated that this potential
correlation is not treated.

3.C.1.4. Ongoing research and development
Survey respondents identified a range of ongoing research and development activities pertaining toLevel 3 PSA and off-site radiological consequence analysis. Specific activities are identified in theindividual responses provided in Appendix C. Notable examples of research and development areasthat are being addressed through ongoing efforts include: 1) optimisation of emergency response andpost-accident recovery decisions; 2) development of Level 3 PSA guidance documents; and
3) enhancing the modelling capabilities of probabilistic consequence analysis codes(e.g. incorporating additional atmospheric transport and dispersion (ATD) and economic models,
incorporating models to account for potential concurrent releases from multiple co-locatedradiological sources with unique radionuclide inventories and accident progression timelines).

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3.C.2.

│ 27

Radionuclide release characterisation

3.C.2.1. Interface between Level 2 analysis and Level 3 analysis
Specific practices vary with respect to radionuclide release characterisation and treatment of theinterface between the Level 2 analysis and the Level 3 analysis. Detailed descriptions of processes
and criteria are provided in the individual survey responses in Appendix C.
Level 2 PSA analysts typically work closely with Level 3 PSA analysts in radionuclide releasecharacterisation to: 1) define radiological release categories or source term groups; and 2) select a
representative accident sequence for each radiological release category or source term group to
estimate the source term characteristics that will serve as an input to the off-site radiologicalconsequence analysis. This practice increases the efficiency of selecting representative accident
sequences and reduces the computational time needed to generate time-dependent radiological releasedata using relevant severe accident analysis computer codes. For most countries that perform Level 3
PSA, on-site severe accident mitigation actions are considered in the Level 2 analysis performed toestimate characteristics of radiological releases (e.g. release fraction timing, release truncation time).
In addition, most countries that perform Level 3 PSA described methods for performing sensitivity or
uncertainty analyses on release categories or release fractions in their individual responses inAppendix C.

3.C.2.2. Set of radionuclides used to characterise off-site radiological consequences
The set of radionuclides used to characterise the off-site radiological consequences attributed toaccidental releases of radiological materials varies across countries with the number ranging from60 to 800. Based on analysis of survey responses, it appears that the computer codes used to perform
off-site radiological consequence analysis is an important driver of the analyst’s decision with respect
to which radionuclides will be considered.

3.C.2.3. Treatment of concurrent releases from multiple co-located radiologicalsources
Practices also vary with respect to consideration of concurrent releases from multiple units orradiological sources co-located at the same site. Survey respondents from a few countries indicatedthey have performed research analyses of the 2011 accident at the Fukushima Daiichi Nuclear Power
Stations that considered multi-unit or multi-source releases. Among countries that have consideredmulti-unit or multi-source releases beyond benchmarking the Fukushima Daiichi accident, modelling
practices vary. For example, Korea has developed an approach in which the release fraction from asingle unit is multiplied by the number of units on a site to develop the multi-unit source term input
for the MACCS computer code. By contrast, Netherlands uses post processing (typically with spread
sheet software) to sum the individual risk contributions attributable to accidental releases frommultiple units. As part of an ongoing Level 3 PSA research project, the United States is exploring theuse of a new multi-source modelling capability in MACCS to model releases from multiple co-locatedradiological sources that can have unique radionuclide inventories and potentially different accident
progression timelines; the objective of this effort is to assess the contribution to total site risk from
accidental releases involving different combinations of two or more major radiological sources at themodelled site.

3.C.3.

Meteorological data

Among countries that perform Level 3 PSA, a typical practice with respect to the use of
meteorological data is the use of site-specific data that includes hourly observations for an entire year

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to perform off-site radiological consequence analyses. Using the probabilistic consequence analysis
codes identified in Section 3.C.1., most countries implement code-specific weather binning andsampling strategies to account for the impact of temporal variability in weather conditions on off-siteradiological consequences. A notable exception to this is the ARANO code used in Finland, which
does not rely on sampling to perform probabilistic analyses. In ARANO, annual weather data fromone meteorological mast is converted into a joint frequency matrix of annual weather statistics. Doses
are calculated for all specified weather conditions and results are then weighted by the probability ofeach weather condition.
More specific details about the treatment of temporal and spatial variability in weather conditions,
including methods for performing sensitivity or uncertainty analyses, are provided in the individual
responses in Appendix C.

3.C.4.

Environmental transport and dispersion

3.C.4.1. ATD Modelling
Most probabilistic consequence analysis codes identified in Section 3.C.1 are limited to Gaussianstraight-line trajectory plume segment or puff ATD models. This is a significant driver typicalpractices with respect to ATD modelling, including the use of one-hour time steps for updating
calculations to be consistent with available site-specific meteorological data. However, some codesused to perform Level 3 PSA are being enhanced to include a Lagrangian particle tracking ATD
model, which – along with the Eulerian ATD model – has already been used within primarily
deterministic consequence analysis codes that can account for changes in off-site weather conditions(e.g. the Finnish VALMA and SILAM codes). For example, a new particle tracking ATD model
based on the HYbrid Single Particle Lagrangian Integrated Trajectory (HYSPLIT) model is underdevelopment for addition to MACCS and will be released in a future version. This particle tracking
ATD model is being developed to provide an alternative ATD model that addresses known limitationsof the Gaussian straight-line trajectory plume segment model. As more ATD model options become
available, a process and criteria will be needed to select which model(s) will be used for a particularstudy. Some factors to consider in making the selection may include: 1) study objectives;
2) computational efficiency; 3) site characteristics; and 4) data availability.
In addition, although it is possible (at least in principle) to account for spatial variability in the ATD
response due to variation in the topography or terrain around the modelled site, this is beyond thecurrent state of practice for Level 3 PSA. In particular, available probabilistic consequence analysis
codes either do not support terrain modelling or rely on the use of simple terrain models.

3.C.4.2. Spatial modelling
A typical practice among countries that perform Level 3 PSA is to use site-specific information
sources (sometimes supplemented with generic information sources) to model various sitecharacteristics, including: 1) population characteristics; 2) agricultural characteristics; and3) economic characteristics.
Another finding based on analysis of survey responses is that the size of the modelled region andspatial intervals around a site are typically application- and site-specific. In general, a larger numberof spatial intervals with finer resolution grid elements are used for the region close to the site, with thenumber of spatial intervals and grid resolution decreasing as the distance from the site increases.

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3.C.4.3. Radiological releases to aqueous pathways
Respondents generally agreed that radiological releases to aqueous pathways should be considered ifthey could result in significant doses to off-site populations. However, radiological releases toaqueous pathways are typically screened out and excluded from Level 3 PSAs for nuclear installations
that have the potential for airborne radiological releases, since airborne releases are expected todominate radiological health risks. In particular, the United States provided two reasons for this: 1)
movement of radionuclides to the accessible environment through aquatic pathways is expected to be
slow relative to atmospheric transport; and 2) releases to groundwater or surface water are consideredeasier to interdict [4]. In addition, the United States response in Appendix C includes information
about a research project the USNRC initiated after the Fukushima Daiichi accident to assess potential
off-site consequences of losing control of highly contaminated water in a severe accident in which thecontaminated water flows to a body of water. [5]

3.C.5.

Protective action (countermeasure) modelling

As previously described in Section 3.B.1, the two countries that require performance of Level 3 PSAs
for comparison with risk acceptance criteria (Korea and Netherlands) conservatively do not model orcredit emergency response protective actions in determining whether prescribed risk acceptancecriteria are satisfied. However, the Korean regulatory authority NSSC recently required confirmationof the effectiveness of protective actions in the licensing process for Shin-Hanul Units 1 and 2(APR1400 series).
Among the remaining countries that perform Level 3 PSA, the following protective actions are
typically modelled for the early (emergency) phase of accident response: 1) evacuation; 2) sheltering;3) dose-dependent relocation; and 4) use of potassium iodide (KI) as a supplementary protective
action to reduce the radiological dose to the thyroid gland by blocking the uptake of radioiodine.
Protective actions that are typically modelled for the intermediate phase include: 1) dose-dependent
relocation; and 2) dose-dependent interdiction or bans on agricultural products (e.g. dairy products,
meats) and drinking water. Finally, protective actions that are typically modelled for the late(recovery) phase include: 1) dose-dependent relocation; 2) decontamination of contaminated land
areas; 3) temporary interdiction or permanent interdiction (condemnation) of contaminated land areas;and 4) dose-dependent interdiction or bans on agricultural products and drinking water.
For those countries that model protective actions, most typically do not model population groups withdifferent protective action behaviours. A notable exception to this is the United States, which uses
site-specific information to select and define multiple evacuation cohorts based on several factors,
including: 1) location or spatial interval around the nuclear installation; 2) delay times for
implementing protective actions; 3) evacuation speeds; and 4) exposure factors and shielding
parameters. Moreover, although Finland indicated it had used a simple probabilistic model to accountfor the probabilities of success or failure for protective actions in a research project, all countries thatmodel protective actions indicated this is not typical practice when performing Level 3 PSAs fornuclear installations.

3.C.6.

Radiological exposure and dose assessment

Among countries that perform Level 3 PSA, practices vary with respect to defining potentially
exposed populations or cohorts, including consideration of potentially sensitive, vulnerable, or criticalgroups (e.g. infants, most exposed individual with otherwise normal characteristics). Some countries
consider potentially sensitive, vulnerable, or critical groups, while others do not differentiate suchgroups from average members of the general public and use generic average values for exposure anddose conversion factors.

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Since the objective of Level 3 PSA is to assess the risks to the off-site public that are attributable to aspectrum of possible accident scenarios involving a nuclear installation, countries that perform
Level 3 PSA typically exclude the on-site worker population from the scope of the linked radiologicalconsequence analysis.
In assessing radiological exposure and dose to the off-site public, countries that perform Level 3 PSA
generally model all relevant exposure pathways that can result in a significant contribution toradiological dose. These exposure pathways typically include: 1) direct exposure to external radiationfrom the plume of released radiological materials (cloud shine); 2) inhalation of radioactivity in theplume; 3) contamination of skin and clothing; 4) direct exposure to external radiation from groundcontamination (ground shine); 5) inhalation of resuspended radioactivity; and 6) ingestion ofcontaminated food and water.

3.C.7.

Radiological health effects

3.C.7.1. Types of radiological health effects considered
All countries that perform Level 3 PSA model and estimate radiological health effects in linked offsite radiological consequence analyses. Individual responses provided in Appendix C include more
specific information about specific health effects considered, dose-response models, and target organs
used in these dose-response models. In general, a range of early radiological health effects are
estimated using deterministic models, including: 1) early fatalities; 2) acute radiation syndrome(radiation sickness); and 3) other early injuries arising from acute doses to the red bone marrow,lungs, gastrointestinal tract, and skin. Fatal and non-fatal cancers involving multiple organs generally
represent the latent radiological health effects that are estimated using stochastic models.

3.C.7.2. Dose-Response models used to estimate excess cancer risk
Survey respondents from all countries that perform Level 3 PSA indicated they use a dose-responsemodel based on the linear no-threshold (LNT) hypothesis to estimate the excess cancer risk
attributable to radiological doses caused by accidental releases from nuclear installations. Althoughprevailing knowledge of biological mechanisms and available epidemiologic evidence continue tosupport the LNT hypothesis, emerging evidence from in vitro radiation biology studies has spurred
heated debate within the scientific community about the validity of LNT-based dose-response models.
However, only the United States indicated that it is evaluating the impact of plausible alternative
dose-response models using sensitivity analyses to account for uncertainty about the true doseresponse relationship for exposures to low levels of ionizing radiation. In recent U.S. studies [6], thedefault LNT-based model was used in the base case analysis, and two models involving dose
truncation levels were used as sensitivity cases. In the first model, the dose truncation was set at
6.2 mSv/year - a level that corresponds to the background radiation exposure an average individual in
the US receives per year, including exposures attributed to medical and other man-made radiologicalsources. In the second model, the dose truncation was set at 50 mSv/year with a 100 mSv lifetime
limit – a level below which the Health Physics Society recommends against quantitative estimation ofhealth risks [7] (though the USNRC has not endorsed this position).

3.C.8.

Economic consequences

Among countries that perform Level 3 PSA, respondents from two countries (Korea and Netherlands)
indicated that economic consequences are not considered in linked off-site radiological consequence
analyses. For the remaining countries, economic models are generally used to estimate the costs
attributed to implementation of modelled protective actions to reduce radiological dose to the off-site
public from the accidental release of radiological materials. Examples of costs considered include:

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1) daily costs of compensation for populations subject to evacuation or short-term relocation arisingfrom food, housing, transportation, lost income, or replacement of lost personal property; 2) costs oflong-term relocation of populations and businesses in interdicted land areas; 3) depreciation costs that
account for loss of value of interdicted property; 4) decontamination costs; and 5) costs arising from
implementation of agricultural countermeasures.
In addition to these economic costs attributed to protective actions that are implemented to reduceradiological dose, Japan indicated that it also translates early and latent radiological health effects into
economic consequences using input parameters that estimate the associated lost income. Although not
performed as part of a Level 3 PSA, the United States may monetise the benefits of averted
radiological health effects in cost-benefit analyses using a conversion factor applied to the averted
population dose within a prescribed distance (typically 50 miles) from the nuclear installation. The
USNRC is currently revising the value of this dollar per person-rem conversion factor and is
proposing a process for routine updates to ensure its value is maintained current. Finally, Korea(KAERI) is currently performing research to evaluate the effects of various factors considered in theeconomic consequence model.

3.C.9.

Consequence quantification and reporting

For the two countries that require performance of Level 3 PSAs for comparison with risk acceptancecriteria (Korea and Netherlands), the consequence metrics selected for quantification and reporting are
driven by the legal requirements associated with these risk acceptance criteria related to radiologicalhealth effects.
For the remaining countries, respondents generally indicated there is no explicit criteria for selecting
consequence metrics for quantification and reporting, except that the objectives of the analysis should
drive the selection of relevant metrics. In general, this means the consequence metrics will be selected
on a case-by-case basis. The United States identified several factors that are typically considered inselecting consequence metrics and the spatial intervals over which they will be quantified. These
factors include: 1) analysis objectives – which are typically linked to application-specificrequirements or questions to be addressed; 2) stakeholder interests; 3) standards requirements or stateof-practice; 4) potential future uses of results and insights; 5) schedule and resource constraints; and6) capabilities and limitations of models and analytical tools.

3.C.10. Risk characterisation
3.C.10.1. Selection of risk metrics
In response to a question about the process or criteria used to select risk metrics for effectively
communicating Level 3 PSA results, survey respondents generally used the same response provided
for the question about selection of consequence metrics described in Section 3.C.9.
Only the United States addressed the question of metrics designed to measure the effects ofconcurrent radiological releases involving multiple units or radiological sources co-located at thesame site. The USNRC indicated that, in principle, risk metrics used to estimate the frequencies ofoff-site public health, economic, and environmental consequences for accidental releases from single
radiological sources can be adapted by adjusting the frequency basis to estimate the same quantities
for accidental releases from multiple radiological sources. For example, instead of quantifying risk
metrics on a per-reactor-year basis, they can be quantified on a per-site-year or per-calendar-yearbasis. The USNRC is exploring options in this area as part of an ongoing Level 3 PSA research
project.

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3.C.10.2. Presentation and communication of risk results
Although practices vary among countries that perform PSA with respect to the presentation andcommunication of risk results, some typical practices were identified. In general, risk results thatcharacterise variability or aleatory uncertainty arising from inherent randomness or stochastic
processes are presented using a variety of formats and graphical displays. Commonly used examples
include: 1) point estimates – especially expected (mean) risk of selected consequence metrics over allweather trials; 2) FN curves that illustrate the frequency of N or more consequences (F) as a functionof total number of consequences (N) to express societal risk; and 3) complementary cumulative
distribution function (CCDF) curves (also termed exceedance frequency curves or risk curves) –which represent the frequencies of exceeding different consequence levels, and are typically used in a
Level 3 PSA to illustrate the effect of variability in off-site weather conditions on off-site radiologicalconsequences.
Commonly used examples of graphical displays used to illustrate uncertainties in risk results include :1) empirical probability distributions for selected risk metrics – typically in the form of probability
density function (PDF) curves and/or cumulative distribution function (CDF) curves; 2) box plots that
illustrate the locations of key summary statistics (e.g. mean value, 50th percentile (median value),
95th percentile, and 5th percentile) for selected risk metrics; and 3) sets of CCDF curves – with thedifferent curves representing different probabilities of frequencies of exceeding different consequencelevels.
Respondents did not differentiate between decision makers and the general public with respect tomethods for presenting and communicating risk results and the uncertainty in risk results. In addition,
most respondents indicated they do not have a communication plan or guidelines for risk
communication. One exception was the United States; the USNRC has developed guidance
documents that aim to enhance risk communication with both internal stakeholders (including
decision makers) [8] and external stakeholders (including the general public) [9].
More specific details about the presentation and communication of risk results – including someexample displays and citations for representative reports – are provided in the individual responses inAppendix C.

3.C.10.3. Treatment of uncertainties
Among countries that perform Level 3 PSA, practices vary with respect to the use of sensitivity or
uncertainty analyses to characterise the uncertainty in risk results. More specific details about thetreatment of uncertainties are provided in the individual responses in Appendix C. The range ofpractices identified in survey responses included use of: 1) information in previous studies toconclude that risk results are accurate to within one order of magnitude; 2) limited scope sensitivity
analyses focused on evaluating the impact on risk results of using alternative parameter values orinput data sets for a selected set of parameters or input data; and 3) Monte Carlo based sampling
methods to propagate parameter uncertainties.
Most respondents who indicated they perform sensitivity analyses by varying parameter values or
uncertainty analyses by specifying uncertainty distributions for propagating parameter uncertainty
stated that they rely on expert judgement to determine which parameters will be varied or will have
uncertainty distributions specified for propagating parameter uncertainty. In addition, respondents
indicated that, while state of knowledge correlation between parameters can be treated in principle: 1)
this is a difficult task due to insufficient information; and 2) the extent to which it is treated will be
driven by the objectives of the analysis.

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3.C.11. Additional topics
As shown in the final survey instrument provided in Appendix B, survey respondents were asked to
identify: 1) any other questions they believed should have been asked in the survey; and 2) any otherinformation, technical challenges, or notable practices they would like to share with the international
community. Although most respondents did not identify any other questions or information to share,
the following additional topics were identified by some respondents:
1.
2.
3.

Consideration of environmental and non-radiological (e.g. psychosocial) impacts.
Development of novel risk metrics and methods for risk communication.
Accuracy of Level 3 PSA results.

If provided, more detailed information about each of these additional topics can be found in theindividual survey responses in Appendix C.

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4. CONCLUSIONS AND RECOMMENDATIONSThis chapter identifies and summarises key conclusions from this activity, including typical practices
and common technical challenges or limitations that were identified through analysis of submitted
survey responses. The chapter also provides recommendations for future activities that follow fromthese conclusions.

4.A. Typical practices
Analysis of submitted survey responses resulted in the identification of typical practices across several
aspects of Level 3 PSA. These typical practices are summarised in this section and organised by Level
3 PSA technical element or aspect.

4.A.1.

Application of Level 3 PSA

Typical practices around Level 3 PSA applications were identified with respect to three aspects:
1) uses of Level 3 PSA; 2) siting of nuclear installations; and 3) establishment of emergency planningor protective action zones. These typical practices are summarised below.

4.A.1.1. Uses of Level 3 PSA
Among countries that are currently performing, planning to perform, or considering performing
Level 3 PSAs for nuclear installations, five broad types of Level 3 PSA applications were identified:
1) applied research; 2) comparison with risk acceptance criteria, safety goals, or other quantitative
objectives; 3) evaluation of protective action effectiveness to inform emergency preparedness andresponse guidelines; 4) inform development of severe accident management plans; and5) environmental assessments. In general, risk metrics used for each type of application and any
corresponding criteria used to evaluate results are driven by the objectives of the Level 3 PSA.
Commonly used risk metrics include: 1) population dose risk; 2) individual risk of radiological healtheffects; and 3) group (societal) risk of radiological health effects.

4.A.1.2. Siting of nuclear installations
Although survey respondents generally acknowledged the potential benefit of using results fromLevel 3 PSAs to inform nuclear installation siting decisions, most indicated that Level 3 PSAs are notbeing used to support such applications. Results from deterministic DBA dose consequenceassessments are typically used to demonstrate that doses to individuals located at any point along theboundaries of defined areas around a nuclear installation for specified periods of time following
postulated fission product releases are within prescribed limits.

4.A.1.3. Establishment of emergency planning or protective action zones
Results from deterministic analyses provide the primary analytical basis for establishing emergency
planning or protective action zones. If Level 3 PSA insights are used in developing requirements forthe sizes and boundaries of emergency planning or protective action zones, such risk insights are

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typically considered as a complement to results from deterministic analyses as part of a risk-informed
approach.

4.A.2.

Level 3 PSA modelling and risk characterisation

This section summarises typical practices that were identified with respect to: 1) modelling choices ananalyst faces in addressing certain Level 3 PSA technical elements; and (2) presentation of Level 3
PSA results as part of risk characterisation. Typical practices related to modelling choices areorganised by Level 3 PSA technical element.

4.A.2.1. Radionuclide release characterisation
Among countries that perform Level 3 PSA, Level 2 PSA analysts typically work closely with Level 3
PSA analysts in radionuclide release characterisation to: 1) define radiological release categories orsource term groups; and 2) select a representative accident sequence for each radiological releasecategory or source term group to estimate the source term characteristics that will serve as an input tothe off-site radiological consequence analysis. This practice increases the efficiency of selecting
representative accident sequences and reduces the computational time needed to generate timedependent radiological release data using relevant severe accident analysis computer codes. Most ofthese countries indicated that they perform sensitivity or uncertainty analyses on release categories orrelease fractions to evaluate the impact on risk results. Another typical practice in this area is thatmost countries that perform Level 3 PSA consider on-site severe accident mitigation actions in theLevel 2 analysis performed to estimate source term characteristics.

4.A.2.2. Meteorological data and other information sources
Most countries that perform Level 3 PSA: 1) use site-specific data that includes hourly observations
for an entire year to perform off-site radiological consequence analyses; and 2) implement codespecific weather binning and sampling strategies to account for the impact of temporal variability inweather conditions on off-site radiological consequences. In addition, most countries use site-specificinformation sources (sometimes supplemented with generic information sources) to model various
other site characteristics, including: 1) population characteristics; 2) agricultural characteristics; and3) economic characteristics.

4.A.2.3. Spatial modelling
Survey respondents indicated that the size of the modelled region and spatial intervals around a site
are typically application - and site-specific. In general, a larger number of spatial intervals with finer
resolution grid elements are used for the region close to the site, with the number of spatial intervals
and grid resolution decreasing as the distance from the site increases.

4.A.2.4. Radiological releases to aqueous pathways
Respondents generally agreed that radiological releases to aqueous pathways should be considered ifthey could result in significant doses to off-site populations. However, radiological releases toaqueous pathways are typically screened out and excluded from Level 3 PSAs for nuclear installations
that have the potential for airborne radiological releases, since airborne releases are expected todominate radiological health risks. Two justifications for this practice were identified: 1) movementof radionuclides to the accessible environment through aquatic pathways is expected to be slow
relative to atmospheric transport; and 2) releases to groundwater or surface water are consideredeasier to forbid [4].

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4.A.2.5. Protective action (countermeasure) modelling
Among most countries that perform Level 3 PSA, a range of protective actions are typically modelled
for the early (emergency), intermediate, and late (recovery) phases of accident response. Protective
actions that are typically modelled in each phase include:•

Early (Emergency) Phase: 1) evacuation; 2) sheltering; 3) dose-dependent
relocation; and 4) use of KI as a supplementary protective action to reduce theradiological dose to the thyroid gland by blocking the uptake of radioiodine.

•Intermediate Phase: 1) dose-dependent relocation; and 2) dose-dependentinterdiction or bans on agricultural products and drinking water.

•

Late (Recovery) Phase: 1) dose-dependent relocation; 2) decontamination ofcontaminated land areas; 3) temporary interdiction or permanent interdiction(condemnation) of contaminated land areas; and 4) dose-dependent interdiction orbans on agricultural products and drinking water.

In addition, most countries that model protective actions typically do not: 1) model population groups
with different protective action behaviours; or 2) use probabilistic models to account for the
probabilities of success or failure for protective actions.

4.A.2.6. Radiological exposure and dose assessment
Since the objective of Level 3 PSA is to assess the risks to the off-site public that are attributable to aspectrum of possible accident scenarios involving a nuclear installation, countries that perform Level
3 PSA typically exclude the on-site worker population from the scope of the linked radiologicalconsequence analysis.

4.A.2.7. Radiological health effects
All countries that perform Level 3 PSA model and estimate radiological health effects in linked offsite radiological consequence analyses. In general, a range of early radiological health effects are
estimated using deterministic models, including: 1) early fatalities; 2) acute radiation syndrome(radiation sickness); and 3) other early injuries arising from acute doses to various tissues or organs.
Fatal and non-fatal cancers involving multiple organs generally represent the latent radiological healtheffects that are estimated using stochastic models.
Respondents from all countries that perform Level 3 PSA indicated they use a default dose-responsemodel based on the LNT hypothesis to estimate the excess cancer risk attributable to radiologicaldoses caused by accidental releases from nuclear installations. However, most indicated they do not
perform sensitivity or uncertainty analyses to evaluate the impact of dose-response model uncertainty
or use of plausible alternative dose-response models on estimates of excess cancer risk.

4.A.2.8. Economic consequences
Although there are exceptions, most countries that perform Level 3 PSA consider economic
consequences in linked off-site radiological consequence analyses. These countries use economic
models to estimate the costs attributed to implementation of modelled protective actions to reduceradiological dose to the off-site public from the accidental release of radiological materials. Examples
of costs considered include: 1) daily costs of compensation for populations subject to evacuation or
short-term relocation arising from food, housing, transportation, lost income, or replacement of lost
personal property; 2) costs of long-term relocation of populations and businesses in interdicted land

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areas; 3) depreciation costs that account for loss of value of interdicted property; 4) decontamination
costs; and (5) costs arising from implementation of agricultural countermeasures.

4.A.2.9. Risk characterisation
In general, risk results that characterise variability or aleatory uncertainty arising from inherent
randomness or stochastic processes are presented using a variety of formats and graphical displays.
Commonly used examples include: 1) point estimates – especially expected (mean) risk of selectedconsequence metrics over all-weather trials; 2) FN curves that illustrate the frequency of N or moreconsequences (F) as a function of total number of consequences (N) to express societal risk; and
3) CCDF curves. Commonly used examples of graphical displays used to illustrate uncertainties inrisk results include: 1) empirical probability distributions for selected risk metrics – typically in theform of PDF curves and/or CDF curves; 2) box plots that illustrate the locations of key summary
statistics (e.g. mean value, 50th percentile (median value), 95th percentile, and 5th percentile) for
selected risk metrics; and 3) sets of CCDF curves.

4.B.

Common technical challenges or limitations

In addition to the typical practices summarised in the previous section, common technical challenges
or limitations across several aspects of Level 3 PSA were also identified through analysis of submitted
survey responses. These common technical challenges or limitations are summarised in this section
and organised by Level 3 PSA technical element or aspect.

4.B.1.

Application of Level 3 PSA

Among countries that are not currently performing Level 3 PSAs for nuclear installations, four
potential barriers to performing Level 3 PSA were identified: 1) absence of a technical or legal
framework to perform Level 3 PSA; 2) large uncertainties in Level 3 PSA results, especially whencombined with uncertainties propagated from Level 1 and Level 2 analyses; 3) limited expected
benefit in terms of potential back fits or safety improvements; and 4) additional resources required toperform detailed off-site radiological consequence analyses.

4.B.2.

Level 3 PSA Technical Elements

This section summarises common technical challenges and limitations that were identified withrespect to various Level 3 PSA technical elements.

4.B.2.1. ATD Modelling
Most computer codes for performing probabilistic analyses of off-site radiological consequences arelimited to straight-line Gaussian plume segment or puff trajectory ATD models. However, some codes
are being enhanced for Level 3 PSA applications to include a Lagrangian particle tracking ATD
model (e.g. VALMA and MACCS). Use of weather data and sampling strategies that account forweather variability over defined time intervals is also driven primarily by code limitations andcapabilities. As previously stated in Section 4.A.2.2., most respondents indicated that they use codespecific weather binning and sampling strategies to account for the impact of temporal variability inweather conditions on off-site radiological consequences.
In addition, although it is possible (at least in principle) to account for spatial variability in the ATD
response due to variation in the topography or terrain around the modelled site, countries that perform
Level 3 PSA do not account for this effect. This is partly because available probabilistic consequence
analysis codes rely on the use of simple terrain models.

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4.B.2.2. Risk characterisation
None of the available probabilistic consequence analysis computer codes can directly combine
conditional consequence results with frequency results from the Level 2 analysis to estimate andcharacterise off-site public risks. A common solution to this limitation is to post-process the results bymathematically combining the radiological release category frequencies with the corresponding
conditional consequences using an application external to the probabilistic consequence analysis code.
In addition, while some commonly used examples of risk metrics were identified in Section 4.A.1.1.,
most respondents indicated they do not have a communication plan or guidelines for risk
communication. A notable exception is the USNRC that has developed guidance documents aiming toenhance risk communication with both internal stakeholders (including decision makers) [8] andexternal stakeholders (including the general public) [9].

4.C.

Recommendation for future activities

A key finding from analysis of submitted survey responses is that many typical practices or common
technical challenges identified as part this WGRISK activity are driven by the limitations orcapabilities of available probabilistic consequence analysis codes. Aspects of Level 3 PSA that appear
to be impacted by limitations or capabilities of the probabilistic consequence analysis code used for aparticular analysis include:• The set of radionuclides used to characterise the off-site radiological consequences
attributed to accidental releases of radiological materials.
• Treatment of concurrent radiological releases from multiple radiological sources
co-located at the same site.
• Selection of the ATD model and options for accounting for the impact of temporaland spatial variability in ATD response.
This finding that the choice of probabilistic consequence analysis code can influence typical practices
and technical challenges or limitations underscores the potential need for future research and leads tothe following recommendation stemming from this WGRISK activity.

Recommendation: Consider performing a study to benchmark available probabilistic
consequence analysis computer codes used for Level 3 PSA applications.
The NEA previously performed similar studies to benchmark probabilistic consequence analysis
codes [10,11]. Yet the most recent of these studies was completed more than 20 years ago. Since
probabilistic consequence analysis codes have evolved considerably during the past two decades, it
seems reasonable to perform a follow-on benchmarking study to compare more recent versions of thecomputer codes. However, many of these probabilistic consequence analysis codes are currently beingenhanced to address various modelling issues and technical challenges, some of which were brought
into focus by the 2011 accident at the Fukushima Daiichi nuclear power plant in Japan. Examples ofthese code enhancements include:•••

Addition of the capability to model concurrent releases from multiple co-locatedradiological sources that can have unique radionuclide inventories and potentiallydifferent accident progression timelines.
Addition of Lagrangian particle tracking ATD models.
Addition of improved economic models.

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Based on these ongoing code enhancements, a logical path forward may to be to delay any future
benchmarking study (if one is to be undertaken) until after these enhancements have been
implemented, with verification and validation testing completed, as appropriate.

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5. REFERENCES

1. Working Group on Risk Assessment, Committee on the Safety of Nuclear Installations. Useand Development of Probabilistic Safety Assessment: An Overview of the Situation at the endof 2010. NEA/CSNI/R(2012)11. Paris, France: Nuclear Energy Agency; 2012. Available at:https://www.oecd-nea.org/nsd/docs/2012/csni-r2012-11.pdf.
2. International Atomic Energy Agency. Output of the IAEA Technical Meeting on Level 3
Probabilistic Safety Assessment 2-6 July 2012. Vienna, Austria: International Atomic Energy
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3. Working Group on Risk Assessment, Committee on the Safety of Nuclear Installations.
Probabilistic Risk Criteria and Safety Goals. NEA/CSNI/R(2009)16. Paris, France: Nuclear
Energy Agency; 2009. Available at: https://www.oecd-nea.org/nsd/docs/2009/csni-r200916.pdf.
4. Niemczyk SJ, Adams KG, Murfin WB, Ritchie LT, Eppel EW, Johnson JD. The
Consequences from Liquid Pathways after a Reactor Meltdown Accident. NUREG/CR-1596.
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5. Yabusaki SB, Napier BA, Perkins WA, Richmond MC, Rakowski CL, Snyder SF et al.
Modelling of Radionuclide Transport in Freshwater Systems Associated with Nuclear Power
Plants. NUREG/CR-7231. Washington, DC: U.S. Nuclear Regulatory Commission; 2017.
Available at: https://www.nrc.gov/docs/ML1711/ML17111A578.pdf.
6. Chang R, Stutzke MA, Schaperow J, Ghosh T, Barr J, Tinkler C. State-of-the-Art Reactor
Consequence Analyses (SOARCA) Report. NUREG-1935. Washington, DC: U.S. Nuclear
Regulatory Commission; 2012. Available at: https://www.nrc.gov/reading-rm/doccollections/nuregs/staff/sr1935/.
7. Health Physics Society. Radiation Risk in Perspective: Position Statement of the Health
Physics Society. PS010-3. McLean, Virginia: Health Physics Society; May 2016. Available
at: http://hps.org/documents/risk_ps010-3.pdf.
8. Szabo A, Persensky J, Peterson L, Specht E, Goodman N, Black R. Effective Risk
Communication: The Nuclear Regulatory Commission's Guidelines for Internal Risk
Communication. NUREG/BR-0318. Washington, DC: U.S. Nuclear Regulatory Commission;
2004. Available at: http://www.nrc.gov/reading-rm/doc-collections/nuregs/brochures/br0318/.
9. Persensky J, Browde S, Szabo A, Peterson L, Specht E, Wight E. Effective Risk
Communication: The Nuclear Regulatory Commission’s Guidelines for External Risk
Communication. NUREG/BR-0308. Washington, DC: U.S. Nuclear Regulatory Commission;
2004. Available at: http://www.nrc.gov/reading-rm/doc-collections/nuregs/brochures/br0308/.
10. Nuclear Energy Agency. International Comparison Study on Reactor Accident Consequence
Modelling. Paris, France: Organisation for Economic Co-operation and Development. 1984.
11. Nuclear Energy Agency, Commission of the European Communities. Probabilistic Accident
Consequence Assessment Codes Second International Comparison: Overview Report. Paris,
France: Organisation for Economic Co-operation and Development. 1994.

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APPENDIX A: CSNI ACTIVITY PROPOSAL SHEET (CAPS)WGRISK(2015)31

1

The CAPS provided in Appendix A represents an exact duplicate of the document that CSNI approved in2015. When the CAPS was developed and approved, IAEA was performing a project to develop guidelines
for performing Level 3 PSA, which would be documented in an IAEA TECDOC report. This project has
since been cancelled and will no longer result in publication of a TECDOC.

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CSNI Activity Proposal Sheet (CAPS) for a Proposed New Activity
WGRISK(2015)3
Project/Activity Title

Status of Practice for Level 3 Probabilistic Safety Assessment
1.The mission of the Working Group on Risk Assessment (WGRISK)
is to advance the understanding and utilization of Probabilistic
Safety Assessment (PSA) in ensuring continued safety of nuclear
installations in Member countries. In pursuing this goal, the
WGRISK routinely shares information regarding PSA
methodologies used to identify nuclear power plant risk contributors
and assess their importance.

Objective

Scope

Based on feedback from recent WGRISK meetings and informationin report NEA/CSNI/R(2012)11, “Use and Development of
Probabilistic Safety Assessment An Overview of the situation at theend of 2010,” published in December 2012, an increasing numberof countries are pursuing development of PSAs that explicitly
calculate public health, safety, and economic impacts (Level 3
PSA). The objective of this task is to: (1) survey member andobserver countries to determine current methodological practices inLevel 3 PSA, (2) identify common challenges and notable practices,and (3) summarize the results of this activity in a final task report.
The purpose of this task is to identify current practices for the
conduct of Level 3 PSA (consequence analysis)1 for nuclear power
plants. This activity will be accomplished through the use of a
focused survey questionnaire to gather information on the current
state of practice in member and observer countries and to solicit
views on several challenging technical areas. The questionnaire willserve as an important input to an expanded task group2
meeting/workshop that will provide a forum for more detailed
discussions and information exchange on key topics for Level 3
PSA. The results of the survey and meeting/workshop will besummarized in a final task report. This activity will complement
ongoing work being done by IAEA on Level 3 PSA. The current
IAEA work is focused on developing a TECDOC which will
outline of state-of-the- art guidance on Level-3 PSA for NPPs3.

1

In this CAPS, “Level 3 PSA” refers to only the portion of a PSA from the release categories and source terms
to public health and economic consequences (i.e. the scope does not include initiating events to plant damage
states (Level 1 PSA) or plant damage states to release categories/source terms (Level 2 PSA).

2

The “expanded task group” includes task core group members and other WGRISK participants familiar withboth the topic and the development of the questionnaire. The expanded task group workshop will include an
expanded agenda that will include sufficient time for participants to conduct a more detailed analysis anddiscussion of the insights obtained from the survey questionnaire.

3

The IAEA TECDOC on Level 3 PSA is expected to be published in 2016. Following issuance of the
TECDOC, IAEA anticipates the publication a safety guide to provide state-of-the-art guidance for Level 3
PSA to replace the outdated guidance contained in IAEA Safety Series 50-P-12 (1996), “Procedures for
Conducting Probabilistic Safety Assessments of Nuclear Power Plants (Level 3) Off-Site Consequences andEstimation of Risks to the Public”.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1

Scope (continued)│ 43This WGRISK activity will complement the ongoing IAEA workby providing more detailed current practices information for
specific technical areas (e.g., those noted in the “Justification”
section).
Although this task is expected to identify areas requiring future
research and development, the resolution of these issues is beyond
the scope of this task (though these issues will be identified anddiscussed in the task report). Approaches and methods that areconsidered notable practices by the task group will be highlighted.
The use of Level 3 PSA (consequence analysis) appears to be
increasing in WGRISK member and observer countries. A Level 3
PSA is used to calculate public health and safety impacts (e.g., earlyfatality and latent cancer risk) as well as economic impacts. This
information supports a more complete characterization of nuclear
power plant risk and provides an important input into cost-benefit
analysis of proposed regulatory changes. Based on the experiencefrom several WGRISK members, there are a numbering ofchallenging issues that would benefit from increased informationexchange and sharing. This information would help harmonize
approaches currently being used for Level 3 PSA and assist
countries in developing new Level 3 analyses.
If approved, this task will examine a number of challenging issues
associated with Level PSA consequence analysis. Examples ofissues that may be examined include the following:

Justification• Risk metrics for effectively communicating Level 3 PSA
results, including metrics appropriate for capturing multiunit
issues.
• Modelling practices, including distance truncation, dose
response (linear no-threshold, low dose truncation, etc.),
exposure pathways (aquatic, airborne), scenario binning(identifying appropriate plant damage states and releasecategories), and radiological transport.
• Release truncation time, including consideration of severe
accident mitigation and offsite response capability.
• Determination of economic impacts due to property and healthimpacts, including valuation of life and property impacts, costs
associated with emergency preparedness and radiologicalcountermeasures (with emphasis on economic impacts that may
influence public safety).
• Consideration of site radiological sources including multiple
units, spent fuel, and other sources.
• Communication practices for the public and decision makers,
including use of expected values; frequency/consequence
curves; uncertainty quantification; and placing low likelihood
events into an understandable context.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT44 │ NEA/CSNI/R(2018)1

Users

Relation to other projects

Safety significance/ priority(see priority criteria in
Section IV)

Expected users include industry, regulators, and technical support
organizations interested in current approaches for PSA methods,
tools, and applications. This task would also be of interest to other
NEA working groups and organizations interested determination of
the health, safety, and economic impacts from nuclear power plantsevere accidents.
It is expected that other CSNI and CNRA working groups with aninterest in severe accident modelling (e.g., WGAMA) and PSA
applications related to public health will have an interest in activity.
Additionally, it is expected that this task will serve to better informongoing activities to improve the estimation of severe accident
health impacts and costs (e.g., severe accident cost study currently
being conducted by the NEA Nuclear Development Division). This
project will coordinated with the CNRA Working Group on Public
Communication (WGPC), the Committee on Radiation Protection
and Public Health (CRPPH), and the NEA Nuclear Development
Division Expert Group on Costs of Nuclear Accidents, Liability
Issues and their Impact on Electricity Costs (EG-COSTNA). Thistask will also be coordinated with the IAEA representative to
WGRISK, to ensure proper coordination with the ongoing IAEA
efforts in the Level 3 PSA area.
Regarding the priority criteria set in Section IV of the CSNI
Operating Plan:• Criterion 1: Relevance to CSNI challenges and technicalgoals – As noted in the following section on “Technical Goal(s)Covered,” this task relates to several important CSNI challenges
and technical goals.
• Criterion 2: Better accomplished by international group –The purpose of this task is to collect and summarize current
developments in Level 3 (consequence analysis) PSA
methodologies and applications from member and observercountries. This task is best accomplished by an internationalgroup such as WGRISK.
• Criterion 3: Likely to bring conclusive results in reasonable
time frame – The objective of this task is to identify the current
state of practice for Level 3 PSA in member and observercountries and identify, where possible, notable practices.
Although it is expected that the activity will identify a numberof difficult challenges for consequence analysis, it is not within
the scope of this effort to resolve these complex issues.
Therefore, it is likely that the objectives of this task can beaccomplished in a reasonable time frame.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1

Safety significance/ priority(see priority criteria in
Section IV) (continued)│ 45• Criterion 4: Maintain and preserve strategic safetycompetence – A key objective of this task is to share the stateof current practice for Level 3 PSA methods and applications
among member and observer countries. This information
sharing will facilitate knowledge management andpromulgation of current practices throughout the CSNI
membership.
The following CSNI Challenges and Technical Goals are addressedby this task:• Effectiveness and efficiency of activities related to safety:(2c) To build and harmonize international approaches to safetyissues through the use of multi-national collaborative efforts
CSNI, and cost and resource sharing

Technical Goal(s) covered• Safe operation of current nuclear installations: (3g) To
further review and assess the development of PSA methods; topromote further PSA applications in the operation of nuclear
installations and review risk-informed approaches.
• Safety in New Nuclear Installations: (4c)To review current
analytical tools as well as risk assessment approaches regarding
their applicability to safety assessments of new designs andfurther develop and validate them where needed.
• Safety in Advanced Nuclear Designs: (5c) To review theadequacy of analytical tools and risk assessment approaches forapplication to advanced designs and evaluate and validate new
analytical approaches when called for by specific features of
new designs.

Knowledge Management
and Transfer covered

A main objective of this task is to document current methods andapplications of Level 3 PSA (consequently analysis) among
member and observer countries. As such, the final task report serves
as a valuable information sharing and knowledge management
PSA document.
If approved, the following milestone schedule is proposed:

Milestones(deliverables vs. time)• September 2015: Development of draft survey instrument togather information regarding member and observer country
experience with Level 3 PSA and views on a range of technical
topics.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT46 │ NEA/CSNI/R(2018)1• December 2015: Finalization of survey questionnaire andreview by WGRISK membership and representatives fromWGAMA, the Working Group on Public Communication(CNRA/WGPC), IAEA, CRPPH and EG-COSTNA.
Distribution of survey to WGRISK members and observers. Seenote below.
Milestones(deliverables vs. time)(continued)• April 2016: Survey responses requested from member andobserver countries.
• October 2016: Develop summary of survey responses and holdexpanded task group meeting/workshop to discuss preliminary
results.
• February 2017: Development of draft task report.
• June 2017: CSNI approval of task report.
Note: CSNI has requested a briefing from WGRISK to describe thecoordination and interaction of this activity with other CSNI
groups, the CNRA WGPC, IAEA, and CRPPH.

Lead organization(s) andcoordination

The lead organization for this task is U.S. NRC. A core task grouphas been identified to coordinate this activity and develop the final
task report. The core task group includes representatives from the
NRC (US), NRG (Netherlands), CNSC (Canada), KAERI (Korea),
JAEA (Japan), and NUBIKI (Hungary).

Participants (individuals
and organizations)

All WGRISK members and observers will be asked to participate.

Resources

Leader (US NRC): 2-3 months
Each member of the core task group would require 1 man-month inorder to participate in task group meetings and summarize member
and observer inputs and develop the final task report.
Each participating organization would require approximately 0.5
man-months to respond to the survey questionnaire.

CSNI Decision

Endorsed subject to the following being recognised

CSNI Observation

It was considered to be a complex CAPs which was overambitious
in its outcomes. The topic crossed a number of working groups aswell as all three main safety standing committees CNRA, CSNI and
CRPPH. It was considered important when developing the detailedplans for delivering this CAPS there was clarity regarding
interaction with CSNI groups, the WG on Public Communications,
IAEA and CRPPH. In addition clarity regarding the differences in
the work of the IAEA and the NEA.
The title should be modified to "Status of Practice for Level 3 PSA"
as the methodologies were not seen to be mature enough to define
best practice. When available the clarity requested above should bepresented to CSNI by WGRISK.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1

APPENDIX B:

FINAL SURVEY INSTRUMENTSTATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT│ 47

48 │ NEA/CSNI/R(2018)1

PURPOSEThe purpose of this WGRISK survey is to elicit and characterize the international state-of-practicewith respect to modelling issues and other technical challenges that practitioners encounter inperforming offsite radiological consequence analyses in support of Level 3 probabilistic safetyassessment (PSA)1 applications for nuclear installations. Your response to this survey will contributeto an enhanced understanding of how practitioners across the international PSA community are
addressing specific modelling issues and other technical challenges in performing Level 3 PSA offsite
radiological consequence analyses in support of nuclear installation applications.
Core task group members2 for this activity will collect and analyse survey responses to generate
results and insights for use in a subsequent expanded task group3 workshop that will provide a forum
for more detailed discussions and information exchange on key Level 3 PSA topics.

BACKGROUND AND TASK OBJECTIVES
The mission of WGRISK is to advance the understanding and utilization of PSA in ensuring thecontinued safety of nuclear installations in member countries. In support of this mission, WGRISK
routinely collects and shares information regarding PSA methods and practices used to estimate
nuclear accident risks and to identify significant risk contributors in support of nuclear installation
applications.
An increasing number of countries are pursuing the development and application of Level 3 PSAs that
directly estimate the frequencies of offsite public health, environmental, and economic consequences
attributable to accidental releases of radiological materials from nuclear installations.4,5 The
experience of several WGRISK members indicates there are a number of challenging issues thatwould benefit from increased information exchange and sharing of methods and practices in this area.
CSNI therefore approved a new WGRISK Level 3 PSA activity in June 2015. The objectives of thisactivity are to: (1) survey member and observer countries to determine current methodological
1

2

3

4

5

A Level 3 PSA is an assessment of the offsite public risks attributable to a spectrum of possible accident scenarios
involving a nuclear installation. In the traditional PSA framework for commercial nuclear power plants, a Level 3 PSA
includes three progressive levels of analysis: (1) core damage accident or plant damage state frequency analysis; (2)accident progression, containment performance, and radiological release analysis; and (3) offsite radiologicalconsequence analysis. Although a Level 3 PSA includes all three levels of analysis, this WGRISK activity focuses only
on the offsite radiological consequence analysis and its interface with other levels of analysis.
The core task group includes representatives from the following WGRISK member countries: (1) Canada, (2) Hungary,(3) Japan, (4) Korea, (5) Netherlands, and (6) United States (lead organization).
The expanded task group includes core task group members and other interested NEA participants. In addition, in order toensure effective co-ordination of this activity, other NEA groups will be kept informed of the progress of this task,
including the (1) Working Group on Analysis and Management of Accidents (WGAMA), (2) Committee on Nuclear
Regulatory Activities (CNRA) - Working Group on Public Communication (WGPC), (3) Committee on Radiation
Protection and Public Health (CRPPH), (4) NEA Nuclear Development Division - Expert Group on Costs of Nuclear
Accidents, Liability Issues and their Impact on Electricity Costs (EG-COSTNA), and (5) International Atomic Energy
Agency (IAEA).
Working Group on Risk Assessment, Committee on the Safety of Nuclear Installations. Use and Development of
Probabilistic Safety Assessment: An Overview of the Situation at the end of 2010. NEA/CSNI/R(2012)11. Paris, France:
Nuclear Energy Agency; 2012. Available at: https://www.oecd-nea.org/nsd/docs/2012/csni-r2012-11.pdf.
International Atomic Energy Agency. Output of the IAEA Technical Meeting on Level 3 Probabilistic Safety Assessment
July 2-6 2012. Vienna, Austria: International Atomic Energy Agency; 2012.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 49

practices in Level 3 PSA; (2) identify common challenges and notable practices; and (3) summarize
the results of this activity in a final task report.1

SURVEY STRUCTURE
This survey is comprised of three parts. Part 1 is designed to obtain respondent information that willallow the group to determine whether responses differ across different countries or types oforganizations.
Part 2 is designed to obtain information about whether and how respondents or organizations use orintend to use Level 3 PSA. Since the purpose or intended use of any PSA can influence multiple
aspects of a PSA model (e.g., scope, level of detail, assumptions, or other modelling choices),
responses to Part 2 will provide important contextual information for understanding and evaluating
responses to Part 3.
Part 3 is designed to elicit information about whether and how respondents or organizations are
addressing specific modelling issues and other technical challenges that practitioners encounter inperforming offsite radiological consequence analyses in support of Level 3 PSA applications fornuclear installations. To facilitate identification and organization of this information, survey items inPart 3 are grouped by the associated major technical elements or tasks that typically comprise suchoffsite radiological consequence analyses.

PART 1: RESPONDENT INFORMATION2
Name:Country:
Organization:
Type of Organization:

� Regulatory Authority
� Utility� Vendor� Academic/Research� Other (please specify): ___________________________________

Mailing Address:
E-mail Address:
Telephone Number:

1

Working Group on Risk Assessment. CSNI Activity Proposal Sheet (CAPS) for a Proposed New Activity: Status ofPractice for Level 3 Probabilistic Safety Assessment. WGRISK (2015) 3. Paris, France: Committee on the Safety of
Nuclear Installations; 2015.
2
Many survey items use the term “you” in asking questions about if or how particular issues or challenges are being
addressed. If you are responding to this survey on behalf of your organization, it is assumed that your response will be
representative of practice within your organization. If your practice as an individual differs from that of yourorganization - and if you are providing information about your personal practice - please ensure this is clearly
documented in your response.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT50 │ NEA/CSNI/R(2018)1

PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes, forwhat types of applications1 are Level 3 PSAs required?

2.2

With regard to the use of Level 3 PSA:
2.2.1. Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
2.2.2. What calculated metrics or results from Level 3 PSAs are used in these applications andwhat (if any) requirements, goals, or criteria are used to evaluate these results? What isthe basis for these requirements, goals, or decision criteria?
2.2.3. If you are not currently performing or considering performing Level 3 PSAs, have you
previously considered performing them? If you have, what was the basis for yourdecision to not perform Level 3 PSA? Please list any barriers you perceive toperforming and applying Level 3 PSA in your country.2
2.2.4. What (if any) alternative methods do you use to estimate offsite public risks attributableto accidental releases of radiological materials from nuclear installations, and for whattypes of applications are they used?

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency planningor protective action zone. Are Level 3 PSAs used to support the siting of nuclear installations inyour country? If they are, how are they used?

2.4

With regard to nuclear power plant emergency planning zones:
2.4.1. What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
2.4.2. What stakeholder groups are involved in the process and what are their respective
responsibilities?
2.4.3. Under what conditions (if any) can emergency planning zones be reduced in size? Can aLevel 3 PSA be used to establish a probabilistic cut off for events that need to beconsidered for emergency planning purposes?

2.5

1

2

3

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria.3 Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?

In this survey, the term “application” is used to broadly represent the purpose or intended use of a Level 3 PSA. Example
Level 3 PSA applications for nuclear installations can include, but are not limited to: (1) siting of nuclear installations;(2) establishing emergency planning or preparedness requirements; (3) developing or applying safety goals or risk
acceptance criteria; (4) comparison with public risks attributed to alternative energy technologies; or (5) developing
requirements
nuclear liability insurance
and compensation schemes for potential third-party damages.
Examples of potential barriers include, but are not limited to: (1) cost or limited resource availability; (2) limited
availability of expertise, information, or technology; or (3) legal requirements.
Working Group on Risk Assessment, Committee on the Safety of Nuclear Installations. Probabilistic Risk Criteria and
Safety Goals. NEA/CSNI/R(2009)16. Paris, France: Nuclear Energy Agency; 2009. Available at: https://www.oecdnea.org/nsd/docs/2009/csni-r2009-16.pdf.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1

2.6

1

│ 51

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA.1 Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.

Examples of such applications include, but are not limited to: (1) analysis of design-basis accident (DBA) scenarios tosupport development of safety analysis reports; (2) analysis of a limited set of beyond-design-basis accident (BDBA)
scenarios to support development of severe accident management strategies; or (3) determination of compensation
awarded for third-party damages.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT52 │ NEA/CSNI/R(2018)1

PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES1
General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.
3.1.2.

3.2

With regard to the scope of Level 3 PSAs:
3.2.1.
3.2.2.

3.3

Offsite radiological consequence analyses:
Risk characterization2

What internal and external accident sequence initiating events and hazards are
considered in developing a Level 3 PSA?
Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling?3 If you do, how is thiscorrelation treated?

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.

Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.
3.4.2.

3.4.3.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses? What (if any) difficulties have you encountered in this area?
If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process?4 If you do,
how are they treated?

1

Many survey items in Part 3 aim to elicit whether and how you address specific modelling issues that practitioners havefound to be particularly difficult or challenging. In providing your response to these questions, please feel free to identify
and describe any related issues that you have found to be challenging in developing and applying offsite radiologicalconsequence analyses in support of Level 3 PSA applications for nuclear installations.
2
Risk characterization is the process of developing and evaluating risk triplets comprised of accident scenarios,
frequencies, and conditional consequences to produce: (1) qualitative insights about accident scenarios and significant
risk contributors; and (2) quantitative estimates for risk metrics of interest. These quantitative estimates are typicallycalculated by combining estimates of radiological release category frequencies from Level 2 PSA analyses withconditional consequence metric estimates from Level 3 PSA analyses for each radiological release category.
3
For example, if high winds cause an initiating event that is modelled in the Level 1 PSA, are high winds modelled in theoffsite radiological consequence analysis for the radiological release associated with the accident scenario? Anotherexample could be the correlations between time of year or season (e.g., summer vs. winter) and both initiating event
frequencies and offsite population density.
4
For example, in specifying the release duration, do you consider the time to implement mitigation measures to stabilize
the reactor core and terminate releases?

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│ 53

3.5

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated?

3.7

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?1

Meteorological Data
3.8

How do your account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?

Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?

3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?

3.12

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic parametersof the ATD vary within this resolution?

3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?

Protective Action (Countermeasure) Modelling
3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
3.14.1. Early/Emergency Phase:
3.14.2. Intermediate Phase:
3.14.3. Late/Recovery Phase:1

For this and other survey items pertaining to sensitivity or uncertainty analyses, it is recognized that performing acomplete uncertainty analysis can require extensive resources. One purpose for including these survey items is to identify
and understand how practitioners are addressing uncertainty in light of practical resource limitations.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT54 │ NEA/CSNI/R(2018)1
3.15

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?1

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.

3.17

What information sources2 do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?

Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite (e.g.,worker) populations or potentially sensitive, vulnerable, or critical groups? If you do, what isyour basis for including these groups, and how are they modelled?

3.19

What exposure pathways do you model?3 What is your basis for selecting these pathways?
What exposure duration is assumed in your models?

3.20

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?

Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.

3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?

3.23

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, andwhat is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?

3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?

1

Population variability or heterogeneity with respect to radiological exposure and dose assessment is addressed initem 3.18.
2
Examples of information sources for this and other questions include, but are not limited to: (1) national or international
standards; (2) national or international guidelines; or (3) generic or site-specific studies.
3
Examples of exposure pathways include, but are not limited to: (1) direct exposure to external radiation from the plume ofreleased radiological materials (cloud shine); (2) inhalation of radioactivity in the plume; (3) contamination of skin andclothing; (4) direct exposure to external radiation from ground contamination (ground shine); (5) inhalation ofresuspended radioactivity; and (6) ingestion of contaminated food and water.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 55

Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.

3.26

With regard to potential economic consequences considered within a Level 3 PSA:
3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.
3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.

3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?

Consequence Quantification and Reporting
3.28

What process and criteria1 do you use for selecting consequence metrics for quantification
and reporting, including spatial intervals or distances from release points?

Risk Characterization
3.29

What process and criteria1 do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?

3.30

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results?2 How do you present theresults from low-probability/high-consequence events in an understandable context for eachgroup? Who is responsible for communicating the results to each group?
(1) Decision makers:(2) General Public:

3.31

With regard to the treatment of uncertainties:
3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?

1

Processes and criteria for selecting consequence or risk metrics can come from a variety of sources including, but notlimited to: (1) legal or regulatory requirements; (2) standards or guidelines; or (3) application- or decision-specificcharacteristics (e.g., ability to compare and choose among alternatives).
2
Example methods include, but are not limited to: (1) expected values (means) or other point estimates; (2) frequencyconsequence curves; (3) complementary cumulative distribution function (CCDF) curves; or (4) a combination of these orother methods.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT56 │ NEA/CSNI/R(2018)1
3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
3.31.3. To what extent do you consider the effect of correlation on parameters?
Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 57

APPENDIX C: SUBMITTED SURVEY RESPONSES BY COUNTRYSTATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT58 │ NEA/CSNI/R(2018)1

Belgium: Bel V and Tractebel

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 59

PART 1: RESPONDENT INFORMATIONName:Country:
Organization:
Type of Organization:

D. Gryffroy*
Belgium
Bel V
� Regulatory Authority
� Utility� Vendor� Academic/Researchx Other (please specify): Technical Support Organisation (TSO) of
the Regulatory Authority

Mailing Address:

Bel V
Rue Walcourt 148
B-1070 Brussels(Belgium)dries.gryffroy@belv.be

E-mail Address:
Telephone Number:* The following persons reviewed and supplemented this survey response:•

members of the Bel V technical staff:
Pieter De Gelder
Didier Degueldre•

for Tractebel Engineering S.A.:
Stanislas Mitaillé

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT60 │ NEA/CSNI/R(2018)1

PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes,for what types of applications are Level 3 PSAs required?
Such requirement does currently not exist for nuclear installations in Belgium (see Royal Decreeof 2011-11-301, which covers the transposition of WENRA Safety Reference Levels (2008)2 forexisting NPPs into Belgian regulations and also covers non-NPPs).
Level 3 PSA is not required either in the updated WENRA Reference Levels (September 2014)3.
These updated WENRA Reference Levels will also be transposed into Belgian regulations (bymeans of a new Royal Decree).

2.2

With regard to the use of Level 3 PSA:
2.2.1.

Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
The performance of Level 3 PSA for existing reactors is currently not considered in
Belgium. The focus is currently on the extension of PSA Level 1 and Level 2 for internaland external hazards, which is requiring important resources.

2.2.2.

What calculated metrics or results from Level 3 PSAs are used in these applications
and what (if any) requirements, goals, or criteria are used to evaluate these results?
What is the basis for these requirements, goals, or decision criteria?
Not applicable (see 2.1 and 2.2.1).

2.2.3.

If you are not currently performing or considering performing Level 3 PSAs, haveyou previously considered performing them? If you have, what was the basis for yourdecision to not perform Level 3 PSA? Please list any barriers you perceive toperforming and applying Level 3 PSA in your country.
The performance of Level 3 PSA has not been considered previously.
Currently, a major barrier to performing and applying Level 3 PSA is that all resources
are currently used for PSA Level 1 and Level 2 developments with higher priority (see2.2.1), linked to applicable regulatory requirements (Royal Decree of 2011-11-301).
Potential barriers to performing and applying Level 3 PSA could also be the highuncertainties of the results (considering the propagation of the Level 1 and Level 2
uncertainties as well) and the limited expected benefit in terms of potential backfits orsafety improvements.

1

Royal Decree of November 30, 2011, Safety requirements for nuclear installations. Available at:
http://www.fanc.fgov.be/GED/00000000/3400/3493.pdf
2Western European Nuclear Regulators Association (WENRA). WENRA Reactor Safety Reference Levels (January 2008).
Available at: http://www.wenra.org/media/filer_public/2012/11/05/list_of_reference_levels_january_2008.pdf
3Western European Nuclear Regulators Association (WENRA). WENRA Reactor Safety Reference Levels (September2014). Available at:
http://www.wenra.org/media/filer_public/2016/07/19/wenra_safety_reference_level_for_existing_reactors_september_20
14.pdfSTATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1

2.2.4.

│ 61

What (if any) alternative methods do you use to estimate offsite public risks
attributable to accidental releases of radiological materials from nuclear installations,and for what types of applications are they used?
Not applicable (see 2.1 and 2.2.1).

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency
planning or protective action zone. Are Level 3 PSAs used to support the siting of nuclear
installations in your country? If they are, how are they used?
The latest siting of NPPs in Belgium dates from the early eighties and at that time norequirement existed for performing and applying Level 3 PSA for siting. Since then, no
requirements have been published for Level 3 PSA to support siting of nuclear power plants orother nuclear installations in Belgium. In this context, it should also be noted that the nuclear
phase-out law of 2003-01-31 excludes building new NPPs in Belgium.
However, for new “class I” nuclear installations (including both NPPs and non-NPPs and definedin the Royal Decree of 2001-07-2011), a Guidance on Safety Demonstration2 has been developedby the Belgian regulatory authority (FANC), requiring dose assessments for different Design
Basis Categories (DBC) to demonstrate that doses to individuals located at any point along theboundaries of specified areas around a nuclear installation for specified periods of time following
postulated fission product releases are within prescribed “quantitative radiological safety
objectives”.
Because of the nuclear phase-out law of 2003-01-31, this Guidance is only applied to new nuclear
installations which are not NPPs.

2.4

With regard to nuclear power plant emergency planning zones:
2.4.1.

What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
There is no formalized process implemented to define the size of the EPZ used aroundthe nuclear installations (NPP and other facilities) in the context of the emergency
preparedness & response arrangements. These EPZs are designed as circles ofappropriate radius around nuclear installations and their dimension depends, among
others, on the potential source term specific for each facility/installation and the genericintervention criteria adopted for a specific protective action (evacuation, sheltering,iodine thyroid blocking, …).
The present EPZs (established in 2003) are currently reassessed in the framework of therevision of the Belgian national EP&R. The results and principles defined in the“HERCA-WENRA approach for a better cross-border coordination of protective actions
during the early phase of a nuclear accident” (document issued in Oct-2014) are ofcourse used in that context.

1

Royal Decree of July 20, 2001, General Regulation on the Protection of the Population, the Workers and the Environment
Against the Danger of Ionizing Radiation. Available at: http://www.fanc.fgov.be/GED/00000000/3400/3492.pdf

2

FANC - Class I Guidances - Safety demonstration of new class I nuclear installations: approach to Defence-in-Depth,
radiological safety objectives and the application of a graded approach to external hazards (February 2015). Available at:
http://www.fanc.fgov.be/GED/00000000/3800/3883.pdf.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT62 │ NEA/CSNI/R(2018)1
The feedback from nuclear emergency exercises clearly demonstrates that thetransposition of the actual emergency intervention zone/sector into an operational
response area is not straightforward and leads to lengthy discussions between theexperts in charge of the assessment of the emergency situation and the decision makers
and between the decision makers and the local authorities. In order to circumvent thisdifficulty and avoid unnecessary delays, Belgium recently developed a practical concept
consisting in dividing the EPZ into blocks of reasonable dimensions, not too large but not
too small either, in order to allow enough flexibility when determining the area(s) where
protective actions are needed. These arrangements taken at the preparedness stage are
expected to improve the efficiency of the response actions.

2.4.2.

What stakeholder groups are involved in the process and what are their respective
responsibilities?
The federal and local authorities are involved together with the nuclear safety authority(FANC & Bel V, as supporting TSO) in the definition of the EPZs. In addition, thefederal and local police services have been deeply involved (through an iterative process)in the implementation of the concept of dividing the EPZ into operational blocks (see2.4.1 above).

2.4.3.

Under what conditions (if any) can emergency planning zones be reduced insize? Can a Level 3 PSA be used to establish a probabilistic cut off for events thatneed to be considered for emergency planning purposes?
Not applicable.

2.5

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria. Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?
Not applicable.

2.6

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA. Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.
Offsite radiological consequence analyses are also performed to support incident response
applications and to support emergency planning exercises. In these applications, dose
assessments are performed in real-time to inform or evaluate offsite protective action
recommendations.

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PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES
Since performance of Level 3 PSA for existing reactors is currently not considered in Belgium, questions
3.1 to 3.31 are considered not applicable.

General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.
3.1.2.

3.2

With regard to the scope of Level 3 PSAs:
3.2.1.
3.2.2.

3.3

Offsite radiological consequence analyses:
Risk characterization:

What internal and external accident sequence initiating events and hazards are
considered in developing a Level 3 PSA?
Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling? If you do, how is thiscorrelation treated?

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.

Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.
3.4.2.

3.4.3.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses? What (if any) difficulties have you encountered in this area?
If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process? If you do,
how are they treated?

3.5

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated?

3.7

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?

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Meteorological Data
3.8

How do you account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?

Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?

3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?

3.12

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic parametersof the ATD vary within this resolution?

3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?

Protective Action (Countermeasure) Modelling
3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
3.14.1. Early/Emergency Phase:
3.14.2. Intermediate Phase:
3.14.3. Late/Recovery Phase:

3.15

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.

3.17

What information sources do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?

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Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite(e.g., worker) populations or potentially sensitive, vulnerable, or critical groups? If you do,what is your basis for including these groups, and how are they modelled?

3.19

What exposure pathways do you model? What is your basis for selecting these pathways?
What exposure duration is assumed in your models?

3.20

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?

Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.

3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?

3.23

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, andwhat is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?

3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?

Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.

3.26

With regard to potential economic consequences considered within a Level 3 PSA:
3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.
3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.

3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?

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Consequence Quantification and Reporting
3.28

What process and criteria do you use for selecting consequence metrics for quantification andreporting, including spatial intervals or distances from release points?

Risk Characterization
3.29

What process and criteria do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?

3.30

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results? How do you present the resultsfrom low-probability/high-consequence events in an understandable context for each group?
Who is responsible for communicating the results to each group?
(1) Decision makers:(2) General Public:

3.31

With regard to the treatment of uncertainties:
3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?
3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
3.31.3. To what extent do you consider the effect of correlation on parameters?

Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?

3.33

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Canada: CNSCSTATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT│ 67

68 │ NEA/CSNI/R(2018)1

PART 1: RESPONDENT INFORMATION*Name:Country:
Organization:
Type of Organization:

Mailing Address:
E-mail Address:
Telephone Number:

Chantal Morin/Usha Menon*
CanadaCanadian Nuclear Safety Commission (CNSC)
X Regulatory Authority
� Utility� Vendor� Academic/Research� Other (please specify): ___________________________________
280 Slater St. Ottawa, Ontario, Canada, K1P 5S9chantal.morin@canada.ca and usha.menon@canada.ca
1-613-947-8222

The following members of the CNSC technical staff reviewed and supplemented this example survey
response:
Y. Akl, Probabilistic Safety Assessment and Reliability Division
Melanie Rickard, Directorate of Environment and Radiation Protection Thuy N’Guyen, Reactor
Behaviour Division
Doug Miller, Directorate of Regulatory Improvements and Major Projects

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 69

PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes,for what types of applications are Level 3 PSAs required?
The Canadian Nuclear Safety Commission does not require that a Level 3 PSA be performed.
However, dose consequence assessments are performed as part of environmental assessmentsthat are required, for example, for major projects such as new build licensing or major
refurbishment of existing reactors. The dose consequence assessments examine bounding
accidents and malfunctions to evaluate doses to ensure that they are below the prescribed dose
limits in the Radiation Protection Regulations. Further, dose consequence assessments are also
performed as part of requirements for deterministic safety assessment as described in CNSCREGDOC-2.4.1 (entitled: Safety Analysis: Deterministic Safety Analysis). This document sets out
requirements of the Canadian Nuclear Safety Commission for deterministic safety analysis fornuclear power plants and small reactor facilities.

2.2

With regard to the use of Level 3 PSA:
2.2.1.

Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
CNSC does not currently perform neither requires a Level 3 PSA. Historically the
CNSC (at the time Atomic Energy Control Board) licensing requirements led somelicensees to do some level 3 PSA. Current requirements only necessitate level 2 PSA.

2.2.2.

What calculated metrics or results from Level 3 PSAs are used in these applications
and what (if any) requirements, goals, or criteria are used to evaluate these results?
What is the basis for these requirements, goals, or decision criteria?
Only dose consequence assessments are performed (see response to 2.1.) No level 3 PSA
are currently performed.

2.2.3.

If you are not currently performing or considering performing Level 3 PSAs, haveyou previously considered performing them? If you have, what was the basis for yourdecision to not perform Level 3 PSA? Please list any barriers you perceive toperforming and applying Level 3 PSA in your country.
Yes, the possibility of level 3 PSA was raised by many public interveners following
Fukushima. So far the CNSC position is that safety goals based on Level 2 PSA results aswell as using a deterministic safety assessments are sufficient to ensure safety. Level 3
PSA results are viewed as containing too much uncertainty

2.2.4.

What (if any) alternative methods do you use to estimate offsite public risks
attributable to accidental releases of radiological materials from nuclear installations,and for what types of applications are they used? [DERPA to review]
The Canadian Nuclear Safety Commission does not require that a Level 3 PSA be
performed. However, dose consequence assessments are performed as part ofenvironmental assessments that are required, for example, for major projects such as
new build licensing or major refurbishment of existing reactors. The dose consequenceassessments examine bounding accidents and malfunctions to evaluate doses to ensurethat they are below the prescribed dose limits in the Radiation Protection Regulations.
Further, dose consequence assessments are also performed as part of requirements for

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT70 │ NEA/CSNI/R(2018)1
deterministic safety assessment as described in CNSC REGDOC-2.4.1 (entitled: Safety
Analysis: Deterministic Safety Analysis). This document sets out requirements of theCanadian Nuclear Safety Commission for deterministic safety analysis for nuclear
power plants and small modular reactor facilities.

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency
planning or protective action zone. Are Level 3 PSAs used to support the siting of nuclear
installations in your country? If they are, how are they used?
Currently full Level 3 PSA are not used to support the siting of nuclear installations. See Answer
to question 2.1

2.4

With regard to nuclear power plant emergency planning zones:
2.4.1.

2.4.2.

What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
What stakeholder groups are involved in the process and what are their respective
responsibilities?
Provincial Authorities

2.4.3.
2.4.4.

Under what conditions (if any) can emergency planning zones be reduced insize? Case-by-case assessment
Can a Level 3 PSA be used to establish a probabilistic cut off for events that need tobe considered for emergency planning purposes?
At this time, Level 3 PSAs are not used for this application.

2.5

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria. Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?
Not at this time

2.6

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA. Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.
Dose consequence assessments are performed as part of environmental assessments that are
required, for example, for major projects such as new build licensing or major refurbishment ofexisting reactors. The dose consequence assessments examine bounding accidents andmalfunctions to evaluate doses to ensure that they are below the prescribed dose limits in the
Radiation Protection Regulations. Further, Dose consequence analysis in the context of
accidents/emergencies is performed to confirm deterministic safety analysis acceptance criteria(e.g., Design Basis Accident consequences should be less than 20mSv).

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PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES
General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.

Offsite radiological consequence analyses:
Depending on the application (emergency response and related exercises or safetyassessment) different codes are used. MAACS 2 is generally used by licensees for safetyassessment and RASCAL and ERP (Emergency Response Projection, a code used by theProvince of Ontario) are used in emergency response or related exercises. Note theseapplications are not within the context of PSA III.

3.1.2.

Risk characterization:
Level 1 and 2 PSA use CAFTA

3.2

With regard to the scope of Level 3 PSAs: N/A
3.2.1.

What internal and external accident sequence initiating events and hazards are
considered in developing a Level 3 PSA?
N/A

3.2.2.

Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling? If you do, how is thiscorrelation treated?
N/A

3.3

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.
The CNSC is involved in a number of post-accident/recovery phase initiatives, including
participation in the IAEA’s Modelling and Data for Radiological Impact Assessments
Programme. Working groups within this initiative are studying a variety of topics, including
model testing and comparison for accidental tritium releases and the use of decision-making tools
in the post-accident recovery phase.

Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses? What (if any) difficulties have you encountered in this area?

3.4.2.

If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
Level 3 PSAs are not performed; however, the Level 2 PSA release categories are
generally defined as in the table below, by the licensees.

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72 │ NEA/CSNI/R(2018)1

3.4.3.

What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process? If you do,
how are they treated?
N/A

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3.5

│ 73

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?
Expectations for dose assessment are best described in a Canadian Standards Association
Standard entitled: N288.2 Guidelines for Calculating the Radiological Consequences to the Publicof a Release of Airborne Radioactive Material for Nuclear Reactor Accidents. This standard isintended to be used by Nuclear Power Facilities for safety assessment as well as real time
emergency response. The requirements include that the radionuclides considered in theassessment shall contribute more than 95% to the total dose. A justification shall be provided forthe specific radionuclides selected for inclusion in a given accident

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated?
Multiple units or radiological sources co-located at the same site are not yet considered in level 2
PSA. However multi-unit impacts, such as some cross link on support systems are modelled in
the level 1 PSA.

3.7

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?
Level 2 PSA results contain sensitivity analysis on release categories ***“The objective of the uncertainty analysis is to determine whether the uncertainty associatedwith consequence assessment could potentially result in a change in results that could impact theconclusions of the Level 2 PRA. The uncertainty analysis approach provides a reasonable tool togauge the robustness of the binning of various PDS sequences into the respective releasecategories. For each representative sequence in RC2, RC4, RC6 and RC7 the following
distributions are reported:- Cs137 and I131 releases,- Corium relocation time into the FMD, and- The total mass of hydrogen produced.
The uncertainty analysis study can only provide reasonable confidence to the upper limit of the
distributions. A probability density function to correlate specific output trends with input
distributions is not possible, as the sample size is too small. However, in some cases generaltrends can be formulated on the overall uncertainty effects of the input parameters on the output
distributions.”1

Meteorological Data
All of this (3.8 to 3.12) detail is covered in extensive detail in CSA N228.2 for emergency
response and accident assessment purposes.
3.8

How do you account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?
N/A

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?
N/A

1

From Bruce B level 2 PSA report

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Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?

3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?

3.12

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic
parameters of the ATD vary within this resolution?

3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?
No, we don’t consider it for now.

Protective Action (Countermeasure) Modelling
N/A, since no Level 3 PSAs are performed

3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
The answer here is provided with respect to dose consequence assessment that is not related to
Level III PSA (as it is not required-discussed above), but generally, with regards to assessmentsthat support other initiatives or processes, such as those practiced in emergency response
exercises. Health Canada and the applicable provinces as well as some Nuclear Power Plants,
have models to simulate radiation dispersion, transport and dose consequences and typicallythose assessments focus on the early phase; mainly the first seven days. Broadly speaking,Canada follows guidance in IAEA GSR Part 7 for the emergency phase. The actions to be takenin the transition from the emergency phase to recovery and for recovery are currently being
developed in Canada. To the extent possible, guidance from the ICRP, the IAEA and countries
with well-developed frameworks on recovery are being relied upon.

3.14.1. Early/Emergency Phase:
3.14.2. Intermediate Phase:
3.14.3. Late/Recovery Phase:
3.15

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?
Some models do account for different age classes, typically those defined by the ICRP. The mostconservative results (resulting from assessment of multiple age groups) are typically applied tothe entire population for urgent and early protective actions. While not yet formalized, theexpectations for the recovery phase would follow those of the ICRP on such dose assessment. As

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such, the dose assessment for those individuals or groups examined would depend on different
habits and behaviours and thus heterogeneity would be assumed. That is, to the extent practical,
different ages, sex and lifestyle habits would be considered.

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.
These are not considered at this time. Since this is an area of current development, it is unknown
if this approach would be used or advisable. At this time, dose consequence assessment forurgent/early countermeasures is not based on a probabilistic modelling of protective actions andassociated predictions of success or failure. Notwithstanding, there is an expectation that decision
makers consider risk benefit in taking countermeasures.

3.17

What information sources do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?
The modelling used will depend on the off-site authority involved as each province as well as thefederal authority (Health Canada) have different modelling capabilities. In terms of theprotective measures taken and their basis, this will also vary somewhat depending on the
responsible authority. However, in all cases, the countermeasures and their basis are consistent
with IAEA GSR Part 7. Each province (with a Nuclear Power Plant) as well as Health Canada
has recommendations and plans with regards to off-site countermeasures.

Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite(e.g., worker) populations or potentially sensitive, vulnerable, or critical groups? If you do,what is your basis for including these groups, and how are they modelled?
The concept of receptors and associated requirements are defined in CSA 288.2. The receptor
should be selected based on the safety objective of the calculation. For safety assessments, a
representative person shall be defined and shall include at a minimum, a hypothetical adultlocated at the site boundary in the downwind direction at the time of the event. For safetyassessments involving collective doses and for probabilistic safety assessments, the population
distribution around the reactor shall be used. In terms of protective measures being consideredat the preparedness and response stages, these should be evaluated on the basis of doses
calculated for the vulnerable populations. Vulnerable populations can be defined as members of
the population who have higher radiation sensitivity or additional needs before, during or after arelease. In practice, vulnerable populations could be viewed as expecting, pregnant or nursing
mothers or vulnerable based on age. For example, models run doses based on different age
classes and the protective measure would be based on the highest dose from one age group.

3.19

What exposure pathways do you model? What is your basis for selecting these pathways?
What exposure duration is assumed in your models?
CSA N288.2 requires that cloud shine, ground shine and inhalation pathways be modelled,
including re-suspension, where applicable. The basis for these pathways is, in part, in theassumption that the release is airborne and is less than 30 days in duration. There is also as
assumption that effective food controls are in place and that dose from ingestion of contaminatedfood and water are minimized. The pathways considered represent standard practice.

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3.20

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?
For those most part, DCFs are based on those available from the ICRP and the U.S EPA. Some
codes use DCFs that are not based on the most recent DCFs available but are usually considered
acceptable nonetheless, based on the intended use and the small amount of “error” introduced inusing these DCFs. Where applicable, age specific DCFs are available within some codes(e.g., ERP).

Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.
Standard N288.2 recommends the use of BIER VII risk coefficients that are based on the risk ofcancer mortality. Further, CSA N228.2 recommends that for severe accident assessments,
deterministic effects should be calculated and describes how this is done.
Further, after the accident at Fukushima, the CNSC’s Commission requested staff to undertake
an assessment of health and environmental consequences of accident scenarios that are more
severe than typically considered in environmental assessment screening reports. The resulting
study is titled Study of Consequences of a Hypothetical Severe Nuclear Accident and Effectivenessof Mitigation Measures (referred to in section 3.22 through 3.24 as “the study”.

3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?
In terms of human health, the focus of the study was to examine the possible impact on cancer
incidence. Cancer is described as a stochastic or latent effect where the probability of occurrence
is proportional to exposure or dose. Deterministic effects, such as acute radiation sickness were
not examined as the estimated doses in this study were below the thresholds for these types ofeffects. The target organs considered in this study were the colon, the bone marrow and thethyroid. The colon was the organ used to evaluate the risk of developing a cancer in the categoryof “all cancers combined”. The colon was chosen to represent the absorbed dose to all organs
because it can illustrate changes in dose to deeper tissues that experience shielding from the more
superficial tissues of the body, it is relatively centred in the body from a physiologicalperspective, it is a highly radiosensitive organ, and it is not sex-specific. The category of allcancers combined includes many cancers, some of which are more radiosensitive than others.
This category provides a general indication of whether there is an increased risk overall. Bonemarrow was used to evaluate the risk of developing leukaemia. Leukaemia is a very
radiosensitive type of cancer and radiation-induced cases of leukaemia could appear as soon as
two years after exposure providing a very strong indication of the occurrence of an effect actuallyoccurring. The thyroid was used to evaluate the risk of developing thyroid cancer. Thyroid
cancer was chosen because it was one of the main health effects that resulted from Chernobyl.
Thyroid cancer is also very radiosensitive and radiation-induced cases of thyroid cancer could
appear as soon as five years after exposure.

3.23

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, and

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what is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?
For this study, risk calculations were performed using the U.S National Cancer Institute’s
radiation risk assessment tool known as “RadRAT”. The risk models used by the risk calculator
are broadly based on those developed by the Biological Effects of Ionizing Radiation (BEIR) VII
committee for estimating lifetime risk for radiation-related cancer. A linear dose-response model
was chosen for all cancers combined and thyroid cancer, and a linear-quadratic risk model waschosen for leukaemia. RadRAT is based on a US population. Canada was deemed to have similar
enough life expectancy and cancer rates to the US population. The RadRAT program reports themean risk and 90 per cent uncertainty interval from the resulting distribution. Furthermore, forthe solid cancer risk models, the assumption in RadRAT follows that of BEIR VII by applying an
uncertain Dose and Dose Rate Effectiveness Factor (DDREF) for all chronic exposures and for
acute exposures below 100 mGy described by a lognormal with a geometric mean (GM) of 1.5
and geometric standard deviation (GSD) of 1.35.

3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?
It is important to note that the dose assessment and human health risk assessment were
performed by different organizations. With regards to the human health risk assessment theparameters are fixed by the program. RadRAT is publicly available and designed to be very user
friendly. The input parameters include demographic information (gender and birth year) and
exposure information (number of exposure events, exposure year, organ exposed, exposure rate
(acute or chronic), and the distribution type (lognormal, normal, triangular or log triangular,
uniform or log uniform, or fixed value).

Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.
There is no Level 3 PSA performed currently and no modelling of economic consequences.
However, cost-benefit analyses were performed on Level 2 PSA results by estimating a cost for aLarge Release event and comparing the delta LRF with cost of modifications.

3.26

With regard to potential economic consequences considered within a Level 3 PSA:
N/A since no level 3 PSA performed

3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.
3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.
3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?
N/A

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Consequence Quantification and Reporting
3.28

What process and criteria do you use for selecting consequence metrics for quantification andreporting, including spatial intervals or distances from release points?

Risk Characterization
3.29

What process and criteria do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?
N/A since no level 3 PSA are performed

3.30

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results? How do you present the resultsfrom low-probability/high-consequence events in an understandable context for each group?
Who is responsible for communicating the results to each group?
N/A(1) Decision makers:(2) General Public:
3.31

With regard to the treatment of uncertainties:
See response to question 3.7

3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?
3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
3.31.3. To what extent do you consider the effect of correlation on parameters?
Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?
Not at this time

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80 │ NEA/CSNI/R(2018)1

PART 1: RESPONDENT INFORMATIONName:Country:
Organization:

Ilkka Karanta (VTT)*
Finland
VTT Technical Research Center of Finland Ltd (in cooperation with STUK,
the Radiation and Nuclear Safety Authority in Finland)

Type of Organization:

Regulatory Authority
Utility
Vendorx Academic/Research
Other (please specify): ___________________________________

Mailing Address:
E-mail Address:
Telephone Number:

VTT, Box 1000, FI-02044 VTT, Finland
Ilkka.karanta@vtt.fi+358 20 722 4509

* The following experts at VTT and the Finnish Radiation and Nuclear Safety Authority STUK reviewed
and supplemented this survey response:
Mikko Ilvonen (VTT)
Marko Marjamäki (STUK)
Jukka Rossi (VTT)
Jorma Sandberg (STUK)STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 81

PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes,for what types of applications are Level 3 PSAs required?
Finland does not require the performance of Level 3 PSAs. Only level 1 and 2 PSAs are required.

2.2

With regard to the use of Level 3 PSA:
2.2.1.

Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
VTT does not currently perform full-scope level 3 PSAs. VTT has performed offsite
radiological consequence analyses, methodologically substantially similar to those
performed in level 3 PSA. As a research institute, many of those analyses have beenconnected with applied research. As an example, the applicability of the integrated
deterministic and probabilistic safety assessment (IDPSA) methodology to level 3 studies
has been investigated.

2.2.2.

What calculated metrics or results from Level 3 PSAs are used in these applications
and what (if any) requirements, goals, or criteria are used to evaluate these results?
What is the basis for these requirements, goals, or decision criteria?
Metrics used relate mainly to health effects from ionizing radiation, such as total
population dose, dose to the most exposed individual, number of cancer deaths.

2.2.3.

If you are not currently performing or considering performing Level 3 PSAs, haveyou previously considered performing them? If you have, what was the basis for yourdecision to not perform Level 3 PSA? Please list any barriers you perceive toperforming and applying Level 3 PSA in your country.
As a research institute, VTT performs analyses primarily when commissioned to do so.
Thus, the reason for not performing level 3 PSA analyses is one to which VTT’scustomers are in a better position to reply.
A reason for the fact that Level 3 PSA is not performed and applied more in Finland isthat it is not required by law or the regulator. Nevertheless, both the regulator Radiation
and Nuclear Safety Authority (STUK) and nuclear power companies (Fortum,
Teollisuuden Voima) have performed level 3 PSA analyses.
A major barrier to performing level 3 PSA analyses is the computational burden ofextensive analyses. Also, the tediousness of detailed analyses is a barrier.

2.2.4.

What (if any) alternative methods do you use to estimate offsite public risks
attributable to accidental releases of radiological materials from nuclear installations,and for what types of applications are they used?
Typically, it is demonstrated that even with a large release, some threshold dose is notexceeded in almost any weather condition.

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency

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planning or protective action zone. Are Level 3 PSAs used to support the siting of nuclear
installations in your country? If they are, how are they used?
There is one older siting study conducted in the 1970s. In this study site alternatives for districtheating power plant were compared.

2.4

With regard to nuclear power plant emergency planning zones:
2.4.1.

What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
The zones around the Finnish nuclear facilities are defined in STUK’s Regulation andRegulatory Guide YVL A.2s.
--

the site area comprises the area surrounding the plant, with only activities
related to the plant, it extends to about 0.5 – 1 km from the plant
precautionary action zone surrounds the site area and extends to about 5 km
from the plant, land use restrictions are in force in this area (e.g. no dense
settlement, no hospitals or such, no significant production that could be
impacted by a severe nuclear accident or endanger the plant)
emergency planning zone extending to about 20 kilometres from the plant; thezone is covered by a detailed external rescue plan for the protection of the publicdrawn up by authorities.

The zones are established according to the above deterministic principles.
Decisions on the exact boundaries of the zones are made in connection with the licensing
process and the general land use planning process. In addition, the environmental
impact assessment supports the licensing process.

2.4.2.

What stakeholder groups are involved in the process and what are their respective
responsibilities?

The licensing process is coordinated by the Ministry of Employment and the Economy. The firstlicensing step is the government decision in principle for which a supporting statement from the local
municipal council is an absolute precondition. The decision in principle on a nuclear project requires
ratification by the Parliament. The construction and operating licenses are granted by the government.
STUK’s positive statement is a precondition at all licensing stages. All the processes include public
hearings where statements and opinions are asked from ministries, the Radiation and Nuclear Safety
Authority (STUK), municipalities, local authorities, rescue services, non-governmental organizations
and from the general public. More information on the Finnish licensing process can be found at the
www address http://tem.fi/en/environmental-impact-assessments.

2.4.3.

Under what conditions (if any) can emergency planning zones be reduced insize? Can a Level 3 PSA be used to establish a probabilistic cut off for events thatneed to be considered for emergency planning purposes?
Conditions for reducing the size of the zones have not been set in the regulations. The
dimensions described above are approximate, but in practice, the reduction of the sizes
has not been under discussion. If the boundary of the precautionary action zone goes

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through a relatively densely populated village, in recent cases the zone has rather been
expanded to include the whole village.

2.5

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria. Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?
We are not aware the level 3 PSAs would have been used for such a purpose in Finland.
Probabilistic offsite dose assessments have been carried out with the source term of 100 TBq of
Cs-137. Long term weather data statistics from the power plant sites were used. Results indicatedthat acute health effects are not expected and long-term restrictions on land use in large areas
are avoided, which was the goal of the 100 TBq of Cs-137 limit set in 1991 for severe accidents.

2.6

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA. Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.
Offsite radiological consequence analyses are used to support incident response planning andreadiness at the regulatory agency STUK. Doses caused by the radioactive emissions of normal
operation are calculated. Emergency preparedness cases are based usually on quite short time
period and then weather is better known, easier to estimate and less probabilistic. Offsite
radiological consequence analyses have also been used for research purposes. Also, the topics
listed in footnote 11 are relevant to VTT; in such cases, a consequence analysis is called for.

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PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES
General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.

Offsite radiological consequence analyses:
VTT uses the following codes for offsite radiological consequence analyses. In-housecodes: ARANO (straight-line atmospheric dispersion, dose assessment, healthconsequences, economic consequences), VALMA (atmospheric dispersion in dynamic
weather conditions, dose assessment), DETRA (biospheric transportation). ARANO is
the main tool for Level 3 PSA. Available also MACCS, RADTRAD and RASCAL

3.1.2.

Risk characterization:
At VTT, tools for this are not available in the in-house codes.

3.2

With regard to the scope of Level 3 PSAs:
3.2.1.

What internal and external accident sequence initiating events and hazards are
considered in developing a Level 3 PSA?
Any that may result in large release with significant probability. Initiating events andhazards are specified in PSA level 1.

3.2.2.

Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling? If you do, how is thiscorrelation treated?
Such correlations are usually a part of the model. For example, if an external event isthat ice has blocked a water intake, then only weather data for winter will be used inthat scenario.

3.3

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.
VTT is participating in a Nordic research project on Level 3 PSA called L3PSA. Other partners
in the project are Swedish nuclear consultancies. The objectives of the project are to survey theuse and utility of level 3 analyses, follow international standardization work in the field, conduct
pilot studies to gain experience in level 3 PSA, and develop guidance for performing level 3
analyses. An industry survey has been conducted, two pilots have been constructed (one Swedish,one Finnish), and a guidance document has been written. The project will end this year.
Research on offsite radiological consequences is also carried out in CASA project by VTT.

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Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses? What (if any) difficulties have you encountered in this area?
If the release fraction from Level 2 PSA is available, there is no problem. But MELCOR
primarily outputs the released masses and if MELMACCS is not available, we use the
ORIGEN code to convert masses to activities.

3.4.2.

If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
Usually in Level 2 PSA inventory is specified for nuclide groups (e.g. noble gases),
categorisation of nuclides into release categories is based on chemical properties of
nuclides.

3.4.3.

What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process? If you do,
how are they treated?
Depends on the code. In ARANO release duration is reduced to few hours, because
weather is remaining the same during dispersion. Also, offsite emergency response
actions cannot be changed during dispersion.

3.5

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?
There are 84 nuclides available in ARANO. In the case of LWR release not all are significant.
Cs-137, I-131, other nuclides depending on the analysis. In VALMA, 496 radionuclides can betaken into account. Selection is based on significance from the dose effects point of view.

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated? Adjacent releases are co-located at the same site.
VTT has analysed the Fukushima accident, but beyond that, releases from multiple units havenot been considered.

3.7

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?
Yes, VTT has conducted uncertainty analyses on release fractions. The method used wasdeveloped by VTT and essentially concerns conducting computational experiments with different
fractions, tabulating the results, constructing statistical models based on them, and using thesemodels in simulations which are then analysed. The VALMA code has uncertainty analysis
capabilities, but computation burden is high.

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Meteorological Data
3.8

How do you account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?
At VTT, this is handled differently in different in-house codes. VALMA can handle both
temporal and spatial variability in meteorological conditions. The weather input is given to
VALMA as series of weather observations at different weather masts (i.e. different locations). In
ARANO, annual weather data from one meteorological mast is converted into joint frequency
matrix of annual weather statistics. Doses are calculated in all specified weather situations andthe result is weighted with the probability of the weather situation. Sampling is not needed.

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?
VTT has performed uncertainty analysis on wind speed, wind direction and precipitation.
Statistical models for these have been constructed and used in simulations.

Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?
The basic model ARANO for Level 3 PSA is a Gaussian straight-line dispersion model in which
vertical dispersion is replaced with the Kz model. ARANO is fast and versatile providing e.g.
countermeasures and calculation of health effects. The Lagrangian trajectory model VALMA
was initially prepared for an emergency preparedness tool, but currently it is under development
to enable also risk calculations. VALMA can work with many kinds of weather data, startingfrom single-point measurements at the weather mast of an NPP (or several masts) and ending
with Monte Carlo particles (even a limited number) that can be calculated, based on NWP(numerical weather prediction) models.

3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?
In ARANO, environmental data is given in polar coordinates (segmented by radial lines - r,Ө). In
this annular grid the angle size is 30 degrees (fixed) which means that data is given in 12 sectors
for distance intervals defined by the user. Spatial intervals can be obtained case by case.
Typically, spatial grid is dense near the source point, but sparse at longer distances.

3.12

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
Topographic parameters are not used in the codes ARANO and VALMA used by VTT.
The ARANO code uses site-specific meteorological data as well as site-specificpopulation, agricultural, and economic data estimated on a polar grid. VALMA codeuses weather data from SILAM, a Finnish code; SILAM takes topography into account,and thus topography is ‘in-built’ into its weather data.

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3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic parametersof the ATD vary within this resolution?
Topographic parameters are not used in the codes ARANO and VALMA used by VTT.

3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?
ARANO or VALMA do not have model for water bodies. Releases to water sources and aqueoustransport are handled in the DETRA code from the biospheric transport point of view. DETRA
uses compartment models.

Protective Action (Countermeasure) Modelling
3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
3.14.1. Early/Emergency Phase
In early/emergency phase of an accident short-term countermeasures are: sheltering,evacuation and iodine tablets. In ARANO there are two parameters affecting theevacuation: distance and evacuation time. The distance means that all the inhabitants
are evacuated up to that range instantly after the time given for the evacuation. The time
is calculated since the plume has reached the point under consideration. If the time value
is 0, the population has been evacuated before the plume is spread to the point and doseto the population is 0. The parameter “warning time” affects evacuation. If there is asure knowledge that the release will start after a certain time period, there is a period of
warning time available for the initiation of evacuation before the release to theenvironment begins. Also, it is possible to define combined effects, which means that
population can be divided into groups with different shielding conditions and the
combined result is the sum of all elements with their weights. In ARANO there is not a
direct input for stable iodine tablets, but it can be taken into account in calculation by
reducing internal dose from iodine isotopes.

3.14.2. Intermediate Phase:
It can be assumed that after the early phase of an accident there is sufficiently time toconsider and evaluate different countermeasures and their combinations. Relocation inthis phase refers to moving people away from the contaminated area for a longer time
period (weeks, months, years). Food prohibition is based on ingestion dose levels or
ground concentrations.

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3.14.3. Late/Recovery Phase:
In late/recovery phase possible term countermeasures include relocation, landdecontamination and food ban. In the case that external radiation dose exceeds a limit
value a potential protection measure could be decontamination. If the cleaned area after
decontamination is still too contaminated, staying there shall be reduced or denied.
Criterion for the contaminated area can be based on the external radiation dose fromfallout during 30 years.

3.15

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?
See response to 3.14.1 Age- and gender-specific variability in protective action behaviours are notconsidered.

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.
Not included in ARANO. However, the probability of evacuation success has been modelled in avery simple way in a research project. In that model, the evacuation was a success with a given
probability that depended on the distance of the municipality from the accident site. If theevacuation was considered a success, all people were considered to have been evacuated, and if a
failure, no evacuation was considered to have taken place.

3.17

What information sources do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?
Site-specific data on population amounts in different municipalities or different locations in the
map grid. Specification of the countermeasures are based on the Guides of the Radiation Safety
Authority.

Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite(e.g., worker) populations or potentially sensitive, vulnerable, or critical groups? If you do,what is your basis for including these groups, and how are they modelled?
Off-site doses do not include on-site workers. Normally individual dose represents the most
exposed individual of the population, who has normal living habits. Calculation parameters
should be realistic or conservative if not well-known.

3.19

What exposure pathways do you model? What is your basis for selecting these pathways?
What exposure duration is assumed in your models?
Cloud shine, ground shine, inhalation and ingestion (cow milk and meat, grain, green and root
vegetables) pathways. These are considered to be the most significant pathways. Exposure
duration varies by analysis. STUK’s guides define one year exposure for the dose limits.

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3.20

│ 89

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?
The normalized external dose conversion factor from the plume is calculated with a separate
numerical method due to 3-dimensional dispersion model adapted. Ingestion dose conversion
factors are based on the data of NRPB but applied to Nordic cultivation conditions. In somecases, age-dependent DCFs are implemented but normally adult dose factors are used. Otherwise
generic dose calculation parameters are used. In the VALMA code, D.C. Kocher’s coefficients forexternal dose are used.

Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.
Yes, we model and estimate radiological health effects.

3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?
Early health effects: radiation illness and fatal cases, these are modelled by deterministic models.
Latent health effects: cancers modelled by statistical model. In early health effects the target
organ is bone marrow, in latent health effects effective dose.

3.23

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, andwhat is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?
Linear no-threshold dose response is used based on generic approach; uncertainty is notconsidered. Usually the estimated number of cancer deaths is calculated as 0.05 times populationdose (manSv).

3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?
Generic average data.

Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.
Sometimes this is asked from the commercial utility side.

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3.26

With regard to potential economic consequences considered within a Level 3 PSA:
3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
Economic consequences are based on the civil defines measures which are dependent on
accident severity. Such measures include e.g. evacuation and land use restrictions (lostestate, lost production). Countermeasures are based on the action levels determined by
STUK. Various countermeasures are available in ARANO.

3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.
We have not included health effects into economic consequences.

3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.
Land contamination may restrict the use of the land, prevent ingestion of the foodstuffs
cultivated on the contaminated surface.
The need of evacuation is assessed based on the dose during one week. The area where
50 mSv is exceeded should be evacuated and based on population data, the number ofthe evaluated people can be calculated. Evacuation costs x €/person and this affects costs.
Basically, the same procedure is carried out if land decontamination, relocation or foodban are considered. Only the integration time of the dose is longer e.g. 30 years and costs
are based on other monetary losses.

3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?
Statistics Centre of Finland can provide data, which is region-specific.

Consequence Quantification and Reporting
3.28

What process and criteria do you use for selecting consequence metrics for quantification andreporting, including spatial intervals or distances from release points?
Depends on the needs of the analysis, so it is specified case by case. The Guides of STUK define tosome degree what consequences and how should be reported. There are no strict specifications
for spatial intervals or distances from the release point. Results should be clearly presented.

Risk Characterization
3.29

What process and criteria do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?
Answer is similar to the previous item. Depends on the analysis specification. Generally,
population dose, dose to the most exposed individual, health effects, cancer deaths. Multiple
radiological sources at the same site have not been analysed.

3.30

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results? How do you present the results

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from low-probability/high-consequence events in an understandable context for each group?
Who is responsible for communicating the results to each group?
(1) Decision makers:
The results of VTT’s analyses are communicated by experts at the regulatory authority or NPP
companies. Ordinary methods of statistics are used to present the results and the uncertainties
associated with them.
We present risk metrics such as the expected number of cancer deaths and the probability thatcancer deaths will result, and graphics such as Farmer’s curve.
Uncertainties are presented as the results of uncertainty analyses. The main vehicle ofuncertainty representation is graphics, and more precisely, uncertainty distributions of various
parameters such as the probability of cancer deaths. Our analyses are usually rather simple, andwe have not represented aleatory and epistemic uncertainties separately.
Examples of our way of representing risk results and uncertainties are contained in the following
two reports:
http://www.vtt.fi/inf/julkaisut/muut/2014/VTT-R-05661-14.pdf
http://www.vtt.fi/inf/julkaisut/muut/2015/VTT-R-05819-15.pdf
And the following conference paper:
Tyrväinen, Tero; Karanta, Ilkka; Rossi, Jukka. Finnish experiments on level 3 PRA.
13th International Conference on Probabilistic Safety Assessment and Management, PSAM 13, 2- 7 October 2016, Seoul, Korea. International Association for Probabilistic Safety Assessment andManagement, IAPSAM (2016).

(2) General Public:
VTT is not involved in communicating analysis results to the general public.

3.31

With regard to the treatment of uncertainties:
3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?
Yes. Normally Monte Carlo simulation is used.

3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
VTT has conducted studies in which weather parameters, radionuclide contents of therelease, and countermeasure success probability parameters have been varied. These
have been varied in the VTT code FinPSA where variation of parameters has been
conducted in an event tree model applied at level 3. VTT code ARANO does not allow
variation of parameters. VALMA code has the capability to do so, but computation isburdensome.

3.31.3. To what extent do you consider the effect of correlation on parameters?
Correlations between parameters have not been handled.

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Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?
Questions concerning environmental consequences (metrics, uncertainty analysis etc.) would
have been relevant.

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Hungary: NUBIKI, HAEA, and MTA EKSTATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT│ 93

94 │ NEA/CSNI/R(2018)1

PART 1: RESPONDENT INFORMATIONName:Country:
Organization:
Type of Organization:

Tamás Pázmándi*Hungary
Centre for Energy Research, Hungarian Academy of Sciences
� Regulatory Authority
� Utility� Vendorx Academic/Research� Other (please specify): ___________________________________

Mailing Address:
E-mail Address:
Telephone Number:

1525 Budapest 114., P.O. Box 49., Hungary
pazmandi.tamas@energia.mta.hu+36 1 392 22 22 / 1361

* with contribution of-

Attila Bareith (NUBIKI, private company)
András Siklósi (Hungarian Atomic Energy Authority)
Péter Szántó (Centre for Energy Research, Hungarian Academy of Sciences, Academic)

and reviewed by-

Gábor Lajtha (NUBIKI, private company)
Előd Holló (NUBIKI, private company)

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PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes,for what types of applications are Level 3 PSAs required?
No, there is not any regulatory requirement regarding Level 3 PSA in Hungary.

2.2

With regard to the use of Level 3 PSA:
2.2.1.

Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
No, but it is planned as research project in the short/medium term.

2.2.2.

What calculated metrics or results from Level 3 PSAs are used in these applications
and what (if any) requirements, goals, or criteria are used to evaluate these results?
What is the basis for these requirements, goals, or decision criteria?
Regulatory requirements are for public dose and to avoid countermeasures althoughthese requirements do not refer to results from a level 3 PSA.

2.2.3.

If you are not currently performing or considering performing Level 3 PSAs, haveyou previously considered performing them? If you have, what was the basis for yourdecision to not perform Level 3 PSA? Please list any barriers you perceive toperforming and applying Level 3 PSA in your country.
When the level 2 PSA was performed for NPP Paks in the early 2000s, a limited scope
analysis was also conducted to yield insights into consequences in terms of public doseand expected exceedance frequencies for fatalities. The results were discussed internallywith plant management, but the details of the analysis were not made public in lack ofregulatory requirements or any other obligations.

2.2.4.

What (if any) alternative methods do you use to estimate offsite public risks
attributable to accidental releases of radiological materials from nuclear installations,and for what types of applications are they used?
Deterministic analysis with specified source term(s) and 2-3 meteorological situations are
performed for these purposes.

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency
planning or protective action zone. Are Level 3 PSAs used to support the siting of nuclear
installations in your country? If they are, how are they used?
No. However, safety regulations require that radiological consequence hazards attributable toexisting nuclear installations have to be explicitly considered in siting nuclear installations. It has
recently been realized that level 3 PSA would be much helpful (if not indispensable) toadequately meet that requirement.

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2.4

With regard to nuclear power plant emergency planning zones:
2.4.1.

What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
Dose calculations for Design Basis Accidents.

2.4.2.

What stakeholder groups are involved in the process and what are their respective
responsibilities?
Individual experts were involved and public hearing was organized.

2.4.3.

Under what conditions (if any) can emergency planning zones be reduced insize? Can a Level 3 PSA be used to establish a probabilistic cut off for events thatneed to be considered for emergency planning purposes?
No.

2.5

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria. Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?
Not yet, but it is planned for the new NPP at Paks in the near future.

2.6

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA. Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.
Deterministic analysis and dose calculations for Design Basis Accidents and Beyond-design-basis
accidents.
Offsite radiological consequence analyses are performed for decision support systems forprotective action recommendations.
(In the SINAC decision support system developed in MTA EK in our research group and used as
an interactive expert system in the Hungarian Atomic Energy Authority Centre for Emergency
Response, Training and Analysis)

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PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES
Since we have only limited experience in Level 3 PSA and there are no ongoing activities in place atpresent, we could only partially answer the questions related to modelling issues below.

General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.

Offsite radiological consequence analyses:
PC Cosyma.
Several computer codes developed in our research group or research institute.

3.1.2.

Risk characterization:
None

3.2

With regard to the scope of Level 3 PSAs:
3.2.1.

What internal and external accident sequence initiating events and hazards are
considered in developing a Level 3 PSA?
Not applicable at present, to be specified if research on Level 3 PSA starts.

3.2.2.

Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling? If you do, how is thiscorrelation treated?
Not applicable at present, to be specified if research on Level 3 PSA starts.

3.3

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.
A new guideline for the new nuclear power plant (Paks-2) is being prepared by the Authority
with our participation with offsite radiological consequence analyses.

Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses? What (if any) difficulties have you encountered in this area?
Not applicable at present. Although we do not have detailed experience in this subject,
our experience from performing Level 2 PSA shows that the released activity ofradionuclides characterizing the dose rate can be determined from the released rate offission product groups versus time. We calculated the released rate of fission productgroups (groups: noble gases, I, Te, Cs, Ba, Mo, Sr, La, Ce, Sb, U) versus time for each
representative sequence of the containment states.
Difficulty: how to specify the representative sequences?

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3.4.2.

If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
We selected 2-6 events of each containment state (we have 13 main containment states)
and calculated the release rates for representative sequences of the containment state.
The selection of representative sequences was based on probability (contribution to thecontainment state) and also we tried to select different physical initiating events and/or
containment events resulting in the same containment failure mode.
The containment states were classified according to release categories and severity. The
release categories were determined according to the released fission product groups,
mainly the Cs-group, the Cs-137 released radioactivity.
The released rate was calculated by the MAAP code. We also used the so-called EUR(European Utility Requirements) criteria and calculated the released activity of 9
isotopes versus time and also took into account the height of the release.

3.4.3.

What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process? If you do,
how are they treated?
The MAAP4/VVER severe accident code was used for the calculation of the sequences,for determining release fraction over time. The calculations were performed until thestable state, meaning that no change is expected in the released fraction. Onsite severe
accident mitigation actions were taken into account in Level 2 PSA. These events weremodelled by the MAAP code.

3.5

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?
We used the Level 2 PSA results as a basis for evaluation with considerations to the following
radionuclides: Xe-133, Sr-90, Cs-137, I-131, Te-131M, Ru-103, La-140, Ce-141, Ba-140. It was
based on the EUR requirements referred to above under question 3.4.2.

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated?
Not applicable at present. We consider it a very important issue. Multi-source and / or multi-unit
accidents should be considered in a Level 3 PSA for multi-unit / multi-source sites toappropriately determine risk. This supposes the availability of site risk model and the associatedsite-level Level 2 PSA, which we have not developed yet.

3.7

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?
Detailed uncertainty calculations were not performed, only a limited one to explore the range of
release fractions for a containment state. Severe accident calculations were performed for aselected containment state. 40 parameters of the MAAP code were selected and varied using
Monte-Carlo methods (Latin Hypercube sampling) and we performed 200 calculations to get the
distribution of released rates of the fission product groups.

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Meteorological Data
3.8

How do you account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?
Not applicable. Only 2-3 (the most likely) meteorological situations are used for the calculations(but not for the purpose of a Level 3 PSA).

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?
No, only the 2-3 most likely meteorological situations are used for the calculations (but not for the
purpose of a Level 3 PSA).

Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?
Gaussian plume model without time dependence.

3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?
DBA calculations performed for the reference group.
BDBA and severe accident calculations are performed for 800 m and 3000 m from the release
point (public dose for located at any point along the boundaries around the NPP).

3.12

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic parametersof the ATD vary within this resolution?
Not applicable at present.

3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?
No, presently not.
Research is ongoing on this topic; a new model and computer code are expected in a few years.

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Protective Action (Countermeasure) Modelling
3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
3.14.1. Early/Emergency Phase:
3.14.2. Intermediate Phase:
3.14.3. Late/Recovery Phase:In the SINAC decision support system (developed in MTA EK in our research group and used as
an interactive expert system in the Hungarian Atomic Energy Authority Centre for Emergency
Response, Training and Analysis) countermeasures are taken into consideration.
Short-term countermeasures: sheltering, iodine prophylaxis and evacuation.
Long-term countermeasures: relocation, food ban and pasture ban.

3.15

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?
No.

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.
No.

3.17

What information sources do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?
Generic data are used.

Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite (e.g.,worker) populations or potentially sensitive, vulnerable, or critical groups? If you do, what isyour basis for including these groups, and how are they modelled?
Critical groups are used, living closest to the release point.

3.19

What exposure pathways do you model? What is your basis for selecting these pathways?
What exposure duration is assumed in your models?
Exposure pathways taken into account:- External doses: cloud shine and ground shine doses.
- Internal doses: inhalation and ingestion.

3.20

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?
Generic data are used, without gender-specificity, but specific conversion factors are used forchildren.

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Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.
No.
(Models for calculating early and late health effects were implemented in the SINAC software
many years ago, but it has never been used. These models and parameters are obsolete.)

3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?
Not applicable.

3.23

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, andwhat is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?
Not applicable.

3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?
Not applicable.

Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.
No.

3.26

With regard to potential economic consequences considered within a Level 3 PSA:
3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.
3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.
Not applicable.

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3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?
Not applicable.

Consequence Quantification and Reporting
3.28

What process and criteria do you use for selecting consequence metrics for quantification andreporting, including spatial intervals or distances from release points?
DBA calculations performed for the reference group.
BDBA and severe accident calculations are performed for 800 m and 3000 m from the release
point (public dose for located at any point along the boundaries around the NPP).

Risk Characterization
3.29

What process and criteria do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?
Not applicable at present.

3.30

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results? How do you present the resultsfrom low-probability/high-consequence events in an understandable context for each group?
Who is responsible for communicating the results to each group?
(1) Decision makers:(2) General Public:

3.31

With regard to the treatment of uncertainties:
3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?
No.

3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
Not applicable at present.

3.31.3. To what extent do you consider the effect of correlation on parameters?
Not applicable at present.

Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?
No.

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Japan: JAEASTATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT│ 103

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PART 1: RESPONDENT INFORMATIONName:Country:
Organization:
Type of Organization:

Yu Maruyama, Masanori Kimura, Toshimitsu Homma
Japan
Japan Atomic Energy Agency
� Regulatory Authority
� Utility� Vendorx Academic/Research� Other (please specify): ___________________________________

Mailing Address:
E-mail Address:

2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195, Japanmaruyama.yu@jaea.go.jp,
kimura.masanori@jaea.go.jp,
homma.toshimitsu@jaea.go.jp+81-29-282-5486, +81-29-282-5459, +81-29-282-6862

Telephone Number:

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PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes,for what types of applications are Level 3 PSAs required?
Level 3 PSA is not a regulatory requirement for licensing of nuclear power plants (NPPs).
The Nuclear Regulation Authority (NRA) of Japan is initiating discussions to implement risk
assessment utilizing Level 3 PSA, as one of methodologies to assess offsite risks by nuclear
facilities, in an activity for review of “self-assessment of safety improvement” by licensees.

2.2

With regard to the use of Level 3 PSA:
2.2.1.

Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
Offsite consequence analysis as an element of Level 3 PSA was performed to technicallysupport activities in the former Nuclear Safety Commission (NSC) of Japan for the
discussion on performance goals of NPPs to satisfy quantitative safety goals of nuclear
facilities, and activities for the revision or formulation of guidelines associated withemergency preparedness and response (EPR) in the NSC or the NRA subsequently after
the accident at the Fukushima Daiichi nuclear power station. Similar analysis iscurrently underway, aiming mainly at providing local governments with referencetechnical information for planning of EPR.

2.2.2.

What calculated metrics or results from Level 3 PSAs are used in these applications
and what (if any) requirements, goals, or criteria are used to evaluate these results?
What is the basis for these requirements, goals, or decision criteria?
In the application to the safety and performance goals, health effects of radiation
exposure among the public were used as metrics, which were namely the average
individual risks of early and cancer fatalities. The effectiveness of emergency protective
actions has been examined by comparing calculated dose to the public with the genericcriteria shown in the IAEA general safety requirements in the application to EPR.

2.2.3.

2.2.4.

If you are not currently performing or considering performing Level 3 PSAs, haveyou previously considered performing them? If you have, what was the basis for yourdecision to not perform Level 3 PSA? Please list any barriers you perceive toperforming and applying Level 3 PSA in your country.
What (if any) alternative methods do you use to estimate offsite public risks
attributable to accidental releases of radiological materials from nuclear installations,and for what types of applications are they used?
Not applicable for items 2.2.3 and 2.2.4

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency
planning or protective action zone. Are Level 3 PSAs used to support the siting of nuclear
installations in your country? If they are, how are they used?
In the application of offsite consequence analysis to EPR described in the response to item 2.2.1,
one of principal targets is to estimate reference distances from points of radionuclide release for

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emergency planning zone (EPZ) or precautionary action zone (PAZ) and urgent protective action
planning zone (UPZ).

2.4

With regard to nuclear power plant emergency planning zones:
2.4.1.

What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
In response to the experiences and lessons learned from the accident at the Fukushima
Daiichi Nuclear Power Station and on the basis of international standards such as IAEA
safety requirements, the NRA issued the Nuclear Emergency Response Guidelines. PAZ
and UPZ are introduced and their sizes are typically specified in the guidelines in orderto prevent deterministic effects and to minimize stochastic effects resulting fromradiation exposure, respectively. Size of PAZ is around 5 km and that of UPZ is around
30 km in radius from NPP.

2.4.2.

What stakeholder groups are involved in the process and what are their respective
responsibilities?
The government, the NRA, local governments and licensees are involved in the process
as stakeholder groups. The government is responsible for formulating thoroughgoingmeasures for EPR with overall efforts of its organizations and functions. The prime
responsibility of the NRA is to issue associated guidelines with EPR (the Nuclear
Emergency Response Guidelines). The local governments have responsibility to develop
and implement regional EPR plans with the aid of relevant organizations and other local
governments. The responsibility of licensees is to establish and maintain arrangementsfor onsite EPR plans and operations, and so on.

2.4.3.

Under what conditions (if any) can emergency planning zones be reduced insize? Can a Level 3 PSA be used to establish a probabilistic cut off for events thatneed to be considered for emergency planning purposes?
Not applicable

2.5

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria. Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?
As mentioned in the response to item 2.2.1, offsite consequence analysis was applied to thediscussion on safety and performance goals. A risk metric for performance goals surrogating
quantitative safety goals (the average individual risks of early and cancer fatalities), containment
failure frequency (CFF), was quantified through offsite consequence analysis. Another risk
metric, core damage frequency (CDF), was determined under the consideration of defence in
depth.

2.6

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA. Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.
Not applicable

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PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES
General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.

Offsite radiological consequence analyses:
Offsite radiological consequence is analysed with OSCAAR (Off-Site Consequence
Analysis code for Atmospheric Release in reactor accident) code, which is composed ofseveral modules to calculate atmospheric dispersion and deposition (ADD) ofradionuclides, early and chronic doses, dose reduction by protective actions, healtheffects and economic impacts.

3.1.2.

Risk characterization:
Not applicable

3.2

With regard to the scope of Level 3 PSAs:
3.2.1.

What internal and external accident sequence initiating events and hazards are
considered in developing a Level 3 PSA?
There has been no experience of offsite consequence analysis for accident scenarios
initiated with external events and hazards. It is practically possible, however, to consider
both internal and external events and hazards in offsite consequence analysis withOSCAAR code in a parametric way. Rational modelling could be one of important
challenges for, as an example, the effectiveness of protective actions under conditions
resulting from external events and hazards.

3.2.2.

Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling? If you do, how is thiscorrelation treated?
No, it is not accounted for.

3.3

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.
The improvement of OSCAAR code has been made and is in progress, including modelling ofprotective actions such as sheltering, evacuation and iodine thyroid blocking, and ADD
modelling, in which topographic effects are explicitly taken into account.

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Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses? What (if any) difficulties have you encountered in this area?
See the response to item 3.4.2.

3.4.2.

If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
For source term analysis as an element of Level 2 PSA, THALES2 (Thermal Hydraulics
and radionuclide behavior Analysis of Light water reactor to Estimate Source terms)
code is being developed. The release characteristics of source term groups specified in
THALES2 code are applied to input data of OSCAAR code. Close communication
between analysts for source term and offsite consequence analyses is made so as toconsistently deliver source term information with requirements of OSCAAR code.

3.4.3.

What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process? If you do,
how are they treated?
No clear definition is set for fraction timing and release truncation time. In general,
although those can be evaluated through source term analysis, how they apply to offsite
consequence analysis is supposed to be dependent on its targets.
In source term and offsite consequence analyses with THALES2 and OSCAAR codes,
severe accident mitigation actions and offsite protective actions are taken into account asmuch as possible, depending on the status of modelling in the codes and objectives ofanalysis.

3.5

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?
It is possible to cover more than sixty radionuclides in OSCAAR code. Important radionuclidegroups include noble gases, iodine, caesium, tellurium and ruthenium, which are based on initialcore inventory, half lives, volatility from fuels, mobility in NPPs, physical and chemical forms,
impacts on public health and the environment, and so on.

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated?
Site Level 1 and 2 PSAs are principally required in order to consider the releases from multi
units and/or multi sources in Level 3 PSA. No activities are planned at present for site level PSAs.
However, in the framework of the OECD/NEA BSAF (Benchmark Study of the Accident at
Fukushima Daiichi Nuclear Power Station) project phase 2, an analysis with OSCAAR code isongoing for the estimation of ADD of radionuclides and consequent ambient dose rate coupledwith the outputs from source term analysis with THALES2 code.

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3.7

│ 109

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?
A methodology has been developed for uncertainty and sensitivity analyses of source term. The
methodology is mainly divided into four steps, consisting of screening of uncertain parameters
based on elementary effect model, sampling by LHS (Latin Hypercube Sampling) with theconsideration of dependence among uncertain parameters, code runs for uncertainty analysis,
and global sensitivity analysis using stochastic surrogate model constructed from the results ofuncertainty analysis applying Bayesian nonparametric approach.

Meteorological Data
3.8

How do you account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?
OSCAAR has a multi-puff trajectory model that can take account of changes in wind direction
and variable long-duration releases. Trajectory and dispersion of each re-leased puff arecalculated using two kinds of grid inputs of meteorological data. The hourly surface wind andatmospheric stability fields on the meso-scale system are constructed by using Grid Point Value(GPV) provided by the Japan Meteorological Agency. The meteorological data on the synopticscale system is also available from GPV data every three hours at three standard pressure levelsof 925, 850 and 700 hPa. The wind data at the release point are also used at the first step of the
trajectory calculations. OSCAAR can handle the spatial and temporal distribution of rainfall to
predict wet deposition. OSCAAR calculations can be implemented with a variety kind of
meteorological sampling schemes such as random, stratified and bin sampling schemes. We havea specific bin sampling scheme which is appropriate for the trajectory dispersion model in
OSCAAR.

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?
We have done the uncertainty and sensitivity of radiological consequences on meteorologicalsampling schemes for use in accident consequence assessments. A full set of weather sequences
sampled from one year of hourly meteorological data was used to investigate the sensitivity ofearly health effects on certain meteorological conditions. A stratified sampling scheme wasdesigned for the trajectory model in terms of initial wind direction, rainfall amount, stability
category and travel time to a certain distance. The performance of the scheme developed wascompared with those of random and cyclic sampling schemes. The statistical variability of theprobability distribution of the consequences was also examined. (Homma, T., et al., Proc. 5thInternational Conference on Probabilistic Safety Assessment and Management, 2753-2758,
Osaka, Nov 27-Dec 1, 2000.)

Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?
The Gaussian multi-puff trajectory model is used in OSCAAR code. A typical time step for
updating calculations is one hour so as to be consistent with the time span for meso-scalemeteorological data. It is possible to decrease time step down to 15 minutes typically, depending
mainly on the time scale of the release of radionuclides.

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3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?
Basically, the meteorological data on the synoptic-scale system is used to cover whole area ofJapan. The criteria to define the boundaries depend on the application of the accident
consequence assessment. OSCAAR code uses calculation grids based on a set of annuli centred on
the site and segmented by radial lines. The spatial resolution is less detailed at large distancesfrom the site, compared with the level of detail at closer area to the site. No clear definitions areavailable for the boundaries and intervals of the spatial grid. Those are determined by expert
judgement based on experiences to date. Finer spatial discretization is generally applied at anarea closer to NPPs.

3.12

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
Site-specific data such as population, agricultural and livestock products and economic
data are taken from domestic statistics in the Ministry of Internal Affairs andCommunications, the Ministry of Agriculture, Forestry and Fisheries and local
governments, and so on.

3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic parametersof the ATD vary within this resolution?
Topographic effects are not considered in OSCAAR code at present. As described in the
response to item 3.3, the improvement of ADD model is currently underway to take theminto account.

3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?
No, they are not considered.

Protective Action (Countermeasure) Modelling
3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
3.14.1. Early/Emergency Phase:
Sheltering, evacuation and iodine thyroid blocking

3.14.2. Intermediate Phase:
Relocation and restriction of ingestion of food, milk and drinking water

3.14.3. Late/Recovery Phase:
Relocation and restriction of ingestion of food, milk and drinking water

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3.15

│ 111

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?
No, they are not modelled.

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.
No, they are not used.

3.17

What information sources do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?
Generic data are typically used for protective action modelling and parameters, which are basedon various references including, for example, Regulatory Guide on Emergency Preparedness for
Nuclear Facilities by the NSC, IAEA-TECDOC-225, NUREG/CR-4551 and OECD/NEA-EC
joint report (EUR 15109) for radiation exposure reduction coefficients, and an academic paper
by Jonson, et al. (Journal of Radioanalytical Chemistry. 65, 223-238, 1981) and ICRP documents
for metabolism modelling for iodine intake.

Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite(e.g., worker) populations or potentially sensitive, vulnerable, or critical groups? If you do,what is your basis for including these groups, and how are they modelled?
Only members of the public are considered as potentially exposed populations.

3.19

What exposure pathways do you model? What is your basis for selecting these pathways?
What exposure duration is assumed in your models?
Exposure pathway modelled are (a) external exposure from radioactive plume, (b) internal
exposure due to inhalation of radionuclides in radioactive plume, (c) external exposure fromradionuclide deposited on the ground, (d) internal exposure due to inhalation of radionuclides
resuspended from the ground, and (e) internal exposure due to ingestion of contaminated food,milk and drinking water. The exposure duration assumed are one day, seven, 14, 21, 30, 200 and
365 days for (a) through (c), and 10, 20, 30, 40, 50 and 70 years for (c) through (e).

3.20

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?
Generic but age- and gender-specific data listed in documents by ICRP and USEPA, and so on,
are used for conversion factors.

Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.
Yes, radiological health effects are modelled and estimated.

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3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?
OSCAAR code considers both early and latent health effects in the population using method
recommended by NUREG/CR-4214. The risk of early health effects is calculated using hazard
function approach in which cumulative hazard is given by a two parameter Weibull function.
The early fatal effects comprise the potentially hematopoietic, pulmonary, and gastrointestinal
syndromes. The risk of latent health effects is given by linear dose response function for eachcancer type. For estimating the lifetime risk in the population, the absolute or relative risk
projection models are available for each cancer type such as leukemia, bone cancer, breast
cancer, lung cancer, G. I. cancer, skin cancer and other cancers.

3.23

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, andwhat is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?
Linear dose response functions based on NUREG/CR-4214 Rev.2 and EPA 402-R-99-001 areapplicable. The uncertainty is not accounted for the true dose-response relationship for
exposures to low levels of ionizing radiation.

3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?
Generic but age- and gender-specific data listed in documents by USNRC and USEPA, JAERIReview 2000-029 (2000), and Japanese demographic data are used for input parameters.

Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.
Yes, economic consequences are modelled and estimated.

3.26

With regard to potential economic consequences considered within a Level 3 PSA:
3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
Modelling is made for economic consequences associated with protective actions such as
evacuation and relocation of members of the public, and restriction of ingestion of food,milk and drinking water, and early deterministic and latent stochastic health effects.

3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.
Giving an example, calculated early and latent fatalities are multiplied by input values
for income and lost duration of income to translate health effects to economic
consequences.

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3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.
The land contamination is taken into account for radiation exposure as described in the
response to item 3.19 and protective action modelling such as relocation as described in
the response to item 3.14, which becomes a part of source in the evaluation of healtheffects. The economic consequences due to the land contamination are taken into accountas cost of capital services.

3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?
OSCAAR code basically uses an approach to estimate the contribution to the Gross Domestic
Product (GDP) except for cost of radiation-induced health effects as described in the response toitem 3.26.3. Applicable data are taken from domestic statistics.

Consequence Quantification and Reporting
3.28

What process and criteria do you use for selecting consequence metrics for quantification andreporting, including spatial intervals or distances from release points?
There is no clear process or criteria for selecting consequence metrics. The analysis withOSCAAR code results in land contamination, health effects due to early and chronic radiation
exposure as a function of distances from release points with the consideration of emergency
protective actions, and economic impacts. These outputs are valuable and flexible to betranslated into various types of metrics for offsite consequences depending on objectives.

Risk Characterization
3.29

What process and criteria do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?
There is no clear process or criteria for selecting risk metrics. In general, appropriate risk
metrics should be selected according to objectives.

3.30

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results? How do you present the resultsfrom low-probability/high-consequence events in an understandable context for each group?
Who is responsible for communicating the results to each group?
(1) Decision makers:(2) General Public:
Risk results for representative accident scenarios are typically presented by using mean ormedian value and 5 and 95 percentile values, considering aleatory and/or epistemic uncertainties
as necessary (Homma T, et al., Nuclear Engineering and Technology, 37(3) 245 (2005), Kimura
M, et al., Journal of Nuclear Science and Technology, 50(3) 296 (2013)). No specific guidelines are
currently available for risk communication.

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3.31

With regard to the treatment of uncertainties:
3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?
The uncertainty and sensitivity methodologies have been implemented for the OSCAAR
code. The parameter uncertainty propagation analyses performed with OSCAAR code
to provide quantitative information on the uncertainties of individual fatality risks using
the distributions on the input parameter values obtained from a joint EC/USNRC
project (Goossens, L.H.J. et al., Radiat. Prot. Dosim., 90 (2000)). The study provided therange of uncertainty for the expected values of individual risks of early and latent cancer
fatalities at area close to the site. In the sensitivity analyses, the correlation/regression
measures were useful for identifying those input parameters whose uncertainty makes
an important contribution to the overall uncertainty for the consequence. (Homma, T, etal., Nuclear Engineering and Technology, 37(3) 245 (2005))

3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
It is considered that selecting and specifying uncertainty distributions of parameters
play important roles in uncertainty analysis. A large part of those have to practicallydepend on expert judgement based on state-of-the-art knowledge. Screening ofparameters could be possible in a systematic way with elementary effect model such as
Morris method (Technometrics, Vol. 33, No. 2, pp. 161-174, 1991).

3.31.3. To what extent do you consider the effect of correlation on parameters?
In the framework of uncertainty and sensitivity analyses described in the response toitem 3.31.1, sampling of parameters are based on LHS with consideration of correlation
between parameters. However, the quantification of correlation on parameters is adifficult task because of insufficient information and knowledge.

Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?

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Korea: KAERI and KINSSTATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT│ 115

116 │ NEA/CSNI/R(2018)1

PART 1: RESPONDENT INFORMATIONName:Country:
Organization:
Type of Organization:

Kwang-Il Ahn (KAERI*) / Taesuk Hwang (KINS)
Rep. of Korea
KAERI / KINS
■ Regulatory Authority (KINS)� Utility� Vendor■ Academic/Research (KAERI)� Other (please specify): ___________________________________

Mailing Address:

Korea Atomic Energy Research Institute (KAERI), Daedeok-Daero989-111, Yuseong-gu, Daejeon, KOREA, ZIP: 34057
kiahn@kaeri.re.kr (KAERI), tshwang@kins.re.kr (KINS)+82-42-868-2657 (KAERI), +82-42-868-0653 (KINS)

E-mail Address:
Telephone Number:(*)

The corresponding organization(**) The following members of the KAERI and KINS technical staff reviewed andsupplemented this survey response:- Sung-yeop Kim, KAERI- Dohyoung KIM, KINS
- Yongjin LEE, KINSSTATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 117

PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes,for what types of applications are Level 3 PSAs required?
In 2001, Korean Government announced the Severe Accident Policy to impose conducting PSA
on all domestic nuclear power plants to identify the vulnerability against severe accident and toassess public risk.
The implementation plan for the policy, inter alia, addressed clearly that newly built NPP with
APR1400 system, should conduct Level 3 PSA and demonstrate to meet safety goal. The policy
defined “Level 3 PSA” as the assessment of off-site radiological consequences to evaluate publicrisk due to severe accident.

2.2

In 2015, the Nuclear Safety Acts were amended to accommodate the severe accident
management. The Acts clearly require conducting the Level 3 PSA tied with “AccidentManagement Plan” to be submitted. According to the aforementioned AMP, PSA results shallsatisfy the risk target values (early fatality risk (or equivalent performance goal), cancer fatality
risk (or equivalent performance goal), frequency of Cs-137 release of more than 100 TBq as
below 1.0⨉10-6/yr.

With regard to the use of Level 3 PSA:
2.2.1.

Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
The Korean utility, KHNP (Korea Hydro and Nuclear Power Co.), carries out the Level
3 PSA on newly built NPPs for operating license. A submitted Level 3 PSA results arereviewed by KINS (Korea Institute of Nuclear Safety, Regulatory Support
Organization).

2.2.2.

What calculated metrics or results from Level 3 PSAs are used in these applications
and what (if any) requirements, goals, or criteria are used to evaluate these results?
What is the basis for these requirements, goals, or decision criteria?

In Korea, a QHO (Quantitative Health Object) for determining the Level 3 PSA to meet the goal is
as follows:

Metrics

Requirements
Goal
Early fatality risk from nuclear accidents
Earlyshall not exceed 0.1% of the total early 5.0E-7/yr
Fatality Risk
fatality risk arising from other accidents
Cancer fatality risk from nuclear accidents
Cancer Fatality
shall not exceed 0.1% of the total cancer 1.0E-6/yr
Risk
fatality risk
2.2.3.

Basis
QHO
0.1% rule
QHO
0.1% rule

If you are not currently performing or considering performing Level 3 PSAs, haveyou previously considered performing them? If you have, what was the basis for yourdecision to not perform Level 3 PSA? Please list any barriers you perceive toperforming and applying Level 3 PSA in your country.
N/A

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2.2.4.

What (if any) alternative methods do you use to estimate offsite public risks
attributable to accidental releases of radiological materials from nuclear installations,and for what types of applications are they used?
N/A

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency
planning or protective action zone. Are Level 3 PSAs used to support the siting of nuclear
installations in your country? If they are, how are they used?
According to the recommendation of EPZ (emergency planning zone) established by IAEA, wehave recently re-established the EPZ in terms of precautionary action zone (PAZ) as within 3 to 5
km and urgent protective action planning zone (UPZ) as within 20 to 30 km. In the process, wehave assessed both PAZ and UPZ ranges through a comparison of exposure doses to emergency
protection criteria which have been assessed using MACCS (MELCOR Accident Consequence
Code System). RASCAL (Radiological Assessment Systems for Consequence Analysis) has been
also used to verify the decision basis of PAZ and UPZ range of IAEA. In consideration of theanalysis results by code and safety margin, the PAZ and UPZ compliance with the regulatory
provision was established.

2.4

With regard to nuclear power plant emergency planning zones:
2.4.1.

What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
The Act on physical protection and emergency planning requires for licensee to establish
emergency planning zone (EPZ) where intensive protective and other response action
shall be implemented in case of abnormal release of radionuclides. The EPZ ischaracterized by radius of being effected preparedness of prompt emergency actions tolessen population exposure dose and specified by licensee consulting with local
government taking into account the network of road and geopolitical circumstance. Afterapproval by the regulatory authority NSSC (Nuclear Safety and Security Commission),
the licensee shall reflect the approved EPZ into their emergency planning.

2.4.2.

What stakeholder groups are involved in the process and what are their respective
responsibilities?

Stakeholder Group
Division of disastermanagement related in Local
Governments
NSSC
KINS
Korea Institute ofRadiological & Medical
Sciences (KIRAMS)
KHNP

Respective Responsibilities
Consultation with licensee
Inter-local government coordination and cooperation
on boundaries
Integration of review and assessment by local
government and relevant agencies or organization
Providing legal basis for EPZ
Providing Technical consultation on EPZ andemergency preparedness
Providing Technical consultation on emergency
medical response and Potassium Iodide (KI)-Providing information on EPZ

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2.4.3.

│ 119

Under what conditions (if any) can emergency planning zones be reduced insize? Can a Level 3 PSA be used to establish a probabilistic cut off for events thatneed to be considered for emergency planning purposes?
In general, the radius of EPZ is estimated more conservatively than the radius derived
from the Level 3 PSA. However, we have currently no experience to curtail the size of
EPZ. The Level 3 PSA would be partially used if it is required to reduce the size of EPZ.

2.5

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria. Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?
The metric of QHO (safety goal) is driven from applying a 0.1% rule to number of annual fatal
casualty by all types of accident based on statistics from the National Statistical Office. The tworisk acceptance criteria, CDF (Core Damage Frequency) and LERF (Large Early Release
Frequency), are used as surrogate of QHO. In this step, the Level 3 PSA is partially utilized todetermine the CDF and LERF. Detailed methodology using the Level 3 PSA was described in
KINS/ER-190 (Performance Goals for the Korean NPPs). A summary of the report is as follow,(1) Perform conservative calculation for early and late fatality risk at specific distance using
MACCS.
(2) Evaluate the final individual risk combining recommended LERF value and pre-calculated
early and late fatality risk.
(3) Verification the risk acceptance criteria through comparison between final individual risk
and safety goal.

2.6

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA. Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.
In Korea, offsite radiological consequence analyses have been performed to support theidentification of risk impact due to the extension of surveillance test interval for ILRT (Integral
Leak Rate Test) of the containment. In addition, Radiological Environment Report (RER) has
been provided as a licensing document, which is based on the design basis accidents (DBAs) andnormal operation procedures.

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PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES
General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.

Offsite radiological consequence analyses:
The MACCS (MELCOR Accident Consequence Code System) is used for the offsite
radiological consequence analyses. HotSpot and ADMS are supportively used in somecases.

3.1.2.

Risk characterization:
Currently, there are no specific codes for the risk characterization

3.2

With regard to the scope of Level 3 PSAs:
3.2.1.

What internal and external accident sequence initiating events and hazards are
considered in developing a Level 3 PSA?
The internal and external initiating events to be considered in the Level 3 PSA may becharacterized by the initiating events considered to perform the Level 1 PSA. For
example, a set of initiating events considered in the Level 1 PSA is shown in the below
table. Then the Level 3 PSA takes into account the initiating events listed in the table.

Type

Group

LOCA

Internal
Transient

External

Initiating Event
Large LOCA
Medium LOCA
Small LOCA
Reactor Vessel Rupture
Steam Generator Tube RuptureInterfacing System LOCA
Large Secondary Side BreakLoss of Feed Water
Loss of Condenser Vacuum
Loss of a CCW System Train
Loss of a 4.16KV Bus
Loss of a 125V DC Bus
Loss of Off-site Power
Station Blackout
General Transients
Anticipated Transients Without Scram

Internal FireInternal Flood
Seismic

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3.2.2.

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Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling? If you do, how is thiscorrelation treated?
We do not take into consideration of correlation between causes of accident sequence
initiating events and offsite phenomenological and consequence modelling for the Level 3
PSA. Whereas, different models are being developed for seismic emergency situationconsidering delay time of evacuation, evacuation speed, ratio of the cohort successfully
evacuated, and etc.

3.3

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.
The research organization, KAERI, has been performing the Level 3 PSAs as a research project,
whose purpose is to apply to the site risk assessment and emergency preparedness & response. In
parallel with this, KAERI is currently preparing a framework to develop a domestic Level 3 PSA
code which will be launched as a government-sponsored 5-year project in March 2017.

Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses? What (if any) difficulties have you encountered in this area?
We have performed Level 2 PSA severe accident analyses with MAAP (mainly in theutility side) or MELCOR code (mainly in the research and regulatory sides), and with
Level 3 PSA with the MACCS code. One of essential parts in the transition process fromLevel 2 PSA to Level 3 PSA is to determine accident sequence-specific source terms. We
have made the transition through regrouping and classification of radionuclides basedon characteristics and frequencies of source term categories resulted from the Level 2
PSA.
When MELCOR is used for the Level 2 PSA, MELMACCS is preferred to be used asinterfacing software. When MAAP is used for the Level 2 PSA, we use KOSCASOURCE (Korea Off-site Consequence Analysis – Source term module) as interfacing
software. KOSCA-SOURCE has been developed by KAERI to convert the results of
MAAP into the input for MACCS. Plume segmentation into 4 segments is possible.
However, there is currently an asymmetry between two codes (for Level 2 PSA and forLevel 3 PSA) in terms of classification of radionuclides. For example, the MAAP code
applies 12 groups of radionuclides (version 4.06) and 18 groups (version 5.0 above) while
MACCS considers 9 groups. Following table provides the scheme of reclassification ofradionuclides for coping with the asymmetry between MAAP 4.06 and MACCS2 for the
Level 3 PSA.

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3.4.2.

Number

MACCS Source-term Group

MAAP Source-term Group1
2
3

Noble Gas
I
Cs

Noble Gas
CsI, RbI
CsI, CsOH

4

Te

TeO2, Te2, Sb

5
6
7
8
9

Sr
Ru
Ba
La
Ce

SrO
RuO2
BaO
La2O3
CeO

If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
The grouping of source term is closely linked with the Level 2 PSA accident scenario andplant-specific characteristics. We have utilized 9 category binning parameters (listed infollowing table) to group source terms into each relevant source term release category(STC) whose number ranges from 19 to 21. Then a representative accident scenario foreach source term group (derived from either frequency-base or consequence-base) is
utilized to assess the characteristics of each source term (using MAAP or MELCOR).

Category
Number binningparameter

Description

1

CONBYPASS

Containment bypass

2

CONISOL

Containment isolation

3

MELSTOP

In-vessel melt retention

4

TIMECFTime of containment failure

5

MODECF

Mode of containment failure

6

CSS

Containment spray system

7

CAVCOND

Cavity condition

8

SCRUB

FP scrubbing for bypass

9

SG

Secondary heat removal (early)

BranchNOBYPASSISL
SGTRISOLATEDNOTISOCSNOTISOLNOCSMELSTOPRVRUPTURE
CFBRB
NOCF
EARLY
LATE
BMT
LEAK
RUPTURE
CS
NOCS
DRY
WET
NOT-SCRUB
SCRUB
SG
NOSG

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What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process? If you do,
how are they treated?
The off-site release point is identified through the MAAP or MELCOR analysis for eachsource term group that is derived in the Level 2 PSA.
Then, we just use it as input parameter (PDELAY of MACCS). When performing
conservative analysis, 72 hours’ total release fraction is assumed to be released in 1 hour
duration (1 plume segment). If more realistic analysis is needed, plume segmentation iscarried out. For example, 50% of 72 hours’ total release fraction is assumed to bereleased in 1 hour and the rest 50% of it is assumed to be released during next 9 hours
The onsite severe accident mitigation actions including cavity flooding strategy orcontainment depressurization strategy are taken consideration into the Level 2 CET(Containment Event Tree)/DET (Decomposition Event Tree) models or MAAP model inthe process of Level 2 PSA. However, an offsite emergency response action is notconsidered in the domestic Level 3 PSA for conservative purpose.

3.5

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?
60 radionuclides are taken into account to characterize the offsite radiological consequences
attributed to accidental release. The maximum inventory of each radionuclide can be obtained bycomputer code under the conservative assumption (e.g., ORIGEN). The chosen 60 radionuclides
for the analyses stems from WASH-1400 and other previous research.

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated?
Currently, the utility side does not consider releases from multiple units. For reference, whenusing MACCS, KAERI has recently developed a framework to use a release fraction of single
unit multiplied by the number of units.

3.7

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?
In the utility side, sensitivity and uncertainty analyses are not included in the scope of theanalyses due to the consideration of most conservative release category or fraction.
However, diverse sensitivity analyses on plume segmentation, plume height, and heat content ofplume have been carried out in the research organizations (mostly being performed in KAERI).
Our experience shows that effective plume height and surface roughness are very importantfactors influencing the consequence.

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Meteorological Data
3.8

How do you account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?
In Korea, plant-specific and hourly-based annual meteorological data are used for the offsite
consequence analysis. The meteorological data are acquired from meteorological tower locatedperipherally around the plant. Whereas, the MACCS code in connection with the Level 3 PSA
has no capability to deal with spatial variability due to moving plume.
To select representative meteorological data set, we adopt a uniform-bin sampling method basedon allocated probability of data bin in terms of wind speed, meteorological stability and degree ofprecipitation.

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?
For the Level 3 PSA, we do the sensitivity study based on plant-specific and hourly-based annual
meteorological data which are collected over 5 years or 10 years.

Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?
As the ATD model for Level 3 PSA, we use the Gaussian Plume model embedded in the MACCS
code.
In Gaussian Plume model, however, it is noted that a time dependency is hard to consider since
the trajectory of released radioactive material plume cannot be changed.

3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?
The boundaries for off-site consequence calculation in terms of Level 3 PSA are comprised ofthree evaluation distance (or radius), namely EAB (Exclusive Area Boundary), early fatality(1.6km, 1mile) and latent fatality (16km, 10mile). The distance interval is set in consideration of
the site area and characteristics.
According to our experience, enough number of spatial grids was found to be very importanteven in short range for early fatality estimation due to steep gradient of plume concentration. For
the purpose of detailed segmentation, detailed data of population distribution should be
supported. KOSCA-POP, which has been recently developed in KAERI, is used to develop
population distribution data in the research level. Three kinds of Korean population data are
installed in KOSCA-POP: area population, centre population, and point population.

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3.12

│ 125

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
The SITE module of MACCS is configured to input the data that reflects the sitecharacteristics such as population and land fraction in each grid. Those site-specific data
corresponding to each grid can be obtained from the National Statistical Office.
For reference, a surface roughness may be one of few topographical characteristics
which can be applied in MACCS. As only one surface roughness length could be applied
as MACCS input, it makes difficulty to decide only one value which represent complex
terrain mixture.

3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic parametersof the ATD vary within this resolution?
The topographic parameter (land fraction, region index, watershed index) in the
MACCS code is normally assigned as single constant value to each grid. However,
MACCS uses a mean value that does not vary with ATD response during the calculation.

3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?
N/A

Protective Action (Countermeasure) Modelling
3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
For the Level 3 PSA, protective actions are not considered for a conservative purpose. However,
we can generally take into account the followings:

3.14.1. Early/Emergency Phase:
In-house sheltering, evacuation, dose-dependent relocation, distribution of potassium
iodide.

3.14.2. Intermediate Phase:
Dose-dependent relocation (based on radiation criteria).

3.14.3. Late/Recovery Phase:
Decontamination, temporal interdiction, condemnation.

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3.15

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?
In Korea, the Level 3 PSA for domestic NPPs doesn’t consider credit to the emergency response(evacuation or sheltering) for the purpose of conservatism. Meanwhile, an offsite consequence
analysis for extension of ILRT test interval had taken the all-out evacuation within EPZ with theassumption of single population group.

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.
As mentioned above, we do not take into account emergency response for doing the Level 3 PSA
for domestic NPP. However, in case of offsite consequence analysis for ILRT, notwithstanding
the recommendation of all-out evacuation, 5 % residual (sheltering) and 95% evacuation are
assumed. Expert judgment is used to assign the probabilities of cohorts.

3.17

What information sources do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?
In terms of emergency response, protective actions are based on the plant-specific radiologicalemergency plan and its evacuation time estimation which is a legal requirement of nuclear
regulation, and international guideline.

Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite(e.g., worker) populations or potentially sensitive, vulnerable, or critical groups? If you do,what is your basis for including these groups, and how are they modelled?
In Korea, the exposed populations (or cohorts) can be interpreted as the habitant who lives
nearby target nuclear power plant, which is benchmarking point to evaluate ingesting pathway.
Meanwhile, on-site personnel are not considered in the case of Level 3 PSA because the target
audience for radiological exposure and dose assessment is only off-site resident people.

3.19

What exposure pathways do you model? What is your basis for selecting these pathways?
What exposure duration is assumed in your models?
For exposure pathways, we utilize two MACCS2 models in the following table.

Exposure Duration
EARLY(~7days)

LATE(~few years)

Exposure pathway
Cloudshine
Groundshine
Cloud Inhalation
Resuspension Inhalation
Groundshine
Resuspension Inhalation
Ingestion from Contaminated food & water

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│ 127

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?
We apply the dose conversion factor for each radionuclides and exposure pathway embedded in
the MACCS code (i.e., cloud shine, cloud inhalation, ground shine, resuspension inhalation, skindose, and ingestion), which are recommended by ICRP without consideration of type of genderand age. For reference, KAERI has recently developed a domestic food chain model for ingestion.

Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.
The radiological health effects have been evaluated with the built-in model of the MACCS code.

3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?
While the organs closely related with early and late fatality are taken into consideration for the
analysis, the organs related to non-fatal effects (injury effect) are out of consideration. Following
table provides the relationship between health effect and target organ associated with early andlatent fatality.

Category
Early fatality

Latent fatality

3.23

Health Effect
Hematopoietic Syndrome
Pulmonary Syndrome
Gastro-intestinal Syndrome
Leukemia
Bone Cancer
Breast Cancer
Lung Cancer
Thyroid Cancer
Gastrointestinal Cancer
Other Cancers

Target Organ
Rad Marrow
Lung
Lower Large Intestine
Rad Marrow
Bone Surface
BreastLungs
Thyroid
Lower Large Intestine
Other

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, andwhat is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?
For the conservatism, we use the LNT (Linear No-Threshold) model as the dose-response model.
Most of international Agencies including ICRP support the LNT model for cancer and genetic
disorder induced from low-dose exposure. The uncertainty analysis in regard with theapplication of LNT has not been made.
For reference, dose and dose-rate effectiveness factor (DDREF) employed in MACCS2, which isa multiplicative adjustment that results in lower tangent of risk for low dose, are supporting toaccount for uncertainty in dose-response in case of low dose level and low dose-rate.

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3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?
We use a generic data for health effect provided in NUREG/CR-4214, which is assigned as single
value regardless of age and gender.

Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.
N/A

3.26

With regard to potential economic consequences considered within a Level 3 PSA:
3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.
3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.
N/A

3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?
N/A

Consequence Quantification and Reporting
3.28

What process and criteria do you use for selecting consequence metrics for quantification andreporting, including spatial intervals or distances from release points?
The main objective of domestic Level 3 PSA is to evaluate the off-site health effect, early and latefatality risk for individual as consequence metrics and to demonstrate its compliance with thesafety goal. Other metrics in MACCS code are used as supporting measures. We had alreadystated the criteria on spatial intervals or distances from release points at the questionnaire 3.11.

Risk Characterization
3.29

What process and criteria do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?
Two risk metrics, early and latent fatality risk are selected for effectively communication on
Level 3 PSA in accordance with the safety goal. No experience of Level 3 PSA for collocatedmulti-unit at the same site has been made.

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3.30

│ 129

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results? How do you present the resultsfrom low-probability/high-consequence events in an understandable context for each group?
Who is responsible for communicating the results to each group?
A communication with decision makers and general public has been made by the regulatory
authority NSSC. Related to risk results and their uncertainties, the mean (expected) values areused to compare early and late cancer fatality risk with safety goal which is set out in question
2.2.2. In the currently submitted Level 3 PSA report, only the mean values of the risk results arelisted and the output of the MACCS code which is submitted as an attachment shows the
distribution that occurs due to the weather changes. That is, the distribution of only random
uncertainty is expressed in CCDF (Complementary Cumulative Distribution Function) form.
High-Consequence-Low-Probability events are generally screened out during performing theLevel 1 PSA, on the basis of the frequency of occurrence. 'Screened out' means whether detailed
analysis is performed or not. For example, detailed analysis of internal flooding event isperformed where CDF exceeds 1E-09/yr for APR-1400.
(1) A communication with decision makers can be made through the risk metrics of early andlate fatality risk which makes possible to explain in terms of safety goal.
(2) A communication with general public in terms of health effect has been made by thesunshine law or public hearing by the regulatory authority NSSC.
Although laws related to information disclosure exist, there are no specific guidelines for risk
communication on PSA results.

3.31

With regard to the treatment of uncertainties:
3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?
We do sensitivity or uncertainty analyses as a part of risk estimation in terms of annual
meteorological data and demography. The sensitivity analyses for the meteorological
data are made based on 5-year data retrospectively from the timing of estimation. In caseof demographic data, the sensitivity on variation demographic data is analysed with 10year data for 50 years from the standpoint of the date of starting licensing processadministratively. All analyses result should be included into submitted report to makepossible tracking changes of input data.

3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
For avoiding uncertainty propagation of parameters adopted for the assessment, wedelineate outstanding parameters by expert engineering judgment and do the analyses
corresponding to the identified outstanding parameter. In recent analysis, 9 outstanding
parameters were identified including dispersion parameter; inverse layer; plume riseeffect; surface roughness; deposition model; source term; dose conversion factor; healtheffect model; emergency response and shielding factor.

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3.31.3. To what extent do you consider the effect of correlation on parameters?
N/A

Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?
N/A

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Netherlands: ANVS & NRG

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PART 1: RESPONDENT INFORMATIONName:Country:
Organization:
Type of Organization:

Mailing Address:
E-mail Address:
Telephone Number:

ANVS: Gert Jan Auwerda, Rob Jansen
NRG: Jacques Grupa, Hans Brinkman
Netherlands
ANVS (Regulator), NRG
X Regulatory Authority
X Utility (research reactor)
� Vendor� Academic/Research
X Other (please specify): TSO_________________________________
ANVS: Bezuidenhoutseweg 67, Postbus 16001, 2500 BA, Den Haag
NRG: Westerduinweg 3, 1755 LE Petten, PO Box 25, 1755 ZG Petten
gertjan.auwerda@anvs.nl,
rob.jansen@anvs.nl, grupa@nrg.eu,
brinkman@nrg.eu
ANVS: ++31 6 11376922
NRG: ++31 224 56 4957

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PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes,for what types of applications are Level 3 PSAs required?
Yes. A level 3 PSA is required to show the nuclear installation containing fissile materials and/or
ores complies to certain risk acceptance criteria when applying for a license (or a change of thelicense) for establishing, constructing, commissioning, operating or decommissioning of a nuclear
installation.

2.2

With regard to the use of Level 3 PSA:
2.2.1.

Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
Power reactor, research reactors, hot cell facilities, waste
decommissioning facilities, waste storage facilities, enrichment facility

2.2.2.

treatment

andWhat calculated metrics or results from Level 3 PSAs are used in these applications
and what (if any) requirements, goals, or criteria are used to evaluate these results?
What is the basis for these requirements, goals, or decision criteria?
Risk criteria are explicitly included in the Nuclear Installations, Fissionable Materials
and Ores Decree (Bkse) as assessment principles for licences to be granted to nuclear
reactors. The outcomes of a level-3 PSA must be compared with these risk criteria andobjectives. The requirements are formulated as follows:a) That the probability that an individual, whom resides permanently and unprotected
outside the facility grounds, dies as a result of a beyond-design accident is less than
10-6 per year. To calculate this ‘individual risk’, the habits, location, age, etc. of thisindividual are chosen in such a way as to result in the worst-case scenario that would
still be reasonable, but does not have to be a real person. In practice this is a 1-year
old infant that would live next to the edge of the installation. In determining the total
dose resulting from an accident, one has to take into account long-term (stochastic)
effects for at least 50 years, including ongoing exposition over that time-period due
to contamination of the environment.
b) That the probability that a beyond-design accident leads to at least 10 direct
fatalities (within a few weeks, due to deterministic effects) is below 10-5 per year, andthe probability on n times more fatalities is n2 times smaller. (e.g., 100 fatalities,
probability < 10-7 per year; 1000 fatalities, probability <10-9 per year; etc.). For the
calculation of this ‘group-risk’ actual distribution of the population is to be taken
into account.
In demonstrating compliance with the risk criteria, it has to be assumed that only theusual forms of mitigative action (i.e. fire brigades, hospitals, etc.) are taken. Evacuation,
iodine prophylaxis, sheltering or decontamination of the environment may not,therefore, be included in the calculation.

2.2.3.

If you are not currently performing or considering performing Level 3 PSAs, haveyou previously considered performing them? If you have, what was the basis for yourdecision to not perform Level 3 PSA? Please list any barriers you perceive toperforming and applying Level 3 PSA in your country.

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2.2.4.

What (if any) alternative methods do you use to estimate offsite public risks
attributable to accidental releases of radiological materials from nuclear installations,and for what types of applications are they used?
For “design basis accidents”, licensees have to show, using a deterministic approach, thatfor representative postulated events the dose for an individual, whom resides
permanently and unprotected outside the facility grounds stays below certain thresholdvalues, depending on the expected frequency of the event. Frequency – dose limit values
are: F≥10-1 per year à dose ≤0.1 mSv; 10-1>F≥10-2 per year à dose ≤1 mSv; 102>F≥10-4 per year à dose ≤10 mSv; F<10-4 per year à dose ≤100 mSv. For individuals
younger than 16 years, the dose limit is multiplied by 0.4. Additionally, the effective
thyroid-dose has to be ≤500 mSv.
The NPP Borssele uses a 'Living PSA', which include a spread sheet tool with strategiclevel-3 results, to easily recalculate the offsite risks when accident frequencies change.

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency
planning or protective action zone. Are Level 3 PSAs used to support the siting of nuclear
installations in your country? If they are, how are they used?
In practice, the ‘group risk’ mentioned under 2.2.2 takes this role. Additionally, for new build
reactors, the newly developed Dutch Safety Requirements (DSR) set some additional goals. The
following table contains the design requirements for a core-melt accident that cannot bepractically excluded:
Protective
actionPermanent
evacuation
Evacuation
Sheltering
Iodineprophylaxis

Evacuationzone(< 3 km)

Shelteringzone(< 5 km)

Beyondsheltering
zone

No

No

No

No

No

May be neededNo

May be neededNoMay
neededMayneededMayneededbebebe

The zones serve as design requirements in combination with the Dutch intervention levels. In that
context, the following intervention levels apply: for sheltering, the intervention level is aneffective dose of E ≥ 10 mSv; for evacuation, the intervention level is an effective dose of E ≥ 100
mSv and for the distribution of iodine prophylaxis, the intervention level for children is a thyroid
dose of Hthy, <18 yr ≥ 50 mSv and the intervention level for adults is a thyroid dose of Hthy,≥18 yr ≥ 100 mSv.

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Figure 2: Zones and intervention levels for postulated core-melt accidents.
Site boundary, 0 km

3 km

5 km

Distance > 3 km,
E < 100 mSv

Distance > 5 km, E < 10 mSv,
Hthy, <18 yr < 50 mSv

The zones specified in the requirements are intended for use during the design of a new reactor,
not in emergency response. The National Plan for Nuclear and Radiological Emergencies(“Nationaal Crisisplan Stralingsincidenten”) and the regional emergency response plans specify
the action to be taken in the event of an accident and the associated preparation zones. The
requirements apply to the design. Hence, an applicant shall demonstrate that, in the event of any
postulated core-melt accident, the requirements will be met. The analyses shall of course takeaccount of the local weather conditions.

2.4

With regard to nuclear power plant emergency planning zones:
2.4.1.

What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
A number of DBA's are selected, and it is calculated to which distance dose thresholds
for various measures (shielding, immediate evacuation, evacuation or relocation, foodcontrol measures) can be exceeded with a (weather statistics dependent) probability ofmore than about 5%.

2.4.2.

What stakeholder groups are involved in the process and what are their respective
responsibilities?
Plant owner, plant operator, local community Mayor, regional emergency organisation(translated from 'veiligheidsregio'), national response organisation.

2.4.3.

Under what conditions (if any) can emergency planning zones be reduced insize? Can a Level 3 PSA be used to establish a probabilistic cut off for events thatneed to be considered for emergency planning purposes?
The emergency response is an independent level of defence, and reducing this motivatedby considering other levels of defence (plant safety systems as in a PSA) could
compromise the 'defence in depth' approach. The motivation of reducing emergency
planning zones should not only rely on a PSA. Ideally, it should be a risk informed
approach.

2.5

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria. Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?
In NL, the PSA (nuclear) and QRA (for conventional plant: coal / gas fired, chemical, oil & gas
industry; QRA is comparable to PSA) is used to determine the size of the unmitigated risks(nuclear and conventional), and this is used to show compliance with the legal limit, mentioned
under 2.2.2. These limits are safety goals / risk acceptance criteria.

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Additionally, it is required to meet the safety goals/ risk acceptance criteria mentioned under
2.2.4. There is a distinction between these deterministic analyses of design base accidents, where
it is shown that each safety system is capable of mitigating the risk of a stylized, but foreseeable
accidents (i.e. a safety goal), and PSA, where the remaining (unmitigated) risk is calculated (i.e.
risk acceptance). In showing compliance to these requirements, calculation methods very similar
to those used for the Level 3 PSA are used.

2.6

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA. Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.
As mentioned under 2.2.4 and 2.5, similar calculation methods as for level-3 PSA are used toshow compliance to the legal limits (safety goals) for design base accidents.
Additionally, PSA-like methods are used for emergency planning and by the licensees for
insurance applications.

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PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES
General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.

Offsite radiological consequence analyses:
License holders use COSYMA as well as codes based on the Dutch New National Model(NNM) such as NUDOS2. MACCS2 is rarely used.

3.1.2.

Risk characterization:
Spread sheet post processing, GIS-(contour)-mapping

3.2

With regard to the scope of Level 3 PSAs:
3.2.1. What internal and external accident sequence initiating events and hazards are
considered in developing a Level 3 PSA?
This is part of the level-1 and -2. Starting point are the international IAEA lists
completed with site specific PIE. The Level 3 PSA carried out for NPP is full scope andincludes all plant states, both core and spent fuel, internal and external events, internaland external hazards, human errors, common cause errors, etc.

3.2.2.

Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling? If you do, how is thiscorrelation treated?
To take into account these correlations is only required if otherwise results would not be
conservative. So a conservative approach is taken to account for such correlations.
However, we have no systematic process to evaluate this. For situations where we
explicitly know that the risk dominating accident sequence is initiated by an event that
correlates to one or more PSA-3 stochastic parameters, ignoring this correlation is - inour experience - in practice conservative. E.g. for the PIE 'impact of a military airplane
during an exercise', we know that this only takes place during daytime. That rules out
most (stable) mixing layer conditions (occurring at night). Including such correlations in
the calculations is perfectly allowable. Similar, high wind speeds can cause damage, but
at the same time give a high dilution factor in air.

3.3

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.
We are in the final stages of the development of a new guidance document for performing PSA-3
calculations. Part of the development of this new guide included studies related to the applicationof the dispersion models in COSYMA versus the dispersion models in the ‘NNM’ model (Nieuw
Nederlands Model = New Dutch Model) and studies related to the applicability of the NNM foremissions related to accidents, as the NNM was originally developed for continuous emissions ofhazardous substances (not necessarily related to nuclear activities).

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Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses?

3.4.2.

What (if any) difficulties have you encountered in this area?
Level-2 provides masses of species that are released as a function of time, and the massesof the chemical elements in the core. Special attention must be given to chemicalelements in a species that is altered during its migration through the plants systems (e.g.
iodine conversion in water), chemical elements that occur in more than one species (I2,
organic-I, CsI, CsOH, Cs2MoO4, other Mo-species), and radioactive daughter productsthat grow during the migration through the plant

3.4.3.

If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
That is based on an expert judgment by the Level-2 specialists, that pick out arepresentative accident for each source term group. The practical limitation is in theLevel-2: each representative accident can take up to some weeks of computing time in
the Level-2 to calculate the release data vs. time. (In level-3 this is a few hours per source
term at max)

3.4.4.

What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process? If you do,
how are they treated?
We carefully study the release fraction timing by visual inspection of the graphs. Early
fatality risk depends on this timing, but until now we observed that accident releases thatcause early fatality risks take only a few hours, so it can be covered by a short sequenceof hourly releases. Long lasting releases relate to partial functioning of one of thebarriers, e.g. a containment. In those cases the release is smaller, and thresholds forearly health effects are not exceeded, in our experience. The stochastic dose-risk isslightly conservatively modelled as linear, no threshold and no DDREF. In that case, it
can be mathematically shown that the release can be assumed to occur in one hour
without getting wrong results, as long as the stochastic risk is smaller than 1. (The coreingredient of the mathematical proof is that the risk is linear with the dose.)
Onsite severe accident mitigation actions should be considered in the PSA-2 to calculate
the source term. For offsite mitigating actions it has to be assumed that only the usualforms of mitigative action (i.e. fire brigades, hospital care, decontamination of persons,etc.) are taken. Evacuation, iodine prophylaxis, sheltering may not be included in thecalculation. Indirect mitigating actions such as restrictions on the food-chain should beconsidered.

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3.5

│ 139

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?
The present database contains more than 800 nuclides. To our knowledge this covers all nuclides
that may be relevant for radiological consequences. The selection of the 800 nuclides is in the end
an expert judgment. In the 1990s, the database was limited to about 50 nuclides, also selected byexpert judgment. We have observed that the 750 nuclides that have been added since then do notcontribute significantly (except in some very exotic accidents).

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated?
Sometimes this is required, it depends on the license type. If the initiating event or hazard resultsin multiple releases, these should be considered together. In that case the risks are summed bypost-processing, most often in a spread sheet.

3.7

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?
No

Meteorological Data
3.8

How do you account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?
Only temporal variability. The risk metrics only need short distance calculations, and somemedium distance averages. These results are not sensitive to the spatial variability in the
Netherlands.
Hour by hour meteorological data sequences for the whole year are available.

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?
A large enough sampling of weather sequences is used. This is code specific. For NUDOS2 thesample size is 8000 to 80000 weather sequences. Straightforward statistics are sufficient.

Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?
This is code specific. NUDOS2 uses one hour averaged Gaussian plume, aerosol size and friction
velocity dependent deposition, aerosol size and rain rate dependent wet deposition (standard
Dutch air pollution model NNM or SRM3).

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3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?
This is code specific. NUDOS2 uses a polar grid of five kilometre for the location dependent
result (max risk, max dose), up to 25 km for regional average food contamination estimates.
There is no hard limit on the number of result grid point (normally we use about 10000 grid
points) in the code. After a calculation, we do a visual inspection of the results to see if the grid
was sufficiently dense. Moreover, the code uses an internal Cartesian grid that adapts
automatically to the plume shape (explicit method). At the end of each weather sequence, theresult on the Cartesian grid is converted to the result output grid (polar).

3.12

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
Site specific data up to five km, combined with generic data for the Netherlands, which isadequate given the size and flat topography of the Netherlands.

3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic parametersof the ATD vary within this resolution?
No
3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?
In principle, yes, should aqueous transport result in significant doses to the public. However,usually aqueous transport and dispersion can be screened out for the required risk metrics if aplant has the potential of causing airborne releases.

Protective Action (Countermeasure) Modelling
3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
3.14.1. Early/Emergency Phase:
3.14.2. Intermediate Phase:
3.14.3. Late/Recovery Phase:
Direct offsite protective actions such as evacuation, sheltering and iodine prophylaxis are notconsidered in the PSA-3. Decontamination of people (after 24 hours) and indirect actions such as
food control measures, flushing of contaminated water and protection of drink water areas areallowed to be taken into account. Onsite protective actions should be part of the PSA-2 modelling(SAMGs).

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3.15

│ 141

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?
As mentioned above, direct offsite protective actions are not to be included in the risk calculation.
The risk should be calculated for the most vulnerable group that is still identifiable as a group. In
practice, this is almost always the group of 1-year old infants.

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.
No

3.17

What information sources do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?
As mentioned above, no credit is taken for direct offsite protective actions. As such, this type ofdata is not needed for the risk metrics used in the PSA. Decontamination is assumed to take place
24 hours after exposure and food control measures are assumed to be taken should foodcontamination be above the legal limits.

Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite (e.g.,worker) populations or potentially sensitive, vulnerable, or critical groups? If you do, what isyour basis for including these groups, and how are they modelled?
Onsite exposures are not considered. Only individuals outside the perimeter of the nuclear
installation are considered. For the individual risk, critical groups are considered. The legal risk
level must be met for each member of a hypothetical group of people. Guidance states that themost critical group that can still be considered a group has to be considered. In practice this is 1year old infants.
For the group risk (CCDF of the number of early fatalities) real data for the number of
inhabitants, workers (in nearby companies) and other volatile population has to be included. this
implies for example that at night the population number is different from the numbers during
day time.

3.19

What exposure pathways do you model? What is your basis for selecting these pathways?
What exposure duration is assumed in your models?
All exposure pathways that are relevant must be included. This includes the 'usual' pathways
cloud-dose, ground-dose, inhalation dose, ingestion dose, resuspension dose. For research
reactors, we found typically that 'external dose' from the contamination inside the reactor
building is dominating the dose rate at the reactor site boundary, since the site of a research
reactor is usually relatively small (e.g. a university or a research site with more than one research'companies'.
The external radiation requires a separate calculation with other tools, such as e.g. micro Shieldor MCNP.
The exposure duration is 50 to 70 (for children) years.

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3.20

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?
All site independent data are prescribed in the legal documents; the data are age specific. Sitedependent data have to be applied. Some site data are difficult to assess and generic values areprovided, at the cost of increasing the conservatism. For example, we use a generic aerosol size
distribution since aerosol size distributions for the releases are not provided by the level-2 PSA.

Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.
Yes

3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?
For early health effects, the dose to the red bone marrow, lungs, GI-tract and skin is assessed.
For stochastic effects the ICRP effective dose is calculated and an age-dependent (linear) risk
factor is used.

3.23

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, andwhat is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?
𝑟 = (1 − 𝑒𝑣−ln(2)�𝐷�𝐿𝐷 �50

LD50 are taken from ICRP, threshold have been reported by ICRP and UNSCEAR. The shapefactor v is calculated using the LD50 and the T, such that r=1% if the dose is equal to thethreshold in the above hazard function.
For latent health effects, for the probability of mortality a risk factor of 5% per Sv is assumed for
adults and 15% per Sv for children, based on ICRP-103 and ICRP-60 as well as UNSCEAR
2012.

3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?
We have a guideline that provides in acceptable parameter values. The values are based on ICRPand UNSCEAR publications. For the early health effects, averaged data are provided; for thestochastic health effects, age-dependent data are provided

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Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.
Economic consequences are not assessed by default in the PSA-3. For selected source terms,
assessments have been made on request of the plant owner. Furthermore, the ‘group risk’ (see2.2.2) can be considered as a measure of the societal impact of a mayor accident.

3.26

With regard to potential economic consequences considered within a Level 3 PSA:
3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
To estimate the costs, areas with high contamination are distinguished, and the costs of
temporarily abandoning (e.g. if a nearby plant is abandoned it costs 'n' million euros perday) this area and cleaning it are calculated by very rough estimates.

3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.
We have no experience, for the source terms we considered, the major costs where in thetemporarily closing of nearby plants.

3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.
The costs are based on yearly turnover data and other rough estimates.

3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?
Regional maps, annual reports of companies and generic data from literature.

Consequence Quantification and Reporting
3.28

What process and criteria do you use for selecting consequence metrics for quantification andreporting, including spatial intervals or distances from release points?
There are well defined, legal endpoints. In addition, licensees provide contour plots on
geographic maps, and statistics of distances up unto which effects of interest can occur.

Risk Characterization
3.29

What process and criteria do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?
PSA-3 is used to show compliance to legal limits. There have been attempts to develop additional
metrics, but it is difficult to find natural reference values, acceptable probability levels, to be able
to frame a new metric.

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3.30

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results? How do you present the resultsfrom low-probability/high-consequence events in an understandable context for each group?
Who is responsible for communicating the results to each group?
(1) Decision makers:(2) General Public:
The individual risk is a single number. However, often a graph is included to show the individual
risk as a factor of the distance from the facility.
The group risk is shown as a graph with on the Y-axis the probability and on the X-axis thenumber of direct casualties (log-log scale). A line representing the requirement (prob <10-5 peryear for 10 direct casualties, and n2 times lower probability for n times more casualties) is
included as a reference.
The licensee/applicant is responsible for communicating the results to the decision makers (theregulatory body). The decision makers (ANVS) is responsible for communicating their decisions
to the general public, based on the information of the licensee/applicant. However,communication to the public is also done by the licensee/applicant as part of the environmental
impact assessment and the license application, as this information is made available to the public
to facilitate public participation.
In The Netherlands, all hazardous (industrial) activities, including non-nuclear activities, have tocomply with the legal risk criteria as part of the licensing process. Although for each site the risks
are addressed separately by the site owner in a safety report, the authorities have compiled allrisk information from the safety reports in the Dutch risk map ("Risicokaart"), which is publicly
accessible online. For example, a map of the risk levels from all hazardous activities (nuclear andnon-nuclear) within 25 km from the nuclear research site in Petten can be found at:
http://www.risicokaart.nl/en/voorvertoning?rd=98753.33-526424.58-118426.73539499.8&zoom=10This figure is provided below as an example display.

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3.31

│ 145

With regard to the treatment of uncertainties:
3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?
We provide a qualitative estimate of the accuracy of the provided risk numbers. Basedon the EC-US NRC uncertainty study [1993 - 2000] we claim that the accuracy is about
one order of magnitude.
We would be interested to acquire (qualitative) uncertainty estimates for the full PSA,
not the consequence assessment on its own.

3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
Not

3.31.3. To what extent do you consider the effect of correlation on parameters?
Not

Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?
We are most interested in:- an overview of international attention for level-3- an international cooperation to calibrate our PSA-3 skills- (new) risk metrics and communication- the qualitative overall accuracy of the full PSA results

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Sweden: SSM

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PART 1: RESPONDENT INFORMATIONName:Country:
Organization:
Type of Organization:

Per Hellström
SwedenSSM
x Regulatory Authority
� Utility� Vendor� Academic/Research� Other (please specify): ___________________________________

Mailing Address:

SE-171
Solna strandväg 96
per.hellstrom@ssm.se+46 8 799 4038

E-mail Address:
Telephone Number:

16

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Stockholm

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PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes,for what types of applications are Level 3 PSAs required?
Level 3 PSA is not a regulatory requirement. PSA level 2 is required. The safety analysis reportsby the utilities also have sections on radiological consequences in case of accidents (incidents)
presenting standard radiological consequence analyses for a set of predefined cases. There are
ongoing discussions on potential level 3 requirement to be implemented in new regulations
currently being developed. One reason for the new regulations project was to consider potentially
new NPPs, where it might have been justified to include level 3 requirements.

2.2

With regard to the use of Level 3 PSA:
2.2.1.

Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
This discussion is ongoing, see text above. One specific issue is what the objective should
be in case of a level 3 PSA requirement, what should be the added value depending on
how such a requirement would be formulated. Questions that are being addressed are“When and how can results from the level 3 PSAs be used and provide additional
insights compared with PSA level 1 and 2?”

2.2.2.

What calculated metrics or results from Level 3 PSAs are used in these applications
and what (if any) requirements, goals, or criteria are used to evaluate these results?
What is the basis for these requirements, goals, or decision criteria?
See previous response.

2.2.3.

If you are not currently performing or considering performing Level 3 PSAs, haveyou previously considered performing them? If you have, what was the basis for yourdecision to not perform Level 3 PSA? Please list any barriers you perceive toperforming and applying Level 3 PSA in your country.
In Sweden, there was in the beginning of the 70ties a study on urban siting of NPPs and acomparison between different power production options. Methods used were similar to
those used in the US at that time. One conclusion was that it is preferred to place NPPs
outside highly populated areas.

2.2.4.

What (if any) alternative methods do you use to estimate offsite public risks
attributable to accidental releases of radiological materials from nuclear installations,and for what types of applications are they used?
As mentioned above, off site radiological consequence analyses are performed for a set of“accidents” and reported in the safety analysis reports. It is necessary to show that theconsequences are below certain limits. One such analysis is for a design release from thescrubber (filtered vented containment) in case of a severe accident. Results are expressedin terms of doses.

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency

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planning or protective action zone. Are Level 3 PSAs used to support the siting of nuclear
installations in your country? If they are, how are they used?
See earlier response with some information on the urban siting study from the early 70ties.

2.4

With regard to nuclear power plant emergency planning zones:
2.4.1. What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
Existing zones are currently being reevaluated. More information can be provided later2.4.2.

What stakeholder groups are involved in the process and what are their respective
responsibilities?
A number of authorities, e.g. local, SSM, MSB (Engelska), more information can beprovided later.

2.4.3.

Under what conditions (if any) can emergency planning zones be reduced insize? Can a Level 3 PSA be used to establish a probabilistic cut off for events thatneed to be considered for emergency planning purposes?
No answer

2.5

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria. Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?
No.

2.6

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA. Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.
Offsite radiological consequence analyses are performed to show that events, conditions andscenarios with certain frequencies meet the related dose criteria. These criteria are the “usual”
criteria for the design of the NPP SSCs with regard to performance and environmental
qualifications.
Certain information on this is collected in the L3 PSA project and in the final report for 2015 text
below from this report that can be made available when completed)

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Table 1 Reference values in Sweden
Frequency range of accident

Source term

Effective dose1

H2
10-2 < fH3
10-4 < f < 10-2
H4
10-6 < f < 10-4

0,001xFILTRA

1 mSv

Equivalent dose,
thyroid2
1 mSv

0,01xFILTRA

10 mSv

10 mSv

0,1xFILTRA

100 mSv

100 mSv

1

Sum of external radiation from radionuclides in air, internal radiation from inhalation(50 years) and external radiation during 30 days from radionuclides on ground.
2
One-year old child, inhalation.
The decision further gives guidance on standardized calculations; it outlines standardized
calculation considerations for dispersion model (Gaussian), source term (total release time is set
to one hour, radioactivity evenly released during this time), release height, plume rise, deposition
velocities and meteorological data (depending on release height). Internal doses from inhalationand effective doses should be shown not to exceed the reference values, without any
countermeasures or normal living factors included.
The decision also outlines considerations for even more realistic calculations, so called adjusted
calculations. Here the release is not restricted to one hour and countermeasures considerations
are included in the assessment.
In the same decision, it is stated that US NRC RG 1.183 [35] should be followed altogether inconservative consequence analysis, except for dose conversion factors and weather in which casethe same conditions as for realistic assessments should be used. Doses should be calculated at a
distance of 200 m from the release point.

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PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES
General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.

Offsite radiological consequence analyses:
LENA is a simple stand-alone tool developed in Sweden during the late 1980s for
emergency preparedness organizations to quickly assess ongoing or postulated accidents.
The program has some limited probabilistic capabilities. The program is still used byemergency response organizations since it does not require significant input or overhead.
There is also ongoing work with a tool for RApid Source Term Pedicion (RASTEP). Thisis financed by SSM.

3.1.2.

Risk characterization:
The consequences are not combined with frequencies

3.2

With regard to the scope of Level 3 PSAs:
3.2.1.

What internal and external accident sequence initiating events and hazards are
considered in developing a Level 3 PSA?
Not applicable

3.2.2.

Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling? If you do, how is thiscorrelation treated?
Not applicable

3.3

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.
Yes, The Nordic PSA group has been financing (together with SSM) a level 3 project. This
project also has links to the Finnish SAFIR program. The project has several parts:
The objective of the project has been to further develop understanding within the Nordic
countries in the field of Level 3 PSA, the scope of its application, its limitations, appropriate risk
metrics, and the overall need and requirements for performing a Level 3 PSA. In the short term
this experience will be valuable for adding quality to Level 1 and 2 PSA. In the longer term, thework will set the foundation for performing a state-of-the-art Level 3 PSA. The final report for2015 is still being edited and will be made available when completed. Part of this work will
include ASME/ANS L3 PSA standards development.

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Project phases
Industrial
SurveyRisk Metrics

Regulations &standards

Finnish pilot
study
Swedishpilot study

Level 3 PSA Practical Guide

Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses? What (if any) difficulties have you encountered in this area?
Not applicable

3.4.2.

If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
Not applicable

3.4.3.

What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process? If you do,
how are they treated?
Not applicable

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3.5

│ 153

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?
Not applicable, however focus on Cs for long term (ground contamination) impacts and noble
gases and Iodine for short term impacts.

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated?
The individual plant analyses do not currently consider multiple sources.

3.7

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?
Not applicable.

Meteorological Data
3.8

How do you account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?
Not applicable. However, radiological consequence analysis uses conservative assumptions.

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?
See previous answer.

Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?
Not applicable

3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?
Not applicable.

3.12

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
Not applicable.

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3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic parametersof the ATD vary within this resolution?
Not applicable

3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?
Not applicable, however radiological consequence analyses consider gas and water release thattogether makes up a certain source term released at specific elevations.

Protective Action (Countermeasure) Modelling
3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
3.14.1. Early/Emergency Phase:
3.14.2. Intermediate Phase:
3.14.3. Late/Recovery Phase:
Not applicable

3.15

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?
Not applicable

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.
Not applicable

3.17

What information sources do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?
Not applicable

Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite(e.g., worker) populations or potentially sensitive, vulnerable, or critical groups? If you do,what is your basis for including these groups, and how are they modelled?
Not applicable, however, the radiological consequence analysis use a set of standard groups.

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3.19

│ 155

What exposure pathways do you model? What is your basis for selecting these pathways?
What exposure duration is assumed in your models?
Not applicable, however the radiological consequence analysis takes into account several
pathways, e.g. direct radiation, from ground contamination and shining (clouds)

3.20

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?
Radiological consequence analysis use generic factors (NRC?)

Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.
Not applicable

3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?

3.23

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, andwhat is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?

3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?

Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.
Not applicable, the urban siting study from the 70ties focused on the public and early and latefatalities.

3.26

With regard to potential economic consequences considered within a Level 3 PSA:
3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.

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3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.
Not applicable

3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?
Not applicable

Consequence Quantification and Reporting
3.28

What process and criteria do you use for selecting consequence metrics for quantification andreporting, including spatial intervals or distances from release points?
Not applicable

Risk Characterization
3.29

What process and criteria do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?
Not applicable

3.30

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results? How do you present the resultsfrom low-probability/high-consequence events in an understandable context for each group?
Who is responsible for communicating the results to each group?
(1) Decision makers:(2) General Public:
Not applicable

3.31

With regard to the treatment of uncertainties:
3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?
3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
3.31.3. To what extent do you consider the effect of correlation on parameters?
Not applicable

Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?

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PART 1: RESPONDENT INFORMATIONName:Country:
Organization:
Type of Organization:

Rainer Hausherr
Switzerland
Swiss Federal Nuclear Safety Inspectorate, ENSIX Regulatory Authority
� Utility� Vendor� Academic/Research� Other (please specify): ___________________________________

Mailing Address:

Swiss Federal Nuclear Safety Inspectorate, ENSI
Industriestrasse 19
CH-5200 Brugg
Switzerland
Rainer.Hausherr@ensi.ch+41 56 460 86 15

E-mail Address:
Telephone Number:

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PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes,for what types of applications are Level 3 PSAs required?
ENSI: There is no regulatory requirement to perform a Level 3 PSA.

2.2

With regard to the use of Level 3 PSA:
2.2.1.

Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
ENSI: No2.2.2.

What calculated metrics or results from Level 3 PSAs are used in these applications
and what (if any) requirements, goals, or criteria are used to evaluate these results?
What is the basis for these requirements, goals, or decision criteria?
ENSI: not applicable

2.2.3.

If you are not currently performing or considering performing Level 3 PSAs, haveyou previously considered performing them? If you have, what was the basis for yourdecision to not perform Level 3 PSA?
ENSI: The legal requirements do not ask for a Level 3 PSA. The Level 3 PSA isconsidered less effective to identify specific plant safety improvements than a Level 1 or
2 PSA since the level 3 PSA incorporates phenomena like wind direction which increases
the uncertainty of the results and hinder the identification of specific plant potential
improvements from the PSA point of view.

Please list any barriers you perceive to performing and applying Level 3 PSA in yourcountry.
ENSI: There are no legal requirements to perform a Level 3 PSA.

2.2.4.

What (if any) alternative methods do you use to estimate offsite public risks
attributable to accidental releases of radiological materials from nuclear installations,and for what types of applications are they used?
ENSI: The regulatory guideline ENSI-A05 requires the computation of two Level 2 PSA
risk metrics:• Large early release frequency (LERF): The LERF addresses the risk of a large
release (more than 2∙1015 Bq I-131) within the first 10 hours after core damage.
According to the Swiss emergency organization concept, a major part of theaffected population would be protected against immediate effects of the accident
within 5 hours. To consider uncertainties a time frame of 10 hours is used for
LERF.
• Large release frequency (LRF): The LRF is defined as the number of events per
year with a release of more than 2∙1014 Cs-137. Long-term soil contamination iscaused by releases of nuclides with long half-life periods. Cs-137 with a half-life ofabout 30 years is the most representative of these nuclides.

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency

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planning or protective action zone. Are Level 3 PSAs used to support the siting of nuclear
installations in your country? If they are, how are they used?
ENSI: Switzerland has decided to phase out nuclear power, so siting of new NPPs is not an issue.
After the Fukushima accident, Switzerland revisited the emergency plans and the protective
action zones. In a first step, the NPPs had to calculate the frequency of exceeding various source
terms. However, in the end it was decided that PSA insights shall not be used for this revision but
to base the revision on worst case scenarios.

2.4

With regard to nuclear power plant emergency planning zones:
2.4.1.

What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
ENSI: The bases for the emergency planning zone concept were established in the late
70s and beginning of the 80s of last century. The few PSA-analysis results (WASH-1400)
that were available at that time were taken into account. The boundary of the inner zone
(zone 1 that is the PAZ) was set on the basis of a radiological criteria, the outerboundary of the adjacent zone 2 (corresponding to the UPZ) was set using engineering
judgement. A heightened preparedness compared to the levels of preparedness elsewherein the country was not deemed necessary beyond the outer boundary of zone 2. The
radiuses of the zones were not changed since their establishment.

2.4.2.

What stakeholder groups are involved in the process and what are their respective
responsibilities?
ENSI: All levels are involved to various degrees, the federal instances, the cantons, thecommunes and the licensees.

2.4.3.

Under what conditions (if any) can emergency planning zones be reduced insize? Can a Level 3 PSA be used to establish a probabilistic cut off for events thatneed to be considered for emergency planning purposes?
ENSI: As there is no internationally agreed cut off or risk beyond which increased
preparedness is not deemed reasonable, a reduction of zone sizes is politically notenforceable.

2.5

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria. Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?
ENSI: No

2.6

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA. Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.
ENSI: For design basis accidents, the licensees have to assess doses to individuals located atcertain distances of the plant for specified periods of time after a postulated fission productrelease.

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PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES
General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.

Offsite radiological consequence analyses:
ENSI: Codes used depend on the purpose they are used for. For design basis accidents
and the purpose of verifying that the resulting doses remain below levels set by theradiation protection ordinance, a Gaussian plume model with a constant weather
pattern (Pasquill-Gifford classes with fixed wind direction and velocity) and flat terrain
is used. The aim is to provide a conservative assessment of the consequences; details canbe found in the regulatory guideline ENSI-G14. For emergency response purposes, acomplex Lagrangian model called LASAT using periodically updated numerical weather
prognosis input from MeteoSwiss as well as complex terrain is used instead. This coderuns within the decision support tool JRODOS and aims to provide a best-estimate
prognosis as a basis for suggesting protective actions to the decision makers.

3.1.2.

Risk characterization:
ENSI: Not applicable

3.2

With regard to the scope of Level 3 PSAs:
3.2.1. What internal and external accident sequence initiating events and hazards are
considered in developing a Level 3 PSA?
ENSI: Not applicable

3.2.2.

Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling? If you do, how is thiscorrelation treated?
ENSI: Not applicable

3.3

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.
ENSI: The system used for emergency purposes (JRODOS/LASAT) is used by several othercountries as well. Research and development is ongoing with Switzerland participating mainly insteering and funding. The code used for analysing radiological DBA consequences is mature andhas experienced few if any developments for years.

Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses? What (if any) difficulties have you encountered in this area?
ENSI: Not applicable

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3.4.2.

If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
ENSI: Similarities regarding accident progression and source term, at least containment
status (open, vented, isolated, non-isolated, bypassed, ruptured, base mat penetrated),
time of release (early or late), mode of ex-vessel releases (dry or submerged core concreteinteraction), and containment fission product removal mechanisms (scrubbing bycontainment spray or by an overlaying water pool). The source terms are usuallycalculated by using MELCOR. Sometimes there are different calculations available for arelease category, allowing the definition of a mean source term for this release category.
Generally, no uncertainty analysis regarding source terms is required.

3.4.3.

What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process? If you do,
how are they treated?
ENSI: Release fraction timing is done by using the corresponding information of the
MELCOR calculations. The PSAs consider accident mitigation actions, among them,filtered containment venting and containment spray. For example, filtered venting is
activated by rupture of a burst disc at a certain containment pressure.

3.5

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?
ENSI: In the Level 2 PSA, at least the following radiological groups shall be considered: Xe, I
including CsI, Cs, Te, Ba, Mo, Ru, Ce, and La.
For the computation of radiological consequences for emergency purposes up to 100 nuclides areused as set by an ENSI directive. The selection of a set of nuclides depends on the goal aimed at;in an emergency case fewer nuclides are considered, depending on knowledge of the situation andavailable simulation run time.

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated?
ENSI: There is only one twin unit site in Switzerland. In an emergency, the tool used foratmospheric dispersion calculations JRODOS is able to simulate the effects of multiple sources atdifferent heights released in parallel and with different time structure.

3.7

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?
ENSI: No.

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Meteorological Data
3.8

How do you account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?
ENSI: As no PSA-L3 are performed in Switzerland, no such variations are performed.

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?
ENSI: No.

Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?
ENSI: (1) see 3.1.1. (2) In an emergency, as soon as new and/or more detailed source term
information is available (potentially having an impact on the dose prognosis) a new simulation
will be performed. As for time dependence, our numerical weather prognosis (NWP) data has atime resolution of 10 min and dispersion simulations are calculated with this time step by default.
Every three hours we receive a new set of NWP data covering the next 24 hrs. Thus, we have asound data basis for performing dispersion calculations and simulate to sufficient degree any
time variability of a source term.

3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?
ENSI: The NWP data (see 3.10) obtained from MeteoSwiss covers an area of 300 km x 300 km.
In case of an emergency, the first simulations will inevitably be performed on the entire data
area. Subsequent dispersion calculations may be restricted to the area of interest, depending on
the meteorological situation and the results of the previous simulations. On the technical side, thesimulation grid can be configured to use up to fivefold nesting with horizontal resolution reducedby half from one nesting step to the next (obviously, the innermost nesting step has highest
resolution). The grid cell size of the innermost nesting can be chosen from a series of values,
between 50 m and 1 km with default being 250 m. Vertical level spacing is taken from input NWP
data.

3.12

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
ENSI: Not applicable

3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic parametersof the ATD vary within this resolution?
ENSI: Not applicable

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3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?
ENSI: For emergency purposes releases to water sources are only considered within theframework of verification of compliance with maximum releases allowed for normal operations.
In emergency situations, aqueous transport can only be estimated grossly based on rules of
thumb.

Protective Action (Countermeasure) Modelling
3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
3.14.1. Early/Emergency Phase:
3.14.2. Intermediate Phase:
3.14.3. Late/Recovery Phase:
ENSI: Not applicable

3.15

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?
ENSI: Not applicable

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.
ENSI: Not applicable

3.17

What information sources do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?
ENSI: Not applicable

Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite(e.g., worker) populations or potentially sensitive, vulnerable, or critical groups? If you do,what is your basis for including these groups, and how are they modelled?
ENSI: Not applicable

3.19

What exposure pathways do you model? What is your basis for selecting these pathways?
What exposure duration is assumed in your models?
ENSI: Not applicable

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1

3.20

│ 165

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?
ENSI: Not applicable

Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.
ENSI: Not applicable

3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?
ENSI: Not applicable

3.23

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, andwhat is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?
ENSI: Not applicable

3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?
ENSI: Not applicable

Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.
ENSI: No, economic consequences are not considered.

3.26

With regard to potential economic consequences considered within a Level 3 PSA:
3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.
3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.
ENSI: Not applicable

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3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?
ENSI: Not applicable

Consequence Quantification and Reporting
3.28

What process and criteria do you use for selecting consequence metrics for quantification andreporting, including spatial intervals or distances from release points?
ENSI: Not applicable

Risk Characterization
3.29

What process and criteria do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?
ENSI: Not applicable

3.30

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results? How do you present the resultsfrom low-probability/high-consequence events in an understandable context for each group?
Who is responsible for communicating the results to each group?
(1) Decision makers:(2) General Public:
ENSI: Not applicable

3.31

With regard to the treatment of uncertainties:
3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?
3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
3.31.3. To what extent do you consider the effect of correlation on parameters?
ENSI: Not applicable

Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?
ENSI: NoSTATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1

United States: USNRCSTATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT│ 167

168 │ NEA/CSNI/R(2018)1

PART 1: RESPONDENT INFORMATIONName:Country:
Organization:
Type of Organization:

Keith L. Compton*
United States
Office of Nuclear Regulatory Research, USNRC

Mailing Address:

U.S. Nuclear Regulatory Commission
Mail Stop T10-B58
Washington, DC 20555-0001
Keith.Compton@nrc.gov+1-301-415-0651

E-mail Address:
Telephone Number:x Regulatory Authority
Utility
VendorAcademic/Research
Other (please specify): ___________________________________* The following members of the USNRC technical staff reviewed and supplemented this example survey
response:
Elijah Dickson, Office of Nuclear Reactor Regulation
Daniel Hudson, Office of Nuclear Regulatory Research
Tony Nakanishi, Office of New Reactors
Donald Palmrose, Office of New Reactors
Jason Schaperow, Office of New Reactors
Nathan Siu, Office of Nuclear Regulatory Research

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 169

PART 2: APPLICATION OF LEVEL 3 PSA
2.1

Does your country require the performance of Level 3 PSAs for nuclear installations? If yes,for what types of applications are Level 3 PSAs required?
The term Level 3 PSA can refer to both an integrated Level 1/2/3 PSA1 conducted in accordancewith industry standards, or simply to the Level 3 offsite radiological probabilistic consequence
analysis (PCA) element of such an assessment. Some of the applications involving PCA that arecited in this example survey response were not conducted as part of an integrated Level 1/2/3
PSA, but the offsite radiological consequence analysis was performed using methodologies
substantially similar to those of a Level 3 PSA. In this example survey response, the acronym
PCA will refer to either the Level 3 offsite radiological PCA element of an integrated Level 1/2/3
PSA or to an offsite radiological PCA performed for other purposes; the term Level 3 PSA willrefer to an integrated Level 1/2/3 PSA.
Level 3 PSA is not a regulatory requirement for licensing of U.S. commercial nuclear power
plants. However, information about the offsite consequences of severe accidents is required forcompliance with environmental assessment requirements specified in national-level legislative
statutes,2 and this information is often generated using PCA. For more information, see theresponse to item 2.2.

2.2

With regard to the use of Level 3 PSA:
2.2.1.

Do you perform, or are you considering performing Level 3 PSAs for nuclear
installations? If you do, for what types of applications do you perform them?
The USNRC is currently performing a full-scope integrated site Level 3 PSA for a U.S.
commercial nuclear power plant site as part of a research project that is described in
more detail in the response to item 3.3.3,4
In addition to this and other USNRC-sponsored Level 3 PSAs that have been performed
for as part of research projects (e.g., WASH-14005 and NUREG-11506), information

1

An integrated Level 1/2/3 PSA is an assessment of the offsite public risks attributable to a spectrum of possible accident
scenarios involving a nuclear installation. In the traditional PSA framework for commercial nuclear power plants, such a
Level 1/2/3 PSA includes three progressive levels of analysis: (1) core damage accident or plant damage state frequency
analysis; (2) accident progression, containment performance, and radiological release analysis; and (3) offsite radiologicalconsequence analysis. In this context, the term “integrated” means the three levels of analysis are linked through the useof logic or mathematical models to support development of qualitative insights and quantitative estimates of selected risk
metrics.

2

Title 42 United States Code, §§4321-4370h, “National Environmental Policy Act.” Available at:https://elr.info/sites/default/files/docs/statutes/full/nepa.pdf.
3
Hudson DW, Stutzke M. Options for Proceeding with Future Level 3 Probabilistic Risk Assessment Activities. SECY-110089. Washington, DC: U.S. Nuclear Regulatory Commission; 2011. Available at: http://www.nrc.gov/reading-rm/doccollections/commission/secys/2011/2011-0089scy.pdf.
4Vietti-Cook, AL. Staff Requirements – SECY-11-0089 – Options for Proceeding with Future Level 3 Probabilistic Risk
Assessment (PRA) Activities. Washington, DC: U.S. Nuclear Regulatory Commission; 2011. Available at:
http://pbadupws.nrc.gov/docs/ML1126/ML112640419.pdf.

5

U.S. Nuclear Regulatory Commission. Reactor Safety Study. An Assessment of Accident Risks in U.S. Commercial
Nuclear Power Plants. WASH-1400 (NUREG 75/014). Washington, DC: U.S. Nuclear Regulatory Commission; 1975.

6

Office of Nuclear Regulatory Research. Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants.
NUREG-1150. Washington, DC: U.S. Nuclear Regulatory Commission; 1990. Available at: http://www.nrc.gov/readingrm/doc-collections/nuregs/staff/sr1150/.

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170 │ NEA/CSNI/R(2018)1
from PCA is typically used to assess the offsite radiological consequences of severe or
beyond-design-basis accidents (BDBAs).1 Applications of PCA at the USNRC include:•••

Regulatory2,3 and backbit analyses4,5,6 to support decisions regarding proposed
regulatory actions;
Environmental assessment reviews with respect to severe accidents and Severe
Accident Mitigation Alternatives (SAMA) analyses for operating power reactor
license renewal7 or Severe Accident Mitigation Design Alternatives (SAMDA)
analyses for new power reactor design stage applications;8 andSupporting applied research studies. Recent examples of such studies include:(1) the State-of-the-Art Reactor Consequence Analyses (SOARCA) Project;9 (2) aconsequence study of a beyond-design-basis earthquake affecting the spent fuel pool
for a U.S. Mark I boiling water reactor;10 and (3) an evaluation of alternative
protective action strategies for emergency response to severe accidents.11

1

The USNRC defines a design-basis accident (DBA) as: “a postulated accident that a nuclear facility must be designed andbuilt to withstand without loss to the systems, structures, and components necessary to ensure public health and safety”(http://www.nrc.gov/reading-rm/basic-ref/glossary/design-basis-accident.html). Beyond-design-basis accidents (BDBAs)
are defined as: “accident sequences that are possible but were not fully considered in the design process because they
were judged to be too unlikely.” BDBAs are thus considered to be beyond the scope of DBAs that a nuclear facility must
be designed and built to withstand (http://www.nrc.gov/reading-rm/basic-ref/glossary/beyond-design-basisaccidents.html).

2

Division of Systems Analysis and Operational Effectiveness, Office of Nuclear Regulatory Research. Regulatory Analysis
Guidelines of the U.S. Nuclear Regulatory Commission. NUREG/BR-0058, Revision 4. Washington, DC: U.S. Nuclear
Regulatory
Commission;
2004.
Availableat:
http://www.nrc.gov/reading-rm/doccollections/nuregs/brochures/br0058/br0058r4.pdf.

3

Office of Nuclear Regulatory Research. Regulatory Analysis Technical Evaluation Handbook. NUREG/BR-0184.
Washington, DC: U.S. Nuclear Regulatory Commission; 1997. Available at: http://www.nrc.gov/reading-rm/doccollections/nuregs/brochures/br0058/br0058r4.pdf.

4

A back fit is a generic or plant-specific modification that becomes effective after specified dates. Examples of back fits
include modification of or addition to: (1) facility structures, systems, components, or design; (2) the design approval or
manufacturing license for a facility; or (3) the procedures or organization required to design, construct or operate afacility. Any of these modifications or additions may result from a new or amended provision in USNRC regulations or
the imposition of a regulatory staff position interpreting USNRC regulations that is either new or different from apreviously applicable staff position.11

5

U.S. Nuclear Regulatory Commission. Title 10 Code of Federal Regulations Part 50, §50.109, “Back fitting.” Available
at: http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0109.html.

6

Allison DP, Conran JM, Trottier CA. Back fitting Guidelines. NUREG-1409. Washington, DC: U.S. Nuclear Regulatory
Commission; 1990. Available at: http://pbadupws.nrc.gov/docs/ML0322/ML032230247.pdf.

7

U.S. Nuclear Regulatory Commission. Standard Review Plans for Environmental Reviews for Nuclear Power Plants:
Environmental Standard Review Plan (with Supplement 1 for Operating Reactor License Renewal). NUREG-1555.
Washington, DC: U.S. Nuclear Regulatory Commission; 2000. Available at: http://www.nrc.gov/reading-rm/doccollections/nuregs/staff/sr1555/.

8

U.S. Nuclear Regulatory Commission. Title 10 Code of Federal Regulations Part 51, §51.30, “Environmental
Assessment.” Available at: http://www.nrc.gov/reading-rm/doc-collections/cfr/part051/part051-0030.html.

9

Chang R, Schaperow J, Ghosh T, Barr J, Tinkler C, Stutzke M. State-of-the-Art Reactor Consequence Analyses(SOARCA) Report. NUREG-1935. Washington, DC: U.S. Nuclear Regulatory Commission; 2012. Available at:
http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1935/.

10

Barto A, Chang Y J, Compton K, Esmaili H, Helton D, Murphy A et al. Consequence Study of a Beyond-Design-Basis
Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor. NUREG-2161. Washington, DC: U.S.
Nuclear Regulatory Commission; 2014. Available at: http://pbadupws.nrc.gov/docs/ML1425/ML14255A365.pdf.

11

Sullivan R, Jones J, Schelling FJ, Bixler N, Walton F. Review of NUREG-0654, Supplement 3, “Criteria for Protective
Action Recommendations for Severe Accidents”: Technical Basis for Protective Action Strategies. NUREG/CR-6953,
Vol. 3.
Washington,
DC:
U.S.
Nuclear
Regulatory
Commission;
2010.
Availableat:
http://pbadupws.nrc.gov/docs/ML1023/ML102380087.pdf.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 171

PCAs for design basis accidents (DBAs) are more prescriptive than PCAs for BDBAs.
For example, U.S. regulations require offsite radiological consequence analyses for DBAs
to assess doses to individuals located at any point along the boundaries of specified areas
around a nuclear power plant for specified periods of time following postulated fission
product releases.1,2

2.2.2.

What calculated metrics or results from Level 3 PSAs are used in these applications
and what (if any) requirements, goals, or criteria are used to evaluate these results?
What is the basis for these requirements, goals, or decision criteria?
A high-level summary response to this survey item is provided below inTable 1. Additional information about application-specific metrics, criteria, and theirbases is provided in a more detailed response after the table and figure.
Table 1. Application-Specific Level 3 PSA Metrics and Criteria

Application
Safety Goal Screening Evaluation
for Regulatory and Backfit
AnalysesRegulatory and Backfit Analyses

Level 3 PSA Metric
Criteria
Change in Core Damage Frequencya
Conditional
Containment
Failure See Figure 1
Probabilityb
Averted Population Dose (0-50 miles)
Averted Offsite Property Damage (0-50 Net value
SAMA and SAMDA Analyses
miles)a
The USNRC uses a subsidiary numerical objective related to Level 1 PSA core damage frequency(CDF) results as a surrogate for a quantitative health [effects] objective related to average individual
latent cancer fatality risk within 10 miles of a commercial nuclear power plant.
b
The USNRC uses a subsidiary numerical objective related to Level 2 PSA large early releasefrequency (LERF) results as a surrogate for a quantitative health [effects] objective related toaverage individual early fatality risk within 1 mile of a commercial nuclear power plant.

Figure 1. Safety Goal Screening Criteria for Regulatory and Back fit Analyses8

1

U.S. Nuclear Regulatory Commission. Title 10 Code of Federal Regulations Part 50, §50.34, “Contents of Applications;Technical Information.” Available at: http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0034.html.

2

U.S. Nuclear Regulatory Commission. Title 10 Code of Federal Regulations Part 50, §50.67, “Accident Source Term.”
Available at: http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0067.html.

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172 │ NEA/CSNI/R(2018)1
The primary analytical technique used to evaluate alternatives in regulatory, SAMA,
and SAMDA analyses is cost-benefit analysis. In these analyses, the principal metriccalculated to support decision making is the net present value of net benefits (net value),
which represents the difference between the sum of monetized and discounted benefits
and the sum of monetized and discounted costs. However, a number of decision criteria
can be used to select and support an appropriate course of action. For example, decision
criteria that can be used in regulatory analyses include: (1) calculated net value;(2) relative importance of attributes quantified in other than monetary terms;(3) relative importance of qualitative attributes (attributes that are not quantified);(4) relationship and consistency of proposed alternatives with existing USNRC legislative
mandate, policies, and processes; and (5) impacts of proposed alternatives on existing orplanned USNRC programs or requirements.8
As part of regulatory, SAMA, or SAMDA analyses, results from Level 3 PSAs can beused to evaluate and screen proposed alternatives based on the magnitude of theestimated benefit in terms of enhancing safety or reducing risk. For example, regulatory
analysis guidelines8 include guidance for performing an evaluation of proposed
regulatory actions with respect to the USNRC safety goals for nuclear power plant
operations.1 This safety goal evaluation is designed to identify when a regulatory
requirement should not be imposed generically on nuclear power plants because theresidual risk is already acceptably low; it is intended to eliminate some proposed
regulatory actions from further consideration, regardless of whether they could bejustified on the basis of their net value. This safety goal evaluation can also be used todetermine whether a proposed generic safety enhancement back fit that does not meetcertain exemption criteria provides a substantial increase in the overall protection of
public health and safety11 to warrant further evaluation of the benefits and costs todetermine whether they are justified on their net value basis.8
Regulatory analysis guidelines include explicit safety goal screening criteria related to:(1) changes in the frequency of core damage accidents; and (2) conditional containment
failure probabilities. These criteria - which are intended to provide a balanced
consideration of measures to prevent and mitigate core damage accidents - can be used
to evaluate results from Level 1 and Level 2 PSAs to determine conformity withsubsidiary safety goal objectives based on core damage frequency (CDF) and large early
release frequency (LERF).8,9 Although these guidelines do not include explicit screening
criteria related to the safety goal quantitative health [effects] objectives (QHOs) for
average individual early fatality risk and average individual latent cancer fatality risk,
results from Level 3 PSAs for these metrics can be used to evaluate proposed regulatory
actions with respect to the safety goals. Additional information about the basis for thesafety goals and the QHOs is provided in the response to item 2.5.
For those proposed regulatory actions that pass the safety goal screening evaluation, adetailed cost-benefit analysis is performed to estimate the net value. The principal
outputs from a PCA or Level 3 PSA that serve as inputs to the cost-benefit analysis are
the averted population dose (collective effective dose) and averted economic costs,
including offsite property damage. Although these metrics are typically calculated forthe offsite region within 50 miles of the nuclear installation, other spatial intervals can beused on a case-by-case basis.8,9 The averted population dose is then monetized using adollar per person-rem conversion factor that combines the value of a statistical life witha nominal risk coefficient for stochastic health effects per person-rem. The USNRC is1

U.S. Nuclear Regulatory Commission. Safety Goals for the Operations of Nuclear Power Plants; Policy Statement;
Republication. 51 FR 30028. Washington, DC: U.S. Nuclear Regulatory Commission; 1986. Available at:
http://www.nrc.gov/reading-rm/doc-collections/commission/policy/51fr30028.pdf.

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currently revising the value of this dollar per person-rem conversion factor and is
proposing a process for routine updates to ensure its value is maintained current.1

2.2.3.

If you are not currently performing or considering performing Level 3 PSAs, haveyou previously considered performing them? If you have, what was the basis for yourdecision to not apply Level 3 PSA? Please list any barriers you perceive toperforming and applying Level 3 PSA in your country.
Not applicable.

2.2.4.

What (if any) alternative methods do you use to estimate offsite public risks
attributable to accidental releases of radiological materials from nuclear installations,and for what types of applications are they used?
Not applicable.

2.3

One application that could potentially benefit from Level 3 PSAs is the siting of nuclear
installations, including establishing the size and boundary (shape) for each emergency
planning or protective action zone. Are Level 3 PSAs used to support the siting of nuclear
installations in your country? If they are, how are they used?
A Level 3 PSA is not required to support siting of nuclear power plants. However, U.S.
regulations require results from prescriptive DBA dose assessments to demonstrate that doses toindividuals located at any point along the boundaries of specified areas around a nuclear power
plant for specified periods of time following postulated fission product releases are withinprescribed limits.18,19 Additional regulations require development of an environmental impact
statement (EIS) to support siting decisions for specified types of nuclear installations.2 The EIS
includes an assessment of the offsite radiological consequences for BDBAs; either a PCA or Level
3 PSA can be used to provide such an assessment.

2.4

With regard to nuclear power plant emergency planning zones:
2.4.1.

What process is used to establish sizes and boundaries for emergency planning zones
around a nuclear installation in which arrangements shall be made at the preparedness
stage for effectively taking protective and other response actions?
A task force comprised of representatives from the USNRC and the U.S. Environmental
Protection Agency (USEPA) developed the process and technical basis for establishing
the sizes of generic emergency planning zones (EPZs) around U.S. nuclear power plants.3
U.S. regulations require nuclear power plant licensees to develop an emergency response
plan that specifies a range of preplanned protective actions that can be taken withinthese EPZs to reduce dose to emergency workers and to the public in the event of an
accident.4

1

Office of Nuclear Reactor Regulation. Reassessment of NRC's Dollar Per Person-Rem Conversion Factor Policy: Draft
Report for Comment. NUREG-1530, Revision 1. Washington, DC: U.S. Nuclear Regulatory Commission; 2015. Available
at: http://pbadupws.nrc.gov/docs/ML1523/ML15237A211.pdf.

2

U.S. Nuclear Regulatory Commission. Title 10 Code of Federal Regulations Part 51, §51.20, “Criteria for and
Identification of Licensing and Regulatory Actions Requiring Environmental Impact Statements.” Available at:
http://www.nrc.gov/reading-rm/doc-collections/cfr/part051/part051-0020.html.

3

Collins HE, Grimes BK, Galpin F. Planning Basis for the Development of State and Local Government Radiological
Emergency Response Plans in Support of Light Water Nuclear Power Plants. NUREG-0396. Washington, DC: U.S.
Nuclear Regulatory Commission; 1978. Available at: http://pbadupws.nrc.gov/docs/ML0513/ML051390356.pdf.

4

U.S. Nuclear Regulatory Commission. Title 10 Code of Federal Regulations Part 50, §50.47, “Emergency Plans”
Available at: http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0047.html.

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These regulations define two generic EPZs for nuclear power plants that are based on
the recommendations of the USNRC and USEPA task force. In general, the plume
exposure pathway EPZ for large light-water reactor (LWR) nuclear power plants shallconsist of an area about 10 miles (16 km) in radius and the ingestion pathway EPZ shallconsist of an area about 50 miles (80 km) in radius. However, the exact size andconfiguration of the EPZs surrounding specific nuclear power plants are determinedbased on consideration of local emergency response needs and capabilities, which can beaffected by many factors, including: demography, topography, land characteristics,
access routes, and jurisdictional boundaries. The sizes of EPZs for gas-cooled reactors
and for reactors with an authorized power level less than 250 MW thermal may be
determined on a case-by-case basis.24
The USNRC and USEPA task force considered several possible rationales in developing
the sizes for the existing generic EPZs, including: risk, probability, cost-effectiveness,
and accident consequence spectrum. After evaluation of each approach, the task force
decided to use the spectrum of possible accident consequences for different classes ofnuclear accidents as the principal technical basis for emergency planning. The task force
also chose to augment this principal technical basis with consideration of theprobabilities of accidents that could result in offsite doses that exceed USEPA Protective
Action Guide (PAG) levels1,2 and that therefore could require emergency response.23

2.4.2.

What stakeholder groups are involved in the process and what are their respective
responsibilities?
The stakeholder groups involved in the process of establishing sizes and boundaries for
EPZs around a nuclear installation, as well as their respective responsibilities, aresummarized below in Table 2.

Table 2. Stakeholder Groups and Responsibilities in Establishing EPZ Sizes and Boundaries

1

Office of Radiation Programs. Manual of Protective Action Guides and Protective Actions for Nuclear Incidents. EPA
400-R-92-001. Second Printing. Washington, DC: U.S. Environmental Protection Agency; 1992. Available at:https://www.epa.gov/sites/production/files/2014-11/documents/00000173.pdf.

2

U.S. Environmental Protection Agency. PAG Manual: Protective Action Guides and Planning Guidance for Radiological
Incidents. Draft for Interim Use and Public Comment. Washington, DC: U.S. Environmental Protection Agency; 2013.
Available at: https://www.epa.gov/sites/production/files/2015-06/documents/pag-manual-interim-public-comment-4-22013.pdf.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 175

Stakeholder Group

Responsibilities• Establishes federal radiation protection standards.
U.S. Environmental Protection Agency• Establishes PAG levels used in developing EPZ basis.
(USEPA)• Participated in joint USEPA/USNRC task force that
established technical basis for existing EPZs.
• Establishes regulatory requirements for onsite
radiological emergency preparedness and emergency
plans.
U.S. Nuclear Regulatory Commission• Ensures
adequacy ofradiologicalemergency(USNRC)
preparedness and emergency plans and issues operating
licenses.
• Participated in joint USEPA/USNRC task force that
established technical basis for existing EPZs.
• Establishes regulatory requirements for offsite
Federal Emergency Management
radiological emergency preparedness and emergency
Agency (FEMA)
plans.
• Develop and implement onsite radiological emergency
Nuclear Power Plant Licensees
preparedness and emergency plans.
• Develop and implement offsite radiological emergency
State, Local, and Tribal Governments
preparedness and emergency plans.
• Represent respective interests and perspectives anda
Otherprovide valuable input to the process.
a
Examples of other stakeholder groups include, but are not limited to: (1) industry organizations;(2) public interest groups; (3) research organizations; (4) local emergency response organizations;and (5) public citizens.

2.4.3.

Under what conditions (if any) can emergency planning zones be reduced insize? Can a Level 3 PSA be used to establish a probabilistic cut off for events thatneed to be considered for emergency planning purposes?
The USNRC has not established criteria for reducing the size of EPZs for large LWR
nuclear power plants. However, USNRC regulations specify that the sizes of EPZs for
gas-cooled reactors and for reactors with an authorized power level less than 250 MW
thermal may be determined on a case-by-case basis.24
Results and insights from the WASH-14005 Level 3 PSAs were used to provide theprobabilistic perspective that informed the task force’s recommendations with respect tothe sizes of the existing generic EPZs. Although a probabilistic cut off was not explicitly
specified for accidents that need to be considered for emergency planning purposes, thetask force concluded that: (1) DBAs and less severe but more likely core damage
accidents should be considered when selecting the technical basis; and (2) certain
features of more severe but less likely core damage accidents should be considered toassure sufficient response capability exists for even the most severe accidents.23

2.5

Another application that could potentially benefit from Level 3 PSAs is the development ofsafety goals or risk acceptance criteria. Are Level 3 PSAs used to support this type of
application in your country? If they are, how are they used?
The USNRC safety goals and corresponding QHOs were developed in the 1980’s andfundamentally reflect a value judgment about the level of incremental (additional) risk to the
public from nuclear power plant operations relative to background risk that was judged to be
acceptable. Although results and insights from available Level 3 PSAs were not explicitly used toderive the safety goals and QHOs, they were used to determine: (1) the practicability of

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measuring attainment of the QHOs; and (2) the need for consideration of uncertainty inevaluating the results from Level 3 PSAs in safety goal evaluations.
Additional information about the USNRC safety goals and QHOs, as well as the role of Level 3
PSAs in their development and implementation, is provided below. For a more comprehensive
treatment, please refer to the 2009 WGRISK report on probabilistic risk criteria and safety
goals.1
The USNRC safety goal policy statement20 broadly defines an acceptable level of radiological risk
to public health and safety from potential releases of radiological materials from commercial
nuclear power plants.2 This policy essentially addresses the question of “how safe is safeenough?” for regulatory decisions regarding nuclear power plant safety and guides agency
evaluations of whether additional safety enhancements beyond those needed to ensure adequate
protection are no longer justified because the level of residual risk to the public is acceptably low
and limited resources could be better applied to other areas.
The USNRC used a hierarchical framework in establishing its safety goal policy. This framework
is comprised of two high-level qualitative safety goals; one goal encompasses risks to individual
members of the public, while the other addresses societal risk. Each qualitative safety goal is
supported by a QHO that can be used to determine whether and to what extent the qualitative
safety goals have been achieved; one QHO is for individual early (prompt) fatality risk, while theother is for individual latent cancer fatality risk.
Although results and insights from the WASH-1400 study5 and from other industry-sponsoredLevel 3 PSAs were available during the development and trial evaluation of the safety goals, thisinformation was not explicitly used to derive the qualitative safety goals or their corresponding
QHOs. Instead, the safety goals and QHOs reflect a value judgment about the level ofincremental (additional) risk to the public from nuclear power plant operations relative tobackground risk that was judged to be acceptable. Through stakeholder engagement and adeliberative process, the USNRC determined that the risk to an average individual from living
within specified distances of a nuclear power plant should be no more than 0.1% (1/1000) of theother major risks to which they are exposed.3
Results and insights from WASH-14005 and from other available Level 3 PSAs were primarily
used to determine the practicability of measuring attainment of the QHOs and the need forconsideration of uncertainty in evaluating the results from Level 3 PSAs. However, results andinsights from the NUREG-1150 study6 were later used to demonstrate that the subsidiary1

Working Group on Risk Assessment, Committee on the Safety of Nuclear Installations. Probabilistic Risk Criteria and
Safety Goals. NEA/CSNI/R(2009)16. Paris, France: Nuclear Energy Agency; 2009. Available at: https://www.oecdnea.org/nsd/docs/2009/csni-r2009-16.pdf.

2

The scope of the safety goal policy includes risks to the public arising from both routine and accidental radiologicalreleases from commercial nuclear reactors. Public risks imposed by the nuclear fuel cycle were excluded based on earlier
assessments that suggested fuel cycle risks were relatively small in comparison to operating reactor risks. In addition,
EISs performed before existing nuclear power plants had been licensed to operate indicated there would be no measurable
radiological impact on members of the public from routine operations. Moreover, since compliance with nationalradiation protection guidance and USNRC regulations was believed to ensure that public risks arising from routine
emissions were comparatively small, the Commission expressed its belief that such risks need not be routinely analysed
on a case-specific basis to demonstrate conformance with the safety goal policy. The USNRC safety goals therefore apply
primarily to public risks arising from potential accidental releases of radiological materials from operating commercial
nuclear reactors.

3

Office of Policy Evaluation. Safety Goals for Nuclear Power Plant Operation. NUREG-0880, Revision 1. For Comment.
Washington, DC: U.S. Nuclear Regulatory Commission; 1983. Available at:
http://pbadupws.nrc.gov/docs/ML0717/ML071770230.pdf.

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objectives for CDF and LERF are acceptable surrogates for the average individual latent cancer
and early fatality risk QHOs, respectively.1

2.6

Offsite radiological consequence analyses may be performed to support applications that are
not related to Level 3 PSA. Please list any additional applications supported by offsite
radiological consequence analyses that you support or perform.
Offsite radiological consequence analyses are also performed to support incident response
applications. In these applications, dose assessments are performed in real-time to inform or
evaluate offsite protective action recommendations. As described in the response to item 3.1.1,
different computer codes are used to perform offsite radiological consequence analyses or dose
assessments for such non-PSA applications.
In addition to accidental releases of radiological materials, U.S. regulations require dose
assessments for certain routine or off-normal releases to demonstrate compliance with specified
dose limits for members of the public.2 However, the scope of this survey is assumed to be limited
to offsite radiological consequence analyses for accidental releases of radiological materials.

1

Drouin, M. Feasibility Study for a Risk-Informed and Performance-Based Regulatory Structure for Future Plant
Licensing, Volumes 1 and 2. NUREG-1860. Washington, DC: U.S. Nuclear Regulatory Commission; 2007. Available at:
http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1860/.

2

U.S. Nuclear Regulatory Commission. Title 10 Code of Federal Regulations Part 20, §20.1302, “Compliance with Dose
Limits for Individual Members of the Public.” Available at: http://www.nrc.gov/reading-rm/doccollections/cfr/part020/part020-1302.html.

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PART 3: LEVEL 3 PSA MODELING ISSUES AND TECHNICAL CHALLENGES
General3.1

What computer code(s) do you use, or are you developing, to perform the tasks listed below?
3.1.1.

Offsite radiological consequence analyses:
The USNRC uses the MELCOR Accident Consequence Code System (MACCS)1 to
perform PCA or the offsite radiological consequence analysis element of a Level 3 PSA.
Other codes are used to perform offsite radiological consequence or dose assessments for
non-PSA applications. For example, the Radiological Assessment System forConsequence AnaLysis (RASCAL) code2 and the HotSpot Health Physics codes3 support
incident response applications, while the RADionuclide, Transport, Removal, And Doseestimation (RADTRAD)4 code supports DBA dose assessment.

3.1.2.

Risk characterization:
The USNRC does not currently maintain a code specifically designed for performing
risk estimation or risk integration by combining frequency estimates from Level 1 andLevel 2 PSA analyses with conditional consequence metric estimates from Level 3 PSA
analyses to calculate risk metrics. Although such a code was developed to support the
NUREG-1150 Level 3 PSAs,5 the USNRC is planning to use available commercial offthe-shelf software (MATLAB)6 to perform the risk estimation task as part of its ongoing
Full-Scope Integrated Site Level 3 Probabilistic Risk Assessment (PRA) Project;3,4 abrief description of this project is provided in the response to item 3.3.

3.2

With regard to the scope of Level 3 PSAs:
3.2.1.

What internal and external accident sequence initiating event hazards are considered
in developing a Level 3 PSA?
The scope of accident sequence initiating event hazards considered in developing a
Level 3 PSA is determined as part of the initiating-event analysis technical element of theLevel 1 (core damage accident frequency analysis) PSA. Hazard groups that are
typically considered in developing Level 1 PSA models for U.S. nuclear power plants are
summarized below in Table 3.

1

http://maccs.sandia.gov/.

2

https://www.usnrc-ramp.com/RASCAL Overview.

3

https://narac.llnl.gov/hotspot.

4

https://www.usnrc-ramp.com/content/snapradtrad-overview.

5

Iman RL, Johnson JD, Helton JC. PRAMIS: Probabilistic Risk Assessment Model Integration System. NUREG/CR-5262.
Washington, DC: U.S. Nuclear Regulatory Commission; 1990.

6

http://www.mathworks.com/products/matlab/?requestedDomain=www.mathworks.com.

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Table 3. Hazard Groups Typically Considered in Level 1 PSA Models for U.S. Nuclear Power Plants1
Hazard Type

Hazard GroupInternal Events
Internal

Example Hazard or Initiating Events
Loss of Offsite Power (LOOPan internal event by
convention)
Transients
Loss of Coolant Accidents (LOCAs)
Steam Generator Tube Ruptures (SGTRs)
High Energy Line Breaks (HELBs)
Interfacing Systems LOCAs (ISLOCAs)
Special Initiators (e.g. support system failures)

Internal Floods
Internal Fires
Seismic Events
High Winds
External Floods
External
Other Hazards

3.2.2.

External Fires
Transportation Accidents
Other Industrial Accidents
Extreme Temperatures
Turbine-Generated Missiles

Do you account for correlation between causes of accident sequence initiating eventsand offsite phenomenological and consequence modelling? If you do, how is thiscorrelation treated?
For some PCA studies, accident sequences are selected by technical judgment. For
example, accident sequences were selected based on professional judgment using insightsfrom site-specific Level 1/2 PSA models for sites analysed as part of the SOARCA
Project.15
Some correlations between the causes of initiating events and offsite phenomenologicaland consequence modelling are typically considered. For example, the impact of seismic
events on evacuation networks, evacuation speeds, and shielding parameters is typically
explicitly evaluated and modelled for seismically initiated accident sequences.

3.3

Are you participating in any ongoing research and development activities related to offsite
radiological consequence analyses? If you are, please provide a brief description of these
activities.
Full-Scope Integrated Site Level 3 PRA Project3,4
The USNRC is performing a full-scope integrated site Level 3 PSA that aims to achieve thefollowing high-level objectives:•

1

Develop a Level 3 PSA, generally based on current state-of-practice methods, tools, and data,
that: (1) reflects technical advances since completion of previous USNRC-sponsored Level 3
PSA studies; and (2) addresses limitations in the scope of these previous studies - including
consideration of accidents involving multiple site radiological sources;

The American Society of Mechanical Engineers (ASME), American Nuclear Society (ANS). Addenda to ASME/ANS RAS-2008: Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant
Applications. ASME/ANS RA-Sb-2013. New York, NY: The American Society of Mechanical Engineers; 2013.

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180 │ NEA/CSNI/R(2018)1•••

Extract new risk insights to enhance regulatory decision-making and help focus limited
agency resources on issues most directly related to the agency’s mission to protect publichealth and safety;
Enhance PRA staff capability and expertise and improve documentation practices to make
PRA information more accessible, retrievable, and understandable; andObtain insight into the technical feasibility and cost of developing new and morecomprehensive Level 3 PSAs.

The scope of this Level 3 PSA project includes: (1) all major site radiological sources, including
all operating reactor units, spent fuel pools, and dry cask storage installations - with accidents
involving individual sources and accidents involving different combinations of more than one
source included within the scope; (2) all modes of reactor operation; and (3) all internal andexternal hazards, excluding deliberate malevolent acts.
Consistent with the high-level project objectives, the Level 3 PSA is generally being developed
using current state-of-practice methods, tools, and data. However, new research andmethodological development is being pursued to address a limited set of technical issues andchallenges (e.g., development of methods for modelling, quantifying, and analysing multi-source
accident scenarios - including treatment of dependent failures, human reliability analysis,
accident management, multi-source radiological releases, multi-source offsite radiologicalconsequence analysis, and candidate site-level risk metrics for characterizing integrated siterisk).
Additional SOARCA Studies
SOARCA Uncertainty Analysis
Uncertainty analyses are being performed for specific accident scenarios that were evaluated aspart of the original SOARCA study. These analyses are conditioned on the assumed occurrenceof specified conditions in the progression of the modelled accident scenarios. The high-levelobjectives of these conditional uncertainty analyses are to:•••Develop insights into the overall sensitivity of SOARCA results to uncertainty in inputs;
Identify the most influential input parameters for accidental radiological releases andaccident consequences; andDemonstrate the application of an uncertainty analysis methodology that could be used infuture source term, PCA, or Level 3 PSA studies.

The uncertainty analyses involve varying multiple uncertain model parameters using Monte
Carlo sampling of parameter probability distributions. Subject matter experts were consulted todetermine the most important uncertain parameters in accident progression, radiological release,and offsite radiological consequence models for variation. Multiple statistical regression
techniques are then used to quantify uncertainty and to determine which parameters have thegreatest influence on the results.
The results of the uncertainty analysis for the long-term station blackout accident scenario at thePeach Bottom Atomic Power Station are available in a draft report.1 The uncertainty analysis for
the Surry Power Station is still in progress.

1

Sandia National Laboratories. State-of-the-Art Reactor Consequence Analysis Project: Uncertainty Analysis of the
Unmitigated Long-Term Station Blackout of the Peach Bottom Atomic Power Station. Draft Report. NUREG/CR-7155.
Washington,
DC:
U.S.
Nuclear
Regulatory
Commission;
2016.
Availableat:
http://pbadupws.nrc.gov/docs/ML1318/ML13189A145.pdf.

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Sequoyah Nuclear Plant
A follow-on SOARCA study has been initiated to develop best estimates of the offsite radiologicalhealth consequences for select accident scenarios involving a pressurized water reactor (PWR)
nuclear power plant with an ice condenser containment. Accident scenarios are being chosen tochallenge this style of containment, which is smaller than the large dry containment used withother PWR nuclear power plants, including the Surry Power Station - which was evaluated aspart of the original SOARCA Project.
The Sequoyah SOARCA study is applying lessons learned and best practices1,2 from the original
SOARCA Project. In addition, the effects of using diverse and flexible coping strategies involving
portable equipment - which the U.S. nuclear industry implemented in response to challenges
identified by the 2011 accident at the Fukushima Daiichi Nuclear Power Station - are also being
modelled as part of this study to characterize their benefits.
Other Research Activities
The USNRC Office of Nuclear Regulatory Research is also performing or supporting a vastarray of other research activities related to offsite radiological consequence analysis and Level 3
PSA. High-level summary information about the objectives, research approach, and statuspertaining to notable research activities is provided in NUREG-1925.3 Where appropriate,
additional information about specific research activities that directly relate to items in this survey
is provided in the response to the relevant survey items.

Radionuclide Release Characterization
3.4

With regard to the interface between Level 2 and Level 3 PSA:
3.4.1.

How do you handle the transition from Level 2 PSA analyses to Level 3 PSA
analyses? What (if any) difficulties have you encountered in this area?
See response to item 3.4.2.

3.4.2.

If you use representative release categories or source term groups, what criteria doyou use to assign radiological release sequences in the Level 2 PSA model to releasecategories or source term groups? How do you address potential inter-sequencevariability within each representative release category or source term group?
The Level 2 PSA analyst selects radiological release categories in consultation with theLevel 3 PSA analyst. The USNRC uses the MELCOR code4 for modelling severe
accident progression and estimation of source terms. Source term information from theLevel 2 element is contained in MELCOR plot files (*.ptf) that are processed using the
MELMACCS interface software utility5 to extract the data needed for the offsite
radiological consequence analysis and to generate the corresponding inputs for the
MACCS code.

1

Ross K, Phillips J, Gauntt RO, Wagner KC. MELCOR Best Practices as Applied in the State-of-the-Art Reactor
Consequence Analyses (SOARCA) Project. NUREG/CR-7008. Washington, DC: U.S. Nuclear Regulatory Commission;
2014. Available at: http://pbadupws.nrc.gov/docs/ML1423/ML14234A136.pdf.

2

Bixler N, Jones J, Osborn D, Weber S. MACCS Best Practices as Applied in the State-of-the-Art Reactor Consequence
Analyses (SOARCA) Project. NUREG/CR-7009. Washington, DC: U.S. Nuclear Regulatory Commission; 2014. Available
at: http://pbadupws.nrc.gov/docs/ML1423/ML14234A148.pdf.

3

Office of Nuclear Regulatory Research. Research Activities: FY2015-FY2017. NUREG-1925, Rev. 3. Washington, DC:
U.S. Nuclear Regulatory Commission; 2016. Available at: http://pbadupws.nrc.gov/docs/ML1606/ML16060A414.pdf.

4

http://energy.sandia.gov/energy/nuclear-energy/nuclear-energy-safety-technologies/melcor/.

5

http://maccs.sandia.gov/melmaccs.aspx.

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The approach to release category binning can vary by application, but may include
consideration of factors such as: (1) core damage status, (2) containment status,(3) scrubbing or filtration of the release, and (4) timing of the release. Release categorybinning guidance is provided in various information sources, including PSA procedures
guides,1 standards,2 and international documents.3
Some particularly difficult modelling challenges or technical issues that would benefit
from information exchange and sharing of best practices among the international PSA
community include modelling of: (1) operator actions prompted by radiation
measurements; (2) effective size and location of release(s); (3) initial conditions of
releases associated with containment/confinement failure caused by hydrogen explosion;and (4) effects of offsite releases or other conditions on onsite actions.

3.4.3.

What process do you use to define release fraction timing (time and duration of
release) and release truncation time? Do you consider onsite severe accident
mitigation actions or offsite emergency response actions in this process? If you do,
how are they treated?
Releases estimated from MELCOR are usually segmented into approximately hourly
plume segments to match the resolution of typical meteorological data. The MACCS
code allows the analyst to model up to 200 distinct plume segments for each source term.
Development of release termination criteria is typically part of the Level 2 element. It is
common practice to use release truncation times like 24, 48, or 72 hours after event
initiation. Truncation times can also be anchored to the onset of fuel damage - whensevere accident management guidance would typically be invoked - and can include
consideration of onsite or offsite mitigation resources and reliability. Alternatively,
release termination can be implemented by developing a source term that is analysedthrough the time of containment depressurization by either base mat melt-through orcontainment failure. For example, the original SOARCA analyses involved the use oftechnical judgment to select a truncation time of 48 hours after accident initiation; thiswas judged to be appropriate based on the belief that sufficient onsite or offsite
resources would be mobilized to flood containment and halt accident progression within
this time period.15
Research is ongoing to: (1) understand precedents for selecting release truncation times;(2) identify factors the analyst should consider in making this decision; and (3) how the
choice impacts results for different Level 2 PSA metrics. This research suggests theanalyst should consider multiple factors in selecting a release truncation time, including:(1) the modelling uncertainty associated with the severe accident analysis; (2) theavailability of accepted human reliability analysis methods for the domains of interest;(3) the role and maturity of accident management at the facility being studied; (4) the
expected impact of truncation time on deterministic and probabilistic results; and(5) study objectives.

1

Brookhaven National Laboratory. Kalinin VVER-1000 Nuclear Power Station Unit 1 PRA: Procedures Guide for a
Probabilistic Risk Assessment (English Version). NUREG/CR-6572, Rev. 1. Washington, DC: U.S. Nuclear Regulatory
Commission; 2005. Available at: http://pbadupws.nrc.gov/docs/ML0604/ML060450618.pdf.

2

The American Society of Mechanical Engineers (ASME), American Nuclear Society (ANS). Severe Accident
Progression and Radiological Release (Level 2) PRA Standard for Nuclear Power Plant Applications for Light Water
Reactors (LWRs): Trial-Use Standard. ASME/ANS RA-S-1.2-2014. New York, NY: The American Society of MechanicalEngineers; 2014.

3

International Atomic Energy Agency. Development and Application of Level 2 Probabilistic Safety Assessment for
Nuclear Power Plants. Specific Safety Guide No. SSG-4. Vienna, Austria: International Atomic Energy Agency; 2010.
Available at: http://www-pub.iaea.org/mtcd/publications/pdf/pub1443_web.pdf.

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3.5

│ 183

What radionuclides do you use to characterize the offsite radiological consequences attributed
to accidental releases of radiological materials? What is your basis for selecting theseradionuclides?
For LWR offsite radiological consequence analysis applications, current state of practice is to use
results from previous LWR radionuclide screening analyses that used the likelihood of an isotope
being a significant contributor to radiological dose as the screening criterion; factors consideredin these screening analyses included radionuclide inventory, half-life, chemical composition,
release fraction (typically related to volatility for core-melt releases), and dose coefficients. The
process for identifying radionuclides for inclusion in the analysis is described in Section 3.3 and8.2.1 of Appendix VI of WASH-1400.5
A set of 54 radionuclides was used to assess offsite radiological consequences using the
Calculation of Reactor Accident Consequences (CRAC) code for WASH-1400.5 The default input
files for the MACCS code include data for a set of 60 radionuclides - the original 54 considered in
WASH-1400 and 6 additional radionuclides (Strontium-92, Yttrium-92, Yttrium-93, Barium-139,
Lanthanum-141, and Lanthanum-142). In addition to these 60 radionuclides, the MACCS input
file data implicitly include 11 daughter products by summing the dose conversion factors (DCFs)
for both parent and daughter radionuclides to allow for consideration of the contribution toradiological doses from progeny.
Based on previous studies, it is common practice to present results for Iodine-131 and Cesium137 release fractions to further characterize the short-term and long-term offsite radiologicalconsequences, respectively.

3.6

Do you consider releases from multiple units or radiological sources co-located at the samesite? If you do, how are these treated?
The current state-of-practice in the U.S. is to perform offsite radiological consequence analyses
for releases from single reactor units or radiological sources - even if they are co-located withmultiple radiological sources at a shared site.
However, the MACCS code was recently enhanced to include the capability to model releases
from multiple, co-located radiological sources with potentially different accident progression
timelines. This capability was introduced with the release of version 3.10. In practice, the analyst
implements the new multi-source model by specifying multiple MELMACCS-generated source
term input files to be combined, as well as the timing offset for each of the source terms.
As part of its ongoing Full-Scope Integrated Site Level 3 PRA Project,3,4 the USNRC is exploring
the use of this added capability to assess the contribution to total site risk from accidental
releases involving different combinations of two or more major radiological source at themodelled site.

3.7

Do you perform sensitivity or uncertainty analyses on release categories or release fractions?
If you do, what methods are used and what is the basis for using them?
Although more formal approaches are available (e.g., phenomena identification and ranking
technique (PIRT), global sensitivity analysis), sensitivity analyses are typically performed withvariations in source term parameters identified by professional judgment to develop a set ofalternative source terms. For example, alternative source terms reflecting different levels ofcontainment filter effectiveness were used in sensitivity analysis cases for the analyses supporting

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an evaluation of containment protection and release reduction for Mark I and Mark II boilingwater reactor rulemaking activities.1
As described in the response to item 3.3, uncertainty analyses on source terms for specificaccident scenarios have been performed as part of follow-on studies for the SOARCA Project. In
these analyses, Monte Carlo sampling methods are used to generate multiple MELCOR outputs
that are then coupled with sampled MACCS input parameters to evaluate the distribution of
accident consequences arising from uncertainties in selected MELCOR and MACCS input
parameters.

Meteorological Data
3.8

How do you account for temporal and spatial variability in meteorological conditions? If you
use a subset (sampling) of available meteorological data to serve as a representative data set,
what criteria do you use for selecting these representative data?
Temporal variability in meteorological conditions is typically addressed by sampling from one
year of hourly weather data. The base year is selected by the analyst based on professional
judgment and typically is intended to reflect a representative year rather than a year withextreme weather conditions. The sampling scheme employed is a form of importance samplingthat assigns each hourly weather sequence to a meteorological bin based on the wind speed,
stability class, magnitude of precipitation, and onset of precipitation. Sufficient samples are
drawn from each bin to ensure that lower frequency, higher consequence meteorologicalconditions (i.e., stable low wind speed conditions or periods of precipitation) are adequately
represented in the overall sample.
Spatial variability in meteorological conditions is typically addressed by specifying a limiting
spatial interval for the use of recorded meteorological data. All spatial intervals beyond thisinterval will use analyst-specified boundary weather conditions. These boundary weather
conditions will also be applied if the duration of recorded meteorological data is not enough totransport the last plume through the limiting spatial interval for measured weather conditions.

3.9

Do you perform sensitivity or uncertainty analyses on meteorological conditions? If you do,what methods are used and what is the basis for using them?
Sensitivity analyses are sometimes performed with alternate years of weather data or by varying
the number of samples drawn for the selected weather year. As described in the response to item
3.8, the analyst uses professional judgment to select one year of hourly weather data to serve as arepresentative year. Sensitivity analyses may therefore be used to evaluate the impact on resultsof using alternative years of hourly weather data.

Environmental Transport and Dispersion
3.10

What atmospheric transport and dispersion (ATD) model(s) do you use? What process andcriteria do you use to select: (1) an ATD model; and (2) a time scale for updating calculations
to account for time dependence?
The MACCS code uses a Gaussian straight-line trajectory plume segment model that can handlechanges in wind speed, wind direction, stability, and precipitation, but not changes in plume
trajectory. Earlier analyses typically employed a limited number of plume segments that could
represent multiple hours of releases, but more recent analyses include plume segments of1

Beall, R H. Evaluation of the Containment Protection & Release Reduction for Mark I and Mark II Boiling Water
Reactors Rulemaking Activities (10 CFR Part 50) (RIN-3150-AJ26). SECY-15-0085. Washington, DC: U.S. Nuclear
Regulatory Commission; 2015. Available at: http://pbadupws.nrc.gov/docs/ML1502/ML15022A218.html.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENTNEA/CSNI/R(2018)1│ 185

approximately hourly duration to match the hourly resolution of the underlying meteorologicaldata.
A new particle tracking ATD model based on the HYbrid Single Particle Lagrangian Integrated
Trajectory (HYSPLIT) model1 is under development for addition to MACCS and will be released
in a future version. This particle tracking ATD model is being developed to provide analternative ATD model that addresses known limitations of the Gaussian straight-line trajectory
plume segment model.
As more ATD model options become available, a process and criteria will be needed to selectwhich model(s) will be used for a particular study. Some factors to consider in making theselection may include: (1) study objectives; (2) computational efficiency; (3) site characteristics;and (4) data availability.

3.11

What process and criteria do you use to define the boundaries and intervals of the spatial gridor domain used for performing offsite consequence calculations?
In the MACCS code, the region potentially affected by an accidental release is represented withan (r, θ) polar coordinate spatial grid system centred on the location of the release. The radius r
represents downwind distance. The angle θ represents the angular offset from north in theclockwise direction. The analyst specifies the number of radial divisions as well as their endpoint
distances. Up to 35 divisions may be defined, extending out to a maximum distance of 9999 km.
However, it is common practice to model the region surrounding a nuclear installation out to aradial distance of 1000 miles. The angular divisions used to define the spatial grid are fixed in thecode, and therefore cannot be modified by the analyst. All of the calculations of MACCS are
stored on the basis of this polar coordinate spatial grid.
USNRC guidelines specify that regulatory analyses should typically calculate offsite radiologicalconsequences within 50 miles of the nuclear installation. However, other spatial intervals can beconsidered on a case-by-case basis.8,9The USNRC safety goal policy statement specifies that risk metrics for comparison with the
QHOs are to be calculated over defined spatial intervals:• Average Individual Early Fatality Risk QHO: 0-1 mile beyond the nuclear power plant siteboundary.
• Average Individual Latent Cancer Fatality Risk QHO: 0-10 miles beyond the nuclear power plantsite boundary.20
PCAs or Level 3 PSAs performed to support research studies may calculate offsite radiologicalconsequences over a broader range of spatial intervals. Although Level 3 PSAs performed as partof previous research studies5,6 calculated some consequence metrics over the entire modelled
spatial domain out to 1000 miles around each of the included nuclear power plant sites, current
USNRC-sponsored PCAs and Level 3 PSAs typically calculate offsite radiological consequences
within 50-100 miles to reflect the limitations of the Gaussian plume segment ATD model. Results
may be calculated out to longer distances to determine whether the defined spatial domain ofinterest sufficiently captures the full range of consequences, but these longer distance results are
typically not used for risk estimation or evaluation purposes.

1

http://www.arl.noaa.gov/HYSPLIT_info.php.

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3.12

With regard to the spatial modelling around a nuclear power plant for Level 3 PSA:
3.12.1. What information sources do you use to develop geographical or topographicalparameters? Do you use generic or site-specific data?
The MACCS code uses site-specific meteorological data as well as site-specificpopulation, agricultural, and economic data estimated on a polar grid, typically using
the SECPOP code.1

3.12.2. Does the ATD response vary spatially with respect to varying topographic
parameters? What is the spatial resolution of the ATD and do topographic parametersof the ATD vary within this resolution?
Because MACCS uses a simple terrain model that does not account for terrain elevationdifferences or topographic effects on winds, the use of topographic data is used only
qualitatively to evaluate the applicability of the Gaussian plume segment model to aparticular site.

3.13

Do you consider radiological releases to water sources and aqueous transport and dispersion
phenomena? If you do, how are these treated? If you have previously considered modellingradiological releases to water sources and aqueous transport and dispersion phenomena, but
decided not to do so, what was the basis for your decision?
The MACCS code includes a simplified model for deposition of atmospheric radioactivity onto
water bodies and runoff of deposited material on land into water bodies to support estimation of
the contribution to radiological doses from ingestion of contaminated water.2
However, aqueous releases directly to water bodies (e.g., runoff, pipe discharges, or discharges toaquifers from base mat melt-through) are typically not addressed. There a two principal
justifications for excluding aqueous releases: (1) the airborne pathway is expected to be dominant
for health risks because movement of radionuclides to the accessible environment is expected tobe slow relative to atmospheric transport; and (2) releases to groundwater or surface water are
considered to be easier to interdict.3
After the 2011 Fukushima nuclear accident, the USNRC initiated a research project to assess
potential offsite consequences of losing control of highly contaminated water in a severe accident
in which the contaminated water flows to a body of water. The final project report presents stateof-the-art hydrologic transport modelling results for leakage of waterborne radionuclides
directly into a freshwater body. The study postulated releases to three types of freshwatersettings: (1) a large river, (2) a small river, and (3) a small lake. Two- and three-dimensionalmodelling was used to explore the concentrations of radionuclides as they are transported
through the three freshwater bodies. The approach determined how advection,
dilution/dispersion, radioactive decay, and adsorption/desorption processes affect concentrations

1

http://maccs.sandia.gov/secpop.aspx.

2

Helton JC, Muller AB, Bayer A. Contamination of Surface-Water Bodies after Reactor Accidents by the Erosion of
Atmospherically Deposited Radionuclides. Health Physics; 1985; 48(6): 757-771.

3

Niemczyk SJ, Adams KG, Murfin WB, Ritchie LT, Eppel EW, Johnson JD. The Consequences from Liquid Pathways
after a Reactor Meltdown Accident. NUREG/CR-1596. Washington, DC: U.S. Nuclear Regulatory Commission; 1981.
Available at: http://www.iaea.org/inis/collection/NCLCollectionStore/_Public/35/047/35047210.pdf.

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of each of the transported radionuclides under the hypothetical conditions in each freshwatersetting.1

Protective Action (Countermeasure) Modelling
3.14

How do you define, and what protective actions (countermeasures) do you model for each of
the nuclear or radiological incident or accident response phases listed below?
3.14.1. Early/Emergency Phase:
3.14.2. Intermediate Phase:
3.14.3. Late/Recovery Phase:
USEPA guidance25,26 uses the following definitions for each radiological incident phase:(1) Early Phase: The beginning of a radiological incident when immediate decisions for effective
use of protective actions are required and must therefore be based primarily on the status of
the radiological incident and the prognosis for worsening conditions. This phase may lastfrom hours to days.
(2) Intermediate Phase: The period beginning after the source and releases have been brought
under control (has not necessarily stopped but is no longer growing) and reliable
environmental measurements are available for use as a basis for decisions on protective
actions and extending until these additional protective actions are no longer needed. This
phase may overlap the early phase and late phase and may last from weeks to months.
(3) Late Phase: The period beginning when recovery actions designed to reduce radiation levelsin the environment to acceptable levels are commenced and ending when all recovery actions
have been completed. This phase may extend from months to years.
The MACCS code utilizes a similar framework. The early phase is modelled using the EARLY
module, and the intermediate and late phases are modelled using the CHRONC module.
However, MACCS imposes constraints that prohibit the overlap of phases in consequencemodels; each phase must begin at the end of the preceding phase.
Early phase protective actions modelled in the MACCS EARLY module include: sheltering,evacuation, and dose-dependent relocation. MACCS also includes a capability to implement a
potassium iodide (KI) model that accounts for use of KI as a supplementary protective action to
reduce the radiological dose to the thyroid gland by blocking the uptake of radioiodine.
Intermediate phase protective actions modelled in the MACCS CHRONC module include: dosedependent relocation or interdiction and dose-dependent bans on agricultural products. Latephase protective actions or recovery actions modelled in the MACCS CHRONC module include:
dose-dependent relocation or interdiction (whether temporary or permanent - which is referred
to as condemnation), decontamination, and dose-dependent bans on agricultural products orcondemnation of farmland.

3.15

Do you model population groups with different protective action (countermeasure)
behaviours? If you do, how do you define them? To what extent do you account for
population density or age and gender distribution effects?
Multiple cohorts are typically used to model population groups with different protective action
behaviours. The MACCS code allows the analyst to define up to twenty cohorts. At a minimum,
two cohorts are used to model and estimate offsite consequences for two distinct groups: (1) an

1

Yabusaki SB, Napier BA, Perkins WA, Richmond MC, Rakowski CL, Snyder SF et al. Modeling of Radionuclide
Transport in Freshwater Systems Associated with Nuclear Power Plants. NUREG/CR-7231. Washington, DC: U.S.
Nuclear Regulatory Commission; 2017. Available at: https://www.nrc.gov/docs/ML1711/ML17111A578.pdf.

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evacuating cohort that models the fraction of the offsite population that begins evacuation when
a general emergency is declared at the nuclear installation; and (2) a non-evacuating cohort thatmodels the fraction of the offsite population that is assumed to maintain normal activity unlesstheir projected dose is predicted to exceed USEPA PAG levels.
A number of factors are typically used to select and define cohorts, including special populations.
Examples of these factors include: (1) location or spatial interval around the nuclear installation;(2) delay times for implementing protective actions; (3) evacuation speeds; and (4) exposure
factors and shielding parameters. Age- and gender-specific variability in protective action
behaviours are generally not considered; however, school populations may be assigned to theirown cohort if they constitute a distinct population within or near the EPZ.

3.16

Do you use probabilistic models of protective action (countermeasure) behaviours that model
the probabilities of success or failure for protective actions (countermeasures)? If you do,please describe the models and their bases.
As stated in the response to item 3.15, the current state-of-practice is to specify multiple cohorts
to model population groups with different protective action behaviours, including those that takea longer period of time to evacuate or that choose to not evacuate at all. For each cohort,
protective actions are modelled using deterministic models that implicitly assume protective
actions (if taken) are 100% effective for the entire cohort and result in specified exposure andshielding factors for each exposure pathway. Although the MACCS code has the capability to
specify probability distributions for each of the parameters used to define these cohort-specificprotective action models, these do not provide a quantitative measure of the probability ofsuccess or failure of protective actions.

3.17

What information sources do you use as a technical basis for protective action(countermeasure) modelling? Do you use generic or site-specific data?
Parameters for modelling protective actions are typically based on site-specific data. For
example, parameters for evacuation modelling are typically derived from information in sitespecific evacuation time estimate (ETE) analyses. U.S. regulations require that an analysis of thetime required to evacuate be provided for various sectors and distances within the plume
exposure pathway EPZ for transient and permanent residents.1 This ETE analysis includes
consideration of multiple factors, including: (1) evacuation demand estimation for multiple
population groups (e.g., permanent residents, transient populations, transit-dependent
populations, special facility residents, and schools); (2) roadway characteristics, capacity
analysis, and traffic control; and (3) alternative scenarios comprised of different combinations of
season, day of the week, time of day, weather conditions, special events, roadway impact, or othercircumstances that should be assessed. Multiple scenarios are evaluated to ensure ETE results
encompass a reasonable range of potential evacuation situations for the specific site. Criteria fordevelopment of ETE studies are available for further information.2
U.S. regulations require nuclear power plant licensees to update their ETE analyses on a periodic
basis; in practice, these analyses are updated approximately every 10 years to coincide with therelease of decennial population census data from the U.S. Census Bureau.24,54

1

US Nuclear Regulatory Commission. Title 10 Code of Federal Regulations Part 50, Appendix E, “Emergency Planningand Preparedness for Production and Utilization Facilities” Available at: http://www.nrc.gov/reading-rm/doccollections/cfr/part050/part050-appe.html.

2

Jones J, Walton F, Wolshon B, Laughlin J. Criteria for Development of Evacuation Time Estimate Studies. NUREG/CR7002.
Washington,
DC:
U.S.
Nuclear
Regulatory
Commission;
2011.
Availableat:
http://www.nrc.gov/docs/ML1130/ML113010515.pdf.

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Radiological Exposure and Dose Assessment
3.18

How do you define potentially exposed populations or cohorts? Do you consider onsite(e.g., worker) populations or potentially sensitive, vulnerable, or critical groups? If you do,what is your basis for including these groups, and how are they modelled?
In general, PCAs or Level 3 PSAs are performed to assess the offsite public consequences or risks
attributable to a spectrum of possible accident scenarios involving a nuclear installation.
Therefore, onsite populations are generally not considered.
As described in the response to item 3.15, cohorts are generally selected and defined based on anumber of factors. Although sensitive, vulnerable, or critical groups are generally not
differentiated from members of the general public within MACCS, such populations may bemodelled as distinct cohorts if they have unique exposure factors or shielding parameters.

3.19

What exposure pathways do you model? What is your basis for selecting these pathways?
What exposure duration is assumed in your models?
The MACCS code models the following exposure pathways identified by MACCS module andthe corresponding radiological incident phases:(1) EARLY Module (Early Phase): (1) direct exposure to external radiation from the plume ofreleased radiological materials (cloud shine); (2) inhalation of radioactivity in the plume;(3) direct exposure to external radiation from ground contamination (ground shine);(4) inhalation of resuspended radioactivity; and (5) contamination of skin and clothing.
(2) CHRONC Module (Intermediate and Late Phases): (1) ground shine; (2) inhalation ofresuspended radioactivity; and (3) ingestion of contaminated food and water.
Exposure durations are typically based on site-specific ETE analyses and protective action
models for the early phase, while a lifetime (50-year) exposure duration is typically assumed forthe late phase. Within MACCS, the analyst specifies the duration of the intermediate phase,
which can range from 0 to 365 days. Recent PCA studies15,16 have used an intermediate phaseduration of 0 days based on professional judgment. Sensitivity analyses can be performed toevaluate the impact of using alternative intermediate phase durations.

3.20

What information sources do you use to develop exposure and dose conversion factors? Do
you use generic or site-specific data? Do you use average or age- and gender-specific data?
Exposure factors such as inhalation and intake rates and dose conversion factor (DCFs) arebased on generic average values obtained from USEPA guidance documents. DCFs express therelationship between: (1) environmental concentrations or intakes; and (2) resultant humandoses or dose rates. They are developed for specific exposure pathways, organs, andradionuclides. The MACCS code requires DCFs for the inhalation, ingestion, cloud shine, andground shine exposure pathways for each organ for which doses are to be calculated and for eachradionuclide included in the analysis. DCF files that were created based on USEPA guidance anddosimetric models1 are typically used as input to MACCS.

1

Eckerman KF, Leggett RW, Nelson CB, Puskin JS, Richardson ACB. Federal Guidance Report No. 13: Cancer Risk
Coefficients for Environmental Exposure to Radionuclides. EPA 402-R-99-001. Washington, DC: U.S. Environmental
Protection Agency; 1999. Available at: https://www.epa.gov/sites/production/files/2015-05/documents/402-r-99-001.pdf.

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Radiological Health Effects
3.21

Do you model and estimate radiological health effects in your offsite radiologicalconsequence analyses? If you do, please respond to questions 3.22 through 3.24. If you donot, please describe any other methods you use to address the offsite radiological healtheffects attributable to possible accident scenarios involving nuclear installations and proceed
to question 3.25.
Early (acute) fatalities and injuries are typically estimated using deterministic health effects
models. Latent (delayed/chronic) fatalities and injuries are typically estimated using stochastic
health effects models.

3.22

What types of early and latent health effects (including fatal and non-fatal effects) do you
model? What target organs do you use for the associated health effects models?
Radiological health effects models have been developed for multiple types of fatalities andinjuries for use in offsite radiological consequence analyses.1 The health effects evaluated in a
specific analysis are typically driven by the needs of the application. Metrics used to support
various applications are described in responses to other items. Notable examples include themetrics used to compare results with the safety goal QHOs for average individual risk of earlyand latent cancer fatalities.
Although early injuries and non-fatal latent cancers are typically modelled and estimated in
PCAs or Level 3 PSAs, these results are generally used to provide additional insights andperspective; they are not typically used for risk estimation and evaluation purposes.
The types of early and latent radiological health effects and target organs that are typicallymodelled and estimated using the MACCS code are summarized below in Table 4.
Table 4. Early and Latent Radiological Health Effects and Target Organs in MACCS

Health EffectTarget Organa,bHematopoietic Syndrome
Red Bone Marrow
Gastrointestinal Syndrome
StomachFatality
Pulmonary Syndrome
Lungs
Total Early Fatalities
Not Applicable
Prodromal Vomiting
StomachDiarrhea
Early
Skin Erythema
Skin
Transepidermal Injury
InjuryThyroiditis
Thyroid GlandHypothyroidism
Pneumonitis
Lungs
Total Early Injuries
Not Applicable
Leukemia
Red Bone Marrow
Bone Cancer
Bone
Breast Cancer
Breasts
Lung Cancer
Lungs
LatentFatality or Injury
Thyroid Cancer
Thyroid GlandLiver Cancer
Liver
Colon Cancer
Lower Large Intestine
Residual Cancers
Bladder Wall
Total Cancer Fatalities or Injuries
Not ApplicableaTarget organs available for specifying early health effects models in MACCS include: (1) lower large intestine, (2) lungs, (3) red
bone marrow, (4) skin, (5) stomach, and (6) thyroid gland.
bTarget organs available for specifying latent health effects models in MACCS include: (1) bladder wall, (2) bone, (3) breasts, (4)liver, (5) lower large intestine, (6) lungs, (7) red bone marrow, (8) thyroid gland, and (9) whole-body lifetime effective dose.
Category

1

Type

Evans JS, Abrahamson S, Bender MA, Boecker BB, Gilbert ES, Scott BR, Yaniv SS. Health Effects Models for Nuclear
Power Plant Accident Consequence Analysis. Part I: Introduction, Integration, and Summary. NUREG/CR-4214, Rev. 2,
Part
I.
Washington,
DC:
U.S.
Nuclear
Regulatory
Commission;
1993.
Availableat:
http://pbadupws.nrc.gov/docs/ML0500/ML050030192.pdf.

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3.23

│ 191

What dose-response model(s) do you use to estimate the numbers of latent health effects
attributable to radiological doses caused by accidental releases from nuclear installations, andwhat is the basis for their use? What (if any) methods do you use to account for uncertainty
about the true dose-response relationship for exposures to low levels of ionizing radiation?
In previous Level 3 PSAs performed as part of research studies,5,6 linear no-threshold (LNT) or
linear-quadratic no-threshold dose-response models were used to quantify total and individual
latent cancer fatality risk. However, in recent USNRC-sponsored PCAs,15,16 sensitivity analysis
has been used to evaluate the effects of using plausible alternative dose-response models forquantifying latent health effects. In these PCAs, the default LNT model was used as the base case,and two models involving dose truncation levels were used as sensitivity cases. In the first model,
the dose truncation was set at 6.2 mSv/yeara level that corresponds to the backgroundradiation exposure an average individual in the U.S. receives in a given year. In the second
model, the dose truncation was set at 50 mSv/year with a 100 mSv lifetime limita level below
which the Health Physics Society recommends against quantitative estimation of health risks.1
Supplementary Information
The dose-response models used for quantifying the lifetime cancer risk attributable to possible
nuclear accidents have been based largely on associations observed in epidemiology studies ofwell-defined populations exposed to radiation doses that were orders of magnitude greater thanwhat the majority of the affected public would be exposed to following a nuclear accident. Since
high-quality dose-response data for the general population are not available for the range ofradiation exposures relevant to nuclear accidents, there is considerable uncertainty about thenature of the true dose-response relationship for these exposures; extrapolation from the highdose range of epidemiologic data to the low-dose range of interest can therefore be performed
using a number of plausible alternative mathematical models that provide a reasonable fit to theobserved data.
For decades, scientific advisory and governmental bodies have repeatedly endorsed the default
use of the LNT dose-response model for radiation protection applications - where precaution andconservatism have been judged to be prudent. Although prevailing knowledge of biologicalmechanisms and available epidemiologic evidence continue to support the LNT hypothesis,
emerging evidence from in vitro radiation biology studies has spurred heated debate within thescientific community about the validity of the LNT model. Moreover, due in part to previously
mentioned scientific limitations, available epidemiologic studies have not demonstrated theoccurrence of radiation-induced cancers among populations exposed to either very low levels ofradiation or to radiation delivered over long periods of time. For these reasons, scientificadvisory bodies and professional organizations now caution against using the LNT model in risk
projections to estimate the number of cancer deaths resulting from low-level radiation exposures
to large affected populations in risk assessment applications; instead, they now recommend notquantifying the lifetime cancer risk for radiation exposures below specified levels. In practice,
this approach is implemented using a truncation dose-response model. In this type of model, adose truncation or cut off level is specified, below which lifetime cancer deaths attributable toaccidental releases are not quantified.

3.24

What information sources do you use to develop input parameters for the health effectmodels? Do you use generic or site-specific data? Do you use average or age- and genderspecific data?
Generic average values from multiple information sources are typically used for specifying
health effects model parameters. Parameter values for deterministic models used to estimate

1

Health Physics Society. Radiation Risk in Perspective: Position Statement of the Health Physics Society. 2010;(January1996). Available at: http://hps.org/documents/risk_ps010-2.pdf.

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early health effects are typically based on information from previous studies sponsored by the
USNRC57 and by the Commission of the European Communities.1 Parameter values forstochastic models used to estimate latent health effects are typically based on information
contained in USEPA guidance56 and U.S. National Academy of Sciences reports.2

Economic Consequences
3.25

Do you model and estimate economic consequences in your offsite radiological consequence
analyses? If you do, please respond to questions 3.26 and 3.27. If you do not, please describe
any other methods you use to address the offsite economic consequences attributable topossible accident scenarios involving nuclear installations and proceed to question 3.28.
Economic consequences or offsite property damages are typically estimated to provide input tocost-benefit analyses performed as part of regulatory, back fit, or environmental analyses.
Economic consequences and other socio-economic indicators (e.g., the extent of land or
population impacted by protective actions) are sometimes estimated in research studies toprovide additional insights that can be used to either: (1) check for consistency with othercorrelated results; or (2) evaluate results from a different perspective.

3.26

With regard to potential economic consequences considered within a Level 3 PSA:
3.26.1. What types of economic consequences do you model? What process do you use to
select cost categories and models for estimating economic consequences?
The MACCS code includes economic models for estimating the costs attributed to
implementation of modelled protective actions to reduce radiological dose to the offsite
public from the accidental release of radiological materials. These costs include: (1) daily
costs of compensation for populations subject to evacuation or short-term relocation
arising from food, housing, transportation, lost income, or replacement of lost personal
property; (2) costs of long-term relocation of populations and businesses in interdicted
land areas; (3) depreciation costs that account for loss of value of interdicted property;(4) decontamination costs; and (5) costs arising from implementation of agriculturalcountermeasures.
MACCS does not include models for estimating some cost categories that could be
important for nuclear accident scenarios. Examples of these cost categories include: (1)costs associated with the number of radiation-induced injuries or fatalities; (2)replacement power costs; and (3) costs associated with storage or disposal ofcontaminated material following decontamination efforts.
An alternative economic model that can estimate national and regional gross domestic
product (GDP) losses using regional input-output techniques is under development for
addition to MACCS and will be released in a future version.

1

Haskin FE, Kraan BCP, Harper FT, Grupa JB, Goossens LHJ, Randall J. Probabilistic Accident Consequence Uncertainty
Analysis: Early Health Effects Uncertainty Assessment. NUREG/CR-6545. Washington, DC: U.S. Nuclear Regulatory
Commission; 1997.

2

Committee on the Biological Effects of Ionizing Radiation (BEIR V); Commission on Life Sciences; Division on Earthand Life Studies; National Research Council. Health Effects of Exposure to Low Levels of Ionizing Radiation: BEIR V.
Washington, DC: National Academies Press; 1990. Available at: http://www.nap.edu/catalog/1224/health-effects-ofexposure-to-low-levels-of-ionizing-radiation.

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3.26.2. If you translate radiological health effects into economic consequences, please
describe your process for doing so.
As described in the response to item 2.2.2, to generate commensurable values for use incost-benefit analyses, benefits in terms of averted radiological health effects may bemonetized using a conversion factor applied to the averted population dose within aprescribed distance (typically 50 miles) from the nuclear installation. The USNRC iscurrently revising the value of this dollar per person-rem conversion factor and is
proposing a process for routine updates to ensure its value is maintained current.21

3.26.3. If you model land contamination, please describe your process for estimating thehealth, environmental, or economic consequences attributed to land contamination.
The health and economic consequences attributed to land contamination are captured in
the dose and economic models, but there is no additional consideration of environmental
consequences beyond the previously described dose and cost models.

3.27

What information sources do you use to develop cost parameters for economic consequencemodels? Do you use generic, region-specific, or site-specific data?
There are a variety of inputs to the economic consequence models. Site-specific data pertaining topopulations, property values, personal incomes, agricultural sales, and land fraction values comefrom census data provided by the U.S. Census Bureau and from the U.S. Department ofAgriculture.
Generic inflation data used to escalate cost values to common years are derived from the U.S.
Bureau of Labour Statistics Consumer Price Index for All Urban Consumers (CPI-U) database.
Generic decontamination plan data, depreciation rates, and expected rates of return fromproperty come from literature reviews and updates to previous Level 3 PSAs performed as partof research studies.5,6Consequence Quantification and Reporting
3.28

What process and criteria do you use for selecting consequence metrics for quantification andreporting, including spatial intervals or distances from release points?
There is no specified process or criteria for selecting consequence metrics for quantification andreporting. Factors that are typically considered in selecting consequence metrics and spatial
intervals can include: (1) analysis objectives - which are typically linked to application-specificrequirements or questions to be addressed; (2) stakeholder interests; (3) standards requirements
or state-of-practice; (4) potential future uses of results and insights; (5) schedule and resource
constraints; and (6) capabilities and limitations of models and analytical tools. Metrics used tosupport various applications are described in responses to other items.
Notable examples of consequence metrics that can be quantified and reported using the MACCS
code are summarized below in Table 5. Although MACCS enables the analyst to quantify andreport a large number of consequence metrics, schedule and resource constraints may impose
practical limits on what is quantified and reported for a particular study.

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Table 5. Notable Consequence Metrics Available for Quantification and Reporting Using MACCS1
MACCS
ModuleConsequence MetricDefinition

Number of cases of specified health effects predicted to occur
within specified spatial interval(s).
The greatest distance at which a specified level of early fatality
risk is exceeded. Can be used to provide information about the
Early-Fatality Radius
size of the region in which early fatalities are predicted to occur
by setting the threshold parameter value to 0.
Number of people predicted to receive early phase doses toPopulation Exceeding Thresholdspecified target organ(s) that exceed specified threshold level(s).
Average individual risk of specified health effect(s) for
individuals within specified spatial interval(s). Calculated by
Average Individual Risk
summing health effect cases predicted to occur in all sectors at
specified spatial interval(s) and dividing by the number ofsectors.
The total population dose to specified target organ(s) from
Population Dosemodeled exposure pathways within specified spatial intervals.
Average individual risk of specified health effect(s) for
individuals within specified spatial interval(s). Calculated by
Population-Weighted Risk
summing health effect cases predicted to occur within specified
spatial interval(s) and dividing by the total population within thespecified interval(s).
Sizes of contaminated land area(s) in which dose(s) to specified
Land Area Exceeding Dosetarget organ(s) exceed specified threshold levels.
Land
Area
Exceeding Sizes of contaminated land area(s) in which concentration(s) ofConcentration
specified radionuclide(s) exceed specified threshold levels.
Fraction of population crossing outer boundary of specified
Population Movement
spatial interval(s) over specified time interval(s).
The total population dose to specified target organ(s) from
Population Dosemodeled exposure pathways in specified spatial intervals.
Economic costs associated with implementing protective actions
Economic Cost
within specified spatial interval(s).
Sizes of area(s) or number of people within specified spatial
Impacted Area or Populationinterval(s) that are impacted by interdiction, decontamination,condemnation, or disposal of agricultural products.
Number of people within specified spatial interval(s) that are
Impacted Population
impacted by evacuation or relocation.
Health Effect Cases

EARLY

CHRONC

Risk Characterization
3.29

What process and criteria do you use for selecting risk metrics for effectively communicating
Level 3 PSA results, including metrics designed to measure the effects of radiologicalreleases involving multiple units or radiological sources co-located at the same site, if
applicable?
As with consequence metrics, there is no specified process or criteria for selecting risk metrics forquantification and reporting. Factors that are typically considered in selecting risk metrics can
include: (1) analysis objectives - which are typically linked to application-specific requirementsor questions to be addressed; (2) stakeholder interests; (3) standards requirements or state-ofpractice; (4) potential future uses of results and insights; (5) schedule and resource constraints;

1

This table does not include a comprehensive listing of consequence metrics that are available for quantification andreporting in the MACCS code.

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and (6) capabilities and limitations of models and analytical tools. Metrics used to support
various applications are described in responses to other items.
As part of its ongoing Full-Scope Integrated Site Level 3 PRA project, the USNRC is exploring
candidate integrated site risk metrics that can be used to measure the effects of radiologicalreleases involving multiple radiological sources co-located at the same site. In principle, the risk
metrics used to estimate the frequencies of offsite public health, economic, and environmental
consequences for accidental releases from single radiological sources can be adapted by adjusting
the frequency basis to estimate the same quantities for accidental releases from multiple
radiological sources. For example, instead of quantifying risk metrics on a per-reactor-year basis,
they can be quantified on a per-site-year or per-calendar-year basis.

3.30

For each of the groups listed below, what methods do you use for presenting andcommunicating risk results and the uncertainty in risk results? How do you present the resultsfrom low-probability/high-consequence events in an understandable context for each group?
Who is responsible for communicating the results to each group?
(1)(2) Decision makers:(3) General Public:
In general, risk results that characterize variability or aleatory uncertainty arising from inherent
randomness or stochastic processes are presented using a variety of formats and graphicaldisplays. Notable examples include: (1) point estimates - especially expected (mean) risk of
selected consequence metrics over all-weather trials - which are typically used to inform
regulatory decision making; and (2) complementary cumulative distribution function (CCDF)
curves (also termed exceedance frequency curves or risk curves) - which represent the frequenciesof exceeding different consequence levels over all modelled radiological release categories andweather trials.
Epistemic uncertainty in risk results arising from imperfect knowledge is typically characterizedusing random or pseudo-random sampling techniques (e.g., simple Monte Carlo sampling or
Latin Hypercube Sampling) to develop empirical probability distributions that characterize theuncertainty in risk results. Notable examples of graphical displays used to illustrate thisuncertainty include: (1) empirical probability distributions for selected risk metrics - typically inthe form of probability density function (PDF) curves and/or cumulative distribution function(CDF) curves; (2) box plots that illustrate the locations of key summary statistics (e.g., meanvalue, 50th percentile (median value), 95th percentile, and 5th percentile) for selected risk metrics;and (3) sets of CCDFs (also termed families of risk curves) - with the different curves representing
different probabilities of frequencies of exceeding different consequence levels.
In general, the USNRC technical staff and managers directly involved with a particular analysis
are responsible for communicating results to decision makers and to the general public.
However, the USNRC has an Office of Public Affairs with communications professional who canprovide advice and consultation to USNRC staff on risk communication efforts. In addition, theUSNRC has developed guidance documents that aim to enhance risk communication with bothinternal stakeholders (including decision makers)1 and external stakeholders (including thegeneral public).2

3.31

With regard to the treatment of uncertainties:

1

Szabo A, Persensky J, Peterson L, Specht E, Goodman N, Black R. Effective Risk Communication: The Nuclear
Regulatory Commission's Guidelines for Internal Risk Communication. NUREG/BR-0318. Washington, DC: U.S. Nuclear
Regulatory Commission; 2004. Available at: http://www.nrc.gov/reading-rm/doc-collections/nuregs/brochures/br0318/.

2

Persensky J, Browde S, Szabo A, Peterson L, Specht E, Wight E. Effective Risk Communication: The Nuclear Regulatory
Commission’s Guidelines for External Risk Communication. NUREG/BR-0308. Washington, DC: U.S. Nuclear
Regulatory Commission; 2004. Available at: http://www.nrc.gov/reading-rm/doc-collections/nuregs/brochures/br0308/.

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196 │ NEA/CSNI/R(2018)1
3.31.1. Do you perform sensitivity or uncertainty analyses as part of risk estimation? If you
do, what process and methods do you use?
Sensitivity analyses are frequently performed for regulatory analyses to evaluate thesensitivity of net value estimates to changes in cost-benefit analysis inputs. For Level 3
PSAs, sensitivity analyses can be performed to: (1) identify important parameters fortreatment in an uncertainty analysis; or (2) assess the impacts of uncertainties related tothe existence of plausible alternative models of logic structures and physical phenomena.
Simplified uncertainty analyses using Monte Carlo sampling methods have been
performed for some regulatory analyses.1 In addition, Level 3 PSAs performed as part ofresearch studies5,6 have included uncertainty analyses. Current state-of-practice is toperform integrated uncertainty analyses using Monte Carlo sampling methods by
specifying probability distributions for selected parameters in the Level 1 and Level 2
PSA models; parametric uncertainty analyses are not typically performed forparameters used to specify the Level 3 offsite PCA model.

3.31.2. How do you determine which parameters will be varied or will have uncertainty
distributions specified for propagating uncertainty? What parameters do you
evaluate?
Professional judgment informed by reviews of previous sensitivity or uncertainty
analyses is typically used to select parameters for variation or sampling.

3.31.3. To what extent do you consider the effect of correlation on parameters?
If performed, the extent to which state of knowledge correlation (SOKC) between
parameters is accounted for in performing a parametric uncertainty analysis is driven
by the needs of a particular PSA application with respect to the degree of site-specificity
and model realism.

Other
3.32

Are there any other questions that you believe should have been asked in this survey? Is there
any other information or are there any other technical challenges or notable practices you
would like to share with the international community?
Two additional questions could have been asked in this survey to add value to this WGRISK
activity:(1) Do you include consideration of non-radiological impacts (e.g., injuries or fatalities caused bytaking protective actions to avoid or reduce radiological dose, psychological or mental healthimpacts, or social disruption) in performing Level 3 PSA? If you do, what types of nonradiological impacts do you consider? What models and sources of information do you use tocharacterize these impacts? If you do not, what is your basis for excluding non-radiologicalimpacts from these analyses?
As described in the response to item 2.5, the USNRC safety goal policy statement broadly defines
an acceptable level of radiological risk to public health and safety from potential releases of1

Skeen DL. Consideration of Additional Requirements for Containment Venting Systems for Boiling Water Reactors withMark I and Mark II Containments. SECY-12-0157. Washington, DC: U.S. Nuclear Regulatory Commission; 2012.
Available at: http://www.nrc.gov/reading-rm/doc-collections/commission/secys/2012/2012-0157scy.pdf.

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radiological materials from commercial nuclear power plants. The safety goal QHOs for average
individual risk of early fatality and latent cancer fatality relate only to the risk of radiologicalhealth effects caused by exposure to ionizing radiation from accidental radiological releases.
Therefore, in making determinations about whether a proposed generic safety enhancement orback fit provides a substantial enough increase in public health and safety to warrant a detailed
evaluation to determine whether it is cost-beneficial, the USNRC focuses on measures ofradiological risk or surrogate measures related to CDF or LERF. However, non-radiologicalimpacts could be considered in evaluating the benefits and costs of proposed actions, andtherefore could be considered in performing a Level 3 PSA.
The MACCS code is limited in its capability to quantify non-radiological impacts. In particular,
as shown in Table 5, the code is able to quantify and report the numbers of people affected byvarious protective actions; these results can be used to provide additional perspective about thesocietal impacts of accidental releases. The code does not include models for quantifying othertypes of non-radiological impacts.
(2) Results from recent studies15,16 suggest that latent cancer fatality risk is sensitive to assumptions
about long-term habitability dose criterion that are used to determine when relocatedpopulations are allowed to return home following a modelled accidental release. What long-term
habitability dose criterion do you use in your Level 3 PSAs and what is the basis for this value?
The USEPA intermediate phase PAG level for relocation of the public is 2 rem (20 mSv)
projected dose during the first year and 0.5 rem (5 mSv) per year projected dose during
subsequent years.25,26 While many states in the U.S. have adopted this as the long-term
habitability dose criterion, some have established more restrictive requirements. In performing a
Level 3 PSA or PCA, it is common practice to use the site-specific value for the long-term
habitability dose criterion for the base case analysis. Sensitivity analyses may then be performed
to evaluate the impact on results of using alternative values.

STATUS OF PRACTICE FOR LEVEL 3 PROBABILISTIC SAFETY ASSESSMENT